Thomas T. Martin I - Amazon S3 · due to the seismic event indicated by PP&L to be within the licensing basis, clearly do not meet the design or licensing basis requirements if the

November 27, 1992

M r . Thomas T. M a r t i n Reg iona l A d m i n i s t r a t o r , Region I U n i t e d S t a t e s Nuc lea r R e g u l a t o r y Commission 475 A l l e n d a l e Road K ing o f P r u s s i a , P A 19406-1415

SUBJECT: SUSQUEHANNA STEAM ELECTRIC STATION DOCKET NO. 50-387 LICENSE NO. NPF-14 10CFR21 REPORT OF SUBSTANTIAL SAFETY HAZARD

Dear M r . M a r t i n :

Pursuan t t o t h e requ i remen ts o f 10CFR21, R e p o r t i n g o f D e f e c t s and Noncompliance, t h i s l e t t e r i s s u b m i t t e d t o r e p o r t a " s u b s t a n t i a l s a f e t y h a z a r d " t h a t e x i s t s i n t h e d e s i g n o f t h e Susquehanna Steam E l e c t r i c S t a t i o n (SSES) l o c a t e d nea r Berw ick , Pennsy lvan ia . T h i s r e p o r t i s b e i n g made by M r . D a v i d A. Lochbaum who, t h r o u g h J u l y o f t h i s yea r , worked as a c o n t r a c t eng inee r i n Pennsy l van ia Power & L i g h t Company's ( t h e l i c e n s e e ) Nuc lea r P l a n t E n g i n e e r i n g S e c t i o n , and M r . Dona ld C . P r e v a t t e who i s c u r r e n t l y , and u n t i l t h e end o f t h i s year , w o r k i n g as a c o n t r a c t eng ineer i n PP&L's Nuc lear P l a n t E n g i n e e r i n g S e c t i o n .

The s u b s t a n t i a l s a f e t y hazard i s as f o l l o w s : The SSES d e s i g n f o r a l o s s o f normal spen t f u e l p o o l c o o l i n g f a i l s t o meet numerous r e g u l a t o r y requ i remen ts . As a r e s u l t , f o r a d e s i g n b a s i s a c c i d e n t , t h e r e i s t h e p o t e n t i a l f o r mel tdown o f i r r a d i a t e d f u e l o u t s i d e p r i m a r y con ta inment and t h e f a i l u r e o f a l l s a f e t y - r e l a t e d systems i n t h e r e a c t o r b u i l d i n g .

For an o p e r a t i n g p l a n t , 10CFR50.72 r e q u i r e s l i c e n s e e s t o r e p o r t i n one hou r any i n s t a n c e o f t h e p l a n t ( a ) b e i n g i n an unana l yzed c o n d i t i o n t h a t s i g n i f i c a n t l y compromises p l a n t s a f e t y , ( b ) i n a c o n d i t i o n t h a t i s o u t s i d e t h e d e s i g n b a s i s o f t h e p l a n t , o r ( c ) i n a c o n d i t i o n n o t covered b y t h e p l a n t ' s o p e r a t i n g and emergency procedures. I t a l s o r e q u i r e s t h a t r e p o r t s s h a l l be made w i t h i n f o u r h o u r s o f any c o n d i t i o n t h a t a l o n e c o u l d have p reven ted t h e f u l f i l l m e n t o f t h e s a f e t y f u n c t i o n o f s t r u c t u r e s o r systems needed t o ( a ) s h u t down t h e r e a c t o r and m a i n t a i n s a f e shutdown, ( b ) remove r e s i d u a l hea t , ( c ) c o n t r o l r a d i o a c t i v e r e l e a s e , o r ( d l m i t i g a t e t h e a c c i d e n t . A l l o f t h e s e c o n d i t i o n s e x i s t a t SSES f o r t h e d e s i g n b a s i s a c c i d e n t (DBA) l o s s - o f - c o o l a n t a c c i d e n t (LOCA) o r LOCA w i t h a l o s s - o f - o f f s i t e - p o w e r (LOOP) as a r e s u l t o f t h e hea tup o f t h e spen t f u e l p o o l wh i ch m e c h a n i s t i c a l l y f o l l o w s t h e s e a c c i d e n t s .

On October 9, 1992, after seven months of attempts to convince PP&L1s management to address these concerns as required, the signatories to this letter declared to PP&L management our intent to report these concerns to the NRC ourselves unless they were properly handled by PP&L. I n response to our declaration and the actions it initiated, Pennsylvania Power & Light Company submitted Licensee Event Report (LER) 92-016-00 to the Nuclear Regulatory Commission on November 17, 1992. Although PP&L1s report acknowledged that concerns had been raised, it dismissed them as having minimal safety significance. The LER is incomplete, inaccurate, unbalanced and misleading in its presentation of our concerns, the pertinent technical and licensing information, and its conclusions. The purpose of this letter is to inform the NRC that we still consider these concerns to be a "substantial safety hazard" which should have been reported by PP&L under 10CFR50.72.

The focus of our concerns is the inability to remove decay heat from the spent fuel pools for the various design events which mechanistically incapacitate the normal fuel pool cooling system and the resultant effects from loss of normal cooling on the safety-related systems and components in the reactor buildings.

The heart of the PP&L position stated in LER 92-016-00 is a legalistic argument that the licensing basis of SSES does not require the loss of normal fuel pool cooling to be considered concurrently with other design basis events such as LOCA or LOCA/LOOP. We agree that loss of normal spent fuel pool cooling is not required to be postulated concurrently, but when it follows mechanistically as a result of the design basis events as it does at SSES, it must be considered.

PP&L cites in the LER as support for its position FSAR Section 9.1.3 and Appendix 9A which it contends contain the only design basis requirements for the fuel pool cooling failure which must be considered - basically, failure due to a seismic event. We contend that there.are other conditions within the SSES licensing basis as described throughout the FSAR which will mechanistically cause failure of the non-safety related normal fuel pool cooling system. such as hvdrodvnamic loads associated with a LOCA. - env the fai imp the

We

ronmental condifions -associated with a LOCA, LOOP, failure of non-safety related service water system, and random, single ures. In 1988, PP&L introduced another failure mode when it emented procedures to manually de-energize non-1E loads in reactor building following a LOCA without a LOOP.

contend that as with all other systems described in the FSAR. the design and operation of the fuel pool cooling system cannot be taken out of context of its mechanistic relationships with the other systems, events, and licensing bases without review and approval by the NRC. The current design and operation of SSES for a loss of normal spent fuel pool cooling, even for failure

Page 2

due to the seismic event indicated by PP&L to be within the licensing basis, clearly do not meet the design or licensing basis requirements if the effects on safety related structures, systems and components in the reactor buildings are considered.

The design basis accident for SSES is the LOCA with a concurrent LOOP. For this event, it must be assumed that the normal fuel pool cooling system will fail as described above. Therefore, the removal of decay heat from the spent fuel pools must be accomplished by the design basis method (and only safety related method available) described in Section 9.1.3 of the FSAR; allowing the fuel pool to boil and providing makeup from the safety related emergency service water ( E S W ) system. However, at the present time there are no design provisions, analyses or procedures which adequately define how, within the applicable regulatory requirements, this function will be accomplished. The effects of the boiling spent fuel pools on safety related equipment in the reactor buildings are also unanalyzed. These deficiencies exist even for the loss of spent fuel pool cooling event described in FSAR Appendix 9A.

When these concerns are addressed within the context of the regulatory requirements, it appears that the necessary steps to provide makeup water from the ESW system following a design basis LOCA cannot be performed due to very high radiation in the areas where valves must be manually operated. Additionally, the current EQ temperature limits of virtually all of the safety related equipment in the reactor building will probably be exceeded by a large margin due to the heat and moisture put into the reactor building atmosphere by the boiling spent fuel pools.

To appreciate the significance of these concerns, the magnitude of the potential effects for the design basis accidents m u s t be considered:

1) The currently calculated radiation levels at some of the ESW valves which must be manipulated are in the thousands of R/hour, not including the associated airborne dose which may be in the hundreds of R/hr.

2) The boiling fuel pools will add approximately 20 million BTU/hr of latent heat to the reactor building atmosphere which is not currently accounted for in the calculations. The total heat load in the reactor building that is currently accounted for is only 5.2 million BTU/hr, and even at this heat load there are a number of areas where the accident temperatures slightly exceed the EQ temperatures.

3 ) At the design makeup rate from ESW to the fuel pools, 5.2 million gallons of water are introduced into the reactor buildings (both the accident and non-accident units will be affected) either through e v a p o r a t i o n / c o n d e n s a t i o n or

Page 3

s p i l l o v e r o f t h e p o o l s . None o f t h e p o s s i b l e d e t r i m e n t a l e f f e c t s o f t h i s w a t e r o n t h e s a f e t y r e l a t e d s t r u c t u r e s , sys tems and componen ts i n t h e r e a c t o r b u i l d i n g s have b e e n a n a l y z e d . I n f a c t , P P & L ' s own r e c e n t e n g i n e e r i n g e v a l u a t i o n f o r t h e s e c o n c e r n s d e t e r m i n e d t h a t t h e s t a n d b y gas t r e a t m e n t s y s t e m w o u l d i s o l a t e due t o h i g h i n l e t t e m p e r a t u r e .

I n a d d i t i o n t o t h e s e m o s t s i g n i f i c a n t s a f e t y c o n c e r n s , t h e r e a r e a l s o r e l a t e d c o n c e r n s o f l e s s e r s a f e t y s i g n i f i c a n c e w h i c h n o n e t h e l e s s c o n s t i t u t e " s u b s t a n t i a l s a f e t y h a z a r d s " . These i n c l u d e t h e f o l l o w i n g :

4 ) F u e l p o o l i n s t r u m e n t a t i o n f o r m o n i t o r i n g t h e c o o l i n g o f t h e f u e l p o o l post -LOCA ( a s a f e t y r e l a t e d f u n c t i o n ) i s n o t e n v i r o n m e n t a l l y q u a l i f i e d , and t h e r e a d o u t s a r e l o c a t e d i n an a r e a w h i c h i s n o t a c c e s s i b l e t o t h e o p e r a t o r s post-COCA.

5 ) The d e s i g n h e a t l o a d s a n d t h e c a l c u l a t e d t i m e s t o b o i l f o r t h e s p e n t f u e l p o o l s have n o t b e e n u p d a t e d t o r e f l e c t changes t h a t have b e e n made i n t h e f u e l d e s i g n , f u e l c y c l e l e n g t h , and r e f u e l i n g p r o c e d u r e s .

F o l l o w i n g o u r O c t o b e r 9, 1992 d e c l a r a t i o n o f i n t e n t t o r e p o r t t o t h e NRC o n t h e s e c o n c e r n s , t h e r e e n s u e d l a r g e s c a l e e f f o r t s w i t h i n t h e PP&L N u c l e a r D e p a r t m e n t t o a n a l y z e t h e c o n c e r n s a n d d e f i n e t h e a c t i o n s n e e d e d t o b e t a k e n . T h i s a c t i v i t y p r o d u c e d a n e n g i n e e r i n g r e p o r t , ME-92-002 ( a t t a c h e d ) . T h i s r e p o r t d e s c r i b e d e x t e n s i v e m o d i f i c a t i o n s a n d p r o c e d u r e changes r e q u i r e d f o r SSES t o c o p e w i t h a l o s s o f n o r m a l f u e l p o o l c o o l i n g e v e n t .

A l t h o u g h t h e r e p o r t a d d r e s s e d many o f o u r c o n c e r n s , i t d i d not a d e q u a t e l y a d d r e s s a l l o f them, a n d some o f t h e p r o p o s e d s o l u t i o n s a r e e i t h e r t e c h n i c a l l y i n a d e q u a t e a n d / o r t h e y d o n o t meet r e g u l a t o r y r e q u i r e m e n t s . I n g e n e r a l , t h e p r o p o s e d s o l u t i o n s a r e n o t a c c e p t a b l e f o r t h e f o l l o w i n g r e a s o n s :

1 ) They p l a c e h e a v y r e l i a n c e o n n o n - s a f e t y r e l a t e d e q u i p m e n t a n d f u n c t i o n s .

2 ) They p l a c e h e a v y r e l i a n c e o n p l a n t m o d i f i c a t i o n s w h i c h h a v e n o t y e t b e e n i m p l e m e n t e d o r e v e n d e s i g n e d .

3 ) They p l a c e h e a v y r e l i a n c e on p r o c e d u r e changes wh n o t y e t b e e n made.

i c h h a v e

4 ) They p l a c e h e a v y r e l i a n c e on a n a l y s e s w h i c h h a v e b e e n p e r f o r m e d .

n o t y e t

5) They p l a c e h e a v y r e l i a n c e o n o p e r a t o r and EOF p e r s o n n e l t r a i n i n g w h i c h h a s n o t y e t b e e n d e v e l o p e d o r p e r f o r m e d .

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They p l a c e heavy r e l i a n c e on o p e r a t o r a c t i o n s f o l l o w i n g a LOCA when t h e r e i s a l r e a d y heavy dependence on o p e r a t o r a c t i o n s and m o n i t o r i n g , and t hese a d d i t i o n a l a c t i o n s must be pe r fo rmed under e x t r e m e l y adverse e n v i r o n m e n t a l c o n d i t i o n s .

They r e l y on assessments of o p e r a t o r a c c e s s i b i l i t y t o t h e r e a c t o r b u i l d i n g w h i c h i n t u r n a r e based on assumpt ions o f c o r e damage w h i c h a r e unrev iewed b y t h e WRC and a r e s u b s t a n t i a l l y l e s s t h a n t h e assumpt ions r e q u i r e d b y NUREG-0737 and t h e S S E S l i c e n s i n g b a s i s r e f l e c t e d i n Chap te r 18 o f t h e FSAR. A d d i t i o n a l l y , t h e a c c e s s i b i l i t y p o s i t i o n t a k e n i n t h e r e p o r t w i t h r e s p e c t t o a i r b o r n e r a d i o a c t i v i t y c o n t r i b u t i o n s i s i n c o n s i s t e n t w i t h t h e r e q u i r e m e n t s i n NUREG-0737, 10CFR50 Appendix J, a c t u a l S S E S Appendix J t e s t r e s u l t s , and t h e d e s i g n o f o t h e r p l a n t systems (e .g . secondary c o n t a i n m e n t and t h e s tandby gas t r e a t m e n t s y s t e m ) . For NRC mandated DBA c o n d i t i o n s , as s t a t e d i n FSAR Chap te r 18, t h e r e a c t o r b u i l d i n g i s i n a c c e s s i b l e f o r days f o l l o w i n g a LOCA.

They r e l y on p r o b a b i l i t y arguments w h i c h may be a c c e p t a b l e i n an I n d i v i d u a l P l a n t E v a l u a t i o n and i n a j u s t i f i c a t i o n f o r i n t e r i m o p e r a t i o n , b u t w h i c h a r e n o t a c c e p t a b l e s u b s t i t u t e s f o r comp l iance w i t h r e g u l a t o r y r e q u i r e m e n t s , u n l e s s t h e y a r e r e v i e w e d and approved b y t h e NRC. These have n o t been.

I n some a reas , t h e r e p o r t ' s c o n c l u s i o n s a r e i n c o n s i s t e n t w i t h t h e f a c t s p resen ted . For example, t h e r e p o r t c o n c l u d e d t h a t Zone I11 v e n t i n g i s a c c e p t a b l e , whereas t h e s u p p o r t i n g documen ta t i on i n d i c a t e s t h a t t h e lOCFRlOO and lOCFR50 Appendix A C r i t e r i o n 19 a l l o w a b l e s f o r o f f s i t e and c o n t r o l room doses r e s p e c t i v e l y a r e exceeded.

I t s h o u l d a l s o be c o n s i d e r e d t h a t t h e c o n c l u s i o n s i n t h i s r e p o r t r e p r e s e n t PP&L's v i s i o n o f systems. equ ipment and p r o c e d u r e s i n t h e f u t u r e , n o t as t h e y e x i s t today . A l t h o u g h e s s e n t i a l l y none o f t h e t e c h n i c a l i n f o r m a t i o n f r o m t h e r e p o r t i s c o n t a i n e d i n t h e i r LER, t h i s i n f o r m a t i o n a l o n g w i t h t h e i r l e g a l i s t i c arguments d i s c u s s e d e a r l i e r . has p r o v i d e d t h e u n d e r l y i n g bases f o r t h e i r d e t e r m i n a t i o n s o f o p e r a b i l i t y and r e p o r t a b i l i t y . B u t t h e l a w r e q u i r e s d e t e r m i n a t i o n s o f o p e r a b i l i t y and r e p o r t a b i l i t y t o b e based on t h e p l a n t c o n d i t i o n s as t h e y e x i s t a t t h e t i m e o f d i s c o v e r y as d i s c u s s e d i n c o n s i d e r a b l e d e t a i l i n NURE6-1022.

I n a d d i t i o n t o t h e s e t e c h n i c a l concerns, we a l s o must p o i n t o u t t h e c o n d i t i o n s adve rse t o q u a l i t y t h a t PP&L1s h a n d l i n g o f t h i s case (and o t h e r r e c e n t s a f e t y c o n c e r n s ) demons t ra tes i n v i o l a t i o n o f lOCFR50 Appendix B. S i n c e o u r conce rns were f i r s t d i s c o v e r e d and r e p o r t e d i n March o f t h i s yea r , t h e r e has been a p rog rammat i c f a i l u r e b y PP&L t o p r o p e r l y e v a l u a t e t h e s e concerns . PP&L r e p e a t e d l y a t t e m p t e d t o i m p r o p e r l y d i s m i s s t h e s e concerns ca i n d e f i n i t e l y d e f e r t h e i r e v a l u a t i o n b y methods i n c T t t m

Page 5

classifying them as design basis document issues (in clear violation of the guidelines expressed i n NUREG-1397), selectively applying regulatory requirements to permit favorable conclusions, claiming that the NRC had already reviewed and approved the design deficiencies based on the FSAR/SER text, and even claiming that an informal, undocumented agreement had been made with the NRC at the time of initial licensing of the plant. Our experience and our knowledge of the difficulties encountered by other engineers with nuclear safety concerns for SSES indicates that PP&L1s program for handling nuclear safety issues is itself cause for concern.

While PP&L cites data to support their contention that their discrepancy management system is effective, most of their data points represent relatively minor discrepancies which are easy to resolve. However, for large problems with extensive or uncertain resolution such as in this case, the system lacks the ability to assure proper evaluation and subsequent implementation. In these cases, PP&Lis treatment violates their own administrative procedures controlling discrepancy management.

In the nuclear power industry, organizations such as PP&L and individuals such as ourselves have legal and ethical responsibilities. PP&L has not fulfilled its responsibilities in this case and has forced us to fulfill ours by submitting this letter.

The more detailed technical descriptions for these concerns and the history of their treatment are contained in numerous letters, memos, and documents. A listing of pertinent documents is contained in Attachment 1 to this letter, with copies of these documents provided as the remaining attachments to this letter.

We expect that you may require additional information from us regarding this matter. We will make every effort to support your requirements in a timely manner. We can be reached at the addresses and telephone numbers listed below. We would also greatly appreciate being kept informed of your actions regarding this matter.

Thank you for your consideration.

David It/. Lochbaum

80 Tuttle Road Watchung, NJ 07060 (908) 754-3577

Sincerely,

Donald C. Prevatte

7924 Woodsbluff Run Fogelsville, PA 18051 (215) 398-9277

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Distribution List

Mr. Thomas T. Martin (with all attachments)

Mr. G. S. Barber (with all attachments) Senior Resident Inspector US Nuclear Regulatory Commission P.O. Box 35 Berwick, PA 18603-0035

US Nuclear Reaulatorv Cornrni ssion ~ttention: ~ocument- Control Clerk Mail Station PI-137 Washington, DC 20555 (with all attachments)

Director, Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 (with Attachment 1)

The Honorable Ivan Selin (with Attachment 1 1 Chairman US Nuclear Regulatory Commission Washington, DC 20555

The Honorable Kenneth C. Rogers (with Attachment 1) Commissioner US Nuclear Regulatory Commission Washington, DC 20555

The Honorable James R. Curtiss (with Attachment 1 ) Commissioner US Nuclear Regulatory Commission Washington, DC 20555

The Honorable Forrest J. Remick (with Attachment 1) Commissioner US Nuclear Regulatory Commission Washington, DC 20555

The Honorable Gail De Planque (with Attachment 1) Commissioner US Nuclear Regulatory Commission Washington, DC 20555

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No. 1

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5

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Attachment 1

Attachment List of Attachments

PP&L Memo from Dave L evatte to Mark Mjaatvedt, "Susquehanna Steam Electric Station Spent Fuel Pool Boiling Issues", March 19, 1992 (ET-0149)

List o f Attachments

ochbaum and Don Pr

PP&L Engineering Discrepancy Report, "Loss of Spent Fuel Pool Cooling Event Design Discrepancies" Originated April 16, 1992 and Oispositioned October 6, 199; (EDR G20020)

PP&L Operability Statement, "EDR #G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies", April 23, 1992

PP&L Memo from Dave Lochbaum and Don Prevatte to Joe Zola, "Supplemental Information for EDR G20020 on Boiling Spent Fuel Pool", June 22, 1992 (ET-0471)

PP&L Draft Screening Worksheet prepared by Art White, "EDR No. G20020", July 1 , 1992

Handout, "EDR 620020 References", July 15, 1992

White Paper prepared by David A. Lochbaum and Donald C. Prevatte, "Safety Consequences of a Boiling Spent Fuel Pool at the Susquehanna Steam Electric Station", July 27, 1992

PP&L Memo from 6. D. Miller to G. T. Jones, "Fuel Pool Cooling Deficiencies", August 18, 1992 (ET-0586)

PP&L Memo from D. C. Prevatte to G. T. Jones, "Fuel Pool Cooling Deficiencies", August 20, 1992 (ET-0587)

PP&L Memo from A. Dyszel to T. C. Dalpiaz, "U2 R105 Fuel Pool Decay Heat Evaluation", August 21, 1992 (PLI-72230)

PP&L Memo from J. M. Kenny to G. T. Jones and C. A. Myers, "EDR on Fuel Pool Cooling", August 25, 1992

PP&L Memo from George T. Jones to Glenn D. Miller, "Fuel Pool Cooling EDR's G20020, G00005", August 27, 1992 (PLI-72267)

PP&L Memo from Glenn D. Miller to George T. Jones, "Fuel Pool Cooling EORs G20020, G00005", August 31, 1992 (PLI-72297)

PP&L Memo from Kevin W. Brinckman to Georcle T. Jones, "Review of Fuel Pool Cooling", september 1 , 1992 (PLI-72288)

At tachmen t 1 L i s t o f A t t a c h m e n t s ( c o n t i n u e d )

No. At tachmen t 16 PP&L Memo f r o m J . R . M i l t e n b e r g e r t o G. T . Jones, "Spent

F u e l Poo l C o o l i n g " , September 9, 1992 ( P L I - 7 2 3 6 7 )

17 PP&L L e t t e r f r o m James E. Agnew t o D a v i d A . Lochbaum, " E D R 620020, Spent Fue l Pool D e s i g n D i s c r e p a n c i e s " , O c t o b e r 7, 1992 (ET-0785)

18 PP&L Memo f r o m G . D . M i l l e r t o G . D. M i l l e r , "Ass ignment o f E D R " , Oc tobe r 7, 1992 (ET-0780)

19 L e t t e r f r o m D a v i d A . Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, " R e p o r t a b i l i t y o f B o i l i n g Spent F u e l Poo l Concerns" , Oc tobe r 9, 1992

20 PP&L Memo f r o m D. A . Lochbaum and D. C . P r e v a t t e t o Georae T. Jones, " E D R System Concerns" , Oc tobe r 13, 1952 ( P L I - 7 2 3 6 5 )

21 PP&L Memo f r o m George T. Jones t o G. D . M i l l e r , "Spent F u e l Poo l I s s u e " , Oc tobe r 14, 1992 ( P L I - 7 2 6 4 0 )

22 PP&L Memo f r o m George T . Jones t o G. D . M i l l e r , J. S. S t e f a n k o and M. W. Simpson, "Spent F u e l Poo l C o o l i n g I s s u e " , Oc tobe r 14, 1992 ( P L I - 7 2 6 4 1 )

23 PP&L Memo f r o m George T. Jones t o A l l N u c l e a r E n g i n e e r i n g Managers and S u p e r v i s o r s , " E n g i n e e r i n g D i s c r e p a n c y (EDR) Program", Oc tobe r 14, 1992

24 L e t t e r f r o m D a v i d A . Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, "D isagreement w i t h S c r e e n i n g , R e p o r t a b i l i t y and O p e r a b i l i t y E v a l u a t i o n s f o r EDR G20020", Oc tobe r 14, 1992

25 Memo f r o m C h a r l e s A. Myers t o George T. Jones, " F u e l Poo l C o o l i n g I s s u e s - R e p o r t a b i l i t y / O p e r a b i l i t y " , Oc tobe r 20, 1992

26 PP&L Memo f r o m Glenn D. M i l l e r t o George T. Jones, " E v a l u a t i o n o f EDR 620020 - Spent F u e l C o o l i n g I s s u e " , Oc tobe r 21, 1992 (PL I -72711 )

27 PP&L Memo f r o m D a v i d A . Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, " E v a l u a t i o n o f EDR 620020 Reportability/Operabilityl', Octobe r 26, 1992 ( P L I - 7 2 7 3 9 )

28 PP&L Memo f r o m D a v i d A . Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, "Response t o E v a l u a t i o n o f EDR G2002OU, Oc tobe r 28, 1992 (PL I -72751 )

At tachmen t 1 L i s t o f A t t a c h m e n t s ( c o n t i n u e d )

No. 29

30

3 1

3 2

33

34

3 5

A t tachmen t PP&L Memo f r o m Glenn D. M i l l e r t o Georse T. Jones , " E v a l u a t i o n o f EDR 620020 - Spent F u e l Pool c o o l i n g issue"; October 29, 1992 ( P L I - 7 2 7 6 3 )

PP&L E n g i n e e r i n g R e p o r t , " Loss o f F u e l Poo l C o o l i n g E v e n t E v a l u a t i o n f o r EDR #G20020", Oc tobe r 29, 1992 (NE-92-002 Rev. 0 )

PP&L Memo f r o m G lenn D . M i l l e r t o George T. Jones, " R e v i s e d E v a l u a t i o n o f EDR 620020 - Spent F u e l Pool C o o l i n g I s s u e " , Oc tobe r 29, 1992 (PL I -72764 )

PP&L Memo f r o m D a v i d A . Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, " P o s i t i o n on EDR 620020 and P lanned A c t i o n s " , November 2, 1992 ( P L I - 7 2 7 8 3 )

PP&L Memo f r o m D a v i d G. K o s t e l n i k and Mark R . M j a a t v e d t t o George T. Jones, "Comments on PLI -72783 R e g a r d i n g EDR G20020", November 11, 1992 (PL I -72857 )

PP&L L e t t e r f r o m H. G . S t a n l e y t o t h e U.S . N u c l e a r R e g u l a t o r y Commission, " L i c e n s e e Event R e p o r t 92-016-OO", November 17, 1992 (PLAS-546)

PP&L S a f e t y E v a l u a t i o n Summary, " P r o c e d u r e EO-IP-055" , 1988 (SER NO. 88 -127 )

A t t a c h m e n t 2

PP&L Memo f r o m Dave Lochbaum and Don P r e v a t t e t o Mark M j a a t v e d t , "Susquehanna Steam E l e c t r i c S t a t i o n Spent F u e l Poo l B o i l i n g I s s u e s " , March 19, 1992 (ET-0149)

Note: T h i s memo documents t h e d i s c o v e r y o f t h e p rob lems w i t h t h e l o s s o f norma l s p e n t f u e l p o o l c o o l i n g e v e n t and t h e r e p o r t i n g o f t h e s e p rob lems t o a s u p e r v i s o r i n t h e PP&L N u c l e a r P l a n t E n g i n e e r i n g S e c t i o n . A p p r o x i m a t e l y f o u r weeks l a t e r , t h e a u t h o r s o f t h i s memo were d i r e c t e d t o i n i t i a t e an E n g i n e e r i n g D i s c r e p a n c y R e p o r t on t h e conce rns . PP&L ' s d e c i s i o n t o g e n e r a t e an EDR on t h e s e c o n c e r n s may have been d r i v e n b y schedu le i n t e r e s t s - t h e a u t h o r s , as p r e p a r e r and t e c h n i c a l r e v i e w e r o f r e a c t o r b u i l d i n g h e a t l o a d c a l c u l a t i o n s t o s u p p o r t t h e PP&L Power U p r a t e P r o j e c t , w o u l d n o t s i g n o f f on t h e c a l c u l a t i o n s u n t i l t h e s e conce rns were addressed. Upon g e n e r a t i o n o f t h e EDR, t h e a u t h o r s s i g n e d o f f t h e c a l c u l a t i o n s c o n d i t i o n a l l y w i t h a n o t e t h a t t h e r e s u l t s m i g h t be a f f e c t e d b y t h e d i s p o s i t i o n o f t h e EDR. PP&L needed t h e s e c a l c u l a t i o n s i s s u e d i n o r d e r t o s u b m i t t h e i r e n g i n e e r i n g r e p o r t o n power u p r a t e t o t h e NRC i n June 1992.

TO : WWMjaatvedt DATE: March 19, 1992

FROM : Dave Lochbaum Don Prevatte

JOB : Power Uprate NUMBER: ET-0149 COPIES: Distribution

FILE: P88-1 REPLY: No

SUBJECT: SUSQUEBANNA STEAM ELECTRIC STATION SPENT FUEL POOL BOILING ISSUES

Potential problems resulting from a boiling spent fuel pool have been uncovered during the preparation of the engineering evaluation of the fuel pool cooling system and the reactor building heat load calculations at power uprate conditions. These problems, their brief history and recommendations to resolve these issues are presented in the attachment to this letter.

The consequences of these problems will be made incrementally worse by power uprate and therefore must be addressed before implementation of power uprate. More importantly, however, these problems affect the current operation of SSES and must be evaluated and resolved as soon as possible.

If additional information is needed, please contact Dave Lochbaum at ETN 220-7768 or Don Prevatte at ETN 220-7781.

Ucud U ' David A." Lochbaum

DISTRIBUTION:

\AC .%ci3iK Donald C. Prevatte

J A Bartos A6-3 (w/a) M B Detamore A6-2 (w/a) J M Werley A6-3 (w/a) SRMS Corres File A6-2 (w/a)

Page 2 of 10

Afta~hnent : Boiling Spent Fuel Pool Issues

I. REACTOR BUILDING BEAT LOADS

Problu:

Reactor building design heat loads do not account for the boiling spent fuel pool event.

History:

The calcs of reactor building pre-uprate and uprate heat loads for Zone I, I1 and I11 under normal and accident conditions (calcs M-RAF-052, -053, -054) assume the spent fuel pool temperature remains at 125OF for all cases. This assumption relies upon use of the service water system to remove heat from the fuel pool heat exchangers post-LOCA and the fuel pool cooling assist mode of RHR to remove heat from the fuel pool post-LOOP. Neither of these operating modes is safety related and therefore may not be available.

The design provision for the loss of fuel pool cooling event is to permit the fuel pool to boil and use ESW to maintain the level in the pool above the top of the fuel. ESW provides redundant seismic Category I makeup lines to each of the two spent fuel pools.

If the spent fuel pool is permitted to boil, the heat loads in the reactor building, particularly in Zone 111, increase significantly. These higher heat loads have not been considered in reactor building analyses to date. The equipment qualification of safety-related equipment in the reactor building may therefore be adversely affected if the heat loads from a boiling spent fuel pool are considered.

Recommendation:

An EDR was prepared on this condition. The options available to resolve this problem include:

1) Analyzing the reactor building heat loads for the boiling spent fuel pool case and update associated analyses for equipment qualification.

2) Providing design capability to maintain spent fuel pool temperature ( 125OF using safety-related equipment such that existing reactor building heat load analyses are adequate.

Page 3 of 10

- Attachment : Boiling Spent Fuel Pool Issues

FUEL POOL TIME-TO-BOIL AND RADIOLOGICAL RELEASE ANALYSES

First Problem:

The analytical 25 hour time-to-boil for the spent fuel pool is nonconservative for the maximum normal heat load in the spent fuel pool.

History:

Bechtel calc 200-0048 Rev. 1, "Boiling Spent Fuel Pool" dated May 7, 1982, determined time-to-boil using the equation:

Time-to-Boil = (m * C * OT) / Q, where

m = mass of water in fuel pool, lb

C = specific heat, BTU/lb-OF

OT = difference between final pool temperature (212OF) and initial pool temperature (l2S0F), OF

Q = fuel pool decay heat load, BTU/hr

This calc used a decay heat load of 9.79~10~ BTU/hr for Unit 1 and 7.92~10~ BTU/hr for Unit 2 to determine times-to-boil of 25.15 hours and 31.087 hours respectively.

FSAR 9.1.3.1 establishes the maximum normal heat load as that heat load resulting from 2840 assemblies discharged to the fuel pool by a routine refueling schedule. FSAR Tables 9.1-2a and 9.1-2b report the maximum normal heat load for Units 1 and 2 as 12.6~10~ BTU/hr. These values were determined in Bechtel calc 153-9 Rev. 1, "Fuel Pool Decay Energy and Temperature".

The spent fuel pool decay heat values used in the boiling spent fuel pool calc and for the FSAR discussion were based upon assumptions for cycle operating lengths, fuel exposures, and reactor power level. SSES has subsequently operated differently than had been assumed such that the decay heat loads in the filled spent fuel pools may exceed 9.79~10~ BTU/hr, resulting in a shorter than analyzed time-to-boil.

Calc NFE-B-NA-053 Rev. 0, "Decay Heat from a Full Spent Fuel Pool (ASB9-2 Method)I1, determined decay heat from a filled spent fuel pool using actual fuel operating history through 1991 and assumptions which bound operation after power uprate. This recent calc reported a maximum normal heat load of

Page 4 of 10

AItachrent : Boiling Spent Fuel pool Issues

=17x106 BTU/hr. The methodology use in this calculation is conservative and may over predict actual decay heat loads by ~20%.

Calc M-FPC-009 was drafted to determine the spent fuel pool time-to-boil and required ESW makeup rate for power uprate. Preliminary results from this calc indicate the fuel pool boils 19.4 hours after loss of fuel pool cooling for the design heat removal capacity of the fuel pool cooling system (13.2~10~ BTU/hr). This calc determined a time-to-boil of ~15.5 hours for the ~ 1 7 x 1 0 ~ BTU/hr heat load calculated for the power uprate case.

Recommendation:

The basis for the time-to-boil analysis should not be the maximum normal heat load, since this value is subject to assumptions of reactor operation which are extremely difficult to predict. Two options are proposed:

1) The time-to-boil analysis for the loss of normal spent fuel pool cooling case should use the design capacity of the fuel pool cooling system since this value bounds any normal heat load stored in the fuel pool. If the normal heat load in the fuel pool exceeded 13.2~10~ BTU/hr, then modifications to the fuel pool cooling system would be necessary to enable the system to maintain pool temperature less than 125OF.

2) The time-to-boil analysis for the loss of normal spent fuel pool cooling case should use a range of spent fuel pool decay heat loads up to at least the design capacity of the fuel pool cooling system. This method bounds any maximum normal heat load for the fuel pool while limiting overly conservative times in the years while the fuel pool is partially filled. Basically, this method provides time-to-boil as a function of decay heat load in the spent fuel pool. This relation can be used for more realistic time-to-boil for current conditions if actual decay heat load in the spent fuel pool is known.

Page 5 of 10

Atrtachment : Boiling Bpent Fuel Pool Issues

Becond Problem:

The analytical 25 hour time-to-boil for the spent fuel pool does not account for the emergency heat load in the spent fuel pool.

History:

FSAR 9.1.3.1 establishes the emergency heat load for the spent fuel pool as that heat load following a full core offload which completely fills the fuel pool. The FSAR specifies the emergency heat load to be 32.6~10~ BTU/hr.

Bechtel calc 200-0048 Rev. 1, "Boiling Spent Fuel Pool" dated May 7, 1982, determined time-to-boil for the maximum normal heat load case only. The actual decay heat load in the spent fuel pool exceeds the maximum normal heat load during every refueling outage at SSES in which the core is fully offloaded.

SSES currently imposes administrative controls during refueling outages when the core is fully offloaded into the spent fuel pool to reduce the potential for loss of fuel pool cooling. Decay heat is removed from the spent fuel pool during these periods by RHR shutdown cooling (when the fuel pool to reactor cavity gates are removed) and by cross-tieing the operating unit's fuel pool cooling system to the outage unit's fuel pool. However, a seismic event in this configuration could cause loss of fuel pool cooling at a time when the time-to-boil is significantly less than 25 hours.

Calc M-FPC-009 was drafted to determine the spent fuel pool time-to-boil and required ESW makeup rate for power uprate. Preliminary results from this calc indicate the fuel pool boils 7.9 hours after loss of fuel pool cooling for a decay heat load of 36.2~10~ BTU/hr, which is the currently analyzed emergency heat load.

The time-to-boil analysis should be expanded to include decay heat loads up to at least the design capacity of the RHR fuel pool cooling assist mode. Operating procedures, off-normal procedures and SSES outage management policies should be reviewed and revised as necessary to ensure that appropriate controls are implemented when the fuel pool decay heat load exceeds the capacity of the fuel pool cooling system and proper responses are taken in event fuel pool cooling is lost.

Page 6 of 10

ACfachmont : Boiling Bpont Fuel pool Issues

Third Problom:

The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates.

Bechtel calc 200-0048 Rev. 1, "Boiling Spent Fuel Pool8@ dated May 7, 1982, determined the evaporation rate from a boiling spent fuel pool using the equation:

Evap Rate = Q / h g - h f where

Q = fuel pool decay heat load, BTU/hr

h, = enthalpy of vapor at boiling, BTU/lb

hf = enthalpy of water at boiling, BTU/lb

This calc used a decay heat load of 9.79~10~ BTU/hr to determine evaporation rate. As reported above, the maximum normal heat load specified 'in FSAR Table 9.1-2a is 12.6~10~ BTU/hr and the emergency heat load specified in FSAR 9.1.3.1 is 32.6~10~ BTU/hr. When the decay heat load in the spent fuel pool exceeds 9.79~10~ BTU/hr, the evaporation rate from the boiling pool will exceed the rate assumed in the radiological release analysis.

SSES currently applies the 25 hour time-to-boil determined by calc 200-0048 as the criterion in deciding when to permit common RHR work during an outage. Therefore, when decay heat loads are less than 9.79~10~ BTU/hr, the time-to-boil is longer than 25 hours and the radiological release in event of loss of fuel pool cooling is bounded by the results from calc 200-0048.

The 25 hour criterion for common RHR work prevents this work from beginning prior to =Day 18-21 each outage. Since core offloading typically starts on Day 5 and is completed by Day 10 or 11, this means that for at least 7 days, the decay heat load in the spent fuel pool is significantly higher than the heat load used to derive the evaporation rate used in the radiological release analysis. The consequences from a loss of fuel pool cooling may be offset by a longer time-to-boil if the fuel pool to reactor cavity gate is removed and the fuel pools are cross-tied, but credit for the additional water inventory available cannot be taken without administrative controls and a time-to-boil analysis for this configuration.

Page 7 of 10

Attachrent : Boiling apent m e 1 pool Issues

Recommendation:

The radiological release analysis should use appropriate evaporation rates for the decay heat loads used in the associatedtime-to-boil analysis. The method and results from these analyses should be clearly stated and conveyed to SSES to ensure that adequate administrative controls are implemented during normal operation and in refuelingto ensure the radiological release analysis results bound actual plant conditions for all operating configurations.

Fourth Problem:

The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms.

History:

Bechtel calc 200-0048 Rev. 1, "Boiling Spent Fuel Pool*o dated May 7, 1982, determined the radiological release consequences from a boiling spent fuel pool. This calc assumed 12 month operating cycles and 184 bundle equilibrium reload sizes to determine the activity terms for failed fuel in the fuel pool. SSES currently has 18 month operating cycles with ~ 2 3 0 bundle reloads which will increase to =254 bundles after power uprate. Since calc 200-0048 implies that most of the activity results from the most recent discharge batch, the effect of increasing the discharge size from 184 bundles assumed in the calc to 230 and 254 bundles would appear to be nonconservative with respect to the radiological release analysis.

Reaonmendation:

The basis for the radiological release analysis should not be the projected operating conditions, since these conditions are subject to assumptions which are extremely difficult to predict. The radiological release analysis should assume conditions which will bound future actual operating conditions. For example, a reload batch size of 320 bundles was assumed in calc NFE-B-NA-053, "Decay Heat from a Full Spent Fuel Pool (ASB9-2 Method)I9, because this size represents the maximum reload batch size possible under core design criteria.

Page 8 of 10

Aeachrent : Boiling Spant Fuel Pool Issues

111. ESW I(AI[EUP TO TEE SPENT FUEL POOL

First Problem:

The impact of the ESW makeup water to the spent fuel on equipment in the reactor building has not been evaluated.

History:

The ESW system and the ultimate heat sink are designed to provide adequate makeup to the spent fuel pool for 30 days following loss of normal spent fuel pool cooling. Based on the original ESW makeup flow of 60 gpm to each fuel pool, the spray pond inventory allocates 5 million gallons of water for this purpose. However, the consequences of this quantity of water on equipment in the reactor building has not been evaluated.

EDR GO0005 was written in 1990 to address discrepancies between the spent fuel pool discussion in FSAR Chapter 9 and actual SSES operation. This EDR also questioned the ESW makeup flow to the spent fuel pool since the 60 gprn flow rate had not been demonstrated to be achievable.

Calc M-FPC-009, "Spent Fuel Pool Boiling Analysi~~~, was drafted to determine the ESW makeup flow required for the design heat removal capacity of the fuel pool cooling system (13.2~10~ BTU/hr) and for the heat removal capacity of the RHR fuel pool cooling mode (32.6~10~ BTU/hr) . These ESW makeup flows were determined to be 31.8 gpm and 67.5 gpm respectively. The interim disposition to EDR GO0005 pointed out that the maximum ESW makeup flow case occurs when all of the reactor core is offloaded to the spent fuel pool, so the higher ESW makeup flow rate could be obtained by the reduced ESW system flow required when there is no fuel in the reactor.

When the ESW makeup flow to the fuel pool exactly matches the boil-off rate from the fuel pool, that quantity of water vapor must also either exit the building via the standby gas treatment system, bring the building to 100% humidity or condense somewhere within the reactor building. When the ESW makeup flow to the fuel pool exceeds the boil-off rate, there will also be overflow once the skimmer surge tank fills and level control is lost.

Eventually in either case, the quantity of water added via the ESW system ends up going through the standby gas treatment system or as water in the reactor building. The consequences of up to 2.5 million gallons of water in each reactor building

Page 9 of 10

- Altaahnent : Boiling Spent Fuel Pool Issuea

could include flooding of the ECCS pumps rooms, inoperability of the ECCS pump room coolers, inoperability of safety related equipment due to higher than analyzed humidity and degradation of the standby gas treatment system due to moisture loading.

A comprehensive evaluation for the boiling spent fuel pool event needs to be performed which accounts for the water present in the reactor building due to boil-off and overflow from the spent fuel pool. This evaluation must address the effects of this water on the operability of systems and components in the reactor building.

Second Problem:

The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be permitted under post-LOCA conditions.

History:

Off-normal operating procedure ON-135-001, 8tLoss of Fuel Pool Cooling/Coolant Inventory", requires the operator to manually open the valves in ESW makeup line to fuel pool if all other means of adding water to the fuel pool are lost. The procedure calls for the valves to be left open until the desired water level is obtained. Since these valves are in the reactor building, it may be impossible for them to be manually operated as directed under all conditions including post-LOCA. In addition, even if the throttle valve is initially adjusted so that the ESW makeup flow to the fuel pool exactly matches the boil-off rate, the subsequent exponential decline in fuel pool decay heat load would require the throttle valve to be periodically adjusted to reduce the ESW makeup flow unless the fuel pool is permitted to overflow.

Rooorrandation:

The required operation of the ESW makeup flow valves should be evaluated from the perspective of accessibility and usage over the entire 30 day period of the boiling spent fuel pool event to ensure that all necessary valve manipulations can be made.

Page 10 of 10

- Aeachment : Boiling Bpent Fuel Pool Issues

Third Problu:

The instrumentation available to the operator post-LOCA may not provide adequate indication of spent fuel pool temperature and level to allow proper response to a loss of fuel pool cooling event.

History:

Off-normal operating procedure ON-135-001, nLoss of Fuel Pool Cooling/Coolant Inventory", requires the operator to manually open the valves in ESW makeup line to fuel pool if all other means of adding water to the fuel pool are lost. The operator enters this procedure upon annunciation of low level in the spent fuel pool or high temperature in the fuel pool cooling system. Each spent fuel pool has temperature indication (TE-15333) and level indication (LT-15332). Each skimmer surge tank has level indication (LT-15312). The skimmer surge tank piping to the fuel pool heat exchangers has temperature indication (TE-15313) and each fuel pool heat exchanger outlet piping has temperature indication (TE-15316A,B,C).

The level and temperature instruments providing these alarms may not be qualified for all conditions, such as post-LOCA, in which they would be required to function. In addition, these instruments may not be powered from class 1E sources such that they would be available post-LOOP when the fuel pool heat exchangers would be without service water.

Recommendation:

The spent fuel pool temperature and level instrumentation, as a minimum, should be verified to be or made to be qualified for all reactor building environmental conditions and required accident conditions.

Attachment 3

P P & L E n g i n e e r i n g D i s c r e p a n c y R e p o r t , "Loss o f Spent F u e l Poo l C o o l i n g E v e n t D e s i g n D i s c r e p a n c i e s " , O r i g i n a t e d A p r i l 1 6 , 1992 and D i s p o s i t i o n e d O c t o b e r 6 , 1 9 9 2 (EDR 6 2 0 0 2 0 )

72. /N/T/AAL ASSESSMENT fscreening): EDMG SUPERMSOR: - , & m c m ~ M ' rrj*

d FuGignature Dale UCENSING

%WS: YES- NO:/ PRESUMPTION OF 0PEWUTY:-

SUPERMSOR: N /A - Full Signhture Date

74. N A L * UA TION (D/spos/tion):

76. DISCREPANCY R M W COMPLETE: SUPERVISING ENGINEER, ENGR PROJECTS

/ Full Signature Date

FORM EPM-QA-1224 Rev. 3 Page 1 of 1

*- of &sds rwlred. Us8 ELM Continuation Sheet.

-

9. Potential Engineering Discrepancy (continued)

PPCL BUBQlJEuAUm STEAM ELECTRIC STATIOls EIOOIIOEEILIliGDISCaEPAlICYREPORT

PILE: BIZ-15 (cont i ruat im sheet)

the fuel pool cooling system used for normal operation and the RHR fuel pool cooling assist mode used for abnormal heat loads are not designed to satisfy seismic category I and single failure criteria. The following discrepancies for the loss of-spent fuel pool cooling event were discovered during the system evaluations for power uprate:

1. EDR NO. kJC@o 2. REV N0.O 3. Page 2 of 6

A. Reactor building design heat loads do not account for the boiling spent fuel pool event. The current calculations for reactor building Zone I, I1 and I11 heat loads assume a spent fuel pool temperature of 125OF for all cases. The reactor building heat load analyses and attendant temperature analyses upon which equipment qualification environmental parameters are based do not account for the additional heat load from boiling spent fuel pool(s). -The additional heat load could be as high as 26.4~10~ BTU/hr compared to the current maximum reactor building heat load of 5.5~10~ BTU/hr (Unit 1 LOCA case). Therefore, the design environmental conditions of safety related equipment in the reactor building may be exceeded if the heat load from the boiling spent fuel pool(s) is considered (5eea+ below

B. The impact of the ESW makeup water to the spent fuel pool on equipment in the reactor building has not been evaluated. The ESW system and the ultimate heat sink are designed to provide adequate makeup to the spent fuel pool for 30 days following loss of normal spent fuel pool cooling. Based on the original design ESW makeup flow of 60 gpm to each fuel pool, 5 million gallons of the spray pond inventory is allocated for this purpose. The water added to the spent fuel pool via the ESW system boils off and exits through the standby gas treatment system or condenses in the reactor building, or the water overflows the pool. The consequences of up to 2.5 million gallons of water in each reactor building could include flooding of ECCS pump rooms, inoperability of ECCS pump room coolers, emergency switchgear and load center room coolers and/or other safety related equipment due to higher than analyzed temperature and humidity conditions, and degradation of the standby gas treatment system due to moisture loading. The standby gas treatment system is designed for 100% relative humidity conditions in the reactor building, but a system design calculation

DRM EPM-QA-122B, Rev. 2 age 1 of 1

- - -


PPCL SUSQ- ST= ELECTRIC STATIOIP EIYOINEKBIIPG DISCBKPAUCI REPORT

PILE: BIZ-15 (Cmtinwtim sheet)

(M-SGT-015) which determined that water buildup in the ductwork before the inlet HEPA filter would not degrade system performance does not consider the potential collapse/failure of the ductwork from the weight of this water.

1. EDR No.6-0 2. REV No.0 3. Page 3 of 6

C. The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be possible. The off-normal operating procedure (ON- 135-001) requires the operator to manually open the valves in the ESW makeup line to the fuel pool if all other means of adding water to the fuel pool are lost. The procedure calls for the valves to be left open until the desired water level is obtained. Since these valves are in the reactor building, it may be impossible for them to be manually operated as .directed under all conditions including post-LOCA without unacceptable risk to the operator from the high radiation levels in the building and potentially high temperature and humidity conditions. The maximum gamma dose rates reported for EQ purposes for Unit 1 reactor building elevations 749°-111 to 818'-1" ranged between 140 and 360 R/hr (C-1815 Sh 7-10). In addition, even if the throttle valve is initially adjusted so that the ESW makeup flow to the fuel pool exactly matches the boil-off rate, the subsequent exponential reduction in fuel pool decay heat load would require the throttle valve to be periodically adjusted to lower the ESW makeup flow unless the fuel pool is permitted to overflow.

D. The instrumentation available to the operator post- LOCA does not provide adequate indication of spent fuel pool temperature and level to allow proper response to a loss of fuel pool cooling event. The off-normal operating procedure ON-135-001, "Loss of Fuel Pool Cooling/Coolant Inventoryw, requires the operator to manually open the valves in ESW makeup line to the fuel pool if all other means of adding water are lost. The operator enters this procedure upon annunciation of low level in the spent fuel pool or high temperature in the fuel pool cooling system. Each spent fuel pool has temperature indication (TE- 15333) and level indication (LT-15332). Each fuel pool skimmer surge tank has level indication (LT- 15312). The skimmer surge tank piping to the fuel pool heat exchangers has-temperature indication

ORM EPM-QA-122B. Rev. 2 age 1 of 1

-


PPLL SUSQUXHAUHA ST- ELECTRIC STATIOU EUGXHBLLBI100 DISQUZPAHCY EEPORT

PILE: PI?-15 (continuation sheet)

(TE-15313) and each fuel pool heat exchanger has temperature indication (TE-15316A,B,C) in its outlet piping. The level and temperature instruments providing these alarms may not be qualified for the temperature and humidity conditions, such as post- LOCA, in which they would be required to function. In addition, these instruments are not powered from class 1E sources such that they would be available post-LOOP when the fuel pool heat exchangers would be without service water.

1. EDR No. b 3 ~ 0 = b 2. REV No.0 3. Page 4 of 6

E. The analytical 25 hour time-to-boil for the spent fuel pool is nonconservative for the maximum normal heat load in the spent fuel pool. The original design calculation (200-0048) used a deca6y heat load ,of 9.79~10~ BTU/hr for Unit 1 and 7.92~10 BTU/hr for Unit 2 to determine times-to-boil of 25.15 hours and 31.087 hours respectively.

FSAR 9.1.3.1 establishes the maximum normal heat load as that heat load resulting from 2840 assemblies discharged to the fuel pool by a routine refueling schedule. FSAR Tables 9.1-2a and 9.1-2b report the maximum normal heat load for Units 1 and 2 as 12.6~10~ BTU/hr.

The spent fuel pool decay heat values used in the boiling spent fuel pool calculation and for the FSAR discussion were based upon assumptions for cycle operating lengths, fuel exposures, and reactor power level. SSES has subsequently operated differently than had been assumed such that the decay heat loads in the filled spent fuel pools may exceed 9.79x106 BTU/hr, resulting in a shorter than analyzed time-to- boil.

A recent calculation prepared for power uprate (NFE- B-NA-053) determined decay heat from a filled spent fuel pool using actual fuel operating history through 1991 and assumptions which bound operation after power uprate. This calculation reported a maximum normal heat load of a17x106 BTU/hr.

Another recent calculation (M-FPC-009) determined the spent fuel pool time-to-boil and required ESW makeup rate for power uprate. Preliminary results from this calculation indicate the fuel pool boils 19.4 hours

DRM EPM-QA-122B, Rev. 2 age 1 of 1


PPLL SUSQWEE?BMA ST- ELECTRIC STATIOH EHGINEKBIUG DISWA.UCY REPORT

PILE: R42-15 (continuation sheet)

after loss of fuel pool cooling for the design heat removal capacity of the fuel pool cooling system (13.2~10~ BTU/hr). This calc determined a time-to- boil of s15.5 hours for the ~ 1 7 x 1 0 ~ BTU/hr heat load calculated-for the power uprate case.

1. EDR N O . & > ~ O 2. REV N0.o 3. Page 5 of 6

F. The analytical 25 hour time-to-boil for the spent fuel pool does not account for the emergency heat load in the spent fuel pool. FSAR 9.1.3.1 establishes the emergency heat load for the spent fuel pool as that heat load following a full core offload which completely fills the fuel pool. The FSAR specifies the emergency heat load to be 32.6~10~ BTU/hr.

The original design calculation (200-0048) determined time-to-boil for the maximum normal heat load case only. The actual decay heat load in the spent fuel pool exceeds the maximum normal heat load during every refueling outage at SSES in which the core is fully offloaded.

SSES currently imposes administrative controls during refueling outages when the core is fully offloaded into the spent fuel pool to reduce the potential for loss of fuel pool cooling. Decay heat is removed from the spent fuel pool during these periods by RHR shutdown cooling (when the fuel pool to reactor cavity gates are removed) and by cross-tieing the operating unit's fuel pool cooling system to the outage unit's fuel pool. However, a seismic event in this configuration could cause loss of fuel pool cooling at a time when the time-to-boil is significantly less than 25 hours, which is not reflected in the off-normal operating procedure (ON- 135-001).

Another recent calculation (M-FPC-009) determined the spent fuel pool time-to-boil and required ESW makeup rate for power uprate. Preliminary results from this calculation indicate the fuel pool boils 7.9 hours after loss of fuel pool cooling for a decay heat load of 36.2~10~ BTU/hr, which is the currently analyzed emergency heat load.

ORE4 EPM-QA-122B, Rev. 2 age 1 of 1


PPLL BUBQ- STEAM ELECTRIC BTATIOH mGIISEKBIHG DIBCRBP~CY -OPT

PILE: P42-15 (cantinution .%eat)

G. The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates. The original design calculation (200-0048) used a decay heat load of 9.79~10~ BTU/hr to determine evaporation rate. As reported above, the maximum normal heat load specified in FSAR Table 9.1-2a is 12.6~10~ BTU/hr and the emergency heat load specified in FSAR 9.1.3.1 is 32. 6x106 BTU/hr. When the decay heat load in the spent fuel pool exceeds 9.79~10~ BTU/hr, the evaporation rate from the boiling pool will exceed the rate assumed in the radiological release analysis.

1. EDR NO. ~ d c o a o 2. REV No.0 3. Page 6 of 6

H. The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms. The original design calculation (200-0048) assumed 12 month operating cycles and 184 bundle equilibrium reload sizes to determine the activity terms for failed fuel in the fuel pool. SSES currently has 18 month operating cycles with ~ 2 3 0 bundle reloads which will increase to ~ 2 5 4 bundles after power uprate. since the calculation implied that most of the activity results from the most recent discharge batch, the effect of increasing the discharge size from 184 bundles assumed in the calc to 230 and 254 bundles would appear to be nonconservative with respect to the radiological release analysis.

r 5

-

I. The analysis for maximum time prior to makeup to a boiling spent fuel pool is based upon nonconservative assumptions. The original design calculation (175- 14) determined the time using evaporation of the entire fuel pool water inventory. The maximum time should be based upon a minimum fuel pool water level which is sufficiently above the top of the fuel to provide the shielding required to allow corrective operator actions.

Page 1 of 1

EDR 620020 Block 14. Dis~osition

The EDR Evaluator recommends the following disposition:

+ Establish the original design basis for the fuel pool cooling system

+ Determine the appropriate design basis for the spent fuel pool cooling system

+ Compare the original design basis with the appropriate design basis

Based on the results of the above, the following actions may be necessary:

+ RevLew current plant practices regarding fuel offload and RHR system outages during refuel outages, and provide recommendations accordingly

+ Identify and recommend design changes, if required + Identify and recommend plant operating procedure changes, if required

SCREENING WORKSHEET

UNIT 1 & 2

EDR NO. G20020

SUSQUEHANNA STEAM ELECTRIC STATION

PENNSYLVANIA POWER & LIGHT COMPANY

Revision 2

PISCREPANCY I~LOCATIONISUBJECT;

Loss of Fuel Pool Cooling Event Design Discrepancies

mSCRIPTION OF CONDITION

This EDR raises 9 issues which are related but are separate. The common thread is that there appears to be a lack of suitable documentation of the design basis which has been presented and reviewed by both PPGL and the NRC. This situation is further complicated in that the long term effects of a loss of spent fuel pool inventory are also not well documented.

I. DOES THE ENGINEERING DISCREPANCY BEp-TE A H m m u 7

BASIS: The lack of suitable design documentation by itself would not affect the calculated accident frequencies. The fuel pool cooling system, as designed and installed, appears to be consistent with the criteria specified by both General Electric and Bechtel.

The potential for the fuel pool to boil will exist if and only the fuel pool cooling system becomes inoperative for an

tim . There are many modes of failure for the fuel pool cooling system. Reliability of the normal fuel pool cooling system is assured as the design incorporates redundant components, backup cooling sources, and the capability to cross tie cooling systems between units.

The most common failure mode which yields a -e loss of fuel pool cooling, is a loss of offsite power. Obviously, once offsite power is restored, fuel pool cooling can also be restored. The estimated time to restore offsite power ranges from 15 minutes to 20 hours, which is within the calculated time to fuel pool boil (25 hours after complete loss of cooling). The probability of a complete loss of offsite power is on the order of a DBA LOCA. Therefore, the loss of fuel pool cooling will not create a high calculated accident frequency.

In fact, this discrepancy has no effect on the core itself, which is the source of high radiation during a DBA LOCA. For this discrepancy to affect the plant, severe core damage must be postulated such that the reactor building becomes inaccessible, in addition to an extended loss of offsite power, rendering existing backup cooling systems useless.

11. DOES THE ENGINEERING DISCREPANCY APPEAR TO INVALIDATE DEFINER rn I - - I

ER EQUIPMENT OR PROCEDVBE_BELBTEPZ

BASIS: The timing of fuel pool boil event due to loss of fuel pool cooling, is such that no immediate operator actions are required to restore pool cooling. Therefore, required immediate operator actions to respond to a DBA LOCA are not altered, and defense in depth has not been invalidated.

Defense in depth requires protection of the core, vessel, and containment. This discrepancy affects none of these.

The extent of the scenario (fuel pool boil) is not expected to cause fuel damage, either in the vessel or spent fuel pool. It is expected that operators will respond and restore fuel pool cooling in accordance with established procedures, or at the direction of the Recovery Manager in the event the Emergency Plan is activated.

111. p OES T HE ENGINEERING DISCREPANCY APPEAR TO ADVERSELY IMP ACT A SYSTEM OR COMPONENT EXPLICITLY LISTED IN THE TECHNICAL SPECIFICATIONS?

BASIS: The fuel pool cooling system and related components are not addressed in the Technical Specification. During outage conditions, decay heat removal is provided for via the RHR shutdown cooling function, while the fuel is in the vessel, or by the fuel pool cooling system, when the fuel is transferred to the spent fuel pool, or by a combination of both systems.

This discrepancy will not affect the operation of the RHR system as specified in the Technical Specifications.

IV. pOES THE DISCREPANCY W-ITY OF A SYSTEM OR COMPONENT TO PERF---ED FUNCTION

BASIS: The fuel pool cooling system is designed and installed as stated in the FSAR.

YES

NO X

There is some concern that it is not operated in accordance with the FSAR description, in that the current practice of full core offload during refueling outages is not consistent with the FSAR description. In addition, when the fuel design from an 8 X 8 configuration to a 9 X 9 configuration, the FSAR description has not been revised, although the Fuels Group preformed the appropriate analysis to ensure safety to the public and the operability of the plant. This concern is being addressed separately under EDR G00005, and will not be evaluated further here.

SAR SECTION (S)

Chapter 9, 9A

V. DOES THE DISC-Y IMPACT ANY APP- WCENSING COMMI-

YES REFERENCE

ection 9.1.3 II BASIS: There are no adverse impacts identified that affect licensing commitments. The design of the Spent Fuel Pool Cooling and Cleanup System has been reviewed and discussed with the NRC, as documented in the Safety Evaluation Report. The NRC has accepted the existing design, including the radiological aspect of a fuel pool boil, and have accepted the provision to use ESW make-up, and RHR fuel pool cooling assist to cope with a loss of fuel pool cooling.

VI. v Bddre - ss D- svecific features whlch affect the saf&y

concern.

SAFETY SIGNIFICANCE

NONE

I MINIMAL X

I MODERATE I I

BASIS: As stated above, there are no adverse impacts on the plant due to this concern. It is obvious that a boiling pool is not desirable. It is not reasonable to assume that the operators will take no corrective action and allow the pool to boil.

PP&L emphasizes command and control in training of its operators. In fact, allowing the plant condition to degrade, or allowing the plant safety systems to initiate to control the event (allowing the plant to control itself), is a clear cause for failure during operator training.

Off Normal Procedures are in place to restore offsite power, and to restore fuel pool cooling. Emergency Operating Procedures are in place which provide guidance in dealing with a boiling pool.

Based on the arguments stated herein, this EDR has been judged to have minimal safety significance. However, due to the fact that discrepancies do exist in the documentation, it is recommendedthatresolution of these discrepancies be pursued, and further, that an assessment of operability and reportability be completed.

* This is an initialassessment. The screening function is to be considered a continuous process. A re-evaluation of the screening status (not necessarily formal, except when determined to be ttsignificantn) should take place by referencing this procedure at each stage of EDR processing (e.g. EDR implementation) to determine, if the issue is now a "safety concernn and is subject to Reportability and/or Operability determinations.

REPORTABILITY EVALUATION COVERSHEET

UNIT # 1 & 2

EDR # 620020

SUSQUEHANNA STEAM ELECTRIC STATION PENNSYLVANIA POWER & LIGHT COMPANY

REPORTABILITY EVALUATION

EDR # 620020

I. Discrepancy Item/Location/Subject:


11. Description of Condition

During a review of the Spent Fuel Pool Systems for the Power Uprate Project, several design discrepancies were discovered which form the basis for the subject EDR.

This EDR raises nine discrepancies which are related, but separate. The common thread is that there appears to be a lack of suitable documentation of the design basis which has been presented and reviewed by both PPLL and the NRC. This situation is further complicated in that the long term effects of a loss of spent fuel pool cooling are also not well documented.

Each of the nine discrepancies are addressed separately, and in greater detail, in the nEDR Evaluationn. This evaluation will address the combined effect of the discrepancies. Since this is an assessment of reportability, the focus will be looking backward, starting from a review of the fuel pool cooling system as designed and installed, to determine if the system will perform its intended function as described in the FSAR and other licensing basis documents.

111. Does the condition b r e a s e the ~robeilitv of occurrence of an accident previously evaluated in the FSAR? (Include specific reference to FSAR sections that are applicable.)

X Yes NO

Provide a discussion of the basis and criteria used in arriving at the above conclusion.

The probability of occurrence of an accident will not be changed by this discrepancy.

The current design of the Spent Fuel Pool Cooling and Cleanup System is consistent with the description in the FSAR and SER,

and the commitments to Regulatory Guides 1.13 and 1.29, General Design Criterion 61. The analyses demonstrate that the offsite dose consequence of a boiling pool is within IOCFR~OO limits.

The Spent Fuel Pool Cooling System is designed as a non- seismic category I system, which is consistent with industry standards. There are two (2) safety grade independent sources of water systems for make up and cooling, the ESW and RHR systems. For evaporative losses, each system is capable of providing water at a rate greater than the maximum boil-off rate of a boiling pool. The design also allows for alternate cooling methods, including cool water feed from the ESW makeup line to the spent fuel pool, draining off the warmer water to the Refueling Water Storage Tank.

The potential for the fuel pool to boil will exist if and only If the fuel pool cooling system becomes inoperative for an extended ~eri0d of time.

The spent fuel pool cooling system is not designed to withstand the effects of a safe shutdown earthquake. In this design basis event, there may be no cooling system recovery, and in this case, the fuel pool can be postulated to boil. Current analyses for a boiling fuel pool demonstrate that offsite doses will remain within 1OCPRlOO limits, as long as the fuel remains covered. Therefore, as an added design feature, the seismic emergency service water make up has been installed to replace inventory lost due to boiling. In reality, it is reasonably expected that alternate cooling systems (including fuel pool cooling, RHR cooling assist) can be restored following a seismic event to minimize the boiling potential.

The most common failure mode which yields a comwlete loss of fuel pool cooling, is a loss of offsite power. Obviously, once offsite power is restored, fuel pool cooling can also be restored. As identified in the Individual Plant Evaluation (IPE) report, the estimated time to restore offsite power ranges from 15 minutes to 20 hours, which is well within the calculated time to fuel pool boil.

The current analysis, which includes such conservatisms as fully loaded spent fuel pool, recently removed fuel bundles from the reactor vessel, indicates 25 hours to boil after complete loss of cooling, although a more recent calculation estimates the time to boil to be on the order of 19 hours. However, data obtained from the most recent refuel outage indicates that the expected time to boil is on the order of 120 hours.

The fuel pool instrumentation, including pool and skimmer

surge tank level, and pool temperature, realistically are expected to be available during a loss of offsite power, even though the instruments have not been specifically analyzed to withstand a potential fuel pool boil condition. These instruments are powered by the diesels via instrument AC Distribution panels 1(2)Y218 and 1(2)Y219.


IV. Does the condition jncrease the conseauences of an accident previously evaluated in the FSAR? (Include specific reference to FSAR sections that are applicable.)

X Yes No

Provide a discussion of the basis and criteria used in. arriving at the above conclusion.

The consequences of an accident are not affected by this discrepancy

It is clear from the design documentation that the spent fuel pool boil event was not expected to occur concurrent with a DBA LOCA. The offsite dose consequences due to a DBA LOCA concurrent with a loss of offsite power have been evaluated as an event. Likewise, the offsite dose consequences due to a loss of spent fuel pool cooling have been analyzed as an event. However, the combined effect of a DBA LOCA & a loss spent fuel pool cooling have not been analyzed.

The analyzed design basis accident (DBA) is a LOCA with a concurrent loss of offsite power. The design requirements for the spent fuel pool cooling system are described in the FSAR Section 9.1.3, which is in accordance with the guidelines of Regulatory Guide 1.13 regarding makeup to the spent fuel pool, the guidelines of Regulatory Guide 1.29 regarding design of non-seismic Category I systems, compliance with General Design Criterion 61 regarding the prevention of uncovering the spent fuel, plus the off site dose guidelines of 10CFR100.

The design for the spent fuel pool cooling system is to prevent the fuel pool from boiling. In addition, the RHR fuel pool cooling assist mode has been provided. But should the pool boil, makeup is available from the ESW and RHR systems to maintain the water level above the fuel throughout the event. This design provision has been incorporated since the fuel pool cooling system used for normal operation is not designed to satisfy seismic Category I criteria.

Review of the fuel pool cooling system design indicates that four of the six ESW valves are easily accessible For operation, but the remaining two are located in a high radiation zone as defined by Chapter 18 of the FSAR. However, Chapter 18 is a result of the TMI-related EQ equipment requirements, which postulated extreme worst case core damage conditions, not DBA LOCA core damage conditions. Should access be a concern, back up cooling systems are available from the adjacent unit.

The use of the RHR assist mode for spent fuel pool cooling requires manual operation of seven valves, five of which are readily accessible, and two are in high radiation zones as described in the FSAR Chapter 18 high radiation zone for the severe accident analysis, which does not represent DBA LOCA conditions. Should access be a concern, back up cooling systems are available from the adjacent unit.

The extent of the scenario (fuel pool boil) is not expected to cause fuel damage, either in the vessel or spent fuel pool. It is expected that operators will respond and restore fuel pool cooling in accordance with established procedures, or at the 'direction of the Recovery Manager in the event the Emergency Plan is activated.

In fact, this discrepancy has no effect on the core itself, which is the source of high radiation during a DBA LOCA. For this discrepancy to affect the plant, severe core damage must be postulated such that the reactor building becomes inaccessible, in addition to an extended loss of offsite power, rendering existing backup cooling systems useless.

The timing of fuel pool boil event due to loss of fuel pool cooling, is such that no immediate operator actions are required to restore pool cooling. Therefore, required immediate operator actions to respond to a DBA LOCA are not altered.

V. Does the condition -t ' e 9 malfunction of equipment important to safety previously evaluated in the FSAR? (include specific reference to FSAR sections that are applicable.)

X Yes No


The probability of occurrence of a malfunction will not be

increased. The spent fuel pool is not expected to boil as a matter of practice. It should be noted that an analysis has been performed that concludes the equipment located in the secondary containment can withstand the temperature effects of a loss of fuel pool cooling.

In the event of a loss of fuel pool cooling, it is expected that operators will respond and restore fuel pool cooling in accordance with establishedprocedures, or at the direction of the Recovery Manager in the event the bergency Plan is activated.

The secondary containment heat load analysis, using best available technology at the time, has incorporated the effect of temperature due to a boiling pool, and determined there is no increase in the probability of occurrence of a malfunction of equipment important to safety. However, new technology may be available which may allow for a more precise determination of the effect. Obviously, any new analyses must also evaluate the effect of existing calculational conservatisms to determine impact.

In addition, this discrepancy has no effect on the core itself, which is the source of high radiation during a DBA MCA. For this discrepancy to affect the plant, severe core damage must be postulated such that specific areas of the reactor building becomes inaccessible, in addition to an extended loss of offsite power, rendering existing backup cooling systems useless. As presented in the FSAR Chapter 18, the refuel floor is shown as a low radiation area. On the other hand, the lower elevations of the reactor building are shown as high radiation areas in the affected unit, only due to the postulated source terms in the piping systems assuming a severely degraded core. However, access to the non-LOCA unit should be unaffected.

VI. Does the condition increase the conse qgence of a m f u n c t ~ o n of equipment important to safety previously evaluated in the FSAR? (Include specific reference to FSAR sections that are applicable.)

X Yes No


As presented in the previous Section V, since no new malfunctions are expected, the consequences of such malfunctions are unchanged.

VII. Does the condition rea ate the ~ossibilitv of an acc . . . ident of g gifferent t w e than previously analyzed in the FSAR? (Include reference to specific FSAR sections that are applicable.)

Yes X NO


Under the present licensing basis, the fuel pool cooling system has been designed to the applicable codes and standards of the 1970's and in doing so met all the Federal, State and Local Government requirements. However, the design may not meet current standards and practices, and if designed today, probably would be different. Based on the Individual Plant Evaluation (IPE) methodology, fuel pool cooling system design today may not necessarily be enhanced, since issues such as lose of offsite power, source terms, offsite doses, and operator actions are based on probability, risk, and realistic perspectives.

There are many scenarios, which have not been specifically analyzed. For those cases, the Emergency Plan will be initiated. Under this plan there are site procedures in place to mitigate the event. Procedures such as Off-Normal (ON), Emergency Operating (EO) , and Emergency Support (ES) are written to allow the operators to take proactive steps to protect the public.

To accommodate these unidentified scenarios, PP&L, as part of the Emergency Plan, performs yearly drills to test the response of the organization to new types of severe accident events.

IIX. Does the condition x g t e ent im~ortant to safeto of a different tvve than

previously evaluated in the FSAR? (Include reference to specific FSAR sections that are applicable.)

X Yes No


The probability of occurrence of a malfunction of a different type will not be increased, as no new types of malfunctions are expected. The spent fuel pool is not expected to boil as a matter of practice. As part of the design requirements the

equipment located in the secondary containment must be able to withstand the effects of a loss of fuel pool cooling.

In the event. of a loss of fuel pool cooling, it is expected that operators will respond and restore fuel pool cooling in accordance with established procedures, or at the direction of the Recovery Manager in the event the Emergency Plan is activated.

The secondary containment heat load analysis, using best available technology at the time, has incorporated the effect of temperature due to a boiling pool, and determined there is no increase in the probability of occurrence of a malfunction of equipment important to safety. However, new technology may be available which may allow for a more precise determination of the effect. Obviously, any new analyses must also evaluate the effect of existing calculational conservatisms to determine impact.

In addition, this discrepancy has no effect on the core itself, which is the source of high radiation during a DBA MCA. For this discrepancy to affect the plant, severe core damage must be postulated such that specific areas of the reactor building becomes inaccessible, in addition to an extended loss of offsite power, rendering existing backup cooling systems useless. As presented in the FSAR Chapter 18, the refuel floor is shown as a low radiation area. On the other hand, the lower elevations of the reactor building are shown as high radiation areas in the affected unit, only due to the postulated source terms in the piping systems assuming a severely degraded core. However, access to the non-MCA unit should be unaffected.

IX. Does the -condition r-y as defined in the basis for any Technical Specification? (Include specific reference to specific Technical Specification sections that are applicable.)

X Yes No


The fuel pool cooling system and related components are not addressed in the Technical Specification. During outage conditions, decay heat removal is provided for via the RHR shutdown cooling function, while the fuel is in the vessel, or by the fuel pool cooling system, when the fuel is transferred to the spent fuel pool, or by a combination of both systems.

The RHR fuel pool cooling assist and the ESW make up has been installed to provide an additional margin of safety.

It is not reasonable to assume that the operators will take no corrective action and allow the pool to boil. As stated above, there are no adverse impacts on the plant due to this concern. It is obvious that a boiling pool is not desirable.

PPLL emphasizes command and control in training of its operators. In fact, allowing the plant condition to degrade, or allowing the plant safety systems to initiate to control the event (allowing the plant to control itself), is a clear cause for failure during operator training.


~ase'd on the arguments stated herein, this EDR has been judged to be not reportable. However, due to the fact that discrepancies do exist in the documentation, it is recommended that resolution of these discrepancies be pursued, including training to sensitize the operators to the effects of a boiling fuel pool, and further, that an assessment of operability be completed. It is further recommended that Nuclear Licensing review this reportability evaluation due to its potential generic implications.

* This is an initial evaluation. The reportability function is to be considered a continuous process. A re-evaluation of the reportability status when an EDR is reclassified to be "safety signif icantn, at any stage of EDR processing (e.g. EDR implementation), should take place by referencing this procedure.

OPERABILITY EVALUATION COVERSHEET

UNIT $ 1 & 2

EDR $ 620020

Susquehanna Steam Electric Station

Pennsylvania Power and Light Company

viewed By/Date


During a review of the Spent Fuel Pool Systems for the Power Uprate Project, several design discrepancies were discovered which form the basis for the subject EDR.

This EDR raises nine discrepancies which are related, but separate. The common thread is that there appears to be a lack of suitable documentation of the design basis which has been presented and reviewed by both PPCL and the NRC. This situation is further complicated in that the long term effects of a loss of spent fuel pool cooling are also not well documented.

Each of the nine discrepancies are addressed separately, and in greater detail, in the "EDR Evaluationn. This evaluation will address the combined effect of the discrepancies. Since this is an assessment of operability, the focus will be looking backward, starting from a review of the fuel pool cooling system as designed and installed, to determine if the system will perform its intended function, and thus support safe operation of the plant.

3. Does this concern apply to other plant structures, systems, trains, or components in the other unit? $f ves. the evaluation must envelo~e a~~licable items,

X Yes No

I. Does the condition clearly establish that the plant is outside its design basis?

X Yes No

Does the condition represent an uaumlyaed state that .ianifiaantlv co-es ~ m t safety?

X Yes NO

111. Does the condition clearly establish a condition m t covered bv the Dlant1s Onerat- and Emeraency Procedure?

X Yes No

Basis: The lack of suitable design documentation by itself would not affect the safe operation of the plant. The fuel pool cooling system, as designed and installed, appears to be consistent with the criteria specified by both General Electric and Bechtel. It is clear that the design basis for the plant (including the spent fuel pool) must consider the effects of a DBA MCA, concurrent with a loss of offsite power, with resulting operator actions expected after a reasonable time to mitigate the consequences.

It is clear from the design documentation that the spent fuel pool boil event was not evaluated to occur concurrent with a DBA MCA. The offsite dose consequences due to a DBA LOCA concurrent with a loss of of fsite power have been evaluated as an event. Likewise, the offsite dose consequences due to a loss of spent fuel pool cooling have been analyzed as an event. However, the combined effect of a DBA M C A a a loss spent fuel pool cooling have not been analyzed, and are not included in the design basis.

The analyzed design basis accident (DBA) is a M C A with a concurrent loss of offsite power. The design requirements for the spent fuel pool cooling system are described in the FSAR Section 9.1.3, which is in accordance with the guidelines of Regulatory Guide 1.13 regarding makeup to the spent fuel pool, the guidelines of Regulatory Guide 1.29 regarding design of non-seismic Category I systems, compliance with General Design Criterion 61 regarding the prevention of uncovering the spent fuel, plus the off site dose guidelines of 10CFR100.

The spent fuel pool is normally cooled and cleaned by its own system, the fuel pool cooling and cleanup system. There are alternate cooling modes for the spent fuel. Separate fuel pool headers enable fuel pool cooling assist and/or makeup provided by RHR System, and separate 2" lines additionally

provide makeup from the emergency service water system.

The design for the spent fuel pool cooling system is to prevent the fuel pool from boiling. But should the pool boil, makeup is available from the ESW system to maintain the water level above the fuel through out the event. This design provision has been incorporated since the fuel pool cooling system used for normal operation is not designed to satisfy seismic Category I criteria.

The potential for the fuel pool to boil will exist if the fuel pool cooling system becomes inoperative f-

There are many modes of failure for the fuel pool cooling system. Reliability of the fuel pool cooling system is assured as the design incorporates redundant components, backup cooling sources, and the capability to cross tie cooling systems between units.

The spent fuel pool cooling system is not designed to withstand the effects of a safe shutdown earthquake. In this design basis event, there may be no cooling system recovery, and in this case, the fuel pool can be postulated to boil. Current analyses for a boiling fuel pool demonstrate that offsite doses will remain within lOCFRlOO limits, as long as the fuel remains covered. Therefore, as an added design feature, the seismic emergency service water make up has been installed to replace inventory lost due to boiling. In reality, it is reasonably expected that alternate cooling systems (including fuel pool cooling, RHR cooling assist) can be restored following a seismic event to minimize the boiling potential.

The most common failure mode which yields a com~lete loss of fuel pool cooling, is a loss of offsite power. Obviously, once offsite power is restored, fuel pool cooling can also be restored. As identified in the Individual Plant Evaluation (IPE) report, the estimated time to restore offsite power ranges from 15 minutes to 20 hours, which is well within the calculated time to fuel pool boil.

The current analysis, which includes such conservatisms as fully loaded spent fuel pool, recently removed fuel bundles from the reactor vessel, indicates 25 hours to boil after complete loss of cooling, although a more recent calculation estimates the time to boil to be on the order of 19 hours. However, data obtained from the most recent refuel outage indicates that the expected time to boil is on the order of 120 hours.

In fact, this discrepancy has no effect on the core itself,

which is the source of high radiation during a DBA MCA. For this discrepancy to affect the plant, severe core damage must be postulated such that specific areas of the reactor building becomes inaccessible, in addition to an extended loss of offsite power, rendering existing backup cooling systems useless. As presented in the FSAR Chapter 18, the refuel floor is shown as a low radiation area. On the other hand, the lower elevations of the reactor building are shown as high radiation areas in the affected unit, only due to the postulated source terms in the piping systems assuming a severely degraded core. However, access to the non-MCA unit should be unaffected.

The timing of fuel pool boil event due to loss of fuel pool cooling, is such that no immediate operator actions are required to restore pool cooling. Therefore, required immediate operator actions to respond to a DBA M C A are not altered.


The extent of the scenario (fuel pool boil) is not expected to cause fuel damage, either in the vessel or spent fuel pool. It is expected that operators will respond and restore fuel pool cooling in accordance with established procedures, or at the direction of the Recovery Manager in the event the Emergency Plan is activated.

Plant operating procedure ON-135-001 prescribes corrective action to be taken in the event of a failure of the spent fuel pool cooling system. Obviously, the first course of corrective action is to determine the cause of failure, and restore cooling to the pool to eliminate the potential for boiling. . If this is not successful, the procedure then directs the operator to utilize RHR in the fuel pool cooling assist mode. In the event efforts to restore pool cooling are unsuccessful, the operator is directed to allow the pool to boil and utilize ESW make up inventory lost due to boiling.

Procedure ON-135-01 also addresses the potential for higher radiation on the refueling floor during a loss of fuel pool inventory event. It states: "The Fuel Assemblies will be kept cool by alternate means or by boiling in the pool with ultimate concern of keeping the assemblies covered with water." This procedure also states that entry into this (ON- 135-001) procedure may require implementation ofthe Emergency Plan. Operations personnel are directed by procedure to ensure operation of the fuel pool cooling system to prevent the spent fuel pool from boiling in the first place. A conscious decision to allow the pool to boil is made only after all efforts to restore pool cooling are unsuccessful.

Emergency Operating Procedure EO-100-104, Secondary Containment Control, may be entered during a fuel pool boil event. As this procedure is event-based, the entry conditions for a boiling pool are expected to be secondary containment area high temperature isolation signal, Zone I11 HVAC exhaust radiation level >2.5 mr/hr, or any secondary containment area radiation level >10 times alarm. This procedure directs the operator to monitor the secondary containment area temperature and/or monitor the secondary containment area radiation levels, and take action to reduce secondary containment area temperatures and/or radiation levels to less than maximum normal. In any event, operations personnel cannot exit the procedure until all entry conditions are clear.

The fuel pool instrumentation, including pool and skimmer surge tank level, and pool temperature, realistically are expected to be available during a loss of offsite power, even though the instruments have not been specifically analyzed to withstand a high temperature environment. These instruments are powered by the diesels via instrument AC Distribution panels 1(2)Y218 and 1(2)Y219. These panels are fed by uninterruptable power supplies (UPS), which are fed from one normal and one alternate MCC. These MCC1s are backed up by the diesels as described in FSAR Table 8.3-1.

During outage conditions, decay heat removal is provided for via the -.shutdown cooling function, while the fuel is in the vessel, or by the fuel pool cooling system, when the fuel is transferred to the spent fuel pool, or by a combination of both systems.

Although the fuel pool cooling system is designed and installed as stated in the FSAR, there is some concern that it is not operated in accordance with the FSAR description, in that the current practice of full core offload during refueling outages is not consistent with the FSAR description. In addition, when the fuel design from an 8 X 8 configuration to a 9 X 9 configuration, the FSAR description has not been revised, although the Fuels Group preformed the appropriate analysis to ensure safety to the public and the operability of the plant. This concern is being addressed separately under EDR G00005, and will not be evaluated further here.

There are no adverse impacts identified that affect licensing commitments. The design of the spent fuel pool cooling and cleanup system has been reviewed and discussed with the NRC, as documented in the Safety Evaluation Report. The NRC has accepted the existing design, including the radiological aspect of a fuel pool boil, and have accepted the provision to use ESW make-up, and RHR fuel pool cooling assist from either unit to cope with a loss of fuel pool cooling.

It is not reasonable to assume that the operators will take no corrective action and allow the pool to boil. As stated above, there are no adverse impacts on the plant due to this concern. It is obvious that a boiling pool is not desirable.

PPLL emphasizes command and control in training of its operators. In fact, allowing the plant condition to degrade, or allowing the plant safety systems to initiate to control the event (allowing the plant to control itself), is a clear cause for failure during operator training.


Based on the arguments stated herein, this EDR has been judged to have no impact on the continued operation of the plant. However, due to the fact that discrepancies do exist in the documentation, it is recommended that resolution of these discrepancies be pursued, including training to sensitize the operators to the effects of a boiling fuel pool, and further, that an assessment of reportability be completed.

An affirmative response to Questions 1,II or I11 is a trigger to initiate immediate Reportability and "Heads Upn Notification to Plant and the NRC.

This is an j,&t-n. The operability function is to be considered a continuous process. A re-evaluation of the operability status when an EDR is reclassified at any stage of EDR processing (e.g. EDR implementation) , should take place by referencing this procedure.

7

Attachment 4

PP&L Operability Statement, "EDR $620020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies", April 23, 1992

Note: This unsigned, handwritten evaluation for EDR 620020 was completed by a PP&L engineer or supervisor within the PP&L Engineering Discrepancy Management Group in order to clear t h e way for the startup of one o f t h e Susquehanna Steam Electric Station units from a refueling and inspection outage. By classifying the concerns in EDR 620020 as design bases document items, the concerns could be written off as not affecting the pending startup of the unit.

Onerabilitv Statement Page No. 63 04/23/92

hplementation Cost Suea: 321 - G D Miller

EDR # Bubjeat 620020 LOSS OF SPENT FUEL POOL COOLING EVENT DESIGN DISCREPANCIES

Description: SEE EDR FORM

Impact of Operability

The open deficiency does not impact the operability of Unit 1 for the following reasons:

Attachment 5

PP&L Memo from Dave Lochbaum and Don Prevatte to Joe Zola, "Supplemental Information for EDR 620020 on Boiling Spent Fuel Pool", June 22, 1992 (ET-0471)

Note: This memo to the engineer within the PP&L Engineering Discrepancy Management Group responsible for evaluating the concerns in EDR 620020 provided a clear, concise discussion of the concerns, the specific regulatory requirements in violation, and their potential nuclear safety consequences.

M E M O R A N D U M

TO: Joe Zola DATE: June 22, 1992

FROM: Dave Lochbaum Don Prevatte 44

Mike Detamore JOB : SUS~. SES NUMBER: ET-0471 COPIES: Jim Agnew

John Bartos Mark Mjaatvedt

FILE: P88-1 Nuclear Records

SWJECT: Supplemental Information for EDR G20020 on Boiling Spent Fuel Pool

The meeting on June 18, 1992 to discuss EDR G20020 was effective in outlining the design bases for the boiling spent fuel pool event. To assist you in evaluating the EDR, we are providing supplemental information for each item in the EDR. A partial draft of this supplement was distributed during the meeting.

We have included a brief discussion of the impact for each item on the current operation of SSES. Please note that for the majority of the items described in the EDR, the safe operation of SSES is presently adversely affected by the discrepancy.

Please contact us if we can support your evaluation in any way.

EDR G20020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies

The regulatory requirements for cooling the spent fuel pool are based upon :

10 CFR 50 Appendix A Design Criterion 61 which states that the fuel storage system shall be designed "to prevent significant reduction in fuel storage coolant inventory under accident conditions8# and

Standard Review Plan (NUREG-0800) 9.1.3 for the spent fuel pool cooling and cleanup system which states that the "safety function to be performed by the system in all cases remains the same; that is, the spent fuel assemblies must be cooled and must remain covered with water during all storage ~onditions.'~

The SSES design utilizes non-seismic, non-Class 1E powered fuel pool cooling and cleanup systems for cooling the fuel pools. In the event of a loss of spent fuel pool cooling, the design provision at SSES is to allow the fuel pools to boil with adequate makeup provided to maintain the water level in the pools above the fuel. The SSES design requirements are based upon:

FSAR Appendix 9A which states that it is assumed "a seismic event causes the loss of cooling to both spent fuel poolsn and that "if cooling is not restored before the pool boils, then makeup water from the Category I Emergency Service Water System can be added to the pool to keep the fuel covered at all times," and

FSAR 6 2 1 1 1 ( a ) states that "The LOCA scenario used for containment functional design includes the worst single failure (which leads to maximum coincident containment pressure and temperature), postulated to occur simultaneously with loss of offsite power and a safe shutdown earthquake (SSE) .I1

Since an analyzed design basis accident (DBA) at SSES is a LOCA with a concurrent LOOP and SSE, and either a seismic event or a loss of offsite power will result in a loss of spent fuel pool cooling, the consequences of this DBA include boiling spent fuel pools. The SSES design was reviewed and approved by the NRC in SSES Safety Evaluation Report (NUREG-0776) 9.1.3 which states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss of spent fuel pool cooling accidents."

EDR 620020 was initiated on April 16, 1992 to address the following design discrepancies for the loss of spent fuel pool event:

June 22, 1992 Page I

EDR 620020 Loss of Spent Fuel Pool Cooling Event Design Discrepancies

A. Reactor Building Design Heat Loads DO Not Account for the Boiling Spent Fuel Pool Event

REQUIREMENT: 10 CFR 50.49 requires that electrical equipment must be qualified to the temperature ##for the most severe design basis accidents.

lo CFR 50 Appendix A General Design Criterion 4 states that llstructures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. "

CONCERN: Secondary containment design analyses are required to account for all heat loads in the reactor building including from the boiling spent fuel pool. The existing design reactor building heat load calcs consider sensible heat from the boiling pool, but neglect latent heat. These calcs indicate little margin to design temperatures in many rooms for a maximum heat load in the reactor building of approximately 5.5~10~ BTU/hr. The total design heat load from the spent fuel pools is 26.4~10~ BTU/hr, which would add at least approximately 20.9~10~ BTU/hr to the existing maximum heat load.

This concern affects the present operation of SSES because:

1) the boiling spent fuel pool is a current design bases event,

2) the fuel pools will boil following loss of fuel pool cooling with their existing decay heat loads,

3) the boiling spent fuel pool event has not been fully considered in reactor building heat load calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to room temperatures in the reactor building exceeding design EQ values.

June 22, 1992 Page 2


B. The Impact of the ESW Makeup Water to the Spent Fuel Pool on Equipment in the Reactor Building Has Not Been Evaluated

REQUIREMENTS: 10 CFR 50 Appendix A General Design Criterion 4 states that "struc&es, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

Standard Review Plan (NUREG-0800) 3.4.1 states that the review of "plant flood protection includes all structures, systems and components (SSC) whose failure could prevent safe shutdown of the plant or result on uncontrolled release of significant radioactivity.. .I1 and that this review "also includes consideration of flooding from internal sources.

CONCERN: FSAR 9.1.3.3 states the design ESW makeup function "is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days following the loss of the FPCCS capacity." The ultimate heat sink and ESW are designed to provide 1.5 million gallons of water to each fuel pool over the 30 day period. In the LOCA-LOOP condition, the reactor building HVAC system in Zone I, I1 and I11 isolation mode recirculates refueling floor air throughout all three zones. The water added to the fuel pools ends up in the reactor building following boil-off and overflow. The effects of this water on the structures, systems and components in the reactor buildings have not been included in design analyses. The potential for common mode failures of multiple ECCS and safety-related systems such as the standby gas treatment system exists.






3) the boiling spent fuel pool event has not been fully considered in EQ and flooding effects calcs, and

4 ) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to common mode equipment failures due to water/humidity.



C. The Manual Valve Manipulations Required to Provide ESW Makeup Flow to a Boiling Spent Fuel Pool May Not Be Possible

REQUIREMENTS: 10 CFR 20.1 requires licensees to "make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to unrestricted areas, as low as is reasonably achievable. It

10 CFR 50 Appendix A General Design Criterion 19 requires suitable design features to limit control room radiation exposure to 5 rem. GDC 19 also requires design features for equipment outside the control room to permit operation in accordance with suitable procedures.

10 CFR 50.47(b) (11) states that licensees assure that "means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPAEmergency Worker and Lifesaving Activity Protective Action Guides. " NDI-6.4.3 specifies that the whole body dose for life saving actions "shall not exceed 75 remw and the whole body dose for entry into a hazardous area to protect facilities or equipment "shall not exceed 25 rem."

CONCERN: The ESW system is required to provide makeup to the pools following loss of fuel pool cooling. Either a seismic event or loss of offsite power can lead to loss of fuel pool cooling. Both conditions are assumed to occur concurrent with a LOCA in the DBA for containment analyses. However, the post-LOCA design EQ dose rates in the reactor building areas where the manual valves are located are 140-360 R/hr and will prevent these valves from being accessed without excessive radiation exposure to the operator. In addition, the reactor building temperature, humidity and emergency lighting conditions would not be conducive to the location and manipulation of manual valves which are used infrequently.

10 CFR 20's ALARA provision requires plant design to minimize radiation exposure. Application of the emergency dose guidelines to this manual valve operation is contrary to the intent of 10 CFR 20.1 and 10 CFR 50 App A GDC 19.






3) the boiling spent fuel pool event analysis depends on makeup from ESW to prevent uncovering irradiated fuel and subsequent fuel damage from overheating, and

4) the potential consequences from the boiling spent fuel pool event will significantly increase if adequate makeup cannot be established, or

5) personnel will receive unnecessary radiation exposures which exceed 10 CFR 20.11GDC 19 requirements and probably exceed 10 CFR 50.47 guidelines in order to align the makeup path.



D. The Instrumentation Available to the Operator Post-LOCA Does Not Provide Adequate Indication of Spent Fuel Pool Temperature and Level to Allow Proper Response to a Loss of Fuel Pool Cooling Event

REQUIREMENTS: 10 cFR 50 Appendix A General Design Criterion 63 states that "appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.

Regulatory Guide 1.97 defines accident-monitoring instrumentation to include lrthose variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events."

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design includes "the instrumentation provided for initiating appropriate safety actions."

Standard Review Plan (NUREG-0800) 7.1 states that "information systems important to safety include those systems which provide information for manual initiation and control of safety systems, to indicate that plant safety functions are being accomplished, and to provide information from which appropriate actions can be taken to mitigate the consequences of anticipated operational occurrences and accidents."

CONCERN: The ESW system is required to provide makeup to the pools following loss of fuel pool cooling. A loss of offsite power can result in loss of fuel pool cooling. The loss of offsite power will also disable the fuel pool temperature and level instruments monitored by the operator and used to initiate the safety action of providing ESW makeup to the boiling spent fuel pool.







4 ) the potential consequences from the boiling spent fuel pool event will significantly increase if adequate makeup cannot be established and lack of monitoring could prevent the required safety action from being initiated properly.



E. The Analytical 25 Hour Time-to-Boil for the Spent Fuel Pool is Nonconservative for the Maximum Normal Heat Load in the Spent Fuel Pool

REQUIREMENTS: FSARAppendix 9A states "conservative results showed that the pools would not boil until at least 25 hours after the loss of cooling."

FSAR Table 9A-2 states the total decay heat loads in the Unit 1 and Unit 2 fuel pools assumed in the loss of spent fuel pool cooling analysis are "9.79" and "7.92" BTU/hr x 106.

CONCERN: The maximum normal heat load in the spent fuel pool is presently higher than 9.79x106 BTU/hr and will also increase as a result of power uprate. The fuel pool will boil in less than 25 hours for any fuel pool heat load greater than 9.79~10~ BTU/hr. (See Figure 1 from Calc M-FPC-009 attached).

This concern does affect the present operation of SSES because the existing decay heat loads in the fuel pools are less than 9.79~10~ BTU/hr.

NOTE : The original determination of maximum normal heat load relied on assumed reactor operating parameters such as fuel type, fuel discharge average exposure, and operating cycle length. These parameters have changed since the original calculation and will probably continue to change as fuel design and fuel management evolves. An approach to bound all such variables would consider the maximum normal heat load in the spent fuel pool to be equal to the design capacity of the fuel pool cooling system (13.2~106 BTU/hr). This approach would bound all heat loads capable of being handled by the fuel pool cooling system without depending upon predictions of fuel and core designs.



F. The Analytical 25 Hour Time-to-Boil for the Spent Fuel Pool Does Not Account for the Emergency Heat Load in the Spent Fuel Pool

REQUIREMENTS: Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design are reviewed to determine that "a seismic Category I makeup system and an appropriate backup method to add coolant to the spent fuel pool are provided1' and that tlengineering judgement . . . used to determine that the makeup capacities and the time required to make associated hookups are consistent with heatup times or expected leakage.

SSES Safety Evaluation Report (NUREG-0776) 9.1.3 states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss of spent fuel pool cooling accidents.

FSAR 9.1.3.1 states that "during an emergency heat load (EHL) condition, one RHR pump and heat exchanger are available for fuel pool cooling.

CONCERN: The emergency heat load condition requires an RHR loop to remove decay heat from the spent fuel pool. A single failure of the valve in the RHR line to the fuel pool, even without a concurrent seismic event or loss of offsite power, could initiate a loss of fuel pool cooling in which the time-to-boil would be significantly less than the 25 hours assumed in the radiological release analysis and in plant operating procedures. This potential exists presently during every refueling outage when the full core is offloaded to the spent fuel pool.

This concern affects the present operation of SSES during refueling outages because:


2) the fuel pools will boil in as little as 12 hours following loss of fuel pool cooling with the existing decay heat loads in the pools during refueling, and

3) the spent fuel pool boiling analysis assumes a minimum time to boil of 25 hours.

June 22, 1992 Page 10


G. The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconservative Evaporation Rates

REQUIREMENT: FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

CONCERN: The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the evaporation rate from the pools based upon maximum normal heat loads of 9.79 and 7.92 x lo6 BTU/hr. As discussed in Item (E) above, the present maximum normal heat load exceeds 9.79~10~ BTU/hr and will increase after power uprate. Therefore, the actual rate at which water evaporates from the boiling spent fuel pool is higher than analyzed which introduces nonconservatism into the offsite dose calculation.

This concern does affect the present operation of SSES (except during refueling outages as noted in Item F above) because the existing decay heat loads in the fuel pools are less than 9.79~10~ BTU/hr.

June 22, 1992 Page 11


H. The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconaervative Activity Terms

REQUIREMENT: FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

CONCERN: The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the source terms in the spent fuel pools based upon assumptions for fuel design and cycle operation. SSES has been operated with different fuel types and longer cycles than assumed in the analysis which introduces nonconservatism into the offsite dose calculation.

In addition, the conclusions reported in FSAR Appendix 9A regarding the thyroid doses from FSAR Table 9A-1 are not valid for all cases. FSAR Table 9A-1 only addresses offsite doses from activity released from the two boiling spent fuel pools. Since the boiling spent fuel pools can occur as a result of the LOCA-LOOP with SSE DBA, these thyroid doses should be added to the doses resulting from the LOCA.




3) the potential consequences from the boiling spent fuel pool event may significantly increase due to higher source term activity associated with 9x9 fuel, larger discharge batch sizes, and higher bundle exposures, and

4) the offsite dose resulting from the boiling spent fuel pool is not considered in the total offsite dose resulting from the DBA LOCA-LOOP.

June 22, 1992 Page 12


I. The Analysis for Maximum Time Prior to Makeup to a Boiling Spent Fuel pool is Based upon Nonconservative Assumptions

REQUIREMENT: Calc 175-14 determined the maximum time available before makeup to a boiling spent fuel pool is required.

CONCERN: The time determined by this design calculation is based upon how long it would take to completely evaporate the entire spent fuel pool water inventory. Allowing the entire spent fuel pool to evaporate prior to makeup would have severe and unanalyzed consequences:

a) reactor building radiation doses would significantly increase,

b) offsite radiological doses would significantly increase due to skyshine, and

c) fuel integrity of the irradiated fuel would be challenged as it was uncovered.

This concern does not appear to affect the present operation of SSES because no document or procedure is known to use the results of this calc. However, an exhaustive search was not performed.

June 22, 1992 Page 13

Figure 1 Time to Boil vs. FPC Heat Load

FPCCS Design Capacity

I Analytical Limit I

Fuel Pool Cooling Heat Load, MBTUIhr

[ - 125 F Pool Temp 110 F Pool Temp I


PP&L D r a f t S c r e e n i n g Worksheet p r e p a r e d b y A r t Wh i te , " E D R No. G20020", J u l y 1, 1992

Note: The f i r s t n i n e t e x t pages o f t h i s d r a f t e v a l u a t i o n o f EDR 620020 p r e p a r e d by an e n g i n e e r w i t h i n t h e PP&L E n g i n e e r i n g D i s c r e p a n c y Management Group were t a k e n a l m o s t v e r b a t i m f r o m t h e a u t h o r s ' memo d a t e d June 22, 1992 ( A t t a c h m e n t 5). The f i n a l t h r e e pages o f ' a n a l y s i s ' f o r EDR 620020 p r o v i d e ample e v i d e n c e o f PP&L1s r e l i a n c e upon p r o b a b i l i t y arguments, use o f r e a l i s t i c i n s t e a d o f d e s i g n c o n d i t i o n s , and o v e r s i m p l i f i c a t i o n o f i s s u e s w h i l e a s s e s s i n g t h e s a f e t y s i g n i f i c a n c e o f conce rns .

SCREENING WORKSHEET

EDR No.G20020

SUSQUEHANNA STEAM. ELECTRIC STATION

PENNSYLVANIA POWER & LIGHT COMPANY

PREPARED BY REVIEWED BY

DISCREPANCY ITEM/LOCATION/SUBJECT:

Loss of Spent Fuel Cooling Event Design Discrepancies

DESCRIPTION OF CONDITION

The regulatory requirements for cooling the spent fuel pool are based upon: 10 CFR Appendix A Design Criterion 61 which states that the fuel storage system shall be designed to prevent significant reduction in fuel storage coalant inventory under accident conditions~~ and Standard Review Plan (NUREG-800) 9.1.3 for the spent fuel pool cooling and cleanup system which states that the "safety function to be performed by the system in all cases remains the same; that is, the spent assemblies must be cooled and must remain covered with water during all storage conditions.

The SSES design utilizes non-seismic, non-Class IE powered fuel pool cooling and cleanup systems for cooling the fuel pools. In the event of a loss of spent fuel pool cooling, the design provision at SSES is to allow the fuel pools to boil with adequate makeup provided to maintain the water level in the pools above the fuel. The SSES design requirements are based upon:

FSAR Appendix 9A which states that it is assumed "a seismic event causes the loss of cooling to both spent fuel poolsft and that "if cooling is not restored before the pool boils, then makeup water from the Category I Emergency Service Water System can be added to the pool to keep the fuel covered at all tirnes,@l and

FSAR 6 2 1 1 1 ( a ) states that "The LOCA scenario used for containment functional design includes the worst single failure (which leads to maximum coincident containment pressure and temperature), postulated to occur simultaneously with loss of offsite power and a safe shutdown earthquake (SSE) . "

Since an analyzed design basis accident (DBA) at SSES is a LOCA with a concurrent LOOP and SSE, and either a seismic event or a loss of offsite power will result in a loss of spent fuel pool cooling, the consequences of this DBA include boiling spent fuel pools. The SSES design was (NUREG-0776) 9.1.3 which states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss os spent fuel pool cooling accidents.

The following design discrepancies for the loss of spent fuel pool event :

A. Reactor Building Design Heat Loads Do Not Account for the Boiling Spent Fuel Pool Event

Requirement: 10 CFR 50.49 requires that electrical equipment must be qualified to the temperature "for the most severe design basis accidents. If

10 CFR 50 Appendix A General Design Criterion 4 states that ltstructures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. l1

Concern: Secondary containment design analyses are required to account for all heat loads in the reactor building including from the heat load calcs consider sensible heat from the boiling pool, but neglect latent heat. These calcs indicate little margin to design temperatures in many rooms for a maximum heat load in the reactor building of approximately 5.5E6 BTU/hr. The total design heat load from the spent fuel pools is 26.436 BTU/hr, which would add at least approximately 20.936 BTU/hr to the existing maximum heat load.




3) the boiling spent fuel pool event has not been fully considered in reactor building heat load calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to room temperatures in the reactor building exceeding design EQ values.

B. The Impact of the ESW Makeup Water to the Spent Fuel Pool on Equipment in the Reactor Building has not been evaluated

Requirements: 10 CFR 50 Appendix A General Design Criterion 4 states that "structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.'

Standard Review Plan (NUREG-0800) 3.4.1 states that the

review of "plant flood protection includes all structures, systems and components (SSC) whose failure could prevent safe shutdown of the plant or result in uncontrolled release of significant radioactivity .... 11 and that this review 81also includes consideration of flooding from internal sources."

Concern: FSAR 9.1.3.3 states the design ESW makeup function "is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days following the loss of the FPCCS capacity." The ultimate heat sink and ESW are designed to provide 1.5 million gallons of water to each fuel pool over the 30 day period. In the LOCA-LOOP condition, the reactor building HVAC system in Zone I, I1 and I11 isolation mode recirculates refueling floor airthroughout all three zones. The water added to the fuel pools ends up in the reactor building following boil-off and overflow. The effects of this water on the structures, systems and components in the reactor buildings have not been included in design analyses. The potential for common mode failures of multiple ECCS and safety-related' systems such as the standby gas treatment system exists.



2) the fuel pools will boil fol-lowing loss of fuel pool cooling with their existing decay heat loads,

3) the boiling spent fuel pool event has not been fully considered in EQ and flooding effects calcs, and

4) the potential consequences from the boiling spent fuel pool event will significantly and adversely affect the safety of SSES due to common mode equipment failures due to water/humidity.

C. The manual valve manipulations required to provide ESW makeup flow to a boiling spent fuel pool may not be possible.

Requirements: 10 CFR 20.1 requires licensees to "make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to unrestricted areas, as low as is reasonably achievable."

lOCFR 50 Appendix A General Design Criterion 19 requires suitable design features to limit control room radiation exposure to 5 rem. GDC 19 also requires design features for equipment outside the control room to permit operation in accordance with suitable procedures.

10 CFR 50.47 (b) (11) states that licensees assure that

"means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides."

NDI-6.4.3 specifies that the whole body dose for life saving actions "shall not exceed 75 remu and the whole body dose for entry into a hazardous area to protect facilities or equipment "shall not exceed 25 rem."

Concern: The ESW system is required to provide makeup to the pools following loss of fuel pool cooling. Either a seismic event or loss of offsite power can lead to loss of fuel pool cooling. Both conditions are assumed to occur concurrent with a LOCA in the DBA for containment analyses. However, the post-LOCA design EQ dose rates in the reactor building areas where the manual valves are located are 140-360R/hr and will prevent these valves from being accessed without excessive radiation exposure to the operator. In addition, the reactor building temperature, humidity and emergency lighting conditions would not be conducive to the location and manipulation of manual valves which are used infrequently.

lOCFR 20% ALARA provision requires plant design to minimize radiation exposure. Application of the emergency dose guidelines to this manual valve operation is contrary to the intent of 10 CFR 20.1 and 10 CFR 50 App A GDC 19.





4) the potential consequences fromthe boiling spent fuel pool event will significantly increase if adequate makeup cannot be established, or

5) personnel will receive unnecessary radiation exposures which exceed 10 CFR 20.1/GDC 19 requirements and probably exceed 10 CFR 50.47 guidelines in order to align the makeup path.

D. The instrumentation available to the Operator Post-LOCA does not provide adequate indication of spent fuel pool temperature and level to allow proper response to a loss of fuel pool cooling event

Requirements: 10 CFR Appendix A General Design Criterion 63 states that "appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.I8

Regulatory Guide 1.97 defines accident-monitoring instrumentation to include "those variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is prsvided and that are required for safety systems to accomplish their safety function for design basis accident events."

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design includes "the instrumentation provided for initiating appropriate safety actions.I8

standard Review Plan (NUREG-0800) 7.1 states that llinformation systems important to safety include those systems which provide information for actual initiation and control of safety systems, to indicate that plant safety functions are being accomplished, and to provide information from which appropriate actions can be taken to mitigate the consequences of anticipated operational occurrences and accident^.^^

Concern: The ESW system is required to provide makeup to the pools following loss of fuel pool cooling. A loss of offsite power can result in loss of fuel pool cooling. The loss of offsite power will also disable the fuel pool temperature and level instruments monitored by the operator and used to initiate the safety action of providing ESW makeup to the boiling spent fuel pool.





4) the potential consequences from the boiling spent fuel pool event will significantly increase if adequate makeup cannot be established and lack of monitoring could prevent the required

safety action from being initiated properly.

3. The analytical 25 hour time-to-boil for the spent fuel pool is nonconservative for the maximum normal heat load in the spent fuel pool.

Requirements: FSAR Appendix 9A states "conservative results showed that the pools would not boil until at least 25 hours after the loss of cooling.

FSAR Table 9A-2 states the total decay heat loads in the Unit 1 and Unit 2 fuel pools assumed in the loss of spent fuel pool cooling analysis are 9.7936 BTU/hr and 7.9236 BTU/hr .

Concern: The maximum normal heat load in the spent fuel pool is presently higher than 9.7936 BTU/hr and will also increase as a result of power uprate. The fuel pool will boil in less than 25 hours for any fuel pool heat load greater than 9.7936 BTU/hr. (See Figure 1 from Calc M-FPC-009 attached).

This concern does not affect the present operation of SSES because the existing decay heat loads in the fuel pools are less than 9.79E6 BTU/hr.

Note: The original determination of maximum normal heat load relied on assumed reactor operating parameters such as fuel type, fuel discharge average exposure, and operating cycle length. These parameters have changed since the original calculation and will probably continue to change as fuel design and fuel management evolves. An approach to bound all such variables would consider the maximum normal heat load in the spent fuel pool to be equal to the design capacity of the fuel pool cooling system (13.236 BTU/hr) . This approach would bound all heat loads capable of being handled by the fuel pool cooling system without depending upon predictions of fuel and core designs.

F. The analytical 25 hour time-to-boil for the spent fuel pool does not account for the emergency heat load in the spent fuel pool.

Requirements: Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design are reviewed to determine that a seismic Category I makeup system and an appropriate backup method to add coolant to the spent fuel pool are provided" and that "engineering judgement. . .used to determine that the makeup capacities and the time required to make associated hookups are consistent with heatup times or expected leakage.

SSES Safety Evaluation Report (NUREG-0776) 9.1.3 states "makeup from the Seismic Category I emergency service water systems would keep the fuel covered during loss of spent fuel pool cooling

accidents."

FSAR 9.1.3.1 states that "during an emergency heat load (EHL) condition, one RHR pump and heat exchanger are available for fuel pool cooling.

Concern: The emergency heat load condition requires an RHR loop to remove decay heat from the spent fuel pool. A single failure of the valve in the RHR line to the fuel pool, even without a concurrent seismic event or loss of offsite power, could initiate a loss of fuel pool cooling in which the time-to-boil would be significantly less than the 25 hours assumed in the radiological release analysis and in plant operating procedures. This potential exists presently during every refueling outage when the full core is offloaded to the spent fuel pool.

This concern affects the present operation of SSES during refueling outages because:


2) the fuel pools will boil in as little as 12 hours following loss of fuel pool cooling with the existing decay heat loads in the pools during refueling, and

3) the spent fuel pool boiling analysis assumes a minimum time to boil of 25 hours.

G. The Radiological Release analysis for a boiling spent fuel pool uses nonconservative evaporation rates

Requirement: FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

Concern: The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the evaporation rate from the pools based upon maximum normal heat loads of 9.7936 BTU/hr and will increase after power uprate. Therefore, the actual rate at which water evaporates from the boiling spent fuel pool is higher than analyzed which introduces nonconservatism into the offsite dose calculation.

This concern does not affect the present operation of SSES (except during refueling outages as noted in Item F above) because the existing decay heat loads in the fuel pools are less than 9.7936 BTU/hr . H. The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms.

Requirement: FSAR Appendix 9A reported that the radiological release analysis for the boiling spent fuel pools event were within the thyroid dose guidelines of 10 CFR 100 and the 1.5 rem thyroid dose requirement of Reg Guide 1.29.

Concern: The design calculation which performed the radiological release analysis for the boiling spent fuel pools event determined the source terms in the spent fuel pools based upon assumptions for fuel design and cycle operation. SSES has been operated with different fuel types and longer cycles than assumed in the analysis which introduces nonconservatism into the offsite dose calculation.

In addition, the conclusions repor-ted in FSAR Appendix 9A regarding the thyroid doses from FSAR Table 9A-1 are not valid for all cases. FSAR Table 9A-1 only addresses offsite doses from activity released from the two boiling spent fuel pools. Since the boiling spent fuel pools can occur as a result of the LOCA-LOOP with SSE DBA, these thyroid doses should be added to the doses resulting from the LOCA.




3) the potential consequences from the boiling spent fuel pool event may significantly increase due to higher source term activity associated with 9x9 fuel, larger discharge batch sizes, and higher bundle exposures, and

4) the offsite dose resulting from the boiling spent fuel pool is not considered in the total offsite dose resulting from the DBA LOCA-LOOP.

I. The analysis for maximum time prior to makeup to a boiling spent fuel pool is based upon nonconservative assumptions

Requirement: Calc 175-14 determined the maximum time available before makeup to a boiling spent fuel pool is required.

Concern: The time determined by this design calculation is based upon how long it would take to completely evaporate the entire spent fuel pool water inventory. Allowing the entire spent fuel pool to evaporate prior to makeup would have severe and unanalyzed consequences:

1) reactor building radiation doses would significantly increase,

2) offsite radiological doses would significantly increase due to skyshine, and

3) fuel integrity of the irradiated fuel would be challenged as it was uncovered.

This concern does not appear to affect the present operation of 88E8 because no document or procedure is known to use the results of this calc. However, an exhaustive search was not performed.

ANALYSIS

I. DOES THE ENGINEERING DISCREPANCY APPEAR TO CREATE A HIGH CALCULATED ACCIDENT SEOUENCE FREOUENCY?

BASIS: No, the postulated concern is based on postulating a' DBA LOCA, a LOOP and an SSE all simultaneously. The probability of such an event approaches zero, it is so vanishingly small.

11. DOES THE ENGINEERING DISCREPANCY APPEAR TO INVALIDATE DEFINED STAGE IN "DEFENSE-IN-DEPTH" AGAINST AN ACCIDENT SEOUENCE, WHETHER EOUIPMENT OR PROCEDURE RELATED?

BASIS: No, the postulated concern takes no credit for manual action. From a realistic point-of-view, there is no basis to assume that fuel damage will occur to the extent that manual actions can be taken to line up ESW and RHR in the spent fuel area.

111. DOES THE ENGINEERING DISCREPANCY APPEAR TO ADVERSELY IMPACT A SYSTEM OR COMPONENT EXPLICITLY LISTED IN THE TECHNICAL SPECIFICATIONS?

TECHNICAL SPECIFICATION SECTION (S)

BASIS: This discrepancy has no basis in fact, and takes no credit for expected operator action.

IV. DOES THE DISCREPANCY APPEAR TO COMPROMISE THE CAPABILITY OF A SYSTEM OR COMPONENT TO PERFORM ITS SAFETY RELATED FUNCTION AS DESCRIBED IN THE SAFETY ANALYSIS REPORT?

11 YES SAR SECTION (S)

BASIS: This concern has no effect on any safety related function as described in the Safety Analysis Report. The EDRrs basis appears to be an invalid application of design philosophies to (realistic) sost accident manual actions.

V. DOES THE DISCREPANCY APPEAR TO ADVERSELY IMPACT ANY APPLICABLE LICENSING COMMITMENTS?

REFERENCE I

BASIS: The discrepancy does not appear to adversely impact any applicable licensing commitments. In fact, the SER specifically addresses the Spent Fuel Cooling Function, and its ESW makeup and RHR cooling function.

VI. SAFETY SIGNIFICANCE ASSESSMENT

Address plant-s~ecific features which affect the safetv sianificance of the concern. Provide a realistic assessment of the actual safetv conseauences and im~lications of the concern.

SAFETY SIGNIFICANCE

MODERATE

CONSIDERABLE

BASIS: There is no safety significance to this EDR since it has no basis in fact if one does not accept the premise that no action can or will be taken by operations personnel to stop a boiling pool from boiling, or to inhibit it from boiling in the first place. Basically, it is a misapplication of plant design parameters, such as postulated fuel melt, to post accident operator actions.

* This is an initial assessment. The screening function is to be considered a continuous process. A re-evaluation of the screening status (not necessarily formal, except when determined to be eesignificantlt) should take place by referencing this procedure at each stage of EDR processing (e.g. EDR implementation) to determine if the issue is now a @Isafety concerne1 and is subject to Reportability and/or Operability determinations.

Attachment 7

Handout , "EDR 620020 R e f e r e n c e s " , J u l y 1 5 , 1992

Note : T h i s handout was p r e p a r e d b y t h e a u t h o r s and d i s t r i b u t e d d u r i n g a m e e t i n g on J u l y 1 5 , 1992 t o d i s c u s s EDR 6 2 0 0 2 0 . The handout summarizes t h e documents r e s e a r c h e d by t h e a u t h o r s w h i l e p r e p a r i n g t h e EDR and s u b s e q u e n t l y d e f e n d i n g i t s m e r i t s .

EDR G20020 References

Design Bases and Related Issues

FSAR 6 . 2 . 1 . 1 . 1 s t a t e s t h a t the "LOCA s c e n a r i o u s e d for c o n t a i n m e n t f u n c t i o n a l d e s i g n i n c l u d e s the w o r s t s i n g l e f a i l u r e ( w h i c h l e a d s t o maximum c o i n c i d e n t c o n t a i n m e n t p r e s s u r e and t e m p e r a t u r e ) , p o s t u l a t e d t o o c c u r s i m u l t a n e o u s l y w i t h loss o f o f f s i t e power and a s a f e shutdown e a r t h q u a k e (SSE) . I 1

FSAR Append ix 9A s t a t e s t h a t I t i t was assumed t h a t a s e i s m i c event c a u s e s the loss o f c o o l i n g t o both s p e n t f u e l p o o l s . "

FSAR Append ix 9A s t a t e s t h a t I t i f c o o l i n g i s not r e s t o r e d before the pool boi ls , then makeup w a t e r from the C a t e g o r y I Emergency Service W a t e r S y s t e m c a n be added t o the pool t o k e e p the f u e l c o v e r e d a t a l l t imes.

S S E S S a f e t y E v a l u a t i o n R e p o r t 9 . 1 . 3 s t a t e s t h a t "makeup f rom the seismic C a t e g o r y I emergency service w a t e r s y s t e m s would k e e p the f u e l c o v e r e d d u r i n g loss o f s p e n t f u e l pool c o o l i n g a c c i d e n t s . " Bechtel Spec M-192 f o r the High D e n s i t y S p e n t Fuel S t o r a g e Racks ( J u n e 1977) s t a t e s t h a t the l l se l l e r s h a l l p e r f o r m a n a l y s i s t o d e t e r m i n e the makeup f l o w r a t e r e q u i r e d t o m a i n t a i n the poo l w a t e r level u n d e r c o n d i t i o n s o f maximum h e a t l o a d , none o f the c o o l i n g systems a v a i l a b l e and pool w a t e r b o i l i n g . "

L e t t e r PLI-7457 f rom A. M . Male t o R. J . S h o v l i n ( J u l y 1 9 7 9 ) s t a t e s t h a t " the s p e n t f u e l pool c o o l i n g s y s t e m i s d e s i g n e d t o m a i n t a i n t e m p e r a t u r e a t or b e l o w 125OF. T h e system i s f u r t h e r b a c k e d u p by the S e i s m i c C a t e g o r y I Append ix B q u a l i f i e d emergency s y s t e m s w h i c h h a v e s u f f i c i e n t c a p a c i t y t o h a n d l e t h i s l o a d . I f a l l o f these r e d u n d a n t systems a r e somehow u n a v a i l a b l e , i t w i l l s t i l l t a k e more t h a n one d a y b e f o r e b o i l i n g b e g i n s . T h i s i s more t h a n s u f f i c i e n t t i m e for o n s i t e p e r s o n n e l t o p r o v i d e from many a l t e r n a t e w a t e r s o u r c e s enough make u p w a t e r t o k e e p the pool f rom b o i l i n g . "

T e c h n i c a l R e p o r t N P E - 8 4 - 0 0 2 (December 1 9 8 3 ) s t a t e s t h a t "SSES i s d e s i g n e d t o a c c e p t and m i t i g a t e a loss o f c o o l a n t a c c i d e n t (LOCA) c o n c u r r e n t w i t h a c o m p l e t e loss o f o f f s i t e power (LOOP) and " the a s s u m p t i o n was made, i n the d e s i g n o f SSES, t h a t the LOCA and LOOP would o c c u r s i m u l t a n e o u s l y , and the s i m u l t a n e o u s o c c u r r e n c e o f LOCA and LOOP becomes the d e s i g n b a s i s event."

J u l y 1 5 , 1 9 9 2 Page 1


EWR MIS 86-0637 determined that the RHR fuel pool cooling assist mode lines could be deleted from the IS1 program since the "present design uses the ESW makeup line as the ultimate heat removal source" with this source being "sufficient to cover the maximum boiloff of a full core offload.I1

EWR MIS 85-0740 stated that "the RHR assist mode to fuel pool cooling is a non-safety function and therefore may be deleted from the ISI program boundaries" and this mode is "non-safety and adequate cooling is still available form boiling and ESW makeup."

NSAG 4-90 (September 1990) reported that in the RHR system design "the fuel pool cooling assist and the shutdown cooling modes share a common suction line. Therefore, the system can not operate in both modes at the same time."

EPRI Report NP-2301 (March 1982) reported that 27% of loss of offsite power events at nuclear plants had been caused by weather related problems. This report also stated that in 5% of all the loss of offsite power events at nuclear plants, the duration exceeded 24 hours.

NSAC Report 182 (March 1992) reported 21 loss of offsite power events lasting longer than one hour at nuclear plants between 1980 and 1991, with the longest event lasting 18:58.

Telecon from Michael Rose (PP&L) to Mort Renslo (Bechtel) of November 9, 1981 states that "according to Bechtel's Civil and Structural Design Criteria for the Susquehanna Steam Electric Station...This criteria states Fuel Pool Structure shall be designed for water boiling during accident condition.I1

July 15, 1992 Page 2


A. Reactor Building Design Heat Loads Do Not Account for the Boiling Spent Fuel Pool Event

FSAR 6.2.2.1(d) states that the safety design bases for the containment removal system is that the system "shall maintain operation during those environmental conditions imposed by the LOCA . " EWR 830658 (March 1983) noted "the initial boiling rate corresponds to ~ 3 0 0 0 cfm of 100% water vapor at one atm. Is the equipment which will be exposed to this atmosphere qualified for it?"

SEA-ME-099 (December 1987) analyzed reactor building temperatures for LOCA, LOCA/LOOP and LOCA/false LOCA cases assuming spent fuel temperatures remained at 125"F, but listed as a nonconservatism that fuel pool heatup in the LOCA/LOOP case would result in higher heat loads from the RHR systems, fuel pool walls and fuel pool surf ace.


EDR 620020 References

B . The Impact of the ESW Makeup Water to the Spent Fuel Pool on Equipment in the Reactor Building Has Not Been Evaluated

FSAR 6.3.1.1.3 s t a t e s t h a t s e p a r a t i o n b a r r i e r s f o r ECCS " s h a l l be c o n s t r u c t e d b e t w e e n the f u n c t i o n a l g r o u p s a s r e q u i r e d t o a s s u r e t h a t e n v i r o n m e n t a l d i s t u r b a n c e s s u c h a s f i r e , p i p e r u p t u r e , f a l l i n g o b j e c t s , e tc . , a f f e c t i n g o n e f u n c t i o n a l g r o u p s w i l l not a f f e c t the r e m a i n i n g g r o u p s . In a d d i t i o n , s e p a r a t i o n b a r r i e r s s h a l l be p r o v i d e d a s r e q u i r e d t o a s s u r e t h a t s u c h d i s t u r b a n c e s d o not a f f e c t both RCIC and HPCI."

FSAR 9.1.3.3 s t a t e s t h a t " the d e s i g n makeup r a t e from e a c h ESW l o o p i s b a s e d on r e p l e n i s h i n g the bo i l -o f f from the M N H L i n e a c h f u e l poo l for 30 d a y s f o l l o w i n g the loss o f FPCCS c a p a c i t y . "

EWR 830658 (March 1983) n o t e d " c o n d e n s a t i o n may be e x p e c t e d from th i s e v a p o r a t i o n w h i c h w i l l r u n down t o l o w e r levels o f the R .B . W i l l t h i s c a u s e loss o f e s s e n t i a l e q u i p m e n t , p a r t i c u l a r l y e l e c t r i c a l ? Has a n e v a l u a t i o n been performed?"

Minutes f rom B e c h t e l m e e t i n g o n HVAC s y s t e m s ( F e b r u a r y 1980) s t a t e s t h a t o r i g i n a l r e q u i r e m e n t f o r SGTS was " t o h a n d l e fumes from a b o i l i n g f u e l p o o l , " but t h a t SGTS w i l l n o t be a b l e t o h a n d l e t h i s m i x t u r e since t h e room w i l l become t o o h o t . " T h i s r e q u i r e m e n t w i l l be d e l e t e d f rom the FSAR."

J u l y 15, 1992 Page 4


C. The Manual Valve Manipulations Required to Provide ESW Makeup Flow to a Boiling Spent Fuel Pool May Not Be Possible

FSAR 9 . 1 . 3 . 2 s t a t e s t h a t Itthe manual s u p p l y valves i n these emergency makeup l ines a r e a c c e s s i b l e a p a r t f rom the r e f u e l i n g floor.11

FSAR 1 8 . 1 . 2 0 ( N U R E G - 0 7 3 7 I t e m 11. B . 2 ) s t a t e s t h a t " e a c h licensee s h a l l p r o v i d e for a d e q u a t e a c c e s s t o v i t a l a r e a s and p r o t e c t i o n o f s a f e t y e q u i p m e n t b y d e s i g n c h a n g e s , i n c r e a s e d permanent or t e m p o r a r y s h i e l d i n g , or p o s t a c c i d e n t p r o c e d u r a l c o n t r o l s . The d e s i g n r e v i e w s h a l l determine w h i c h t y p e s o f corrective a c t i o n s a r e needed for v i t a l a r e a s t h r o u g h o u t the f a c i l i t y . 1 1

FSAR 1 8 . 1 . 2 0 . 3 . 3 . 4 . 1 d e f i n e s v i t a l a r e a s a s t h o s e I fwhich w i l l or may r e q u i r e o c c u p a n c y t o p e r m i t a n o p e r a t o r t o a i d i n the m i t i g a t i o n o f or r e c o v e r y from a n a c c i d e n t . I 1

FSAR 1 8 . 1 . 2 0 . 3 . 2 . 1 s t a t e s t h a t " a r e v i e w was made t o determine w h i c h s y s t e m s c o u l d be r e q u i r e d t o o p e r a t e a n d / o r be e x p e c t e d t o c o n t a i n h i g h l y r a d i o a c t i v e m a t e r i a l s f o l l o w i n g a p o s t u l a t e d a c c i d e n t where s u b s t a n t i a l core damage h a s o c c ~ r r e d . ~ ~

FSAR 1 8 . 1 . 2 0 . 3 . 2 . 5 s t a t e s s t e x p o s u r e s for a r e a s not c o n t i n u o u s l y o c c u p i e d ( f r e q u e n t and i n f r e q u e n t o c c u p a n c y ) m u s t be d e t e r m i n e d c a s e by c a s e , t h a t i s , m u l t i p l y the t a s k d u r a t i o n by the a r e a d o s e r a t e a t the t i m e o f e x p o s u r e . "

FSAR 1 8 . 1 . 2 0 . 3 . 3 . 3 s t a t e s t h a t "GDC 19 i s a l s o u s e d t o g o v e r n d e s i g n b a s e s f o r the maximum p e r m i s s i b l e d o s a g e t o p e r s o n n e l p e r f o r m i n g a n y t a s k r e q u i r e d p o s t - a c c i d e n t . These r e q u i r e m e n t s t r a n s l a t e r o u g h l y i n t o the o b j e c t i v e s t o be m e t i n the p o s t - a c c i d e n t r e v i e w a s g i v e n b e l o w .

R a d i a t i o n Exposure G u i d e l i n e s Occupancy Dose R a t e O b j e c t i v e s Dose O b j e c t i v e C o n t i n u o u s 15 mR/hr 5 Rem for d u r a t i o n F r e q u e n t 100 mR/hr 5 Rem for a l l a c t i v i t i e s I n f r e q u e n t 500 mR/hr 5 Rem p e r a c t i v i t y Acces sway 5 R / h r I n c l u d e d i n a b o v e doses1 '

FSAR 1 8 . 1 . 2 0 . 3 . 4 . 3 s t a t e s t h a t t h e r e v i e w resul t s "show t h a t the r e a c t o r b u i l d i n g w i l l be g e n e r a l l y i n a c c e s s i b l e for s e v e r a l d a y s a f t e r the a c c i d e n t due t o c o n t a i n e d r a d i a t i o n s o u r c e s . "

FSAR F i g u r e 18 .1 -4 shows Room 1-105 where ESW v a l v e s 153500/153501 a r e l o c a t e d t o be i n Rad Zone V I I I w i t h d o s e r a t e s over 5000 R / h r .

J u l y 1 5 , 1 9 9 2 Page 5


FSAR Figure 18.1-6 shows Room 1-514 where ESW valves 153090A&B and 153091A&B are located to be in Rad Zone V with dose rates between 5 and 50 R/hr.

NDI-6.4.3 establishes the whole body dose for life saving to be 75 Rem, with a dose limit of 25 Rem for less urgent measures to protect equipment.



D. The Instrumentation Available to the Operator Post-LOCA Does Not Provide Adequate Indication of Spent Fuel Pool Temperature and Level to Allow Proper Response to a LOSS of Fuel Pool Cooling Event



E. The Analytical 25 Hour Time-to-Boil for the Spent Fuel Pool is Nonconservative for the Maximum Normal Heat Load in the Spent Fuel Pool

FSAR 9.1.3.1 states that "the pool will begin to boil 25 hours after loss of cooling."

FSAR Appendix 9A states that "the conservative results showed that the pools would not boil until at least 25 hours after the loss of cooling.

FSAR Table 9A-2 reports the total decay heat load to be 9.79~10~ BTU/hr in the Unit 1 SFP and 7.92~10~ BTU/hr in the Unit 2 SFP for the boiling spent fuel pool analysis.

Bechtel Calc 200-0048 (July 1977) determined a 25 hour time to boil for the Unit 1 SFP and a 31 hour time to boil for the Unit 2 SFP based upon 12 month operating cycles and 184 bundle reload sizes.

PP&L Calc NFE-B-NA-053 (February 1992) determined a spent fuel pool maximum normal heat load of ~ 1 4 . 6 ~ 1 0 ~ BTU/hr and emergency heat load of ~ 3 0 x 1 0 ~ BTU/hr using actual SSES operating history through 1991 and projected operation until the pool is filled.

NSAG Report 4-90 states that "Appendix 9A of the FSAR states that at least 25 hours would be required to boil the spent fuel pool under worst case loading."



F. The Analytical 25 Hour Time-to-Boil for the spent Fuel Pool Does Not Account for the Emergency Heat Load in the Spent Fuel Pool



G. The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconservative Evaporation Rates

July 15, 1992 Page 10


H. The Radiological Release Analysis for a Boiling Spent Fuel Pool Uses Nonconservative Activity Terms

July 15, 1992 Page 13


I. The Analysis for Maximum Time Prior to Makeup to a Boiling Spent Fuel Pool is Based Upon Nonconservative Assumptions

NSAG Report 13-84 (December 1984) reported that water level in the spent fuel pool dropping to within five inches of the top of the irradiated fuel Itwould cause radiation levels on the 818' elevation of the reactor building in excess of 100,000 rem/hour.lt

July 15, 1992 Page 12

Figure 1 Fuel Pool Cooling

Unit 1 Spent Fuel Pool F Unit 1 FPCCS, if

AC power available piping intact Ul sW available decay heat < 13.2~10

NO

Unit 2 FPCCS, if AC power available piping intact U2 SW available pools crosstied (or U2 SFP not handled) decay heat < 13.2~10 6

NO

Unit 1 RHR FPC Assist, if . U1 RHR SDC not required manual valves aligned . U1 RHRSW/UHS available U1 RHR Uncontaminated

- Yes

Unit 2 RHR FPC Assist, if . U2 RHR SDC not required manual valves aligned . U2 RHRSW/UHS available U2 RHR Uncontaminated

Loss of U1 SFP Cooling

Yes

Yes

Unit 2 Spent Fuel Pool F Unit 2 FPCCS, if

AC power available piping intact . U2 SW available decay heat < 13.2~10

Yes

Yes

Unit 1 FPCCS, if . AC power available - piping intact - U1 SW available pools crosstied (or U1 SFP not handled) decay heat < 13.2~10~

I

Unit 2 RHR FPC Assist, if . U2 RHR SDC not required . manual valves aligned U2 RHRSW/UHS available . U2 RHR Uncontaminated

Figure 2

NO

Figure 2 c l

- Yes

June 25, 1992

Unit 1 RHR FPC Assist, if . U1 RHR SDC not required manual valves aligned U1 RHRSW/UHS available ul RHR Uncontaminated

NO

Loss of U2 SFP Cooling

Figure 2 Fuel Pool Heatup

Figure 1 '7 Loss of U1 SFP Cooling +

Time to Boil function of initial pool temperature pool water inventory - decay heat

Unit 1 Spent Fuel Pool Boiling . RB Heat Load Impact EQ Impact SGTS Impact RB-HVAC Impact RE Flooding

I

Figure 3 11

Figure 1 Lr' Loss of U2 SFP Cooling '

Time to Boil function of initial pool temperature . pool water inventory - decay heat

Unit 2 Spent Fuel Pool Boiling RB Heat Load Impact EQ Impact , ; SGTS 1mpa;t RB-HVAC Impact RB Flooding

Figure 3 u

June 25, 1992

Figure 3 Fuel Pool Boiling 5 Makeup

Figure 2 '7 Unit 1 Spent Fuel Pool Boiling - Figure 2 '7

Unit 2 Spent Fuel Pool Boiling - June 25, 1992

Cond Xfer Water, if Cond Xfer Water, if AC power available - AC power available

piping intact . manual valves aligned NO NO -

RWST to Cask Storage Pit, if RWST to Cask Storage Pit, if AC power available . AC power available piping intact . piping intact manual valves aligned manual valves aligned

NO

U2 ESW System, if manual valves aligned manual valves aligned

NO NO

U1 ESW System, if manual valves aligned . manual valves aligned

NO NO

Loss of U1 SFP Water Level - RB Dose Impact Offsite Dose Impact Fuel Damage Impact

Loss of U2 SFP Water Level RB nose Impact Offsite Dose Impact . Fuel Damage Impact


Wh i te Paper p r e p a r e d b y D a v i d A . Lochbaum and Dona ld C . P r e v a t t e , " S a f e t y Consequences of a B o i l i n g Spent F u e l Poo l a t t h e Susquehanna Steam E l e c t r i c S t a t i o n " , J u l y 27, 1992

Note : T h i s paper was handed t o t h e PP&L Manager o f N u c l e a r P l a n t E n g i n e e r i n g i n a m e e t i n g r e q u e s t e d b y t h e a u t h o r s . T h i s pape r was p r e p a r e d when t h e a u t h o r s became c o n v i n c e d t h a t t h e PP&L E n g i n e e r i n g D i s c r e p a n c y Management Group and t h e PP&L S u p e r v i s o r , E n g i n e e r i n g P r o j e c t s were u n a b l e t o p r o p e r l y e v a l u a t e EDR 620020. The t i m i n g o f t h i s paper was d i c t a t e d b y t h e end o f M r . Lochbaum's c o n t r a c t a t PP&L.

SAFETY CONSEQUENCES OF A

BOILING SPENT FUEL POOL AT THE

SUSQUEHANNA STEAM ELECTRIC STATION

July 27 , 1992

Prepared by:

b,c .\A, Donald C. Prevatte

EXECUTIVE SUMMARY

Engineering Discrepancy Report (EDR) G 2 0 0 2 0 was written in April 1992 after nine potential problems associated with the boiling spent fuel pool event were identified during system evaluations to support the power uprate project for PP&Lfs Susquehanna Steam Electric Station. The major concerns raised in EDR G20020 are:

1. Regulations require that instrumentation shall be provided for the fuel storage systems to detect conditions that may result in loss of heat removal capability and to initiate appropriate safety actions.

Contrary to this requirement, the water level and temperature instrumentation for the spent fuel pools do not satisfy Class 1E criteria and are not included in the equipment qualification program. These instruments will fail following a loss of offsite power and may fail following a loss of coolant accident. The ultimate consequence of such failure could be an irradiated fuel meltdown outside primary containment.

2. Regulations require that nuclear power plant designs limit personnel radiation exposures to 5 5 Rem per individual for control room occupation and actions required to mitigate or recover from an accident.

Contrary to this requirement, the manual ESW valve manipulations required to provide makeup to a boiling spent fuel pool following a loss of coolant accident could require a radiation exposure significantly higher than 5 Rem. The ultimate consequence could be significant radiation overexposure or inability to provide ESW makeup and an irradiated fuel meltdown outside primary containment.

3. Regulations require that structures, systems and components important to safety be designed to accommodate the effects of the environmental conditions associated with postulated accidents.

Contrary to this requirement, the effects of ESW makeup water to a boiling spent fuel pool have not been considered in the SSES design. The effects include flooding, high temperature, and high humidity. The ultimate consequences could include failure of multiple ECCS and other safety related systems.

4. Regulations require that electrical equipment be qualified to the temperature for the most severe design basis accidents.

Contrary to this requirement, the SSES reactor building temperature analyses used in equipment qualification evaluations do not account for the heat load from a boiling spent fuel pool. The ultimate consequences could include failure of multiple ECCS and other safety related systems.

Safety Consequences of a Boiling Spent Fuel Pool

SYSTEM DESCRIPTION

Each of the two operating nuclear power plants at the Pennsylvania Power and Light (PP&L) Company's Susquehanna Steam Electric Station (SSES) has a spent fuel pool. Each spent fuel pool is designed to store up to 2,840 irradiated fuel bundles discharged from the reactor core after approximately four and a half years of operation. As of July 1992, the Unit 1 spent fuel pool contained 1400 irradiated fuel bundles and the Unit 2 spent fuel pool held 1004 irradiated fuel bundles.

The irradiated fuel bundles stored in the spent fuel pools generate heat from the nuclear decay of fission products. The amount of heat generation exponentially decreases with time as a function of the half life of the fission products.

The spent fuel pools are located in a common refueling area within the secondary containment structure. Each spent fuel pool is connected to a reactor cavity and to the other spent fuel pool. The reactor cavity is the area above the reactor pressure vessel which is flooded during a refueling outage after removing the drywell shield blocks, drywell head and reactor pressure vessel head to permit fuel transfer between the reactor core and the spent fuel pool. These connections are normally isolated, except during refueling outages, using gates.

Unit 1 Unit 1 Unit 2 Unit 2

Reactor Spent Spent 1 Reactor Cavity Fuel Fuel Cavity

Pool PO01

Each spent fuel pool has a fuel pool cooling and cleanup system (FPCCS) which circulates water from the fuel pool through a heat exchanger and demineralizer to maintain proper fuel pool water chemistry and to keep its temperature 5 125OF. The FPCCS has a design capacity of 13.2~10~ BTU/hr. As of July 1992, the decay heat load in the Unit 2 spent fuel pool was =2.1x106 BTU/hr while the decay heat load in the Unit 2 spent fuel pool was ~ 2 . 9 7 ~ 1 0 ~ BTU/hr. Heat from the FPCCS heat exchangers is transferred to the service water (SW) system which in turn dissipates the energy to the atmosphere via the cooling tower. The FPCCS and the SW system are non-safety related systems which are not designed to satisfy seismic, Class 1E power, equipment qualification and single failure criteria. The FPCCS is designed such that it cannot fail in a way which drains water from the spent fuel pool.



If the FPCCS is unavailable, the fuel pool cooling assist mode of the residual heat removal (RHR) system is designed to circulate water from the spent fuel pool through a heat exchanger to keep the fuel pool from boiling. The fuel pool cooling assist mode of RHR is manually initiated by opening valves in the reactor building. The fuel pool cooling assist mode of RHR has a capacity of 3 2 . 6 ~ 1 0 ~ BTU/hr. Heat from the RHR heat exchanger is transferred to the RHR service water (RHRSW) which in turn dissipates the energy to the atmosphere via the spray pond. The fuel pool cooling assist mode of RHR is a non-safety related function which is not designed to satisfy seismic, Class 1E power, and single failure criteria. The fuel pool cooling assist mode of RHR is designed such that it cannot fail in a way which drains water from the spent fuel pool.

If both the FPCCS and the fuel pool cooling assist mode of RHR are unavailable, the spent fuel pool water will boil unless cooling is re-established. The time required to reach boiling is a function of the decay heat load in the spent fuel pool, the initial temperature of the water, and the volume of water available. The volume of water available is primarily dependent upon the presence or absence of the gates between the spent fuel pools and the reactor cavity. The emergency service water (ESW) system is designed to provide makeup to the boiling spent fuel pool to compensate for water lost through boil-off and evaporation. The ESW makeup supply is manually initiated by opening three valves in the reactor building. The ESW system uses water from the spray pond. The ESW system and the spray pond are safety related systems which are designed to satisfy seismic, Class 1E power, and single failure criteria as applicable. The design provision at SSES is for the ESW system to provide adequate makeup to a boiling spent fuel pool if cooling is lost.

The reactor building heating, ventilating and air conditioning (RB- HVAC) system circulates tempered air through each reactor building and the refueling zone during normal operation. The RB-HVAC system maintains these areas at a slight negative pressure relative to the outside environment to prevent leakage of potentially airborne radioactivity to the atmosphere. The exhaust Prom the potentially contaminated areas is filtered to remove radioactive materials. In an emergency, the supply and exhaust lines are isolated and the RB- HVAC system recirculates air throughout the reactor building affected by the emergency and the refueling zone. During a loss of offsite power (LOOP), the supply and exhaust lines are isolated and the RB-HVAC system recirculates air throughout the both reactor buildings and the refueling zone. The RB-HVAC system in recirculation mode does not provide any cooling function, so the reactor building and refueling zone air temperatures increase based upon piping, lighting, transmission and equipment heat loads.



The standby gas treatment system (SGTS) is designed to maintain the secondary containment at a negative pressure relative to the outside environment in an emergency. The SGTS takes suction on the recirculation plenum of the RB-HVAC system and processes this air through a filter train to remove radioactive materials. The SGTS is normally in standby except during testing. The SGTS is designed to satisfy seismic, Class 1E power, and single failure criteria.

The emergency core cooling systems (ECCS) and the reactor core isolation cooling (RCIC) system are located in the lower elevations of each reactor building. These systems provide water to the reactor pressure vessel during transients and accidents. These systems are normally in standby except during testing. The ECCS are designed to satisfy seismic, Class 1E power, and single failure criteria.

BOILING SPENT FUEL POOL DESIGN ANALYSIS

SSES Final Safety Analysis Report (FSAR) Appendix 9A reports the results of an analysis performed to quantify the radiological consequences of a loss of spent fuel pool cooling event. The analysis assumed the initiating event was an earthquake which resulted in the failure of the FPCCS on both units. The analysis concluded that the secondary containment design with SGTS operation kept offsite doses to a small fraction of 10 CFR 100 limits even with conservative assumptions of initial fuel failures in the spent fuel pools.

CONTAINMENT DESIGN ANALYSES

SSES Final Safety Analysis Report (FSAR) Chapter 6 reports the results of analyses performed to demonstrate the capability of the safety related systems to mitigate the consequences of postulated accidents such that the containment design parameters are not exceeded. The postulated accidents included main steam line breaks and loss-of-coolant accidents (LOCAs) with and without concurrent loss of offsite power. A design basis accident (DBA) for SSES is defined as a LOCA with a simultaneous LOOP and safe shutdown earthquake and the worst case single failure which results in the maximum containment pressure and temperature conditions. SSES FSAR Chapter 6 indicates margin to containment design parameters for the analyzed postulated accidents.

Reactor building room temperatures following postulated accidents were analyzed for equipment qualification. A procedure to manually shed all the non-Class 1E power loads in the reactor building z 2 4 hours after a LOCA without a LOOP was developed to prevent room temperatures from exceeding equipment qualification limitations.



CONCERNS OVER BOILING SPENT FUEL POOL EVENT

Engineering Discrepancy Report (EDR) G20020 was written in April 1992 after nine potential problems associated with the boiling spent fuel pool event were identified during system evaluations to support the power uprate project. The four major concerns raised in EDR G20020 are:

1. Inadequate Instrumentation

A. Resulatory Requirements

10 CFR 50 Appendix A General Design Criterion 63 states that "appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions."

Regulatory Guide 1.97 defines accident-monitoring instrumentation to include "those variables to be monitored that provide the primary information required to permit the control room operators to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety function for design basis accident events."

Standard Review Plan (NUREG-0800) 9.1.3 states that the review of the spent fuel pool cooling and cleanup system design includes "the instrumentation provided for initiating appropriate safety actions."

Standard Review Plan (NUREG-0800) 9.1.3 for the spent fuel pool cooling and cleanup system states that the "safety function to be performed by the system in all cases remains the same; that is, the spent fuel assemblies must be cooled and must remain covered with water during all storage conditions . " Standard Review Plan (NUREG-0800) 7.1 states that "information systems important to safety include those systems which provide information for manual initiation and control of safety systems, to indicate that plant safety functions are being accomplished, and to provide information from which appropriate actions can be taken to mitigate the consequences of anticipated operational occurrences and accidents."



B. Concerns

The ESW system is required to provide makeup water to the spent fuel pools following loss of fuel pool cooling to keep the irradiated fuel covered thus preventing fuel damage from overheating. A loss of offsite power or LOCA can result in loss of fuel pool cooling since the FPCCS and the SW system are not normally supplied by Class 1E power. The loss of offsite power will also disable the spent fuel pool temperature and level instruments monitored by the operator and used to initiate the safety action of providing ESW makeup to the boiling spent fuel pool. The post-LOCA environment in the reactor building may disable the spent fuel pool temperature and level instruments since they are not covered under the equipment qualification program. Therefore, the existing spent fuel pool temperature and level instrumentation is inadequate to ensure the required safety action of providing adequate makeup to a boiling spent fuel pool is properly initiated and monitored under all postulated accident conditions.

If the spent fuel pool is permitted to boil without adequate makeup, its water level will drop. A study by the PP&L Nuclear Safety Assurance Group (NSAG Report 13-84, December 1984) reported that water level in the spent fuel pool dropping to within five inches of the top of the irradiated fuel "would cause radiation levels on the 818' elevation of the reactor building in excess of 100,000 rem/hour." At that dose rate, an individual on the refueling floor would receive a lethal radiation exposure in approximately 16 seconds. This severe condition is just the beginning of the adverse consequences of spent fuel pool boiling without adequate makeup. At this point, the radiation source term results in offsite doses exceeding 10 CFR 100 limits and in dose rates within the reactor building that prevent any personnel access. The situation progresses ultimately to uncovering irradiated fuel bundles in the spent fuel pool and fuel damage from overheating. The situation has the potential for a substantial meltdown of irradiated fuel outside the primary containment.



2. Manual ESW valve Operation

A. Resulator~ Requirements and Licensinq Commitments

10 CFR 20.1 requires licensees to "make every reasonable effort to maintain radiation exposures ... as low as is reasonably achievable.I1

10 CFR 50 Appendix A General Design Criterion 19 requires suitable design features to limit control room radiation exposure to 5 rem. GDC 19 also requires design features for equipment outside the control room to permit operation in accordance with suitable procedures.

10 CFR 50.47(b)(ll) states that licensees assure that "means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for control ling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides. "

SSES FSAR 18.1.20 in response to NUREG-0737 Item 1I.B. 2 states that "each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or postaccident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility . " SSES FSAR 18.1.20.3.3.4.1 defines vital areas as those "which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident."

SSES FSAR 18.1.20.3.2.1 states that Ira review was made to determine which systems could be required to operate and/or be expected to contain highly radioactive materials following a postulated accident where substantial core damage has occurred.

SSES FSAR 18.1.20.3.2.5 states "exposures for areas not continuously occupied (frequent and infrequent occupancy) must be determined case by case, that is, multiply the task duration by the area dose rate at the tzme of exposure."

SSES FSAR 18.1.20.3.3.3 states that "GDC 19 is also used to govern design bases for the maximum permissible dosage to personnel performing any task required post-accident. These requirements translate roughly into the objectives to be met in the post-accident review as given below.



Radiation Exposure Guidelines occupancy Dose Rate Objectives Dose Objective Continuous 15 mR/hr 5 Rem for duration Frequent 100 mR/hr 5 Rem - all activities Infrequent 500 mR/hr 5 Rem per activity Accessway 5 R/hr Incl in above doses1I

SSES FSAR 18.1.20.3.4.3 states that the review results "show that the reactor building will be generally inaccessible for several days after the accident due to contained radiation sources. " SSES FSAR Figure 18.1-4 shows Room 1-105 where ESW valves 153500/153501 are located to be in Rad Zone VIII with dose rates over 5000 R/hr. SSES FSAR Figure 18.1-6 shows Room 1-514 where ESW valves 153090A&B and 153091A&B are located to be in Rad Zone V with dose rates between 5 and 50 R/hr. These valves must be manually opened to initiate ESW makeup to the spent fuel pools in the loss of fuel pool cooling event.

PP&L administrative procedure NDI-6.4.3 specifies that the whole body dose for life saving actions 'shall not exceed 75 rem" and the whole body dose for entry into a hazardous area to protect facilities or equipment "shall not exceed 25 rem." 10 CFR 20's ALARA provision requires plant design to minimize radiation exposure. Application of the emergency dose guidelines to a design which requires manual valve operation is contrary to the intent of 10 CFR 20.1 and 10 CFR 50 App A GDC 19.

B. Concerns

The ESW system is required to provide makeup to the spent fuel pools following loss of fuel pool cooling. Either a seismic event or loss of offsite power can lead to loss of fuel pool cooling. Both conditions are assumed to occur concurrent with a LOCA in the DBA for containment analyses. However, the post-LOCA dose rates in the reactor building areas where the manual valves are located are 5 to 5,000+ R/hr and will prevent these valves from being accessed without excessive radiation exposure to the operator. In addition, the reactor building temperature, humidity and emergency lighting conditions would not be conducive to the location and manipulation of manual valves which are used infrequently. Therefore, the manual ESW valve manipulations required for makeup to boiling spent fuel pools may not be accomplished for all postulated accident conditions.



In addition, since the boiling spent fuel pool analysis reported in SSES FSAR Appendix 9A assumed a seismic event initiated the loss of fuel pool cooling, the intentional shedding of non-Class 1E power loads in the reactor building following a LOCA without a LOOP represents either the creation of a new kind of accident or the increased probability of a previously analyzed accident.

3. Effects of ESW Makeup Water on Reactor Building Systems

A. Requlatorv Requirements, Licensinq Commitments and Desiqn Bases

10 CFR 50 Appendix A General Design Criterion 4 states that "structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents."

Standard Review Plan (NUREG-0800) 3.4.1 states that the review of "plant flood protection includes all structures, systems and components (SSC) whose failure could prevent safe shutdown of the plant or result on uncontrolled release of significant radioactivity ..." and that this review "also includes consideration of flooding from internal sources."

SSES FSAR 6.3.1.1.3 states that separation barriers for ECCS "shall be constructed between the functional groups as required to assure that environmental disturbances such as fire, pipe rupture, falling objects, etc:, affecting one functional groups will not affect the remaining groups. In addition, separation barriers shall be provided as required to assure that such disturbances do not affect both RCIC and HPCI . " SSES FSAR 9.1.3.3 states that "the design makeup rate from each ESW loop is based on replenishing the boil-off from the MNHL in each fuel pool for 30 days following the loss of FPCCS capacity . Minutes from Bechtel meeting on HVAC systems (February 1980) states that original requirement for SGTS was "to handle fumes from a boiling fuel pool," but that SGTS will not be able to handle this mixture since the room will become too hot. "This requirement will be deleted from the FSAR."

An internal PP&L engineering work request (EWR 830658, March 1983) noted Itcondensation may be expected from this


safety Consequences of a Boiling Spent Fuel Pool

evaporation which will run down to lower levels of the R.B. Will this cause loss of essential equipment, particularly electrical? Has an evaluation been performed?" The response to these questions was "This is an inappropriate format to ask questions. Comments were requested and none received. Furthermore, no budget exists with which to fund the engineering time required to respond to these questions."

B. Concerns

The ultimate heat sink and ESW are designed to provide 1.5 million gallons of water to each spent fuel pool over the 30 day period. In the LOCA-LOOP condition, the reactor building HVAC system in Zone I, I1 and 111 isolation mode recirculates refueling floor air throughout all three zones. The water added to the spent fuel pools ends up in the reactor building following boil-off and overflow. The effects of this water on the safety-related structures, systems and components in the reactor buildings have not been included in design analyses. The ECCS and RCIC room coolers are known not to be designed for latent heat effects. Dampers in the SGTS and RB-HVAC system close when the entering air temperature exceeds 165OF, while the boiling spent fuel pool was calculated to produce air temperatures of =180°F. The potential for common mode failures of multiple ECCS and safety-related systems such as the standby gas treatment system exists. Failure of one or more of these safety-related systems could increase the consequences of postulated accidents.

4. Reactor Building Heat Loads

A. Resulatorv Requirements. Licensinq Commitments and Desisn Bases

10 CFR 50.49 requires that electrical equipment must be qualified to the temperature "for the most severe design basis accidents . " 10 CFR 50 Appendix A General Design Criterion 4 states that "structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents."

An internal PP&L engineering work request (EWR 830658, March 1983) noted "the initial boiling rate corresponds to -"3000 cfm of 100% water vapor at one atm. Is the equipment which will be exposed to this atmosphere qualified for it?" The response

July 27, 1992 Page 10


to this question was "This is an inappropriate format to ask questions. Comments were requested and none received. Furthermore, no budget exists with which to fund the engineering time required to respond to these question^.^^

A PP&L engineering report (SEA-ME-099, December 1987) analyzed reactor building temperatures for LOCA, LOCA/LOOP and LOCA/false LOCA cases assuming spent fuel temperatures remained at 125OF, but listed as a nonconservatism that fuel pool heatup in the LOCA/LOOP case would result in higher heat loads from the RHR systems, fuel pool walls and fuel pool surf ace.

B. Concern

Secondary containment design analyses are required to account for all heat loads in the reactor building including from the boiling spent fuel pool. The existing design reactor building heat load calcs consider sensible heat from the boiling pool, but neglect latent heat. These calcs indicate little margin to equipment qualification temperature limits in many rooms for a maximum heat load in the reactor building of approximately 5 . 5 ~ 1 0 ~ BTU/hr. The total design heat load from the spent fuel pools is 26.4x106 BTU/hr, which would add at least approximately 20.9~10~ BTU/hr to the existing maximum heat load. Even the current heat loads in the spent fuel pools could increase the maximum heat load in the reactor building by ~50%.

The remaining five concerns raised in EDR G20020 involved nonconservatisms in analyses for the boiling spent fuel pool event.

DISCUSSION OF OPPOSING VIEWPOINT

The discussions and meetings which have occurred since EDR G20020 was initiated have yielded one primary argument against the issues raised in EDR G20020 having nuclear safety significance. This argument is that the licensing bases LOCA/LOOP accident for SSES does not assume a boiling spent fuel pool resulting from the event. In order for this assumption to be valid, spent fuel pool cooling must either not be lost or must be restored prior to boiling. There are several faults in this assumption:

1) Since the FPCCS and the SW system are non-safety related systems, their components are not included in the equipment qualification program and may not survive the pressure, temperature, humidity and radiation environment in the reactor building following a postulated accident. Therefore, the

July 27, 1992 Page 11


FPCCS which is definitely lost following a LOCA/LOOP may also be lost following a LOCA without a LOOP.

2) Since the fuel pool cooling assist mode of RHR is a non-safety related function, its components are not included in the equipment qualification program and therefore may not survive the pressure, temperature, humidity and radiation environment in the reactor building following a postulated accident. In addition, the fuel pool cooling assist mode of RHR has not been utilized since the initial startup testing program and its valves were removed from the inservice inspection program several years ago and the valves may have experienced failures which have not yet been detected which would prevent their successful operation. Therefore, the fuel pool cooling assist mode of RHR may be lost following a LOCA/LOOP and a LOCA without a LOOP.

3) The fuel pool cooling assist mode of RHR requires the manual opening of valves in the reactor building which may be inaccessible following a postulated accident due to radiation levels.

4) For the LOCA/LOOP case, it has been argued that the SSES design implicitly assumes restoration of offsite power typically within 24 hours and essentially always within 48 hours after event initiation. SSES FSAR Chapter 8 reports PP&L grid experience in support of these restoration times. However, no documentation was found which states that PP&L has defined the LOOP duration for design bases events. As EDR G20020 and EDR GO0005 both address, the spent fuel pool may begin boiling in less than 24 hours. In any case, the reactor building temperature analyses for equipment qualification purposes presently counter any such credit for restoration of offsite power since non-Class 1E power loads in the reactor building may be shed z24 hours after offsite power is restored in order to satisfy room temperature limitations.

EDR G20020 identified concerns with the SSES design provisions for the boiling spent fuel pool event. The SSES design, coupled with current operating procedures, would have significant nuclear safety consequences if a loss of spent fuel pool cooling occurred. Therefore, these concerns must be resolved for SSES. In addition, many of these concerns are applicable to other BWRs and possibly even PWRs in the United States. Therefore, these concerns must be reported to INPO/NRC in order for the adverse condition to be remedied throughout the industry.

July 27, 1992 Page 12


PP&L Memo f r o m G . D . M i l l e r t o G . T . Jones, " F u e l Poo l C o o l i n g D e f i c i e n c i e s " , Augus t 18, 1992 (ET-0586)

Note : T h i s memo b y t h e PP&L S u p e r v i s o r , E n g i n e e r i n g P r o j e c t s p r o v i d e s an i n d i c a t i o n of how PP&L na r rowed t h e i r scope o f e v a l u a t i o n f o r t h e conce rns i n EDR 620020 t o j u s t t h e d e s i g n o f t h e f u e l p o o l c o o l i n g and c l e a n u p system. W i t h t h e e x c e p t i o n o f t h e i n s t r u m e n t a t i o n f o r t h e f u e l p o o l s , t h e d e s i g n o f t h e f u e l p o o l c o o l i n g and c l e a n u p sys tem has n o t been c h a l l e n g e d i n EDR G20020 and i t s subsequent s u p p o r t i n g documents. The conce rns a r e t h a t t h e e f f e c t s o f b o i l i n g s p e n t f u e l p o o l s on other systems and components i n t h e r e a c t o r b u i l d i n g have n o t been a d e q u a t e l y a n a l y z e d .

Page 1

M E M O R A N D U M

DATE: 8/18/92

TO: 6. T. Jones A6-2

FROM: G. D. M i l l e r A6-3

JOB: Engineering NUMBER: ET-0586 COPIES: D i s t r i b u t i o n Techno1 ogy Corres. F i l e A6-2

Engr Tech F i l e A6-3

FILE: A45-1A REPLY: Not appl i cab1 e

SUBJECT: Fuel Pool Cool i n g Def ic iencies

The fol lowing act ions are being taken i n response t o concerns ra ised about the EDR program and the fue l pool cool ing issues described i n EDR 620020 and EDR 600005.

The EDR process governing procedure, EPM-QA-122, w i l l be revised as f o l l ows:

1. The appeal process described i n paragraph 5.4 w i l l be revised w i th the f i r s t step o f the appeal changed t o the Engineering Review Committee. Subsequent appeal may be t o e i t he r the Manager - Nuclear Engineering, Superintendent - SSES, o r Manager - NSAG.

2. The management review described i n paragraph 5.14 w i l l be revised t o include:

a. P e r i o d i c r e v i e w o f a l l inva l idEDRsbytheEngineer ingRev iew Committee, and

b. Per iodic review o f a l l EDRs open greater than s i x months by the Engineering Review Committee.

Fuel Pool Cool i nq Def ic iencv Resolution

Engineering Technology has respons ib i l i t y t o resolve both EDRs (EDR GO0005 was previously assigned t o System Engineering). These EDRs are assigned t o Mark Mjaatvedt and are being worked f u l l time by Michael Crowthers. The EDR

EPM-1016, Rev. 1

Page 3

M E M O R A N D U M (CONTINUATION SHEET)

We p lan t o keep the o r ig ina to rs o f these concerns informed o f our progress as we work our way through t h i s e f f o r t . A formal plan inc lud ing schedule f o r the above a c t i v i t i e s i s under development.

Systems Analysis (Kevin Brinckman) i s i n the process o f conducting an independent design review o f these issues. Addi t ional ly, I have requested t h e i r review o f t h i s issue from an IPE perspective when resources become avai lable.

As a fol lowup a c t i v i t y I plan t o request an assessment o f the EDR process from an independent organization. This assessment w i l l focus s p e c i f i c a l l y on the va l i da t i on and v e r i f i c a t i o n steps o f the process. This should be conducted by NSAG. NQA has once i n the past conducted an audi t o f the process. They do not agree tha t discreDancies are not necessari ly deficiencies. They def ine any discrepancy t o be a condi t ion adverse t o qual i ty , whereas our program recognizes the po ten t i a l f o r discrepancies i n documentation which do not cons t i t u te actual d e f i c i e n t condit ions. Other programmatic audits have taken place on the EDR process, but none have examined the philosophy o r c r i t e r i a used t o determine the v a l i d i t y o f engineering issues.

EPM-lOlC, Rev. 1 (51)

Attachment 10

PP&L Memo f r o m D. C . P r e v a t t e t o G . T . Jones , " F u e l Poo l C o o l i n g D e f i c i e n c i e s " , August 20, 1 9 9 2 ( E T - 0 5 8 7 )

M E M O R A N D U M

TO :

FROM :

JOB :

NUMBER:

SUBJECT:

G. T. Jones A6-2

w D. C. Prevatte A6-3

Engineering Technology cc :

ET-0587 REPLY:

Fuel Pool Cooling Deficiencies

DATE: August 20, 1992

FILE: A45-1A

Distribution Corres. File A6-2 ET File A6-3

This memo is written in response to Mr. G. D. Miller's memo ET-0586 of 8/18/92 concerning the discrepancies ansociated with the fusl pool cooling system described in EDRs G20020 and G00005.

Mr. Miller's memo states that, "Our initial evaluation of the EDRs has concluded that the safety significance is minimal." I strongly disagree with this evaluation and I hereby request that the safety significance of these EDRs, particularly EDR G20020 be reevaluated for the following reasons:

1. The primary basis given for this conclusion is that "...the design of the fuel pool cooling system was specifically reviewed and approved by the NRC with full knowledge of the fact that the FPCCS was not a safety related system and that fuel pool boiling could be expected to occur under a specific combination of hypothetical conditions."

I consider this basis to be invalid for the following reasons:

a. This basis appears to miss most of the main points of EDR G20020. It focuses on the non-safety related FPCCS which is not the concern. The primary concerns are with the NRC mandated (Reg. Guide 1 1 3 ) safety-related backup cooling scheme of allowing the fuel pool to boil and providing makeup water from the safety-related ESW system. The concern is the potential inability of the operators to put this scheme into effect because of inaccessibilityto the associated valves due to post-LOCA radiation levels in the reactor building, and the potential negative effects of a boiling spent fuel pool on virtually all of the safety-related systems in the reactor building, effects which have not been analyzed.

b. The I# . . . specific combination of hypothetical conditions ..." referred to in the memo is LOCA/LOOP. This is not some off-the-wall accident scenario as the response seems to imply. This is the standard, universally recognized, NRC mandated design basis accident (DBA).

G. T. Jones Fuel Pool Cooling Deficiencies

August 20, 1992 Page 2

The conditions of concern are not postulated. They are mechanistic consequences of that DBA.

2. The memo also cites as another basis, confusion concerning what was originally considered in the design, the changes to the fuel design, and outage practices which have not been accounted for in the FSAR analysis. This basis would seem to support the EDRs contentions, not refute them. If indeed there is confusion on these subjects, then, at best, the condition is unanalyzed and therefore by definition in NRC regulations and our procedures, a safety concern.

Although this information is certainly pertinent to a historical perspective of these concerns, to determination of the magnitude of the heat loads involved, and to formulation of the corrective actions that may be effected, it is not pertinent to the ability of the plant to perform as required for the DBA conditions. The information to make this determination is clear and available today.

3. Another basis cited is that 'I.. . the design and procedural features which exist today provide a reasonable level of assurance that the actual safety consequences are minimal." There is no elaboration on what these design and procedural features are. In conversations with Mr. Miller and others who seem to consider the EDRs as having very low safety significance, no design or procedural features have been cited. The only features that have been cited are "heroic action" of the operators, the EOPs, an EOC staff who will understand the concern and do whatever needs to be done, and a low probability of occurrence.

These are not valid features. Heroic operator action is not a valid basis for the design of a plant, nor are the EOPs (even if they were correct in this area) which address many conditions potentially outside the plant licensing and/or design bases.

And, contrary to the memo's contention, the EOPs as they stand today are not correct. They currently tell the operator he has a minimum of 25 hours until the fuel pool boils. Under worst case conditions, it may be less than half that time; and with the LOOP conditions, he has no instrumentation to tell him the condition of the fuel pool. Under DBA conditions, the operator is flying blind using nonconservative information.

Additionally, current EOPs move the plant toward the conditions of concern, not away from them. The current EOPs require deenergizing the non-1E loads in the reactor building


August 20, 1992 page 3

at t=24 hours if reactor building temperatures are as analyzed. This, in effect, imposes a LOOP on the reactor building, thus initiating the conditions of concern.

The knowledge of the EOC staff is also not a valid feature if the conditions of concern are not formally addressed in any official design and/or procedural documents. Although it is claimed that today's staff would understand the concerns, there is no reason to believe this is true since the concerns aren't documented outside the EDRs and there is no training on this eventuality. And ten years from now, if the concerns are not formalized in writing, they will he even less understood.

Additionally, even if the staff does understand, if the conditions are not analyzed, which they are not, the plant could be brought to a condition where recovery is not possible in spite of their full understanding.

Low probability is also not a valid feature. This is discussed in detail further in this memo.

4. The statement in the memo regarding the NRCts It... full knowledge of the fact that the FPCCS was not a safety-related system and that fuel pool boiling could be expected to occur . . ." seems to imply that if the NRC approved it as is, that makes it acceptable even if we discover discrepancies that may not have been originally considered. I am aware of no evidence that indicates the NRC approved of our design with the understanding that: (a) the operator would be exposed to unacceptable radiation levels under design basis conditions in effecting the FSAR described fuel pool boil scheme for alternate cooling; and (b) that the boiling fuel pool might create a myriad of unanalyzed conditions in the reactor building that could threaten the operability of many of the safety-related systems in the building.

5. The memo concludes that the safety significance is llminimal.ff Per procedure EPM-703, Rev. 0, Section 5.3, Is.. .a 'minimal' classification generally signifies a documentation type of discrepancy." In other words, not a real engineering concern, but rather a documentation error that can be resolved by making editorial changes to the documents. Per this procedure, if an EDR's safety significance is classified as llminimalfl, it does not even have to be evaluated for operability and reportability.

These EDRs are not in any reasonable evaluation just a documentation discrepancy. They are fundamental engineering concerns raised by two engineers intimately familiar with the systems after exhaustive research. To dismiss these concerns



by classifying them as just a documentation discrepancy is ludicrous. If concerns such as these don't even get to the stage in the process where they are required to be evaluated for operability and reportability, concerns which involve the safety of the operators and potential threat to virtually every safety-related system in the reactor building, then what does it take to trigger operability and reportability evaluations? The threshold appears to be much too high.

6. The memo's reference to It.. . a specific combination of hypothetical conditions ..." implies a probabalistic argument as to why the safety significance is "minimal." Indeed, in conver:sations witii, Mr. Miller and others this argument has been explicitly raised. This argument is not valid with regard to design bases for several reasons.

First, our design basis conditions of LOCA/LOOP which produce the conditions of concern are mandated by regulation. That, for design purposes, dictates a probability of 1.

Second, even for LOCA without a LOOP, our current EOPs dictate a self-imposed LOOP on the reactor building at 24 hours, again making the probability for LOCA/LOOP equal 1.

Third, even for a LOOP or FPCCS system failure without a LOCA, the consequences of fuel pool boil are unanalyzed.

Fourth, EDR procedure EPM-703, Rev. 0, Section 5.3, has as a caution, capitalized, bold letters and underlined as follows, "The EDMG Evaluator must not put heavy emphasis on the perceived small probability of occurrence or the expected satisfactory outcome of analysis or reanalysis to justify continued operation with the existing discrepancy." Section 5.4 goes on to say, "SAFETY SIGNIFICANCE must be based on the potential adverse consequences of failures, even those of very low probability." Thus, by our procedures, potential consequences should be the dominant factor in evaluating safety significance, not probability. The potential consequences of the concerns raised in these EDRs, and subsequent documentation generated by Mr. Lochbaum and myself, are very grave.

The EDR process at PP&L was developed in response to a 1990 SALP inspection finding that safety significant issues were not being handled in a timely manner. Our EDR procedures are filled with words that reflect this concern; words like quickly, expeditiously, immediately, early, timely. For the step where we are today, the "screeningN step, the procedures1 (EPM-703, Rev. 0, Section 5.2)



intent is that EDRs be llquicklyll screened after a discrepancy enters the EDR process. As of today, the official screening still has not been completed four months after the EDR entered the process, and approximately one month after Mr. Lochbaum and I personally brought these concerns to your attention. Neither the intent of the procedure nor the intent of the NRC are being fulfilled.

The plan outlined in Mr. Miller's memo for the "final e~aluation~~ would appear to further delay the required actions. Although all of the activities in the plan are important to understanding the problems more completely and effecting the most effective solutions, none of them are prerequisites for performing a valid "screening", and most of them are not required to determine operability and reportability. To make these activities prerequisites for a final "screening" evaluation and then for the operability and reportability determinations, is to further delay the process unnecessarily. The information is available to make these determinations today, and they should be made immediately if we are to do what's legitimately required of us.

I have made my own operability and reportability determinations based on extensive research on these concerns. At best, the operability of the fuel pool cooling in the boil and feed mode, the fuel pool instrumentation, and much of the safety-related equipment in the reactor building is unanalyzed with regard to the effects of the boiling fuel pool on this equipment, with strong indications that analysis would show it as inoperable. If this is the case, per 10CFR50.72 and 50.73, it is reportable.

I would welcome any hard, definitive, documentary information indicating that my conclusions are wrong. Both Mr. Lochbaum and I, and for that matter, many others who would like to see different conclusions, have searched for contrary evidence. To the best of my knowledge, none has been found.

This is not to say that the plant should necessarily be shut down. I believe that very credible arguments can be made for a J.I.O. I therefore don't understand why there is such an apparent reluctance in the organization to acknowledge these concerns and move ahead with resolution expeditiously. Although resolution will have a cost, certainly, that cost does not necessarily have to include plant shutdown.

I therefore strongly urge that the formal screening evaluation and the evaluations of the operability and reportability of these concerns proceed without further delay with priority over all other activities in Mr. Miller's plan, and that we expeditiously get on with the process of resolving these concerns.



I sincerely appreciate your continued personal attention in these matters, and I am at your service in addressing these concerns.

DISTRIBUTION:

J. E. Agnew F. G. Butler R. G. Byram M. H. Crowthers G. D. Gogates

G. J. Kuczynski D. A. Lochbaum

A6-3 A6-3 A6-1 A6-3 SSES

SSES, S&A Enercon

Miller Mittenberger Mj aatvedt Myers Ref ling Stef anko Sweeney Zola

A6-3 A6-1 A6-3 A2-4 A6--3 A9-3 SSES A6-3


PP&L Memo f r o m A. D y s z e l t o T. C . D a l p i a z , "U2 R105 F u e l Poo l Decay Heat E v a l u a t i o n " , Augus t 21, 1992 (PL I -72230 )

Note: T h i s l e t t e r t r a n s m i t s i n t e r i m g u i d a n c e t o t h e SSES s i t e p e r s o n n e l f o r use d u r i n g an upcoming r e f u e l i n g ou tage . T h i s gu idance i s n e c e s s a r y because EDR 600005, i n i t i a t e d i n September 1990, has n o t y e t been d i s p o s i t i o n e d and t h e a p p l i c a b l e d i s c u s s i o n s i n FSAR S e c t i o n 9.1 and Append ix 9A a r e no l o n g e r a c c u r a t e .

SUSQUEIII\WWA S T E N ELECTRtC STATION US RlOS FUEL POOL DECAV MEAT EVALUATION CCN 711087 &72230

FILC-

Refmencar: I] PLI-67B13, *Fuel Pool Cooling Heat Loads - LOR 600005." Apr i l 12, 1991.

2) PLI-70395, 'U1 R106 Fuel Pool Decay Heat Evaluation,* February 7, 1992.

3) 0-0686, "Fuel Pool Cool ing Defictencies." 6/16/92.

This memo provides short t a n re l i a f f o r the opan EDR, WM#)S, (Referancs I ) wtth rrspect t o the upcmlnq UZ 5R10 by providinp WFE's avr lur t ion o f the rasul t in ti# constrrlnts f o r per fomncr o f comon MR system outage work,

!O stmitar the U1 RlM avalurtlbn (Itofarenee 2). A long te rn solution te. th ls EDR i s requtred t o ansum future successful, t t u l y outa s. I n short, the U2 r 3R10 ovrluation shows no chamga tn the currrnt ou tqa sc edule (i.r., Sapt. 28 for cwnon RHR systmm wrk) t o r no fuel f r i l u ro r I n Unit O Cyclr 5. However, future evduattonr, which rtll involva hiphrr hart loads urd rddress the p o t m t h l for furl f ~ i l u r o ( r , rill l lkoly r f f ac t the outage schudula. Rofwencr 3 provides r dercr \ pt ion o f the long term solution to th is problem.

fShR Section 9.A descr iks the radloloplcrl ralarse results from a loss o f f u r l pool eoollnp avmt. The asscuptlons urod fo r thr FW analysls include b core ralords, i n Cora fwl rhu f f l i n , a d r uxllwr fwl rxposun o f 211.609 mP/RlU. Tho current o arr t lon at !usquohmnr SES Sncludrs r fwl relead batch slzo o f r o x i u l y b o f the core, r n u a i m fu r l dtschrrgs rnposura 3 fi P, of 10,000 lYD/ , and I full c o n o f f lo rd f o r r rch mutaga. I f OM or more fual f r l l u n s m suspmctod to hrva occurrod duri t h r oplrat inp cycla Just

P "R p r io r to m outrga or dmrin fuel handling af ter s utdom, rnrlyser must br prrformod t o rssurr the rad ologtcrl rolaase frca th. postulrted loss o f full1 pool cw l tnp avant u o l os t than those prasantd in thr FSAR. Thw FSAR anrlyrms am bounding pravidod t h r t conw, WIR r y s t w outage work i s not startad u n t i l the docay h o d l w e l i s low enoqh t o provant fuel pool bal l ing I n less than 05 hours. If no fual h l l u r e s rrr rurprctmd t o hrva occurred durimg tha oparath cycla just p r io r to an outaga o r during fuel hmdl ln 'I r 0 after shutdown, ma ysar must br p a r f o w d to asrurr the fuel 01 water eve1 c m bo utn ta lnod durtng & lobs o f fuel pool cwt tng eumt. T e fuel pool

To address thts issue for t& upcmlng outwe, Nuclear fuels Englnaartng has calculated the t o t r l &cay heat o f the fuel I n tho spent fuel pads for the uz 5Rt0 conslstmt with the rpproaeh i n h fe rmce I , This decoy hen* l e v d dnc1ud.r r to ta l o f th. decay heat frw the fuel i n tk. UI pool, U2 ool. and the fu l l core off1o.d. Figu*. 1 show* the crlculatod spent fuel poo 7 decay heat as r function of ttm after shutdm. The Curve Irbel led wnomlnalD I s r t r l cu l r t lon o f the decay hart based on the methodology tn NUREG-0800. This fueth&Jogy has been show t o roduce higher decay heat fevelc thrn the more rigorous mtbodology I n the AN ! 5.3-1979 decay heat standard. The curve labellad *mwIwmm i s also based on the lmthodolagy +n NUREG-0600 but accounts for uncertainty tn th r reactor powat level (1u) and uncertainty i n the decay h a t 1~1thodo1q)y.

Nucltar Fuels Enginerrin has also uer fomd a calculatlon t o dstermlnn. the 8 t ime after the U X S shut awn that the srprt fuel pool boi l ing rate during a postulated loss of spent fuel pol coollng event i s less than the 60 tPn ESU makeup flow rata. Based on the *.rr1rmnm decay heat curve i n Figure 1. a spray pond temperature o f 90°F, and open fu r l pool pates, the f u r l pool boil ing rate during a postulatrd loss o f fuel pool e o o l l n ~ event i s less than the 60 GPH ESY mrkru rate subsequent t o 14 days after martor shutdown. P Therefore, NfL's ova urt lon hdtcr tes that provided fuel Failure d y r not occur durlng the rrrulnlng U2C5 operation or durtng fuel hand1 ing ~ftr r shutdown, c m RtlR systm outage work should not start unt t l at b a s t 14 days after reactor shutdown. I f r fuel fa i lure occurs. a calculation should be prrformad t o dateminr t f further outage r r s t r l c t~ons rre nrcersary. Note that the above calculrtinns haw barn d o w n t e d i n WFE-B-M-046, h v . 2 sad indopendently reviewed i n rccordance with QA procedurrr EPII-QA-301.

Thr 14 day rmstrlctlon on c n m n c m n t o f cmmn RHR system work does not impact the current U t SRIO outage schedule.

A. Dyszel Nuclear Fuel Hanagement Project Enptneer Nuclear Fuels Enginwrinp

CC: J, ~ . ~ g n e w SSES A. J. R O S C ~ O ~ ~ A94 K. W. Harwanko SSES R. A. Saccono SSES 6. T. Jones A6-Z J. P. Spadaro A*-3 3. n. ~ u l i c k 49-3 J. 5. Stefanko AS-3

A 9 4 J. Zola A64 N4l F i l r


PP&L Memo f r o m J . M . Kenny t o G . T . Jones and C . A . M y e r s , "EDR on F u e l Poo l C o o l i n g " , August 2 5 , 1992

N o t e : T h i s ' c o n f i d e n t i a l ' memo i s t h e f i r s t documented i n d i c a t i o n t h a t t h e NRC had been i n f o r m a l l y n o t i f i e d o f t h e c o n c e r n s r a i s e d i n EDR G20020.

August 25, 1992

G. T. Jones A6-2 C. A. Myers A2-4

EDR ON FUEL POOL COOLING

On August 24, 1992 I briefed both Swtt Barber and Jim Raleigh of the NRC on the status of our review of contractor originated fuel pool cooling wncerns documented on an EDR. I noted that our current position was there were no immediate wncerns with system operability or need for reportability under regulations identified but that our efforts were continuing to address the identified issues. I also noted that George Jones had discussed the concerns with the contractors and was personally involved in resolving the issues.

I had previously briefed Jim Raleigh in July of the fuel pool concerns and reviews being performed. Swtt did bring to my attention an open inspector finding concerning the Haddam Neck fuel pool draindown event and subsequent efforts by NSAG on fuel pool issues. He noted there were 28 open items and that we should review these issues for status. I indicated it was my understanding Engineering would be addressing the NSAG open issues on the fuel pool as part of their effort to resolve the open EDR.

cc: J. E. Agnew G. D. Miller J. R. Miltenberger R. R. Sgarro H. G. Stanley

i

A6-3 A6-3 A6-1 A2-4 SSES

JMK: tah FuelPool.EDR

Attachment 13

PP&L Memo from George T . Jones to Glenn D. Miller, "Fuel Pool Cooling EDR's 620020, G00005", August 27, 1992 (PLI-72267)

August 27, 1992

Glenn D. M i l l e r A6-3

SUSQUEHANNA STEAM ELECTRIC STATION FUEL POOL COOLING EDR'S 620020, GO0005 PLI-72267 FILE A45-1A

Reference: ET-0587, ET-0586

The re fe rence l e t t e r s show me t h a t t h e r e s t i l l i s a d i f f e r e n c e o f p ro fess iona l o p i n i o n r e l a t i v e t o t h e s i g n i f i c a n c e o f t h e sub jec t EDR1s. T h i s d i f f e r e n c e i s t o be resolved. Th is i ssue i s t o be worked e x p e d i t i o u s l y u n t i l we have reso l ved t h e ou ts tand ing quest ions.

I wish t o p o i n t ou t t h a t our b e l i e f t h a t a design has been reviewed and approved by NRC, i s n o t adequate j u s t i f i c a t i o n f o r c l a s s i f i c a t i o n o f t h e s i g n i f i c a n c e o f an issue. The issue must stand on i t ' s own m e r i t s .

I am f u r t h e r concerned t h a t ou r ac tua l c o n f i g u r a t i o n and method o f ope ra t i on apparent ly d i f f e r s f rom t h a t descr ibed i n t h e FSAR. The FSAR i s ou r l i c e n s e d bases and any d e v i a t i o n from t h a t d e s c r i p t i o n i s requ i red t o have a thorough and complete documented eva lua t i on on f i 1 e.

There were twenty -e igh t open i tems r e s u l t i n g from NSAG Review o f Fuel Pool Cool ing. These need t o be inc luded i n t h i s review.

I wish t o have t h e schedule f o r r e s o l u t i o n o f t h i s i ssue acce lera ted and t h e c l a s s i f i c a t i o n o f t h e s i g n i f i c a n c e o f t h i s i ssue reevaluated. I am expect ing a t l e a s t d a i l y updates o f ou r progress. .PFease- - I f you have & questions, p lease c a l l me. A-Y* George 5

Attachment

cc: J. E. Agnew F. G. B u t l e r R. G. Byram M. H. Crowthers G. D. Gogates G. J. Kuczynski D. A. Lochbaum S. M. Hauseman

A6-3 w/a 0. C. P revat te A6-3 w/a J. R. Mi l tenberger A6-1 w/a M. R. Mjaatvedt A6-3 w/a C. A. Meyers SSES w/a J. G. R e f l i n g SSES, S&A w/a T. J. Sweeney Enercon w/a J. A. Zo la A6-2 w/a Nuc. Rec. F i l e s

A6-3 w/a A6-1 w/a A6-3 w/a A2-4 w/a A9-3 w/a SSES w/a A6-3 w/a A6-2 w/o

DATE: 8/18/92

Page 1

TO: -!

FROM: G. D. M i l l e r A6-3

JOB: Engineering NUMBER: ET-0586 COPIES: D i s t r i b u t i o n Techno1 ogy Corres. F i l e A6-2

Engr Tech F i l e A6-3

FILE: A45-1A REPLY: Not appl icab l e

SUBJECT: Fuel Pool Cool i n g Def ic iencies

The fo l low ing act ions are being taken i n response t o concerns ra ised about the EDR program and the fue l pool cool ing issues described i n EDR 620020 and EDR G00005.

EDR Process Chanqes

LJ,~$ The EDR process governing procedure, EPM-QA-122, w i 11 be r e v i red as fo l lows:

1. The appeal process described i n paragraph 5.4 w i l l be revised w i t h the f i r s t step o f the appeal changed t o the Engineering Review Committee. Subsequent appeal may be t o e i t he r the Manager - Nuclear Engineering, Superintendent - SSES, o r Manager - NSAG.

2. The management review described i n paragraph 5.14 w i l l be rev ised t o include:

a. Periodic review o f a l l i n v a l i d EDRs by the Engineering Review Committee, and

b. Periodic review o f a l l EDRs open greater than s i x months by the Engineering Review Committee.

Fuel Pool Cool i nq Deficiency Resolution

Engineering Technology has respons ib i l i t y t o resolve both EDRs (EDR GO0005 was prev ious ly assigned t o System Engineering). These EDRs are assigned t o Mark Mjaatvedt and are being worked f u l l time by Michael Crowthers. The EDR

EPM-lOlB, Rev. I

Page 2

n E n o R A N D u n (CONTINUATION SHEET)

evaluations (safety s igni f icance, operabilitylreportability) are being done by Jim Agnew and Joe Zola.

evaluation o f these EDRs has concluded t h a t the safety s ign i f icance This i s based pr imar i l y on our understanding tha t the design o f the

ool i ng system was speci f i c a l l y reviewed and approved by the NRC w i th f u l l knowledge o f the f ac t tha t the FPCCS was not a safety re la ted system and tha t fue l pool b o i l i n g could be expected t o occur under a spec i f i c combination o f hypothet ical condit ions. However, i t i s not ye t c lear t o what extent the NRC (o r the indust ry) considered the long-term e f fec ts o f the fue l pool b o i l i n g condi t ion. The record on t h i s subject i s confusing and fu r the r complicated by 7 changes t o the fue l design and actual outage pract ices which have not been accounted f o r i n the FSAR analysis. We be1 ieve tha t the design and procedural features which e x i s t today provide a reasonable leve l o f assurance t h a t the I 0 actual safety consequences are minimized. However, procedure enhancements and addi t ional operator t r a i n i n g are c l ea r l y required as p a r t o f the reso lu t ion o f these concerns. This evaluation w i l l be f u l l y documented as pa r t o f the revised EDR package.

A f i n a l evaluat ion o f t h i s concern i s predicated on completion o f a h i s t o r i c a l review o f a l l ava i lab le documentation. Thus, our plan f o r reso lu t i on includes:

Complete invest igat ion o f h i s t o r i c a l design and l i cens ing information, including requests f o r informat ion from the o r i g i n a l design organizations (GE and Bechtel),

Establ ish the fue l pool and fue l pool cool ing system design basis based on the derived design basis and current operating pract ice,

Review fue l pool designs o f other b o i l i n g water reactors,

Complete a new analysis o f the fue l pool and fue l pool cool ing system based on the establ ished design basis,

Prepare a point-by-point comparative descr ip t ion o f our current operating p rac t i ce and design basis against the o r i g i n a l FSAR analysis ( igc lud ing the ind iv idual issues i d e n t i f i e d i n the EDR),

L i s t and assess each dev ia t ion from the o r i g i n a l analysis as described i n i tem 5,

Prepare an ope rab i l i t y evaluation accounting f o r each dev ia t ion as addi t ional information becomes available,

Re-evaluate a l l issues f o r repor tab i l i t y (ongoing),

Prepare recommendations t o resolve each issue described i n i tem 4.

Page 3

H E H O R A N D U H (CONTINUATION SHEET)

We plan t o keep the o r ig ina to rs o f these concerns informed o f our progress as we work our way through t h i s e f f o r t . A formal plan inc lud ing schedule f o r the above a c t i v i t i e s i s under development.

Indeoendent Review bv Svstems Analysis

Systems Analysis (Kevin Brinckman) i s i n the process o f conducting an independent design review o f these issues. Addi t ional ly , I have requested t h e i r review o f t h i s issue from an IPE perspective when resources become avai lable.

Assessment o f EDR Process

As a fol lowup a c t i v i t y I plan t o request an assessment o f the EDR process from an independent organization. This assessment w i l l focus speci f i c a l l y on the va l i da t i on and v e r i f i c a t i o n steps o f the process. This should be conducted by NSAG. NQA has once i n the past conducted an audi t o f the process. They do not agree t h a t d i sc re~anc ies are not necessari ly def ic ienc ies. They def ine any discrepancy t o be a condi t ion adverse t o qua1 i t y , whereas our program recognizes the po ten t i a l f o r discrepancies i n documentation which do not cons t i tu te actual de f i c i en t condi t ions. Other programmatic audits have taken place on the EDR process, but none have examined the philosophy o r c r i t e r i a used t o determine the v a l i d i t y o f engineering issues.


Page 4

D i s t r i b u t i o n

J. R. Mi l tenberger F. G. B u t l e r C. A. Myers R. G. Byram J. S. Stefanko G. J. Kucyznski J. E. Agnew J. A. Zola M. R. Mjaatvedt M. H. Crowthers D. C. Prevat te D. A. Lochbaum

M E M O R A N D U M (CONTINUATION SHEET)

A6- 1 A6-3 A2-4 A6- 1 A9-3 SSES S&A A6-3 A6-3 A6-3 A6-3 A6-3 Enercon

GEORGE T. JONES. MAN4GER


Attachment 14

PP&L Memo from Glenn D. Miller to George T. Jones, "Fuel Pool Cooling EDRs 620020, G00005", August 31, 1992 (PLI-72297)

August 31, 1992

George T. Jones A6-2

SUSQUEHANNA STEAM ELECTRIC STATION FUEL POOL COOLING EDRs 620020, GO0005 PLI-72297 FILE A45-1A

Reference: PLI-72267, ET-0587, ET-0586

I n response t o your l e t t e r PLI-72267 we are con t i nu ing t o work t o r e s o l v e t h e issues i n t h e re ferenced EDRs. As I exp la ined t o you p r e v i o u s l y I met w i t h M r . Prevat te and discussed h i s concerns a t l e n g t h on August 21. We acknowledged our d i f f e r e n c e s and agreed t o cont inue working toward r e s o l u t i o n .

I want t o reemphasize t o you t h a t we are n o t us ing t h e p r i o r rev iew and approval o f ou r system design by t h e NRC as a bas i s f o r t h e c l a s s i f i c a t i o n o f t h e s a f e t y s i g n i f i c a n c e o f t h i s issue. Statements which I made i n ET-0586 r e l a t i v e t o t h e s a f e t y s i g n i f i c a n c e were in tended t o sununarize t h e EDMG p o s i t i o n on screening. The re fe rence t o t h e NRC i s o n l y a statement o f t h e f a c t t h a t our design ch i losophy a t t h e t ime o f l i c e n s i n g was reviewed and approved by t h e NRC. Th is p o s i t i o n i n f a c t c o n s t i t u t e s ou r l i c e n s i n g basis .

The FSAR conta ins re ferences t o analyses regard ing t h e Fuel Pool Cool ing and Cleanup System and t h e Fuel Pool. It does n o t descr ibe t h e exact manner i n which we operate t h e p l a n t . Our c u r r e n t f u e l design and f u e l c y c l e dev ia tes from t h e FSAR d e s c r i p t i o n . Th i s i s t h e sub jec t o f G00005. Our outage p r a c t i c e s d i f f e r from t h e d e s c r i p t i o n i n t h e FSAR. The c u r r e n t outage p r a c t i c e i s t h e sub jec t o f a p e r i o d i c ana lys i s done by Nuclear Fuels f o r each r e f u e l i n g outage.

We have reviewed t h e twenty-e ight "open i tems" from t h e NSAG review. Twenty-six of t h e twenty -e igh t i tems were reso lved t o t h e s a t i s f a c t i o n o f NSAG. The remain ing two i tems r e f e r t o t h e need f o r t h e l e v e l and temperature i n d i c a t i o n t o be a v a i l a b l e i n t h e c o n t r o l room v i a a PMS (computer d i s p l a y ) format and t o add r e f l a s h c a p a b i l i t y f o r alarms from panel OC211. These m o d i f i c a t i o n s are on t h e books b u t n o t being a c t i v e l y worked t o t h e best o f my knowledge.

EDMG completed a r e v i s i o n t o t h e EDR screening f o r 620020 on F r iday August 28, 1992 and requested comments. The s i g n i f i c a n c e was evaluated as min imal . Based on my rev iew o f t h i s screening document i t i s unacceptable as w r i t t e n and I have requested i t be rev ised. We are proceeding t o rev iew t h e i ssue f o r r e p o r t a b i l i t y regard less o f t h e f i n a l s i g n i f i c a n c e l e v e l f rom t h e screening rev iew.

Work on r e s o l v i n g t h e issue i s assigned t o Mark Mjaatvedt . Michael Crowthers has been work ing on t h i s i ssue s ince J u l y 20, 1992. 1 have a l so assigned Dave K o s t e l n i k as o f today. We are p lann ing t o i n v o l v e Bechtel and GE. A schedule i s under devel opment .

Attachment 15

PP&L Memo from Kevin W . Brinckman to George T. Jones, "Review of Fuel Pool Cooling", September 1 , 1992 (PLI-72288)

Note: This engineering report was prepared by a PP&L engineer previously not associated with EDR 620020 at the request of the PP&L Manager of Nuclear Plant Engineering to provide him with an independent appraisal of the concerns raised in the EDR. This independent evaluation basically concludes that a LOCA with a loss of normal fuel pool cooling would put the operators "in a position where they would be required to make decisions on removing ECCS equipment from containment/core cooling service to cool the fuel pool" and points out that it would involve unanalyzed conditions. This report also raises, for the first time, the concern that the hydrodynamic loads of the LOCA might damage the non-seismic, non-safety related fuel pool cooling system piping.

September 1, 1992

George T. Jones A 6 4

SUSQUEHANNA STEAM ELECTRIC STATION REVIEW OF FUEL POOL COOLING PLI - 72288 F i l e A45-lA

Reference : EOR-G20020

I have completed a review o f the concerns over the adequacy o f fue l pool cool ing and the consequences o f f ue l pool b o i l i n g a t Susquehanna brought about by the referenced EDR. The i n ten t o f the review i s t o determine the consequences o f off-normal and accident events on the a b i l i t y t o maintain fue l pool cooling, and t o evaluate the safety impl icat ions o f a loss o f fue l pool cooling. The review i s not an evaluation o f the l i cens ing o r design basis o f fue l pool storage and cool ing a t Susquehanna.

Attached i s a repor t which documents my evaluation and f ind ings. It i s my opinion t h a t f u r t he r evaluation o f the concerns discussed i n the referenced EDR and my ' repor t are warranted t o assure spent f ue l storage a t Susquehanna does not reduce p lan t safety. Section 4.0 o f the repor t provides recommendations and Section 5.0 d e t a i l s my conclusions. Provided below i s a summary o f these conclusions.

The use o f RHR i n the fue l pool cool ing ass i s t mode, post LOCA, occupies ECCS equipment which i s accounted f o r i n the long term cool ing analysis o f FSAR Chapter 6.2. I bel ieve postulated break and s ing le f a i l u r e combinations e x i s t where RHR equipment i s needed f o r core/containment cool ing t o remain w i t h in the analyzed condi t ions o f the FSAR.

A LOCA which r e s u l t s i n core degradation and f i s s i o n product release w i l l make the RHR valves required f o r manual alignment o f RHR fue l pool cool ing ass is t inaccessible. This combined w i th a loss o f normal f ue l pool cool ing w i l l lead t o fue l pool b o i l i n g given s u f f i c i e n t spent fue l decay heat.

I f the fue l pool was allowed t o b o i l , moisture and energy release t o the reactor bu i l d i ng dur ing a LOCA would create a severe environment f o r which the safety-grade equipment may not be qua1 i f i e d . Degradation o f the iod ine removal e f f i c i ency o f the SGTS charcoal beds due t o moisture carry-over could e f f e c t o f f - s i t e release ca lcu la t ions.

The fue l pool t roub le alarm i n the cont ro l room cannot be counted on f o r re1 i ab le ind icat ion. Access t o the re fue l i ng f l o o r i s required t o monitor fue l pool condit ions.

Loss o f fue l pool cool ing dur ing re fue l i ng w i th a f u l l core o f f - load i s a concern since by Technical Speci f icat ions, both RHR loops may be inoperable. A b o i l i n g fue l pool on the re fue l ing u n i t along w i t h a LOCA on the other u n i t causes equipment q u a l i f i c a t i o n and o p e r a b i l i t y concerns on the LOCA un i t .

G. T. Jones Page 2 September 1, 1992 PLI - 72288

I f a la rge break LOCA occurred a t Susquehanna, I am conf ident t ha t our current procedures, equipment, and pract ices would maintain the p lan t i n a safe condit ion. However, combined w i th a l oss o f f ue l pool cooling, the operators would be put i n a pos i t i on where they would be required t o make decisions on removing ECCS equipment from containment/core cool ing service t o cool the fue l pool. It i s my op in ion t h a t these evaluations need t o be done beforehand so t h a t methods are i i p l a c e t o handle the s i t u a t i o n i f it arises.

From the research I have done over the past several weeks, I have found t h a t there are many issues which warrant a more de ta i led evaluat ion than I was capable o f i n t h i s short time. I have t r i e d t o capture the major issues and provide a quick assessment o f each.

z 2 5 i ' w h ~ Kevin W. Brinckman

Attachment

CC: J. E. Agnew F. G. Bu t le r R. G. Byram M. H. Crowthers G. D. Gogates G. J. Kuczynski D. A. Lochbaum S. M. Hausman 0. C. Prevatte 6. D. M i l l e r J. R. Miltenberger M. R. Mjaatvedt C. A. Myers J. G. Re f l ing T. J. Sweeney J. A. Zola Nuclear Records

A6-3 w/a A6-3 w/a A6-1 w/a A6-3 w/a SSES w/a SSES w/a Enercon w/a A6-2 w/a A6-3 w/a A6-3 w/a A6-1 w/a A6-3 w/a A2-4 w/a A6-3 w/a SSES w/a A6-3 w/a A6-2 w/a

REVIEW OF FUEL POOL COOLING DURING POSTULATED -

. ,OFF-NO- AND ACCIDENT EVENTS

Susquehanna Steam Electric Station

Units 1 6 2

August 1992

Prepared by:

Kevin W. Brinckman Project Engineer-Nuclear Systems

Preface

This paper presents the findings of a review of the adequacy

of spent fuel pool cooling at Susquehanna SES during off-normal and

emergency situations. The intent of the review is to determine if plant

safety can be maintained. This paper is not a review of the licensing

requirements for spent fuel storage at Susquehanna SES.

The design of the spent fuel pool cooling system (FPC) was examined

to determine to what extent its operability and integrity would be affected

by postulated off-normal or accident conditions. The availability

of other means of spent fuel pool cooling was researched, and the function

and accessibility of the alternatives was evaluated. The design of the

spent fuel pool instrumentation was reviewed to determine if spent fuel

pool monitoring could be maintained in an off-normal or accident condition.

The consequences of a boiling fuel pool were considered. It was assumed

that sufficient decay heat would exist in the stored spent fuel to produce

pool heat-up and eventually a boil-off condition if cooling was not

provided.

No detailed calculations were performed as part of this effort.

Conclusions are drawn from current design and licensing information.

Analysis or re-analysis of issues such as post-LOCA radiation levels,

probability and consequences of clad and fuel failure, and equipment

qualification could alter some of the conclusions. Based on the

research done to compile this report, it is the author's opinion that

further evaluation of the concerns discussed in EDR GZOOZO and this

report are warranted to assure spent fuel storage at Susquehana does

not reduce plant safety.

TABLE OF CONTENTS . Preface .................................................. i

............................................. 1.0 Introduction 1 . .

2.0 Postulated Events ........................................ 2 3.0 Safety Evaluation ........................................ 5 4.0 Recommendations ......................................... 17 5.0 Conclusions ............................................. 19 6.0 References .............................................. 21

Page 1

1.0 Introduction

The purpose of this paper is to provide a review of spent fuel pool

cooling capability at Susquehanna SES following a postulated off-normal

or accident condition. A series of events was considered to determine

whether potential situations exist at Susquehanna SES which could result

in fuel pool boiling. These events are discussed in Section 2.0.

Section 3.0 a safety evaluation of the postulated events in

Section 2.0 and discusses what aspects could be further evaluated to

assess the consequences of the respective event.

Potential problems with a loss of fuel pool cooling and subsequent

alignment of the RHR fuel pool cooling assist mode are found. Access

to the RHR valves necessary to manually align RBR in the fuel pool cooling

assist mode would be hindered by an event which caused core degradation

and fission product release to the primary coolant. The consequences

of using ECCS equipment for RER fuel pool cooling assist during an

accident need to be considered. A FMEA is warranted to determine if

ECCS can provide fuel pool cooling and long term post-LOCA containment

cooling with postulated single failures and breaks. Operator guidance

for implementing RHR fuel pool cooling assist post-LOCA should be

developed.

If fuel pool boiling were to occur during a LOCA, the energy

and moisture released to the reactor building would create a severe

environment for which much of the safety grade equipment may not be

qualified. The entrained water could overload the SGTS moisture

removal equipment and reduce the iodine removal efficiency of the

charcoal beds.

The effect of a loss of fuel pool cooling during refueling with a

full core offload is evaluated. The lack of operable RER equipment is

the concern in this situation. A fuel pool boiling event along with

a LOCA on the other unit would put the plant in an unanalyzed condition.

Page 2

2.0 Postulated Events

In this section off-normal or accident events are discussed in

which the normal spent fuel pool cooling system (FPC) is lost and

alternate means are required to remove the decay heat from the fuel

pool. The events chosen are considered to provide an envelope of

situations where FPC is lost and the plant is in a degraded condition.

Events 1 to 5 are presented in order of what is considered highest to

lowest probability of occurrence, although frequencies have not been

researched as part of this effort. The last event considered is a

loss of FPC with a full core offload to the spent fuel pool. It is

the only event evaluated which is not concerned'with decay heat in the

reactor vessel.

2.1 ~vent#l : Loss of Offsite Power (LOOP1

Event dl postulates a dual unit LOOP, which causes a loss of power

to the fuel pool cooling system (FPC) pumps and loss of service water

for cooling of the FPC heat exchangers. The event assumes both reactors

are stabilized and no fuel failure occurs.

2.2 EventXZ : LOCA Without Fuel Failure

Event 112 postulates a LOCA o n one unit. It is assumed that

the emergency core cooling systems perform to provide sufficient core

cooling to maintain fuel and clad integrity. No seismic event

is postulated, however, hydrodynamic loads due to steam discharge to the

suppression pool must be considered. Since portions of the FPC are not

seismically designed it is indeterminate whether the FPC piping and

equipment will be damaged by the hydrodynamic loads. It is assumed that

FPC is lost for the duration of the accident. The non-1E fuel pool level

and temperature indication is also assumed lost due to structural damage

resulting from the hydrodynamic loads. If still functional, the fuel

pool cooling pumps would be shut down by the non-essential load shed

24 hours after the LOCA, if implemented per EP-IP-055~. To prevent

fuel pool boiling, RHR fuel pool cooling assist must be initiated.

Page 3

2.3 Eventil3 : LOCA-LOOP Without Fuel Failure

Event 113 postulates a LOCA on one unit coincident with a LOOP.

It is assumed that the emergency core cooling systems maintain adequate

core cooling to prevent fuel or clad damage. FPC is lost for the

duration of the accident due to the mechanisms discussed in Events $1

and #2. It is assumed that hydrodynamic loads will disable the fuel

pool instrumentation'.

2.4 EventY4 : LOCA With Fuel Failure

Event #4 postulates a LOCA on one unit witti fuel failure as

described in the accident analysis of FSAR Chapter 15.6.5, which

specifies release of radioactive material in accordance with the

assumptions of Regulatory Guide 1.3~. Radiation dose rates within

the reactor building for this scenario have been calculated in FSAR

Chapter 18.1.20 per the guidelines of N U R E G - O ~ ~ ~ ~ , which applies

the assumptions of Regulatory Guide 1.3~ to specify the release of

radioactive materials to the primary coolant.

FPC is lost immediately due to the service water LOCA load shed

and it is assumed that the hydrodynamic loads during the LOCA blowdown

cause the non-seismic portions of FPC to fail. FPC is lost for the

duration of the accident and RHR-fuel pool cooling assist must be placed

in service to remove the decay heat from the spent fuel pool.

2.5 Event#5 : LOCA-LOOP With Fuel Failure

Event $5 postulates the LOCA discussed in Event #4 coincident with

a LOOP. FPC is lost for the duration of the accident due to the mech-

anisms discussed in Events $1 and 2. It is assumed that hydrodynamic

loads will disable the fuel pool instrumentation.

Page 4

2.6 Event#6 : Loss of FPC With Full Core Offload

This event postulates a loss of FPC during refueling when the

entire core is offloaded. The emergency heat load as defined in

FSAR Chapter 9.1.3 is considered, as well as a lesser heat load

which would not normally require RER fuel pool cooling assist.

Page 5

3.0 Safety Evaluation

3.1 EventXl : Loss of Offsite Power (LOOP]

3.1.1 Evaluation

In theevent of a LOOP, the spent fuel pool cooling system (FPC) is

disabled due to losb'of power to the FPC pumps and loss of service water

for cooling of the FPC heat exchangers. Assuming the reactor is

stabilized and no fuel failure occurs spent fuel pool cooling can be

provided using the REX fuel pool cooling assist lineup. Operations has at

least 19 hours with the design heat load on the'fuel poolZ to initiate

RBR fuel pool cooling assist before fuel pool boiling is calculated to

occur. Since no fuel failure results from the LOOP, access is available to

the reactor building and the manual operations required to establish RBR

fuel pool cooling assist can be performed without exposing operators to

doses above station limits. RER fuel pool cooling assist has sufficient

cooling capacity to remove the projected spent fuel pool emergency

heat load as defined in FSAR Chapter 9.1.

Per the action statement of Tech Spec 3.8.1.1, the reactor must

be in cold shutdown within 108 hours after the LOOP if at least one

source of offsite power to the Class 1E distribution system in not

operable. The ECCS design contains sufficient redundancy and

flexibility that the reactor could be brought to cold shutdown while

operating RBR fuel pool cooling assist.

Fuel pool temperature and level indication are provided on panel

OCZll located on the refueling floor. Panel OCZll is powered off of

the diesel generators and will have power available during the LOOP.

The fuel pool instrumentation is non-1E and is not environmentally

qualified. While the LOOP environment is not expected to be severe

on the refueling floor and the instrumentation should remain functional,

Page 6

this cannot be guaranteed. A group alarm for OC211 exists in the control

room, but the history of trouble with this alarm is extensive.

Operator access to the refueling floor could be limited due to

airborne radioactivity from evaporation off of the fuel pool surface

or elevated air temperatures due to the loss of HVAC. Therefore, the

operators should be-given some guidance on the amount of time they

have after a loss of' fuel pool cooling to initiate RBR fuel pool cooling

without having to rely on the OC211 instrumentation. ON-135-001 does

inform the operators that they have at least 25 hours until the fuel

pool boils after loss of FPC. This time could be extended considerably

by a cycle specific analysis.

Since the fuel pool will not boil for at least 19 hours, electrical

loads required to support RHR fuel pool cooling assist will not have

any impact on short term recovery from the LOOP.

3.1.2 Summary

A L W P does not present a significant challenge to providing fuel

pool cooling. RBR fuel pool cooling assist can be aligned and has ample

heat removal capacity to handle the spent fuel pool heat load. However,

adequate indication of fuel pool conditions may not be available to the

operators and it would be advantageous to provide additional guidance

to the operators on how long they have to establish RHR fuel pool

cooling assist after a loss of spent fuel pool cooling.

3.2 EventXZ : LOCA Without Fuel Failure

3.2.1 Evaluation

If the recirculation discharge line break is postulated, one loop

of LPCI is lost. With the other loop of LPCI functional a substantial

Page 7

makeup source to the reactor vessel is maintained and extended fuel

uncovery and cladding damage can be avoided. To prevent fuel pool

boiling, RHR fuel pool cooling assist must be initiated. Since no fuel

failure is postulated, access to RHR valves necessary to align RHR fuel

pool cooling assist is available and can be accomplished without large

dose rate exposures to the operator.

With the design'heat load on the fuel pool, Reference 2 calculates

that operators have approximately 19 hours to provide fuel pool cooling

to prevent fuel pool boiling. RHR fuel pool cooling assist is aligned

with RHR loop A suction off of the fuel pool skimmer surge tank and

discharge through the loop A RHR heat exchanger back to the fuel pool.

If the postulated break is in the reactor recircuation loop B discharge

line, RHR loop A LPCI is required for vessel reflood. Therefore, to align

RHR fuel pool cooling assist, RBR loop A discharge to the vessel would

need to be terminated after 19 hours. FSAR Chapter 6.2 analyses for

long term cooling after the containment design basis LOCA assumes at

least one LPCI pump is available for vessel makeup or containment spray.

A potential problem exists here if the break in recirculation loop B

disables loop B LPCI and a single failure of the loop B containment

spray valve is postulated. With this scenario, loop A RBR flow may

be needed for containment pressure/temperature control and would not

be available to provide a closed loop of RHR fuel pool cooling assist.

If the hydrodynamic loads resulting from steam blowdown to the

suppression pool do not damage the fuel pool temperature and level

instrumentation, power is available from the diesel generators to keep

the equipment operable. However, as discussed in Section 3.1, the

operation of this instrumentation cannot be guaranteed. Therefore, it

it is assumed that fuel pool conditions will be indeterminate during a

a LOCA unless access to the refueling floor is possible.

Page 8

3.2.2 Summary

Without fuel failure resulting from the LOCA, access to the reactor

building and refueling floor is available and the necessary operations

required to monitor fuel pool status and initiate RER fuel pool cooling

assist can be performed. However, the availability of fuel pool

temperature and level indication cannot be guaranteed and alternate means

of fuel pool monitoring or specific guidance on time to initiate RER fuel

pool cooling assist is recommended. Consideration must be given to the

ability of ECCS'to provide long term cooling to both the containment

and fuel pool with postulated pipe breaks and single failures. A more

detailed FMEA is warranted for this event.

3.3 EventY3 : LOCA-LOOP Without Fuel Failure

3.3.1 Evaluation

The consequences of this LOCA-LOOP on fuel pool cooling are no

different than those discussed for a LOCA without fuel failure. Fuel

pool cooling is not needed in the short term when recovery from the

LOCA is taking place. By the time the fuel pool approaches boiling

(> 19 hours per Reference 2 ) , the LOCA recovery should be in the

long term cooling phase. RER pumps will be operating with RER heat

exchangers valved in to remove decay heat from the containment. It is

assumed that sufficient ECCS was available to maintain core cooling

and preclude fuel failure. Therefore, the RER equipment area will be

accessible for aligning the RER fuel pool cooling valves, and the fuel

pool can be maintained in a safe condition. This evaluation is again

contingent on sufficient ECCS equipment remaining available to provide

both long term containment and fuel pool cooling.

The availability of fuel pool level and temperature indication is

the same as discussed in Section 3.1.

Page 9

3.4 Event#4 : LOCA With Fuel Failure

The postulated fuel failure for this event changes the situation

considerably. In the previous events without fuel failure, the reactor

building remains accessible. The occurrence and extent of fuel failure

are the major factors in determining the options available for responding

to a loss of fuel pool cooling. This event is evaluated below for

several different situations.

3.4.1 NUREG-0737 Postulated Fuel Failure

For the postulated LOCA scenario, FPC is lost and RHR fuel pool

cooling must be established to avoid fuel pool boiling. To align RBR

for fuel pool cooling five valves must be manually opened (151060, 151070,

153021, and 153070A,B for Unit 1, or their counterparts for a Unit 2

event). Three of these valves are in room 1-514 and two are in room 1-202.

Access to the reactor building is restricted by the post-LOCA radiation

dose rates which are a dependent upon the extent of postulated core

degradation, system operations which transport the radioactivity

throughout the reactor building, and shielding in place to isolate the

sources.

A Susquehanna specific plant shielding analysis was performed in

response to one of the action plan requirements of NUREG-O~~~'.

N U R E G - O ~ ~ ~ ~ requires the use of Regulatory Guide l.j4 assumptions

for release of fission products from the fuel. The shielding analysis,

presented in FSAR Chapter 18.1.20, calculates radiation levels in the

the reactor building and provides the results as radiation zone maps

one hour after the LOCA in Figures 18.1-2 to 18.1-8. Per this analysis

room 1-514 is a radiation zone V ( 5 50R/hr) and 1-202 is a radiation

zone VIII (>5000R/hr) with the postulated fuel failure per Regulatory

Guide 1.3~. These radiation levels make the valves inaccessible

until sufficient decay has occurred. FSAR Figures 18.1-9 and 18.1-10

Page 10

provide decay factors as a function of time which can be applied to the

dose rates calculated in the shielding analysis. These curves show

that an order of magnitude decrease in radiation dose rates can be

expected 19 hours after the LOCA, by which time RBR fuel pool cooling

assist must be established to avoid fuel pool boiling assuming the design

heat load. This reduction in dose rates makes room 1-514 accessible

for the valve operat-ions required within the 5 Rem per activity

guideline of 10CFRSO'Appendix A GDC 19. However, an order of magnitude

reduction in the room 1-202 calcuated dose rate will not allow access

to the RBR valves. Therefore, with fuel failure postulated per the

guidelines of N U R E G - O ~ ~ ~ ~ . the necessary manual valve alignments

cannot be performed to establish RER fuel pool cooling assist and if a

sufficient decay heat load exists fuel pool boiling will occur.

A further examination of room 1-202 shows that it houses much of

the RHR piping running from the RER Loop A pumps to the containment.

The DBA LOCA which is analyzed for peak clad temperature is the

recirculation loop discharge line break with a failure of the LPCI

injection valve on the other loop. With the loss of both LPCI loops

core uncovery results for a sufficient time to expect clad damage and

release of fission products to the primary coolant. Transport of these

fission products throughout the reactor building creates the radiation

sources in the rooms housing the RER valves. Since room 1-202 houses

the RHR Loop A pump discharge pipe, the water drawn from the suppression

pool by Loop A will travel through this room. Fission products within

the pipe will sustain the radiation source in the room. Valves

151060 and 151070 are in close proximity to the RBR discharge pipe

and the operator would have no permanent shielding if required to

manually open the valves. Therefore, without a reduced source term

in room 1-202, access to RHR valves necessary to establish RHR fuel

pool cooling is not possible in time to prevent fuel pool boiling

at the design fuel pool heat load.

Page 11

3.4.2 Fuel Pool Boiling and Makeu~

Without FPC or RBR fuel pool cooling assist available, fuel pool

boiling will occur at a time dependent on the decay heat load in the spent

fuel pool. Emergency makeup water is available by design from the ESW

system by performing manual valve alignments in rooms 1-105 and 1-514.

These rooms are reported as radiation zones VIII and V respectively in the

FSAR Chapter 18 anaiysis discussed above and the same accessibility problem

will exist with room 1-105 as found with 1-202. A better alternative

for fuel pool'makeup with a degraded core condition is via a firehose

on the refueling floor, if it is established prior to the fuel pool

reaching boiling temperature. This alternative could be implemented

with minimal dose to the operator. However, the post-LOCA condition of

the non-Q fire protection system is uncertain and access to the ESW

valves may ultimately be required for emergency makeup to the fuel pool.

Appendix 9A of the FSAR analyzes the consequences of a loss of

spent fuel pool cooling to determine the time to boiling, capability

to add makeup water, and off site releases in the event of fuel pool

boiling. A problem with responding to a loss of spent fuel pool cooling

with makeup only is that the decay heat generated by the spent fuel is

ultimately transferred to Zone I11 and the LOCA unit's reactor building.

The design spent fuel pool heat load is 12.63+6 Btu/hr which is

two to three times the calculated post-LOCA heat load on the LOCA unit

plus Zone 111. The extreme amount of energy and moisture deposited into

the reactor building from a boiling fuel pool would undoubtedly create

an environment which could jeporadize equipment operability and make

the reactor building inaccessible. Moisture carryover to the SGTS

would reduce the iodine removal efficiency of the charcoal beds if the

entrained water overloaded the moisture removal equipment in the SGTS

trains. This would effect offsite releases and the assumed efficiencies

used in offsite dose calculations.

Page 12

3.4.3 ECCS Desiqn Basis LOCA

The ECCS design basis LOCA considers the break in the primary

coolant boundary along with the worst case single failure which results

in the highest fuel cladding temperature. For Susquehanna SES, the

ECCS DBA LOCA is the recirculation discharge line break with a failure

of the LPCI injection valve on the other loop. Loss of both LPCI loops

results in core uncovery for a sufficient time to expect clad damage and

release of fission products to the primary coolant.

The LOCA analysis performed to determine the peak clad temperature

and clad oxidation does not predict the fission 'product release since

the lOCERlO0 offsite dose analysis must be performed using the releases

specified in Regulatory Guide 1.3. However, per NU REG-^^^^^, the fuel fission' product release from the DBA LOCA would be substantially less

than that specified in Regulatory Guide 1.3~. For example, if fuel

clad temperature is maintained between 1300F and 2100F. NUREG-12~8~

assumes 2 percent of the iodine in the core is released. This value is

based on estimated gap release values derived in WASH-1400~ for

situations of core degradation with cladding failure. The position of

NUREG-0737~ is that 50 percent of the core iodine is transfered to

the primary coolant.

If the postulated core damage is limited to cladding failure and

the guidelines of NUREG-1228~ are applied, the radiation levels

within the reactor building due to activity within the primary coolant

could be reduced by an order of magnitude. However, even at the reduced

radiation levels it is uncertain whether access to the RHR valves is

possible while remaining within the post-accident operation dose rate

criteria of M I R E G - O ~ ~ ~ ~ . Analysis of the source term expected

from clad failure and the decay factor would have to be performed to

determine radiation levels for this scenario. Conclusions on the

accessibility of RHR valves could then be drawn.

Page 13

Again, consideration must be given to the ability to provide

long term cooling of the containment. If RHR loop A could be placed

into RIIR fuel pool cooling assist mode, RHR loop B would be required

for reactor/containment decay heat removal. Since the single failure

has already been postulated for the LPCI injection valve, RHR loop B

would be available for containment spray. Case C of the long term

cooling analysis in-FSAR section 6.2.1.1.3.3.1.6 bounds this scenario.

3.4.4 Summary

If a LOCA is postulated which produces core degradation and fission

product release, the valves required to be manually opened to align

RER in the fuel pool cooling mode and ESW for fuel pool makeup will be

inaccessible unless the extent of core damage is small. Analysis does

not currently exist to quantify what "small" is. The plant shielding

analysis in FSAR Chapter 18 predicts LOCA radiation levels which are

too high to permit access to the RHR valves in time to prevent fuel

pool boiling with the spent fuel pool design heat load. Although,

fuel pool boiling is analyzed in Appendix 9A of the FSAR for offsite

dose releases, analysis of the consequences on reactor safety could

not be found.

It is not sufficient to merely makeup the water evaporated from

the pool surface. This mode of operation uses the reactor building

and refueling floor as a heat sink for the spent fuel decay heat and

reservoir for the condensate. In doing so, an environment is imposed

on the plant for which it is not analyzed.

It is not certain if realignment of RHR loop A to fuel pool cooling

mode will leave sufficient ECCS equipment to provide long term

containment heat removal as analyzed in FSAR Chapter 6.2. Further

evaluation of this issue is warranted.

Page 14

3.5 Eventt5 : LOCA-LOOP With Fuel Failure

The LOCA-LOOP with fuel failure causes the same concerns as the

LOCA with fuel failure. The radiation levels currently analyzed in

the rooms where operator access is required to manually align RHR

fuel pool cooling assist are too high to allow the required action.

Since operators'have at least 19 hours to establish RER fuel pool

cooling assist before pool boiling, short term recovery from the LOCA

is not an issue'. However, the effect of removing a loop of RHR from

the long term LOCA recovery and the additional impact of the LOOP on

equipment availability should be evaluated.

3.6 Eventt6 : Loss of FPC With Full Core Offload

3.6.1 Ebnerqencv Heat Load

If the full core offload produces the emergency heat load (EHL)

on the spent fuel pool as described in FSAR Chapter 9.1, then RHR

fuel pool cooling is required to remove the heat load on the pool.

In this situation, RHR would be aligned in the fuel pool cooling

configuration and safety is maintained. Since all of the fuel is

offloaded from the reactor vessel, RER is not required for core cooling

and use of the RHR equipment for fuel pool cooling is the priority.

3.6.2 Maximum Normal Heat Load

If the full core is offloaded for refueling with decay heat loads

at current levels, FPC has sufficient capacity to adequately cool the

fuel pool. Techncical Specifications for refueling operations (3.9.11)

require RBR shutdown cooling be operable when irradiated fuel is in

the reactor vessel. With a full core offload, applicability of this

requirement is removed and both loops of RER may be inoperable. If

Page 15

FPC is lost, operators have at least 19 hours to get a loop of RHR

(one pump and heat exchanger) into service for fuel pool cooling.

Nineteen hours is the time to boil with the maximum normal heat load2.

The emergency heat load need not be considered in this scenario since

a lesser beat load which normal FPC could cool was postulated.

Off normalprocedure ON-135-001 provides operators with actions

for dealing with a loss of spent fuel pool cooling. They are first

instructed to operate RHR in the fuel pool cooling assist mode if

available. If all means of cooling are lost, the procedure instructs

the operators to "ALLOW water in Fuel Pool to boil", and provides a

caution that evacuation of the refuel floor may become necessary due

to increasing radiation levels. A note is also included which informs

the operator that boiling should not occur before 25 hours after loss

of cool'ing.

The major concern with this event is radiation release due to the

boiling fuel pool. Since the core is ofeloaded, degradation of equipment

in the reactor building due to the severe environment imposed by the

heat, moisture, and radiation released from the fuel pool is not a

concern provided a makeup water source can be maintained to the fuel

pool. FSAR Appendix 9A analyzes the consequences of a boiling fuel

pool and finds that the offsite releases are within lOCFRlOO limits.

Another concern is the effect on the equipment of the other unit

which if not also in refueling, may need to respond to a LOCA. A LOCA

on the operating unit would initiate recircualtion with Zone I11 and

the LOCA unit's air space. The effects of the boiling fuel pool could

then be imposed on the LOCA unit and create equipment qualification and

operability questions.

Page 16

3.6.3 Summary

If both loops of RER are rendered inoperable during refueling

with a full core offload, the plant is not placed in an LCO. However,

a loss of FPC would require a loop of RER and RBR service water be placed

back in service for RER fuel pool cooling assist before the fuel pool

boils. If fuel pool-boiling occurs on one unit, equipment degradation

on the other unit could result and place the plant in an unanalyzed

condition.

Page 17

This section provides several actions which are recommended to

provide a better understanding of the consequences of a loss of fuel

pool cooling and provide Operations with guidance on operator response

to the event.

1) A LOCA which resblts in core degradation and releases of radio-

activity to the primary coolant is an event which also jeporadizes

the ability' to cool the spent fuel pool with systems currently intended

for that function. If operators are required to implement RAR fuel

pool cooling during a LOCA, analysis is warranted to determine what

dose rates would be encountered to access the RHR valves for a range

of degraded core conditions. With this information it would be possible

to determine if alignment of RHR in the fuel pool cooling mode is

feasible during a LOCA where core damage is postulated.

2) A FMEA should be performed to determing if alignment of RHR in the

fuel pool cooling mode during a LOCA (or post-LOCA) would place

the plant in an unanalyzed condition with respect to long term

post-LOCA containment heat removal. Operator action should be

developed to respond to a loss of fuel pool cooling with a LOCA.

3) If non-Q equipment is assumed unavailable during a LOCA, then spent

fuel pool temperature and level indication is lost. Therefore, it would

be prudent to provide the operators with the anticipated time to fuel

pool boiling on loss of fuel pool cooling on a cycle specific basis.

This would allow maximum time for source term decay prior to required

access to high radiation zones.

4) The preferred response to loss of FPC is to establish an alternative

method of decay heat removal such that the total energy does not get

transferred to the Zone I, 11, and I11 atmosphere. Otherwise, the

consequences of fuel pool boiling on the plant's ability to safely

Page 18

shut down should be analyzed.

One alternative to be considered is crosstying the Unit 1 and 2

fuel pools by flooding the shipping cask storage pit and removing

the gates to both fuel pools. Then by initiating RFfR fuel pool

cooling assist on the non-LOCA unit, cooling could be provided to

both pools. Further evaluation of this alternative should address

whether power would be available to the overhead crane to remove

the gates in a LOCA or LOOP scenario, the ability of RHR fuel pool

cooling assist to cool both pools, sources of water to flood the

cask pit, and allowable operator response time to initiate this

procedure.

Page 19

5.0 Conclusions

Following is a summary of the major observations and conclusions

resulting from the evaluations performed to compile this report.

1) In a LOCA condition, postulated pipe breaks and single failures

can disable enough RHR equipment that operators may not be able to

align RER Loop A'for fuel pool cooling assist and still maintain

adequate core and containment heat removal. Currently, Operations

has no clear instructions on how to handle this situation and would

likely be hesitant to remove equipment from core/containment cooling

service following a LOCA. While alternate shutdown cooling and other

means of decay heat removal may exist, the emergency operating

procedures contain no instructions directing the operators to redirect

~ c c ~ e ~ u i ~ r n e n t to spent fuel pool cooling service. Removing all RHR

pumps from either LPCI or containment spray duty would place the

plant in an unanalyzed condition for long term post-LOCA cooling.

2) Two of the RBR valves required to manually align RHR fuel pool cooling

assist are in a room which contains much of the RHR Loop A piping.

In order to acces the valves, operators would be in the direct line

of site and in close proximity to RHR piping containing primary

coolant. A LOCA which produces core degradation and fission product

release will make the RHR valves inaccessible due to the source term

from the primary coolant. Decay of this source term is not expected

to be sufficient enough to allow access to the valves before the

predicted time to fuel pool boiling.

3) If the fuel pool was allowed to boil, moisture and energy released

to the reactor building during a LOCA would create a severe environment

for which much of the safety-grade equipment may not be qualified.

This would put the plant in an unanalyzed condition.

Page 20

4) Fuel pool monitoring equipment is located on panel OC211 on the

refueling floor. The instrumentation is non-1E but is powered off

of the diesel generators. The non-1E equipment is not qualified

to be operable in an accident condition. Therefore, the trouble

alarm in the control room cannot be relied on. Access to the

refueling floor is required to monitor fuel pool conditions.

- 5) Loss of FPC during refueling with a full core offload is a concern

if both RHR loops are inoperable. RER equipment must be placed back

into service.quickly enough to cool the fuel pool before boiling

occurs. A boiling fuel pool on the refueling unit along with a LOCA

on the other unit causes equipment qualification and operability

concerns on the LOCA unit.

6 ) Off-normal procedure ON-149-001 notes that operators have 25 hours

before fuel pool boiling occurs on a loss of FPC, and instructs

them to initate RBR fuel pool cooling assist if available. Otherwise,

the operators are told to allow the fuel pool to boil and provide

makeup.

Page 21

6.0 References

1) Bechtel Power Corporation Drawing ElO-1, "Single Line Meter h

Relay Diagram 125MC, 250MC. lZOVAC Systems Units 1 h 2".

Revision 18.

2) Pennsylvania.Power & Light Co. Calculation M-FPC-009 Revision 0,

*Spent Fuel'Pool Boiling Analysis".

3) Pennsylvania Power & Light Co. Procedure EP-IP-055 Revision 0,

"Post Accident Response to LossofReactor Bldg W A C (Reactor

Bldg Non-1E Electrical Load Shed)".

4) U. S. Atomic Energy Commision Regulatory Guide 1.3, Revision 2,

"Assumptions Used For Evaluating The Potential Radiological

Consequences of a Loss of Coolant Accident for Boiling Water

Reactors", June 1974.

5) U. S. Nuclear Regulatory Commission, NUREG-0737, "Clarification

of TMI Action Plan Requirements", November 1980.

6) U. S. Nuclear Regulatory Commission, NUREG-01228, "Source Term

Estimation During Incident Response to Severe Nuclear Power

Plant Accidents", October 1988.

7) U. S. Nuclear Regulatory Commission, WASH-1400 (NUREG-75/014),

"Reactor Safety Study: An Assessmemnt of Accident Risks in U.S.

Commercial Nuclear Power Plants", October 1975.

Attachment 16

PP&L Memo from J. R. Miltenberger to G. T. Jones, "Spent Fuel Pool Cooling", September 9, 1992 (PLI-72367)

Note: This memo was prepared by the PP&L Manager of the Nuclear Safety Assurance Group at the request of the PP&L Manager of Nuclear Plant Engineering. The memo acknowledges the concerns in EDR 620020 need to be resolved and suggests that Engineering go beyond the EDR concerns and conduct an in-depth design review of spent fuel pool cooling operations. By this date, the Manager of Nuclear Plant Engineering had two ( 2 ) independent, in-house evaluations of the concerns raised in EDR 620020 which did not refute the primary safety issues in the EDR.

cc: H. W. K~ismr TW-16 W I O R. 6. fi ram b6-1 W/O H. O. stanley SSES wlo C. A. W r s hZ-4 wlo

0. 7. Jones A6-2 A. J. D%ninyuer SSES wjo sWm11-*

SUSQUEHANNA S T E M ELECTRIC STATION SPENT FUEL PWl COOLING

On'hugust 19 you ruquested that I prowldm you copfes o f the work that NS116 has done on'the fuel pool cooling issue a d that I state pmclsely what my camems arm regarding th. f u r l pool coollng system and I t s employment.

Ny conemrnr and issues are as follaws:

1. Thm conpany does not havr m off icial celculation o f the decy heat loads t o be expected undmr outage conditions. A calculation projected mhrad fo r smwmral years i s namdmd f o r outage planning. The projection 4s necessary t o account for the accumulation o f fuel in the pool.

2. The station i s vulnerablm t o loss o f drccy heat removal capability during the smrvicm water outage pariod of a refueling outage. During th is t ime w rely upon thm o rat ing uni t spent fuel pool cooling system fo r docw heat removal. Bac r up methods o f decay heat removal are: 1) use o f m Optratlng Unit RHR systcn i n tha Fuol Pool Cooling Assist Node and 2) bolting. Neithar back up method i s attracttvm.

o Boiling, although safe, t s not a vlable alternative f o r p o l I t i ~ a 1 reasons.

o Use o f m RHR syrtarn frm the aperating uni t i n the fuel pool cooling r r s I s t mode i s not attractive for operatlonal reasons.

o RHR has not been tested i n the fun1 pool coollng asslst mode.

WSA6 has argued and w i l l continue t o argue that thm r i s k IS accm table. 1 However, a study should be made to determine whethw a more viab e mathod of beckup decy heat removal during servica water outlges i s feasfblm.

3. Calculattons should be made o f the radioactive effects o f loss o f water from thm spent fuel pool. Nm currently do not knaw the effects on the pomr plant or upon thm publlc. Also, we do not have r procedure i n place dealing with the radiological consequences o f losing water from the spent furl pool.

I . hdtcations arm neadcd t o enable thr operators tn the control room t o dtrect ly monitor condttlons i n thm spent fual pools under normal ,

operating condltlons and under casualty conditions.

Thr above items a n b rad upon a body o f work done by NSAo over thm vearr. Thry err also bared on review o f NlfllllRC 91-06, W~~ELINES-FOR it&&jRY- ACTIONS TO ASSESS SHU KWGERENT and uoon our ~a r t t c to r t l on i n va~faur -~ ... indurtry m a t i n g ~ - the- l i tes t o f which wai'the lNPb-0utrgo Hanagar/Operatlons Manager Heettng i n August 92.

Thry paral lel. som but not a l l of the concerns raised by Don Fmvatta end Oaw Lochbaum i n their p)per. SAFE'IY CONSEQUENCES OF k BOILING SPENT FUEL #K)L AT THE SUSqUEHArmk S T W I ELECTRIC STATION.

I beliavm that a11 of the NSAO ltm can k addrrssrd durlng the process o f wro lv in the E n g l ~ e r l Dfscrepancy Report and subsquent correspandmce s u k i t t J by Prevrttm m "% Lochbarn.

f have attached tho r t levmt MSAG Raports m d carrcspondence.

The following rmctionr providr smar te r o f the M A G Reports and other pertinent docmnts and some suppartlng arguasntr.

1.1 Report 13-84. I q l l c a t ~ o n s o f Loss o f Water f r o m the Spmt fual Pool Duo t o Raactor Cavity R t l u r e o r Othcr buses. . This report wu enerated fn response t o the Haddm Neck Cavlty Seal Frilurm. The mQor rwamndatlonr have been 1wplmmonted and NSA6 considerr the raport t o have bomn closed.

The following itma ware not dono and should be raconridewd:

6.2.1 The impttcrtions of loss of spent fual pool lev81 upon radfrtlon rhlelding be thoroughly analyzed. This anrlysls bm addmd t o the FSAR.

5.2.3 Provido additional instruments that would Improve the oprrator's ab i l l ty t o respond t o a loss of lave1 from the spent fuel pool. Lovrl, tamperaturfi and rrdlatIon instr tmnts should k conslderrd.

The essence of both of the abovm recwwndatlons was r e p a t 4 i n HSAB Report 1-88. Sae below.

6.2.4 Instal l a wrtartiqht door betwan the raactor building rum roam and tho dlvtaion 11 core spray r o o m

NPE dtd an analysis and concluded that the r i s k o f f loodlng d fd nat warrant ins ta l la t ion o f l door t o protect the DiviJfan 11 Cora spray r o m (PLf-16640 of August I , lPB6). This was not pursuod fu r ther by NSAG. A study W S done by cmpetrnt e ineerr m d NSA6 awe tmd the rasslts. The issue may be worth rreons d e r i ~ tn l i g h t o f tRe P r m t t m concrrn.

"! 1.2 Repart 2-86, Analysis o f Operstlons With Potential f o r Draining thm

Rlactor Yassel.

Thls n p o r t was a fol low on t o Report 13-84. It analyzed the Trchnlcal Spclcifications, potential drainage paths md industry events. 1t reconnrnded t h r t an instructtan be wrlttmn deltnlng Operations w i th Potential For Dralnlng the Reactor Yessol.

All-QA-326. Oprratlans w i th Potmt la l f o r Drahfng t h r Reactor Vassml/Cavity, was wr i t ten and mt throw* eight rrvlstons. I t was reissurd on Wlrch 30, 1992 as OP-Al?-326. It i s re t l ve l y used durlng outago p lmnf ng and execution.

Thero are no open NSAO tssues.

1.3 Report 1-88, Inadvrrtent Dralnlng o f Water fmra t h r Spent Fur l k to l s on S e p t m h r 12, 1981 vla the Cask Storylr Plt.

This war r response t o I draina e incidmnt. The event o f September 12, 1987 uncovered a vulnerabi l i ty ? o fual pool drainaga tha t had not k m tonridered in NSAO bports 13-81 and 2-84,

Hfscellmeous p r o g r m a t i c ncomrndatlons k e r l i n p l m n t e d includin a pol icy t h r t the gates w i l l ba ke t i n r t a l l r d as much as porsiblm. ' ~ h s P policy, by the way, has Man f o l owed fa t th fu l ly .

Tw recannendatlons regarding the physical ins ta l la t ion rentah open.

5.1.3 Correct the problems with the fuel pool cooling system alarms. These include:

b. Thm hit 2 side o f tho OC211 panel has no r e f l l s h capabil i ty. (OC2ll i s the l o c r l fue l pool cooling p m r l .)

3.5.3.4 Upgradr the ir fornat lon concamlng the fuel pools m d fuel pool cool i n system that i s available t o the control room operator. !onsttier the an inrer ing work request submlttrd II by Operations Uanagmnt t a t ca l l s for the c r f t l c r l parameters t o bu made available t o the operators.

Rmconnnandrtion 3.5.3.4 1s essent ial ly I repeat o f rmcomnendation 5.2.3 o f Report 1344. It t s thm subjnct o f RGBts nmmo t o GTJ dated h e 30. Copy rttachmd.

A t h i r d recoannndrtion I s o f intmrrst. It rerds as follows:

5.2.2 Assess the adepuacy o f the fuel pool drainln malys is i n the FSAR conridmring the conditlon i n whtch ! he cask storage it rtms nr remvmd. Prmprrs I safety evmluation and

fOd50.59 rrvlew bmfore thm gatms are rezmved. Conrldar the p s s l b i l l t y o f uncovering rpont fue l by draining and or siphoning the cask storage p i t .

This reemendation I s a reattack on the recmmnendation f o r such an m r l y s i s i n Report 13-84. Thm safety evaluation was wr i t ten and accepted by the WRC and by thm SRC (PLI-54531). I took issue w i th the fact tha t ths shine from the spent fuel pool was not addrasrmd and wrote r mmm t o Rymrs (dated March 10, 1988) strongly recmendlng tha t a shlne m r l y s i s be done. (See attached.) Homvrr, no action was trknn.

I ctosmd tho Item. There was no qurstion i n my r i n d tha t the evolution i s safe md other than the s h i w question, whlch senmmd t a bother no on. but m, the safety evrluat!on snswred tha mail.

1.4 Report 4-90. Outage Plannlng ln fom*t lm.

The W a n t o f t h l s report was t o consolidrte thm knowledge galned by NSAG during the course o f r a v i h n g outage safety. The pr inc ipal user was intended t o bm N A G . However. the Information was made rvai lab le t o ovoryone concerned vi t h outrge p l mnlng.

The vrr ious heat renaval paths. including spent fuel pool teeling, hrvn bean anrlyzed and the capabilities and l i a l t a t i o n s o f tach have bean l l s t r d . Every e f f o r t was r r d m t o c i t e dastgn base rnfrrmcms, fornal t e s t resul ts and approved calculrt ions. When no calculations axistmd, NSAG d i d t h e i r awn.

A dmtr i lnd analysis has b n n made f o r each mtlestona o f an outage w l th ~uspec t t o dmcly hart remval. A table hrs bran devmloped showing thm requlrmmnts f o r erch condition, thm primary and thm back up coollng nnthods avaflable. A discussfon o f tha r r rv lce water outage and the lmplicatlons o f uslng the $pant f u r l wl cooling systems i s found on pages 23 - 25. The c rpab l l i t l e r o f t e spent fuel pool cooling systams arm found i n Apprndix C.

R There are no rmcoaauendatlons pmr sm.

The p r i nc lp r l problnn w i th the report i s that the derlgn heat loads arm out of drto. At the tim the report war wr l t ten NPE d i d not have an approved heat load cr lcuf ation. Thm hart load figures were takan fran a

Page S Smptawhr 9, 1992 PL1-ILS6t

calculat~on nadm by Jack Rafling for 8 X 8 fuel. NSAG intends t o update the report. We hava'galned knowledge since the re o r t was written. part icularly about the effects o f boiling. and wr R avo found soma minor errors that nemd to ba corrmctmd. Also, we want t o include a chaptmr on containment, which was not ready when the report was published.

Thm principal obstacle t o upgrading NSAG 44-80 i s the absence o f a dmffnitivm dmcw heat lord calculation. Thm Rmfllng calculatlon 1s a

a oad enough rpproxinrtion t o pannit g a ~ r a l outage planning. b e v e r . t understates the heat loads, par t icuhr fy i n the early days o f m

outage.

I would l i k e t o update Report 4-90. It Is usmful t o NSAO and t o other p lmnln entltlms withln WU. I t has barn dlt tr ibutad to the NRC, t o IHW an 8 t o intmrestad prr t ies I n thm industry. t t has bamn ci ted by both the WRC and by INPO ms mvldence of PPYVS leadership i n the outage n m r g m t f ie ld. I war invlted t o addreso the recant IHPO Uutrge/Opermttont Hmagars~ Mctfng largely beauso 1 had gone over tb ls report i n dmtmil wtth the fNPO Outage folks last rumer.

1 do not intmnd t o update thr rrport un t i l I have r calculation stgned by the rpproprtate enplntrrfng authority in hand. NSWs Intentton i s not t o do original enpineerlng work. Raport 1-90 consolidates existif ig knculedgo i n a fom that I s useful to persons planning and recrinrlng outige safmty. I want t o ba sure that any (nferences i n 4-90 are based upon o l f l c i a l PP&L heat load calculatlonr, HSAG, of course, I s aware that a calculatlon i s done b Nuclear Fuels before each outage. Such a calculation was urmd as the r; arts f o r IS% Raport 5-90. ANALYSIS OF ALTERNATE SHUTWYN COOLINB. Howorer, the company does not have an o t f l e f r l u lcu la t fon o f thr dec hart loads to be expected under outagm conditions. A study pmJactad r ead fo r s ~ v e r r l years i s needed for outage plmnina.

'3:

As you know, mK headed r )(UIIARC task force t o address outagm issues. Thr resul t was NWRC 91-06 GUIOELIKS FOR INOUSTRY ACTIONS TO ASSESS SMlTWYN WEMENT. MlKARC 91-66 is, tn my Judgment, 4n excellent piece o f work m d i t behooves us t o implmrnt i t mxpedltlously.

W S A G has r e v i m d the status o f our po l i c i r r vls-a-vls N W R C 91-06. The finding i s t h r t we are i n general eanylliance with the guidmltnes. Howaver. a Iargm mMbrr o f our paf ic let and practicms are not captured i n directives. Also, a nunber o f the issumt are not vcll covered by our procedures. Details arr found I n PLIS-39784 o f July 21, 1992.

NUMARC 91-06 contains thnm srctlons which are pertinent t o t h l r dtrcursion. They are:

4.1 .I Loss of Decay Heat Rpaovrl

4.1.3 Loss o f Spent Fuel Pool Coating

4.2.5 Rmwtor Cavity Smal h f k e

1 will rw a few words about tach.

2.1 t o r s of Decay Hrat Rmoval

ma guidallnm rmads as follows:

A procrdurm should bm establishrd t o address the loss of nonu1 OHR capabll i ty durfnp shutdown Condltlonr. Thm procodurn should prforl t lze thm a l ta rna t r cooling n t h & avrilablm (e.g., pmvit feed md blrtd, i low pressure pump f w d and bleed, high plussum pump lmmd md feed, n f l u x cooling, etc.) and tha t would bm nnployed fo r a givrn smt of conditions that ere plmnod f o r t h m outage. The pwcedura should hrvr a sound techntcal basis that includes thm following:

o fnitlrl mgnitude of decay heat

o tfm t o bofljng

o tim to corm uncovery

o M t f r l RCS water lnvrntory conditfon .. . . . o RCS eonfifluratlons . . . . . . . . ..

The SSES procedurn fo r alternate decay heat r m v r l is ON-119-001, LOSS OF MR SWmoSW CWLING W E . Thl s nrocoduw was in mart bared upon NSAG Rcoorts. r~ - - - - .. . 2-90, 4-90 and 5-90. I t was rrcmntly revlmd'by ~ n g i n e e d n ~ and fouhd toybe aceeptabh (PLI-71670 of June 19, 1992).

#SAG agrees that Ul-149601 I s trchnically corrcet. The necessary lnfornatlon has been included t o rnrblr 0 erators t o mxccute the varlous hmat r m v a l paths. Hwcvar. our rnview s R owd that thm p m d u r n dons not m f l r c t a11 of the qufdance of NlMRC 91-06. I t needs t o k updated t o includc the i n i t i a l mgnitudm o f decay hmrt. ti# t o boiling, EIpabilltfms of the various hart ranoval r thr , otc. Huch of t h r necessary infonnatlon i s found in NW Rrport 4-90, W ~ E PLANNING INFOREUTION. Hwever, NSAG 4-90 is based on a d e w heat calculation rsrulnlng a X 6 fuel. This calculation nceds t o bm updated to rclflect current conditions.

In order t o properly implement NUMARC 91-06 a drery heat crlculrtlon proimcted ahead for sevmral years is needad for input Into ON-149-001.

'

Pigr 7 September 9, 1992 PLI-72367

2.2 Los* o f Spent Furl Pool Cooltng

The guldel lnr stator:

Many u t l l t t i o s hrvr ehosrn t o off-load thr core t o tho spent fuel ( S P ) durlng t h W n f u e l i n outqrs. This practlce shffts decay e: rrmovrl requiremnts frm t 1 r RCS to the SFP. An event that r rsul ts I n tho loss o f SfP coollng may have the ram undrstriblr e f f rc ts as r loss o f DHR event i f appropriate compensatory measurer are not trkon.

1) The out890 rchbdule should provtde a DEFENSE I N DEPTH comensurate wfth the r i sk assoclotad wlth tors of SFP coollng.

2) A procedure should be estrbltshed for nspossr t o r loss o f SFP cool inp want.

YI w i l l constder the procedure f i rs t . 011-135-001, LOSS OF FUEL PGUL Clh3LIMGlC6PUM IWENIORY orovlder uuldmco on how t o r rs to r r fuel Pool ~ . . r ~ - - ~ - - ~~~

tooling l n th r evmt of varlaur cosiialti&. I f fuel pool cooltn can not be R r@storad, the options rrr t o placn RHR i n tha Fuel Pool Coolinq r i s t Hod8 i n recordrnu wfth OP-140-003, RHR DPERATIOFI IM FUEL Wol COOCIWG ASSIST or t o allow the od t o boll. Thr pracedurr stator that bol l lng should not occur bafare 25 R ours af ter toss o f cooling and it s~ec l f los vrr!wr nethods of

The procedure does not dlscurs cross connecting the fuel pools v i a the cask storage p i t a d rmw lng tha hart using the other unl t fuel pool coallng system. h a t can be remved by forced canveetlon (The pmcedurc Is in OP-136-001.) or by natural convection. Cooling by natural convection was dona tor m r i gh t day period during the Unlt 2 ZRIO. (Sea NSAG Report 4-90). I t i s r l r o posstbla t o cool tk unrffectod fuel pool usiag RHR I n Furl Pool Cooltng Arsfrt and r m v e heat from the affected pool by natural convectlon v ia the cask storagm pi t . This method I s not discussed I n ON-135-001.

h s m r y I viable mcmlure addrrrses the loss o f spent fuel pool coollng. i Hmver , a l l of tha horn backup methods are not included.

Z.Z.2 Defensm In Depth

During o u t q r lannlng r e v i m NSM takrs trrdlt for the following methods of ! spent fuel poo cooling:

1) Nomrl operatfan o f the outage unlt fur l pool coollng system.

Hr. 6. T. Jones Page 8 Se tcnbrr 9. 1992 PL!-72367

2) Us4 O f boa fuel pool cooling y s t m s i n parallel. F low betmen pools i s by natural convection via the cask storage pft. Thfs was tested durlng the Unit 2 1RIO. $n NSA6 4-90.

3) Coollng both fuel pools using the operating unft fuel pool coaling systm per 0~-135-0@1.

4 Coollng the outege fuel pool using m outage unlt RHR system i n fuel pool coollng assist.

B) Coolt the operating fuel pool urlng an oprratlng un i t RHR systm i n fu3 pool coollng ~ s s I J I and coaling the autage un l t by natural wnvectlon v h th4 cask storage plt.

6) Boiling

Ourtng refueling outages the station routinely removes the onttrr station deccy heat lo rd usidg th r operatlng unl t fuel pool cooling system. the s l tur t lon i s that the corr has been o f f loaded and both RHR rubsyrt#rs arc out o f rkrvlce. I n ordrr t o do r service w a t e r outage and t o clkan tho coaling twrrs, t h i outage un l t fuel pool cooltn systm nwst be dlaabled. w 9 r r t i m r l i z e doing th is basld upon the fa lowlng:

1) Prior t o disabling the SY syrtarn we run a trrt and prow that the operating fuel pool cwlfng system will handle the heat load. The t a r t rvsults are approved by thr PDRC.

2) Thr appropriate #s havr bem done t o the fuel pool coalfyl systems pr ior t o the outage.

3) Yh4 aplratlng fuel pool coollng system can hold fun1 ool P tentperaturr behw 200.F with one heat erehanger out o ~ W t c e .

4) The opsratlng un i t RHR systm can be used i n the fuel pool coallng assist mdr, i f necessary, t o cool the pool.

5) In the worst case boil ing w i l l occur. I f tht$ happens thrro w l l l be no rl n l f i c m t damlgr t o the station and no hannful e f f r c t upon the pub1 ! c.

6 ) Rmple ttm u i s t s to in i t ia te correct{ue action. Tha FSkR states that over 25 hours wlll elapso beform boiling begins.

I n I typlcal outa e both RHR systems are out of service for about 10 drys I( bqlnnlng i t abou doy 18. Ourin t h b period the back up cool~ng methods rrr 0 use of one o f the other un l t RHR OOPS i n fuel pool caoling assist o r bolltng. Nmlthnr of these to attrrctivn. Boll ln vlolates the Technfcal Sprelflcatlons P m d maker the a16 elevation unlnhabitab e, Vsr of thr oprratlng unl t RHR system Implies taktn a 72 hour LC0 on the operatlng untt and i t reduws the capability a4 the sa 9 r t y systomr nrtdrd t o support operatlonr.

Mr. 6. T. Jones Pwm 9

A point t o consider 1s that a t SSES the RHR system has never been ful ly t s s t d i n the f u r l pool cool ing a t r l s t mode. During the t r s t program flow was established a t about 2000 w. Hfgher flows wmrm not attained because tho r k lmr surge tank kmpt runnln dry. OP-149-003 ca l ls f o r r maxiann flow of S 6000 gp l~ i n the furl pool cool ng r s s l r t wda. RHR was &ctually used i n the f u r l pool coalhg r s r l s t node a t Brunrwlck i n 1983. However. tk.r wrm unrbl. t o n a l n t r h flow abov8 2000 g p because rt higher flow rates thm pool overflowed. A flow o f 2090 g w l l l carry away tha haat load durlns swvlce

delta Ts would k slmllrr. r wrter outage condltlans. Fue pool cooling flow i s about 1800 gpm and the

Based upon tha rbovr I believe that a study should br made t o drtmmlnm whather r more viable aathod of backup decay heat removal durlng smrvice water O U ~ ~ ~ O S I S f ~ a ~ l b l e e

On August 27 GTJ tssued PLI-72267 t o Glenn U l l l e r addressing the Prevattm/Lochbau* concern. The l a t t e r stated,

Thmre are twnty-eight open items rssultlng fm NSA6 Review o f Fuel Pool Caal ing. Thrsm need t o be included i n t h l r revlm.

1 prmsme that thlr statemnt i s brsrd on Jim Kenny% W m o f hu ust Zf 9 docummntinq h is conversation u l t h Scott Barber md Jlm Raleigh o the NRC.

On September 2 Andrr Domlnguez dtrcussed thm NSAG items with Scott Barber. Scott s m r r i z e d his concerns a t follows:

1) PPLL has ramvaluated tha seat l lves from ftve years recanmnded by the vendor t o 12 years Pot t h m lower and 24 yr r rs f o r the upper.

2) Lack o f fuel pool cooling instrunentatlon i n the control M ~ A .

3) Adepurcy of operator tralnim]. I s i t adequate? I s it s t i l l br lns donel

4) The effacts of radtatlon following a drain down event have not been addrersmd.

Dollinguez dlscussrd barber*^ concerns with JW Zolr, Jim Agnw and nark ratvedt l i t e r on Scptwber 2. They w i l l bm addrarsed. Agne* and Domlnguez

Y r n t o dfscuss the i t e m with Barber on October 5.

Barbmr's ltem P I s the subject of NSAG's only a f f i c i r l open ftems i n the spent fur l pool area.

This apmr was wrlttan I n rasponso t o your request that I summrlze the t t m s ralatP*a t o fuel pool cooling t h r t are of concern t o W. The r-ry i s found I n the openhg parrgraph. Tha rucceedlng pagmr provldm background and just i f icr t lon.

1 su stod 8bove t h a t th. USA6 ttems can be rddressrd duri the pm#ss o f rrso P' vlnp the EDtt s u b t t t l d by Pravattm and Lochbarn, I WP "! d also l i k e t o rug r s t tk8t Enqtneerlng go beyond msolvlng tb r specific concerns. T h m f u r l poa ! caallng systems are r v i t a l part of thm SSES wtrfle pronss. Engtnmmring should taka th ls opportunity t o do m fn-depth analysts o f thm fuel pool cwtlnp s l tur t lon including nvlmrlrlng our ~ r i r t l n g prrct icrs and proc ldu re~ En fnmmrlng s h l d then provide thm t d n i c a l laformtion Mas¶ary t o fn !, l l l gan t l y operam th r furl pool cooling c ~ a p l e ~ under nownel mnd upset candt tfonr . Vwy respectfully,

Attachment 17

PP&L L e t t e r f r o m James E. Agnew t o D a v i d A . Lochbaum, "EDR 620020, Spent F u e l Poo l D e s i g n D i s c r e p a n c i e s " , Oc tobe r 7, 1992 (ET-0785)

Note: T h i s l e t t e r t r a n s m i t t e d t h e f o r m a l e v a l u a t i o n p e r f o r m e d b y t h e PP&L E n g i n e e r i n g D i s c r e p a n c y Management Group f o r EDR G20020. T h i s e v a l u a t i o n d e t e r m i n e s t h e i s s u e s i n t h e EDR t o have m i n i m a l s a f e t y s i g n i f i c a n c e w i t h no a f f e c t on p l a n t o p e r a b i l i t y u s i n g t e c h n i c a l r easons w h i c h c o n t r a d i c t t h o s e e x p r e s s e d i n t h e EDR and i n t h e i ndependen t PP&L e n g i n e e r i n g r e p o r t ( A t t a c h m e n t 1 5 ) .

Pennsylvania Power & Light Company Two North Ninth Street*Allentown, PA 18101-1179*215/774-5151

October 7, 1992

Mr. David A Lochbaum Enercon Services, Inc 4115 William Penn Highway One Franklin Centre Murrysville, PA 15668

Engineering Discrepancy Management Group EDR 620020, Spent Fuel Pool Design Discrepancies ET-0785 File A46-1A

Please find enclosed completed copies of the Screening Evaluation, Reportability Evaluation, and Operability Evaluation for the subject Engineering Discrepancy Report (EDR).

The evaluation results and disposition are summarized herein:

Screenjgq: The significance of this discrepancy has been determined to be minimal. However, the priority of implementation has been elevated to ensure prompt resolution of the discrepancies.

pe~ortability: This discrepancy has been determined to be not reportable. However, I have specifically requested an independent review of reportability by Nuclear Licensing.

O~erability: This discrepancy has been determined to have no impact on the operation of Susquehanna SES.

The discrepancy evaluation function is considered a continuous process. A re-evaluation of the Safety Significance (Screening), Reportability, and/or Operability status will be performed at any stage of EDR processing, including implementation, as additional information becomes available, in accordance with the Discrepancy Management Program.

If you have any comments or questions on the attached, please do not hesitate to contact me.

James E. Agnew (215) 774-7777

cc: G.T. Jones A6-2 G.D. Miller A6-3 C.A. Myers A2-4 EDR File A6-3 ET Memo File A6-3 NR File A6-2

Attachment 18

PP&L Memo from G. D. Miller t o G. D. Miller, "Assignment of EOR", October 7, 1992 (ET-0780)

Date: 10/07/92

G D Miller A6-3

Engineering Disarepaney Nanaguont Group Ansigmont of BDR

BT: ~ 9 b rile nr6 a - This is to assign you EDR No. 620020 Rev. 0 for implementation in accordance with EPM-QA-122, Revision 3.

SUBJECT: LOSS OF SPENT FUEL POOL COOLING EVENT DESIGN DISCREPANCIES

ACTION ITEM: Establish original design basis for fuel pool cooling sys, determine appropriate design basis for spent fuel pool coolong sys, compare the design basis & resolve as necessary

INITIAL UNIT ACTION DATE: 11/19/93 {Cycle U17)

The EDR CLOSURE DATE: 05/20/94 {Cycle U26)

The Priority Classification is:S

The INITIAL UNIT ACTION DATE is for EDRs written against both units and reflects the date and cycle for which action is required. All activities associated with this issue for that unit shall be completed before the end of the cycle.

The EDR CLOSURE DATE reflects the deadline for complete EDR closure, includin the disposition of all related actions required to resolve the deficiency.

Enclosed is a copy of the subject EDR and copies of the appropriate evaluations.

Please ensure timely implementation of this EDR and keep me current of developments leading to EDR closure.

The EDMG Planner will visit your assigned engineer/planner on a routine basis for status updating.

You will be alerted at regular intervals (90, 60, 30 and 10 days to closure), if applicable, to allow you time for an orderly implementation of the EDR.

Upon implementation of the EDR, please notify EDMG of completion of required work and of any documentation generated.

Attachments

cc: J E Aqnew A6-3 w/o M R Mjaatvedt A6-3 w/a J A Zola A6-3 w/o D Lochbaum A6-3 w/EDR form D F McGann SSES w/a Engr Tech File (JW) A6-3 w/o EDR File (MSS) A6-3 w/a NR File A6-2 w/a


L e t t e r f r o m D a v i d A . Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, " R e p o r t a b i l i t y o f B o i l i n g Spent F u e l Poo l Concerns" , Oc tobe r 9, 1992

Note : T h i s l e t t e r was hand d e l i v e r e d i n a m e e t i n g r e q u e s t e d b y t h e a u t h o r s . The a u t h o r s had n o t r e c e i v e d t h e PP&L f o r m a l e v a l u a t i o n s ( A t t a c h m e n t s 17 and 3 ) f o r EDR 620020 u n t i l m i n u t e s b e f o r e t h e m e e t i n g . T h i s l e t t e r d e c l a r e d t h e a u t h o r s i n t e n t i o n s t o r e p o r t t h i s m a t t e r t o t h e NRC i f PP&L d i d n o t p r o p e r l y e v a l u a t e t h e conce rns b y November 2, 1992. On J u l y 27, 1992, t h e a u t h o r s had e s c a l a t e d t h e i r conce rns t o t h e PP&L Manager o f N u c l e a r P l a n t E n g i n e e r i n g ( A t t a c h m e n t 8). On t h i s d a t e , t h e a u t h o r s e s c a l a t e d t h e i r conce rns t o t h e S e n i o r V P , N u c l e a r , t h e SSES P l a n t Manager, and fl members o f t h e S a f e t y Review Commit tee b y copy o f t h i s 1 e t t e r .

George T. Jones Pennsylvania Power & Light Company Two North Ninth Street, A6-2 Allentown, PA 18101-1179

October 9, 1992

SUBJECT: REPORTABILITY OP BOILING FUEL POOL CONCERNS

Dear Mr. Jones:

On April 16, 1992, Engineering Discrepancy Report EDR G20020 (Attachment 1) was initiated in accordance with PP&L engineering procedure EPM-QA-122 by the two signatories to this letter, Mr. David A. Lochbaum and Mr. Donald C. Prevatte, to address nine nuclear safety concerns relating to the boiling spent fuel pool event at the Susquehanna Steam Electric Station. This letter is being written to formally reiterate our belief that these concerns are very real and very significant nuclear safety issues, and to convey our concern that they are not being addressed in accordance with either the letter or the intent of PP&L procedures and Federal regulations, and to express our determination that they must be acknowledged, reported, and resolved in a manner commensurate with their significance.

These concerns developed from Mr. Prevatte's evaluation of the reactor building ventilation systems for power uprate, Mr. Lochbaum's evaluation of the spent fuel pool cooling system for power uprate, and our technical reviews of each other's work. These concerns were initially documented in our memo dated March 19, 1992 to our supervisor, Mr. Mark Mjaatvedt (Attachment 2) . Mr. Mjaatvedt routed this memo to his supervisor, Mr. Glenn Miller. On April 15, 1992, Mr. Mjaatvedt directed us to initiate an EDR based upon Mr. Miller's review of our memo.

It was not until June 11 or 12, 1992 that the Engineering Discrepancy Management Group (EDMG) engineer, Mr. Joe Zola, contacted one of us, Mr. Lochbaum, to report that from his preliminary assessment of EDR G20020, the concerns had no safety significance. Mr. Lochbaum indicated at that time that he sincerely considered each of the nine concerns to have adverse nuclear safety significance both at the present time and in the future.

Mr. Zola arranged a meeting on June 18, 1992 to discuss the concerns. Attendees at this meeting were Mr. Zola, the EDMG supervisor Mr. Jim Agnew, Mr. Charlie Brown, Mr. Kevin Browning, Mr. Dave Pai, and both of us. We felt the meeting was successful in that the system engineers (Mr. Brown and Mr. Browning) conceded that the spent fuel pool would boil following reasonable scenarios

within the SSES design bases. Mr. Pai reported that the standby gas treatment system was not designed for the conditions resulting from a boiling spent fuel pool and would isolate on high temperature. We issued a supporting document for EDR G20020 on June 22, 1992 (Attachment 3) providing additional information on the nine issues. This document explicitly stated the regulatory requirements which are not being satisfied for each of the nine concerns in EDR G20020 along with the associated adverse safety implications.

An onsite meeting was held on July 8 or 9, 1992 to discuss EDR G20020. We learned about the meeting the day before and asked to attend, but this request was denied. We were told that our position at the meeting would be represented by the EDMG engineer (Mr. Zola) . Based upon feedback from Mr. Mjaatvedt , Mr. Chris Boschetti, and Mr. Paul Weaver, this meeting was not productive.

After Mr. Mjaatvedt told us about the onsite meeting, we looked at the EDMG file on EDR 620020. Other than the EDR itself, the only contents in the file were a page from a memo dated April 23, 1992 (Attachment 4) in which EDR G20020 was determined not to affect Unit 1 operability due to %o apparent impact on plant. DBD issue" and a draft screening worksheet for EDR G20020 prepared by Mr. Art White (Attachment 5). The screening worksheet, which stated "this discrepancy has no basis in fact" and other blatant falsehoods, convinced us that EDR G20020 was being dismissed improperly. We protested immediately to Mr. Mjaatvedt, who promised to consult with Mr. Agnew on the status of EDR 620020.

On July 10, 1992, Mr. Miller came to us and asked if we had problems with the EDR process in general and EDR G20020 in particular. After a discussion lasting several hours in which we expressed our strong concern with both, Mr. Miller pulled EDR G20020 from its file and Verifiedf8 the identified concerns. Mr. Miller also directed the EDMG to arrange a meeting with all concerned parties to review EDR G20020 and obtain its resolution.

This large meeting was conducted on July 15, 1992. We prepared supplemental information which was distributed at this meeting (Attachment 6). This information addressed the regulatory requirements, licensing requirements, and design evolution history for the items in EDR 620020. Attendees at this meeting included Mr. Mike Detamore, Mr. Miller, Mr. Jim Kenny, Mr. Rocky Sgarro, Mr. Zola, Mr. John Bartos, Mr. Mjaatvedt, Mr. Tony Roscioli, Mr. White, Mr. Agnew and both of us. The meeting was not productive. Basically, we were told that our concerns were unfounded because the SSES design bases were not required to handle a LOCA/MOP event coupled with a loss of spent fuel pool cooling event, and operators would take the necessary appropriate corrective measures anyway. We strongly disagreed with both of these positions.

On July 1 7 , 1992, Mr. Lochbaum was informed that his contract would not be extended beyond July 31, 1992. On July 2 3 , 1992, Mr. Lochbaum asked Mr. Mjaatvedt for a final meeting with Mr. Agnew to update him with the status of EDR 620020 prior to his termination. When the scheduled meeting was cancelled and not rescheduled, we came to you on July 29 , 1992 to express our concerns with the EDR process and EDR 620020. We prepared another summary of the boiling spent fuel pool concerns which we provided to you at that time (Attachment 7 ) . This summary covered the four major safety concerns in EDR 620020 with their requirements and consequences. Mr. Lochbaum met with Mr. Jim Miltenberger later that day at your request.

On August 1 8 , 1992, Mr. Miller issued a letter addressing EDR 620020 (Attachment 8 ) and stating that its safety significance was minimal because the NRC had reviewed and approved the fuel pool cooling system design at SSES. Mr. Prevatte responded to Mr. Miller's letter (Attachment 9 ) and Mr. Lochbaum called you to register disagreement with the position outlined by Mr. Miller. On August 27 , 1992, you issued a letter (Attachment 10) to Mr. Miller directing .him to reconsider the safety classification for EDR G20020 and provide a schedule by August 31, 1992. On August 31, 1992, Mr. Miller responded to your letter (Attachment 11) by clarifying his position and reporting that a schedule was still under development. In the meantime, Mr. Kevin Brinckman completed his independent appraisal of the boiling spent fuel pool issues and released his report (Attachment 1 2 ) . Mr. Brinckman ' s study essentially endorses every concern identified in EDR 620020 and even points out that the probability and/or consequences of some concerns may be greater than presented in the EDR.

A number of informal' discussions between Mr. Miller and Mr. Prevatte and Mr. Lochbaum on October 5 and 6, 1992 addressed PP&L1s reportability and operability determinations. Mr. Miller stated that these determinations were about to be formally issued. Mr. Miller indicated that PP&L determined the concerns in EDR 620020 not to be reportable under 10CFR50.72 and not to affect operability. Both Mr. Prevatte and Mr. Lochbaum registered strong objections to the justification offered by Mr. Miller for these determinations. Mr. Miller indicated the issue would be discussed with the NRC at the quarterly meeting on October 7 , 1992. Mr. Lochbaum asked Mr. Miller if he could attend this NRC meeting and was told it would be inappropriate. Mr. Lochbaum additionally requested that Mr. Miller in his presentation before the NRC clearly state that the originators of EDR 620020 have not yet been provided with documentation of PP&L1s operability and reportability determinations and have strongly disagreed with the justifications offered informally by Mr. Miller. Mr. Miller told Mr. Prevatte that he would provide the NRC Resident Inspector with a copy of EDR 620020, its reportability/operability determination, and all related correspondence.

Page 3

Mr. Lochbaum discussed this matter with Mr. George Jones by telephone on October 8, 1992. Mr. Jones stated the fuel pool concerns had been discussed with the NRC during the previous day's meeting and that the SSES NRC Resident Inspector would be given information on EDR 620020 that day.

While there have been some informal discussions since August 31, 1992, there has not been any documented progress made on this issue since the end of August 1992. In EDR G20020 and the supplemental information, we provided PP&L with a comprehensive package detailing nine problems with the boiling spent fuel pool event at SSES along with their associated adverse nuclear safety implications. To date, none of these nine items has been formally refuted by PP&L. It appears to us that PP&L is unwilling to concede these are problems until it has completely defined the measures required to resolve the problems. This course of action does not satisfy the procedural requirements of EPM-QA-122 or the reporting requirements of 10CFR50.70. Nuclear safety concerns must be formally reported to the NRC in order for other sites with similar conditions to be alerted.

We consider the problems identified in EDR G20020 to have significant adverse nuclear safety implications. A design bases event at SSES is a LOCA with a concurrent LOOP. Even in the case of a LOCA without a MOP, SSES procedures may initiate a shedding of non-Class 1E loads inside the reactor building in order to limit room temperatures. Since it is a non-safety related system not powered from Class 1E sources, the fuel pool cooling system will not operate. Without the fuel pool cooling system, either the fuel pool cooling assist mode of RHR (a non-safety related, non-single failure proof function) must be initiated to provide fuel pool cooling or ESW makeup to the fuel pools must be initiated to maintain water level in the boiling spent fuel pool. Either operation would require operator entry into radiation fields significantly higher than reported in FSAR Chapter 18 and permitted by 10CFR50. If the spent fuel pool boils, the effects of the latent heat load on reactor building room temperatures and of the condensation/overflow on reactor building equipment operation have not been evaluated. The boiling spent fuel pool therefore represents the potential for providing the means for the common mode failure of all ECCS and safety related equipment in the reactor building. If ESW makeup to the boiling spent fuel pool cannot be achieved, there is also the potential for the meltdown of irradiated fuel outside of the primary containment with the concurrent failure of the standby gas treatment system.

The nine concerns identified in EDR G20020 indicate that PP&L has not performed an integrated engineering evaluation of the boiling spent fuel pool design event. The resolution to EDR 620020 must

Page 4

include a thorough engineering assessment of the loss of normal spent fuel pool cooling event on component, system and plant levels to ensure that adverse consequences such as pressurization within the reactor building due to spent fuel pool boiloff do not jeopardize secondary containment integrity and ECCS performance.

We are very concerned that the nuclear safety issues raised in EDR 620020 are not being evaluated for operability and reportability in a proper manner. You have stated that the NRC Resident and Region I personnel have been informally notified about this issue. However, we have not yet seen a document prepared by PP&L, other than Mr. Brinckman's study, which presents this issue in a complete and accurate manner. To date, we have seen no action taken on Mr. Brinckman's study. Mr. Miller and the EDMG have repeatedly attempted to narrow the scope of EDR 620020 by focusing solely on the fuel pool cooling system design bases. During our only contact with Nuclear Licensing, Mr. Kenny claimed our concerns were unfounded since SSES was not required to design for both a LOCA/LOOP and a loss of spent fuel pool cooling event. Therefore, we doubt that PP&L has presented the NRC with a thorough understanding of the issues raised in EDR 620020. Consequently, we discount any claim by PPhL that the NRC has indicated that these issues are not reportable.

Because EDR 62002.0 was initiated over five months ago, because we meet with you over nine weeks ago, because Mr. Lochbaum is no longer working at PP&L and because Mr. Prevatte may not be working at PP&L for much longer, we respectfully request that PP&L provide to us in accordance with EPM-QA-122 Sections 4.4 and 4.5 written documentation of:

la) A technical justification for gg& of the nine items identified in EDR G20020 indicating why item is not a nuclear safety concern, - OR -

lb) A copy of the report made to the NRC addressing the items raised in EDR G20020.

2) A final approved screening worksheet for EDR G20020 per EPM-QA-122 and EPM-703.

3) A final approved Reportability Evaluation for EDR 620020 per EPM-QA-122 and EPM-704.

4) A final approved Operability Evaluation for EDR G20020 per EPM-QA-122 and EPM-705.

We request your response by no later than November 2, 1992.

Page 5

If we consider the technical justification to be inadequate, the NRC report to be incomplete, or if PP&L fails to respond by the specified date, we intend to proceed with our own report to the NRC on this subject. This report would cover the nine concerns identified in EDR 620020 and indicate that several of these concerns had been raised numerous times in the past but never resolved by PP&L. It would also express our concerns that PP&L does not have an effective program for handling and resolving questions of nuclear safety as evidenced by PP&L1s treatment of EDR 620020 and other recent EDRs and safety issues. We acknowledge your statement that you are not satisfied with the EDR program, but the existing program is severely faulted and no substantive corrective measures have been instituted since our meeting with you on July 29, 1992.

We deeply regret having this issue reach this stage, but we know of no legitimate alternate actions we could have taken to avoid this point. In fact, we sincerely feel we have been extremely patient, professional and open-minded in our dealings with PP&L on this issue. We urge PP&L to properly resolve this issue so that our next step.need not be taken.

Sincerely, - David A. Lochbaum \ b & Q !I%&

Donald C. Prevatte

Attachments:

1) Engineering Discrepancy Report G20020, V.~ss of Spent Fuel Pool Cooling Event Design Discrepanciesn8, April 16, 1992

2) Memo from Dave Lochbaum and Don Prevatte to Mark Mjaatvedt, InSusquehanna Steam Electric Station Spent Fuel Pool Boiling Issuesnq, March 18, 1992 (ET-0149)

3) Memo from Dave Lochbaum and Don Prevatte to Joe Zola, "Supplemental Information for EDR 620020 on Boiling Spent Fuel PoolM, June 22, 1992 (ET-0471)

4) Operability Statement Page 63, EDR #G20020, April 23, 1992

5) Screening Worksheet, EDR G20020, Draft by Art White

6) EDR G20020 References, July 15, 1992

Page 6

7) White Paper prepared by David A. Lochbaum and Donald C. Prevatte, losafety Consequences of a Boiling Spent Fuel Pool at the Susquehanna Steam Electric Stationn, July 27, 1992

8) Memo from G. D. Miller to G. T. Jones, "Fuel Pool Cooling Defi~iencies*~, August 18, 1992 (ET-0586)

9) Memo from D. C. Prevatte to G. T. Jones, "Fuel Pool Cooling Def icienciesIo , August 20, 1992 (ET-0587)

10) Memo from G. T.-Jones to Glenn D. Miller, "Fuel Pool Cooling EDR's G20020, G00005", August 27, 1992 (PLI-72267)

11) Memo from Glenn D. Miller to George T. Jones, "Fuel Pool Cooling EDRqs G20020, G00005", August 31, 1992 (PLI-72297)

12) Memo from Kevin W. Brinckman to George T. Jones, 08Review of Fuel Pool Coolingoo, September 1, 1992 (PLI-72288)

Copies: H. W. R. G. H. G. W. R.

Keiser By ram Stanley Corcoran

Kemper

Doty Iorf ida Sabol Licht Stef anko Miltenberger Myers Kenny Butler Miller Agnew Zola Mj aatvedt Kuczynski Boschetti Sweeney Gogates Manski Richardson

A6-1 SSES 21 Broadleaf Circle Windsor, CT 06095 115 Polecat Road Glenn Mills, PA 19342 A9-3 SSES A2-5 A6-1 A9-3 A6-1 A2-4 A2-4 A6-3 A6-3 A6-3 A6-3 A6-3 SSES SSES SSES SSES Enercon Enercon

Page 7

Attachment 20

PP&L Memo from D. A. Lochbaum and D. C. Prevatte t o George T . Jones, "EDR System Concerns", October 13, 1992 (PLI-72365)

Note: This memo followed up on the concerns voiced by the authors in a meeting with the PP&L Manager o f Nuclear Plant Engineering on the overall handling o f nuclear safety issues by PP&L.

October 13, 1992


SUSQUEHANNA STEAM ELECTRIC STATION EDR SYSTEM CONCERNS PLI-72635 FILE A45-1A

This letter is being written to follow up on the meeting on Friday, October 9, 1992 between yourself, Mr. Chuck Myers, Mr. Glenn Miller and the signatories to this letter to discuss the evaluation of EDR G 2 C 3 2 P .

In this meeting and other previous conversations with you and Mr. Miller, we have expressed our concerns that the EDR process is not working as required by the procedures and federal regulations. The following is a listing of the most significant concerns we have developed in attempting to work with the system along with some suggestions on how we feel the system could be improved:

CONCERNS WITH CURRENT EDR SYSTEM AND APPROACH

1. The EDR Group is a part of the same organization which has the primary responsibility for the plant design, thereby creating a basic conflict of interests. The EDR Group should be independent.

2. Inadequate resources are earmarked for the disposition of EDRs .

3. The Itpresumption of operability" philosophy is carried to the illogical extreme, to the end that, for many EDRs an a~.-+p.- -~. - . ;a: - rclationnhip roszlts bsti:=-,r: the ECZ izitiztcr cn2 the system. Although the "presumption of operabilityt1 should exist until proven otherwise, hand-in-hand with this philosophy should be a tspresumption of validityv1 in any concern raised. The EDR process should be a search for the truth, not a process to make problems go away. The discoverer of a problem should not have to make an air-tight case for the process to work. All he should have to do is have reasonable indication and belief that a problem exists, and the process should then research the issue on both sides, and then let the issue stand or fall on its own technical merits or lack thereof.

4. Every step in the process should be performed to a clock. EDRs must not be allowed to languish as they sometimes do today.

G. T. Jones Page 2 October 13, 1992 PLI-72635

5. The solutions to the reported problem, the anticipated solutions, or the lack thereof should be entirely divorced from the evaluations of the EDRs. It should not be the responsibility of the EDR Group to resolve EDRs, only to evaluate them..

6. Although not sanctioned by procedure, the present presumption seems to be that if an item is reportable, it is not operable. This is not necessarily true. A J.I.O. can be generated to contime plant sperztix ir. msst cafzc ,&her? an iterr. is reportable. The two determinations must be independent.

7. The determination of operability and reportability should have less reliance on the legalistic aspects of the issue and more on the technical validity or lack thereof. If there is a conflict between the legal and the technical evaluation, the technical should prevail. There is overwhelming evidence that this is the intent of the NRC in all of the CFR reporting requirements.

8. Potential cost of resolution should have absolutely zero consideration in operability/reportability determinations.

9. Responses to concerns in EDRs should be made on a point-by- point technical basis, not motherhood type statements that are more appropriate for press releases than technical documents.

10. The following reasons and other similar reasons for dismissing an EDR concern should be absolutely disallowed:

The NRC approved it. We have (had) an understanding (undocumented) with the NRC . There are ( x ) number of backups. Therefore, the weakness in this item will be made up for in the backups. We're allowed a single failure. The operators will take whatever action is necessary to make up for the weakness. The EOPs, EPs and other similar features of our defense- in-depth make up for the weakness. PRA (appropriate for J.I.O.s, not for reportability/operability determination). The issue will be addressed in an upcoming DBD.

11. The EDR process is too complex and convoluted. The process and forms should be simplified. There are too many gates that EDRs must pass through.

G. T. Jones Page 3 October 13, 1992 PLI-72635

12. The effectiv ess of th e EDR process should not b e judged on the volume of EDRs that pass through the system, but rather on the technical depth and quality of the evaluations.

13. The argument has been raised that the NRC does not want us reporting too much, and this position has been used to rationalize not reporting conditions which probably should have been reported. It should be borne in mind that the penalty for reporting too much is essentially nothing; the penalVl- foi reporticg tco littls cad$ be catastrephic, for the company, for the individual, for the customer, and in the extreme case for the communities around the plant. We have everything to lose and nothing to gain by not reporting when we should. From every point of view - cost, reputation, safety, ethics, credibility - we should err on the side of reporting.

14. A step in the EDR evaluation process should be the written concurrence or non-concurrence of the originator with the evaluations.

We are submitting this critique with the hope that it can be used constructively to improve the quality and effectiveness of the system. If we can provide any further input, please do not hesitate to call upon us.

: J. E. Agnew, Jr. A6-3 G. D. Miller A6-3 C. A. Myers A2-4 EDR File A6-3 N R File A6-2

\ALP\& D. C. Prevatte

Attachment 21

PP&L Memo from George T. Jones t o G. 0. Miller, "Spent Fuel Pool Issue", October 14, 1992 (PLI-72640)

Note: In this memo, t h e PP&L Manager of Nuclear Plant Engineering directs a coordinated engineering effort t o be initiated to address the concerns in EDR 620020. This action comes approximately six (6) weeks after receipt o f an in-house engineering report (commissioned solely t o assess the concerns in EDR 620020) which did not refute the primary safety issues raised in EDR 620020.

You are t o h d i a t e l y form 4 team whlah includsrr r ap ruen t a t i on from Tmohnology, Yloditications, Fuels, Bystam Enginemring, Oporatiom and Liceruing.

Develop and u-ta of the o p r a b i l i t y and the repor tab i l i ty determination t h a t apecif ioal ly addra.8 a l l of thm is8um8 rainad i n tha oubj.crt EDR. Thim action is t o k complated by wtober 21, 1991.

. . Dava l~p a jumtification for intmrim op8ration t h a t c l . a r l y addrm8sem a l l of the isrue8 r a i d i n IIDB-20010. This a c t i v i t y i m to 1N ~lqpleted by Octobu 11, 1992.

Develop any long tnrm actionn naeded t o oonplately rmsolve thm imsua to tho point of l . q l a a ~ t a t i o n , i.e. p ~ u r m s rfflmion r u g for WRC roviow, lod i f i ca t io lu ready f o r start of dosiqn. The initial 8oop. of any aationm is to be ample to8 by Octobmr 28, 1992. This act ion i m t o ba completed by W w a p b . r 11, 1991.

By copy of this l a t t e r to E . A. Nyors, 3. 6 . Stefanko, C. J. Kuczynrki, R. G. 8t.nZey urd n. W. Slrp.on, you are requaeted to tlupply individual part icipants i n t h i s orrort.

A

Attachment 22

PP&L Memo from George T. Jones t o G. D. Miller, J. S. Stefanko and M. W. Simpson, "Spent Fuel Pool Cooling Issue", October 14, 1992 (PLI-72641)

Mr. G. b. Miller m. J. 6 . Stofanko . M a. Bimpmon

While w e haw analyard tho haat load. i n tho Spent Fuel Poole due t o n m ruloada and operating mtratqiee, w havo not updated c u t a i n deslgn analymom to r Bpent Fuml Pool Cooling System Design Bamim m t m . I a l w need t o know i f them a m oUur r i n i l a r mituationr. You ara armigmd the following actioner

Ddcurrurt tho haat loadm that nmad t o be coneiderad during normal operation, singlo unit (full ooro offload) outagrm, and a 2 uni t (2 f u l l core otflaadm) outmgmm fo r uae by mclear Tachnolagy i n updating damiqn - m i 8 analymfm. Loud - a. B. mtotanko.

Updata our damign b a d 8 analpom for tha Spent Fuel Pools (and the appropriate FSAR sactiom) wing tho abovn Information. &ad - 9. n. mil l* .

U p Q a t r our radiological analyoim for the spmt Fuml Pools (and the approplirta FSAR oactionc) w i n g th. above

Identify a l l c n n m wore changes in fuol dmmign or core loading mtratagy m y bpaot wbatantial ly th. demign of tho plant inoluding ohangom i n deeign analymrs, operation prw~durmm or hudwarm. Idmntify d e h l t e m m a re n o t handled M a normal part of rolead analymim and domign. - a 6. mtofallko

RoviBa prooodurmm rovimiern that provide clear instruction for review of madification by Puolm Group mad adequate input plant d u i g n modifications. b u d - H. W. 8hpmon

Jarry 8tmf.nko ham the w a r a l l lead for thi8 o f to r t and fo r amsurinp timt PPLL t a k u a l l of the menmurea required t o r e f l e c t tho i m a o t of fuel reloads into our Wim, pr~e rdura and

At tachmen t 23

PP&L Memo f r o m George T. Jones t o A l l N u c l e a r E n g i n e e r i n g Managers and S u p e r v i s o r s , " E n g i n e e r i n g D i s c r e p a n c y (EDR) Program", Oc tobe r 14, 1992

Note : T h i s memo d i r e c t s PP&L s u p e r v i s o r s and managers t o r e v i e w t h e PP&L EDR p rog ram w i t h t h e i r p e r s o n n e l . I t s t a t e s t h a t f o r EDRs i n i t i a t e d , " t h e p r e s u m p t i o n i s V a l i d i t y u n t i l p r o v e n o t h e r w i s e . " The PP&L S u p e r v i s o r o f t h e E n g i n e e r i n g D i s c r e p a n c y Management Group t o l d one o f t h e a u t h o r s (Lochbaum) t h a t t h e r e a s o n he d i d n o t a c t i v e l y p u r s u e t h e EDR f r o m A p r i l 1992 t o J u l y 1992 was t h a t t h e consequences o f EDR 620020 were so l a r g e t h a t i t was i n c o m p r e h e n s i b l e t h a t t h e y c o u l d have been m i s s e d d u r i n g d e s i g n .

ti- U b b , P

Date: October 14, 1992

To : TO ALL NUCLEAR ENGINEERING MANAGERS AND SUPERVISORS

From: GEORGE T. JONES

Subject: ENGINEERING DISCREPANCY (EDR) PROORM

I would like to share some thoughts with you on the Engineering Discrepancy Program and I want you to share them with your people within the next two days and supply me with the attendees and resulting comments.

It is important to keep in mind that an essential element of any safety culture is ensuring that conditions adverse to quality, plant safety and reliability are promptly identified, reported and corrected.

Our purpose in establishing the Engineering Discrepancy Program ,

and the Group which supports it is to assure continued attention to this essential element.

In addition to establishing the group, we established a periodic review of the progress in close ovt of EDRs with both PORC, SRC and ERC. I also review our progress in this area with Senior Management. It is an item of their continuina interest and will remain so. We also conducted detailed training in the process, which most of you attended. If you or your people have not attended training, please contact Walt Rhoades and schedule this training. It is your responsibility that your people are trained.

We have achieved success in identifying potentially safety significant items and have improved our performance in resolving these issues in a timely fashion. We have also achieved success at identifying and separating those items not considered to have safety significance but important to do.

It is important to understand everyone's obligation to this Program.-In Policy Letter 90-003, signed by Harry Keiser, it states:

"Our expectations of Nuclear Department Personnel are that they will:

A c ~ l o r J s Identify and report any known or perceived deficiency - in the design or operation of Susquehanna, or any ~ R A I I ~ DIM RWOP-TS significant event which occurs at Susquehanna in

A accordance with established procedure^.^^

In section 4.2 of EPM-QA-122 it describes your responsibilities as an originator of an EDR.

Initiate the EDR form, making an initial assessment of whether the potential discrepancy is a Safety Concern or impacts a Technical Specification Action Statement (TSAS), determining if a SOOR is warranted, and notifying his or her Supervisor of the potential engineering discrepancy.

Provide a clear, concise statement of the discrepancy by describing both the requirement and the existing condition so that the difference is clear. The discrepancy is to be described in a concise manner such that it may be understood by an individual who is not intimately familiar with the task, special process, item, etc., which constitutes or is associated with the cited discrepant condition. The description shall provide for direct reference back to the material, equipment, systems, activities or services associated with the discrepant condition.

Inherent in the responsibility of the originator is the obligation to review any potential adverse condition with his supervisor, or the EDMG. If possible this should be done before '

initiating an EDR. This permits information to be exchanged when it is most current in the originators mind and is an aid to those who need to resolve the issue. It should not be a cause of delay in preparing the EDR. The originakor is in the best position to fully describe the condition, and thus the organization can benefit from the research already performed and allow for more timely evaluations of significance. The originator will be contacted by EDMG and requested to participate in EDR evaluations to ensure the concern is adequately addressed.

Once the originator completes the form it is brought to their supervisor. The supervisor will, within one day, determine the validity of the EDR (process is called validation) and document the basis of the determination on an EDR Continuation Sheet. The supervisor will then perform the following:

Explain his/her determination to the Originator and assure the individual fully understands the basis for the determination.

~egcbly sign, date and print full name.

Transmit to NE-Engineering Discrepancy Management Supervisor (James E. Agnew). The presumption is Validity** until proven otherwise.

The Engineering Discrepancy Group is responsible to:

Verify the accuracy of the validation process (called Verification) .

Accept and document the EDR.

Track, screen, perform initial and subsequent operability/ reportability determination with Licensing, SSES.

Compliance and plant support, prioritize, evaluate the EDR.

Assure assignment and acceptance of that assignment, of evaluation and resolution.

Prepare SOORs when appropriate and notify the Originator of the results of validation and resolution.

Review and report status of outstanding EDRs commensurate with their safety significance and age.

Monitor the accuracy of the Engineering Discrepancy List and Database.

Drive the implementation and close out of EDRs.

Transfer EDR records to SRMS.

The process has recently been revised to: Provide an appeal path whenever the originator disagrees with the results of the validation evaluation, the operability evaluation or the reportability evaluation.

Each originator and supervisor is expected to utilize this appeal process. Whenever the appeal process is utilized, I expect the supervisor to assist and support the originator in making the appeal.

Let me emphasize my commitment to this program. It is important for all personnel to recognize the use of this procedure (EPM-QA- 122, Rev. 3) is not optional. Our values demand we utilize procedures because they represent the consensus best way to do the job. Please reinforce with your personnel this procedure is expected to be used in identifying discrepancies.

Attachment 24

Letter from David A. George T. Jones, Reportability and G2002OU, October 14,

Lochbaum and Donald C. Prevatte t o "Disagreement with Screening,

Operability Evaluations for EDR 1992

Note: This letter transmits the authors' point by point rebuttal of the technical reasons formulated by PP&L in determining that the concerns in EDR 620020 had minimal safety significance.

Mr. George T. Jones Pennsylvania Power & Light Company Two North Ninth Street, A6-2 Allentown, PA 18101

October 14, 1992

SUBJECT: Disagreement with Bcreening, Reportability and Operability Evaluations for EDR G Z O O Z O

Dear Mr. Jones:

In the meeting on October 9, 1992 between yourself, Mr. Glenn Miller, and Mr. C. A. Myers of PP&L and the signatories to this letter, we provided PP&L with several technical problems with the screening, reportability and operability evaluations completed by PP&L for EDR G20020. This purpose of this letter is to formally transmit our comments on these evaluations and to clearly st'ate that we consider the technical justifications offered in these documents to be inadequate with respect to the items outlined in our letter dated October 9, 1992 to you.

PP&L1s position to date has been that restoration of offsite power can be accomplished in time to permit operator actions to provide adequate cooling to the spent fuel pools and prevent fuel pool b o i l i n g . EDR G20020, its supplemental information and Mr. Kevin Brinckman's report all clearly identify existing SSES design conditions in which normal fuel pool. cooling is lost and when manual valves in the reactor building used to initiate ESW makeup or RHR fuel pool cooling assist are inaccessible. Loss of offsite power is only one of these cases. PP&L has not yet shown that the NRC has reviewed and approved a duration of 520 hours for the loss of offsite power design event. By comparison, PP&L defined the design basis LOCA/LOOP event to be the LOCA with a concurrent LOOP and consistently applied this design assumption in calculations, reports, FSAR discussions, and licensing correspondence.

PP&L1s position also relies upon operator actions to prevent fuel pool boiling if normal fuel pool cooling is lost. While alternate fuel pool cooling methods may be utilized under certain conditions, PP&L has not yet shown that the NRC has reviewed and approved methods to cover all of the operating and postulated accident conditions required within the SSES design basis. The licensing basis for SSES in the event of loss of fuel pool cooling as described in the SSES FSAR and in the NRC's SER is to permit fuel pool boiling and use ESW makeup to maintain water level. As detailed in EDR G20020 and in Mr. Brinckman's report, the boiling spent fuel pool condition represents an unanalyzed state with potential severe adverse consequences.

In addition, the use of RHR fuel pool cooling assist in case normal spent fuel pool cooling is lost is not described in the SSES FSAR or NRC1s SER and could adversely affect core and containment cooling following an accident. The boiling spent fuel pool condition is clearly not adequately analyzed in the SSES design.

Therefore, we consider the concerns identified in EDR G20020 to warrant a 88Considerablet8 safety significance determination and to be reportable under 1 0 CFR 5 0 . 7 2 . In addition, the operability of SSES is presently adversely affected by the problems. We are confident that a justification for continued operation could be written to permit SSES operation until necessary modifications are implemented.

We remain hopeful that the ongoing reviews by the SRC and Nuclear Licensing will result in a proper resolution to EDR G20020. We are available to respond to any questions regarding our position on this issue.

Thank you for your continued personal attention to this matter..

Sincerely,

0~&W David A.dLochbaum

Attachment

\%sg\k&& Donald C. Prevatte

Distribution: C. A. Myers @/a) A2 -4 G. D. Miller (w/a) A6-3 J. E. Agnew W a ) ~ 6 - 3

October 13, 1992 Attachment

Document/Section Comment

General

General

General

The screening, reportability and operability evaluations lack references to cited information which makes verification difficult. Considering the level of detail that went into documenting the concerns in EDR 620020, it is inappropriate to justify determinations on nuclear safety issues with vague, uncited references.

The screening, reportability and operability evaluations place undue emphasis on the fuel pool cooling system design. Eight of the concerns identified in EDR G20020 had nothing to do with the design requirements of the fuel pool cooling system. PP&L has repeatedly attempted to divert attention from the real problems identified in EDR G20020 by focusing solely on the fuel pool cooling system design.

The screening, reportability and operability evaluations rely primarily upon timely restoration of offsite power and operator actions to prevent fuel pool boiling. The basis for limiting the duration of the LOOP has never been reviewed and accepted by the NRC. Operator actions cannot be performed for the postulated DBA LOCA with the radiation levels reported in SSES FSAR Chapter 18, which do not include airborne contributions. If airborne activity is considered, the conditions are significantly worse.

Screening/Page 2 The characterization of the concerns in EDR G20020 is too simplistic. Rather than a 'lack of suitable documentation1, EDR 620020 reported several cases in which existing documentation was non-conservative.

Screening/Page 2 The five questions in the screening are not addressed for each of the nine concerns raised in EDR G20020. The five questions in the screening are not even addressed for the most severe or most limiting of the nine concerns raised in EDR G20020.

Screening/Page 2 Although no selection is specified, it is assumed that PP&L1s answer to Item I is l'NO'l.

Screening/Page 2 Last paragraph states that the most common failure mode for a complete loss of spent fuel pool cooling is a loss of offsite power. SSES FSAR Appendix 9B analyzed a complete loss of spent fuel pool cooling due to a seismic event. What is the basis for a LOOP being the 'most

Page 1


common failure mode?' In addition, even if the LOOP case is the most common failure mode, all other failure modes must be covered in the design.

Screening/Page 2 Last paragraph states that "the estimated time to restore offsite power ranges from 15 minutes to 20 hours...". No basis is provided for this estimate. In reality, numerous examples can be cited of LOOPS that have lasted longer than 20 hours, most recently Turkey Point as a result of Hurricane Andrew. Additionally, three credible causes of LOOP which are required by Federal regulations to be designed for can cause LOOPS which could last longer than 20 hours. They are tornado, earthquake and sabotage.

Screening/Page 2 Last paragraph specifies a 25 hour time to boil for the spent fuel pool. EDRs GO0005 and G20020 and Kevin Brinckman's report (PLI-72288) all challenge the validity of the 25 hour time.

Screening/Page 2 Last paragraph concludes that if offsite power is restored in a timely manner, then spent fuel pool boiling will not occur. This response does not address any of the other failure modes for the non-safety related, non-seismically designed fuel pool cooling system. A seismic event, random failure of the non-safety related equipment in the system, or common mode failure of the non- safety related equipment in the system has the potential to incapacitate the fuel pool cooling systems for longer than the time required for the pools to achieve boiling. In addition, current SSES emergency procedures following a LOCA require the operators to de-energize the non-1E reactor building electrical loads (which includes the fuel pool cooling equipment) if the reactor building temperatures are as currently analyzed.

Screening/page 3 First paragraph misrepresents the concerns of EDR G20020. The severe core damage which renders the reactor building inaccessible is required by Federal regulations to be included in the design basis as documented in SSES FSAR Chapter 18. With the reactor building inaccessible for days after a LOCA, spent fuel pool boiling will occur following a loss of fuel pool cooling caused by a LOOP, a non-1E reactor building load shed required under emergency procedures, a seismic event, or a failure due to the consequences of the LOCA itself (PLI-72288). This discrepancy can then adversely affect cooling of fuel in the core and in the spent fuel pools as well as secondary containment integrity.

Page 2


Screening/Page 3 Although no selection is specified, it is assumed that PP&Lgs answer to Item I1 is "NOgg. The correct answer is ggYESgg.

Screening/Page 3 The response to Item I1 is technically incorrect. The discrepancy has the potential for causing the common mode failures of ECCS equipment, the standby gas treatment system, and other safety related equipment required to mitigate the consequences of design basis accidents.

Screening/Page 3 The last paragraph in Item I1 misrepresents the concerns of EDR G20020. EDR G20020 does not state that fuel pool boiling will cause fuel damage in the core. However, EDR 620020 does point out that for the postulated design basis LOCA with the core damage conditions documented in SSES FSAR Chapter 18, the operator actions required to align either ESW makeup or RHR fuel pool cooling assist to the fuel pools will not be possible and damage to the fuel in the fuel pools may result. Kevin Brinckmangs report (PLI-72288) supports the position in EDR 620020.

Screening/Page 3 The response to Item I11 is overly restrictive in that it only addresses the fuel pool cooling system. With the exception of tenperature anc? level instrumentation in the fuel pools, EDR G20020 does not question or challenge the design and operation of the fuel pool cooling system. The concerns in EDR G20020 are that the boiling spent fuel pool is an inadequately analyzed event with the real potential for causing the failure of every safety related system in the reactor building. These systems are explicitly listed in the Technical Specifications. Therefore, the answer to Item I11 should be ggYESgg.

Screening/Page 4 The response to Item IV does not address the concerns identified in EDR G20020. With the exception of temperature and level instrumentation in the fuel pools, EDR G20020 does not question or challenge the design and operation of the fuel pool cooling system. EDR G20020 specifically questions the ability of the ESW system to provide makeup flow to the boiling spent fuel pool and of other safety related systems in the reactor building to operate in the conditions resulting from a boiling spent fuel pool. Therefore, the answer to Item IV should be I1YESw.

Screening/Page 4 It should be noted that EDR GO0005 was written in 1990 and has yet to be resolved. A timely and complete resolution to EDR GO0005 would have

Page 3


removed some of the 'complexityo of the issues surrounding EDR G20020.

Screening/Page 4 The response to Item V is incorrect. The SSES FSAR and the NRC's SER do not discuss the use of RHR fuel pool cooling assist to cope with a loss of fuel pool cooling. The SSES FSAR and the NRC's SER only discuss the use of RHR fuel pool cooling assist to handle the condition of a fuel core offload during a refueling outage. As stated in Kevin Brinckman's report (PLI-72288), the use of RHR fuel pool cooling assist for fuel pool cooling following a LOCA is an unanalyzed event. The SSES FSAR and the NRC's SER state that the design provision for the loss of spent fuel pool cooling event is for the ESW system to provide adequate makeup to the boiling fuel pools. EDR G20020 identifies concerns that such makeup may not be available, and if made available may adversely affect performance of other safety related equipment. Therefore, the answer to Item V should be ooYESot.

Screening/Page 5 The response to Item VI is based on operator actions to prevent fuel pool boiling. The response states Init is not reasonable to assume that the operators will take no corrective action and allow the pool to boi1.I' The regulatory basis for designing to satisfy postulated accidents is to provide assurance that actual plant responses are bounded. EDR G20020 does not state or imply that every loss of spent fuel pool cooling event must result in spent fuel pool boiling. However, the only safety related design provision to cope with a loss of spent fuel pool cooling at SSES is use of the ESW system to provide makeup to the boiling fuel pools. The other non-safety related methods of cooling the spent fuel pool have not been fully analyzed for use under all required conditions. Therefore, the concerns identified in EDR G20020 represent at least 'moderate' safety significance if not 'considerable' safety significance.

Reportability/Pg 2 The last paragraph of Section I1 states that each of the nine discrepancies in EDR G20020 are addressed separately and in greater detail in the "EDR Evaluationv. If this document is the screening document, the nine concerns were not addressed separately in Revision 1 of the screening document and we did not see Revision 0 of the screening document. If this document is another document, we have not seen it.

Page 4


Reportability/Pg 3 The response to Item I11 is misleading and technically inaccurate. EDR G20020 raised questions concerning the radiological release analysis for a boiling spent fuel pool and identified several non-conservatisms in this analysis. Therefore, the 10 CFR 100 limits are challenged.

Reportability/Pg 3 The response to Item I11 states that the spent fuel pool cooling system has "two (2) safety grade independent sources of water systems for make up and cooling, the ESW and RHR systems." The RHR fuel pool cooling assist mode is non- safety related, non-single failure proof. As detailed in Kevin Brinckman's report (PLI-72288), use of RHR fuel pool cooling assist following a MCA is an unanalyzed condition with the potential for adversely affecting core and containment cooling. As detailed in EDR G20020, use of ESW makeup to a boiling spent fuel pool is an inadequately analyzed condition with the potential for adversely affecting performance of safety related equipment in the reactor building.

Reportability/Pg 4 The response to Item I11 describes an alternate cooling method using ESW supply to the fuel pool with draindorqn of water to the refueling water storage tank. The recent ESW flow balance concluded that sufficient flow was available to makeup to compensate for boiloff, but did not indicate the flow margin necessary to maintain the fuel pool below boiling. Additionally, there is no evaluation indicating the RWST can handle the heat rejected by this means. It is highly inappropriate to justify away concerns about an unanalyzed condition by relying upon another equally unanalyzed condition.

Reportability/Pg 4 The response to Item I11 states that the fuel pool instrumentation are expected to be available during a loss of offsite power. As detailed in Kevin Brinckman's report, this instrumentation would probably not be available following a LQCA. This instrumentation is non-lE, non-safety related and non-seismically designed.

Reportability/Pg 4 The response to Item IV is technically inaccurate. The response states that the "analyzed design basis accident (DBA) is a LQCA with a concurrent loss of offsite power." Since either a M O P or the consequences of the LQCA can produce a loss of spent fuel pool cooling, the combined effect of a DBA MCA and a loss of spent fuel pool cooling must be analyzed. Hence, EDR G20020 was written.

Page 5


Reportability/Pg 5 In the response to Item IV, the use of RHR fuel pool cooling assist is discussed. As detailed in Kevin Brinckmangs report (PLI-72288), the use of this RHR mode following a U)CA is an unanalyzed conditions with potential adverse impact on core and containment cooling functions. Reliance upon this unanalyzed RHR function is inappropriate.

Reportability/Pg 5 In the response to Item IV, it is stated that the post-MCA radiation levels reported in SSES FSAR Chapter 18 are for EQ requirements. However, these Chapter 18 requirements are for DBA MCA conditions covering personnel access. The ESW and RHR system manual valves in the reactor buildings on both units would be inaccessible post-DBA MCA due to the contained radiation dose which is reflected in SSES FSAR Chapter 18 and the airborne contribution which is not addressed in SSES FSAR Chapter 18.

Reportability/Pg 6 The paragraph at the top of Page 6 stated that "an analysis has been performed that concludes the equipment can withstand the temperature effects of a loss of fuel pool cooling." However, no such valid analysis has ever been found. Some of the existing analyses of reactor building temperatures following z. L.OCA do assume a fuel pool temperature of 212'F and account for the sensible heat from a boiling pool, but none of these calculations account for the significant latent heat transferred from a boiling fuel pool. In addition, there are no analyses which show the safety related equipment in the reactor building can withstand effects such as humidity, flooding, condensation, etc from a boiling spent fuel pool.

Reportability/Pg 6 The third paragraph on page 6 is misleading. As stated above, the existing calculations at best only considered the sensible heat from a boiling spent fuel pool. The neglected latent heat represents almost 5 times the calculated reactor building heat load and would clearly adversely impact reactor building room temperatures.

Reportability/Pg 7 The response to Item VII is inaccurate and misleading. The SSES plant design for the boiling spent fuel pool condition does not meet all the Federal regulations if systems beyond the fuel pool cooling system are considered. In addition, design evolutions at SSES since the initial design, such as for high density spent fuel storage racks, 9x9 fuel, and the non-1E shedding of reactor building loads, should have been opportunities to detect and correct the deficiencies identified in EDR G20020.

Page 6


Reportability/Pg 7 In the response to Item VII, the Emergency Plan is relied upon heavily. Heroic operator actions are appropriate to mitigate events outside the plant's design basis such as ATWS. The post-U)CA scenarios are analyzed and documented, but they are deficient as detailed in EDR 620020. It is inappropriate to rely on operator actions to fulfill requirements imposed on the plant design.

Reportability/Pg 7 In the response to Item IIX (sic VIII), it is stated that l1as part of the design requirements the equipment located in the secondary containment must be able to withstand the effects of a loss of fuel pool cooling.11 As stated previously, EDR G20020 was written to report several conditions in which these design requirements are not satisfied. Once again, operator actions cannot be relied upon to satisfy the design requirements under postulated DBA LOCA conditions.

Reportability/Pg 9 EDR G20020 is determined not to be reportable based primarily upon reliance on operator action and restoration of offsite power within the time to boil. Sufficient technical justification has not been provided which would support this determination for the DBA LOCA with the radiation levels reported in SSES FSAR Chapter 18.

Operability The comments for the screening and reportability evaluations apply to the operability evaluation as well.

Page 7

Attachment 25

Memo from Charles A. Myers t o George T . Jones, "Fuel Pool Cooling Issues - Reportability 1 Operability", October 20, 1992

To: George T. Jo~cr Ad-2 Copy: Glenn Millet A6-3

Rocky Sgarro A2-4

J h urn^ 112-4

From: Chatles AMPS A 2 4

Subject: Fuel Pool Cooling Irmsr - ReportabfU.tylOpanM1ity Per your request, I have rwfmoal the subject matter. I did not idantify any sfngle matter hat, at this time with tha informaticm available, appmed to micat the requimmts for rapor t i to NRC, Sonis of issues are. however, of &y significance; they should be gurmed more expeditidy than has beem tha case d they h M be formally b r o e t to NRC's attention.

The scope of my review included review of donrmantation associated with the subject and dilrcusslonswkhmystaff(RackySgatfoandDanM~). ldbinotperfomaayin-depth intarview but did obtain some additional data and input from some of the people involved. Since 1 mted the effart to bve a degtee of h d e g d w e from the work done by your staff, I did not discus their assessment with them nor rely an the lo@ they had used in their dstwminations. Rocky S p r m has rwialrred thelf dstsrmlnationo in detail and will provide feedback directly to Jh Alprsv. My focus was on whether tha mattees lnvold were reportable or net and thek impact on the operability of plant sy.8tsms.

In detm&@ Mpartabiiity. I have bused my review on the r ~ r ~ t s contained in iaCFRS0.72 md .73 andon 1OCFR50.9 (references 1-11.

In doiw thls review, I have not given specific consideration to the impacts of the Pawr Uprate projeut since we have not yet received NRC approval. I did conelude that whlle ths project would haw some impact on the numbers produced by various analyses. it was unlikely that ths change5 add change the basic course of my kip basis evants. That is the impact a m to be one of sligbtly redudng tima margins or small fncnms in dosa conssplsoces on the order of 5%.

Orsnisw EDR G2Mm3 end supPorune documentation give a geneidly accurate idautification of the issues. I've grouped thass issuss into 5 general concerns as dssctibed below (pmnthetiMI1 refer- anr to tha specific paragraphs in seftion 9 of the EDRJ.

1. fhe design bash for tha ptant and the FSAR have not been sufficiently updated to reflect changa In fuel design and operatian, particularly in tegards to the Spent Fusl Pool Cooling system design analysis (Reference 11) and tb~ Spent Fuel Pool events such m doscribed in FSAR Appendix 9A (Reference 121. There m y ba addltW c~nsapwnces bemuse of these fuel deslgn and operational chirnges that haam not ban adqmtely analyzed. Tb phipal c h p ~ in the fuel desm and op~ation include incraad burnup, chm$ng from a 12 month to an 18 monthM cycle, and wing 9x9 fuel desip (EDR Ssction 9E, 9F, 9G and 9H)

2. The Wgn basis for the plant snd the FSAR have not been updated to reflect the c h a m in haw the Spent Fuel Pool k opsiatsd dur@ outages. There may be additional Pailure modes and/or additional c q m w , because of thess chmgs, that have not ban &quataly analyzed. The prindpal c h g ~ are routinely offloading the entire CON m h outage, intertleiag ths two fuel pools (using ons fUel pool cooling system fat both d u r i i servle mter outages), and using ths RHR systsm in cl*cay hmt removal mode RS an assist to fuel pool eoollng to handla pool heat lads p t e r than the fuel pool cooling system dm@. [EDR Sectlon 9E, 9F, 9G. 9H, and 91)

3. Tha deign be& of thd plrrnt and the FSAR (Par normal opsr~tlon) do not address the effect d Spmrt Fusl Pool boiling on ths equipment, prhcipdly the Standby Gas Tceatmant System as ibtiPLed in NRC Rdatory Qulda 1.29, that ml& bs u4ad to reduce ths eowequacw of loss of Spent Fuel Pwl cool@ sptgns. [EDR Won 9B).

4. Ths design bash of the plant and the FSAR (for normal opmtion) do not eddrass the effect of Spent Fuel Pool boiling on tbn equipment d to &eve safe shutdown, pnrticularly in regard to moisture emtent in the alr, hmdtw af condensutlon (including flooding ptotsetion), and the additional heat load on ths secondary cartejnmnat ventilation systems. (EDR Section 9 4 9B)

6. The deslgn basis of the plant and tba FSAR address neither the ef!Fect OF Spent Fuel Pool boiling on equipant nwdad to m i w t e the deslg~ beds LOCAILOOP event (awne hwards as above) nor measures to restom coolin# to prevent the Spent Fuel Pool from boiling during the LOCAILOOP. (The Spent Fuel Pool Cmliig Systems b one of the loads shed durhg the event.) (EDR Section 9A, 9B, 9C, 9D)

Discussion

In my review. I carefully considered each of the 9 Items (EDR G2Ml20, items 9A through 91). For easa of handling, I will adrlrase then in ths 5 grormps identified above.

As indicated in the EDR item 9G and 9H a m b e t of c h g a in tha d e s i i of the fuel or haw it operates have been madn sinca the tables associated with BAR Section 9.1.3 and FSAR Appendix 9A m e prapared. These include a tlmnga to 9x9 h l , him h p , a longer fUal cycle, arnd fixing of the outsges in spring arrd fall. Clearly these FSAR portions repuke updating to a c c e t s thsss clmngs. My review, howwat, indicated tht nona of these changes haw a substantial impact on the sedety of tho spent fuel. The principd isotope of coneern is 1-131. This isotope builds ta an aquilibriurn within a few waaks of the start of a cycle 4 the inmtopy in thc spnt fuel is relatively insensitive to burnup, cycle leagth, ot the size of tho fuel pellet. Thm m y be soma d l changes in ptoductlan rats due to additional Plutonium Rsslon with lncmased burnup, or scrme chsmgas in the Ueoretlcd rel- of fission products from leaking fusl rods (the esap rate &cient) but, here too, 1 expect the Impact to be smatl. fhaa is good reason to bslive that the existing analysis rasults are still tmndb#. Longer refireline cycles lead to larger discharger of f'wl outage; thls is uldnssed below with 0th fuel pool operational issues. The reload applicntim mads Mch cycle appeat to address properly the other lsswws (i.e. issues other thaa the impact an ths s p n t fuel pool). --WethespsdOc MchaFacterisucsa~notthesamsag ahat isaddtesssd in the FSAR in rsgrrd to spent bl dotage, the results of thc FSAR anrlyias are judged to remain accumte md the &tion L ju@d to n&Qniffcurtly canpromi= plant safety.

Item 9E, 9F, 9G and 9H discuss c k q s in how the spout Rae1 is opefated that have not been ddtesscd in FSAR Section 9.1.3 and Appendix QA. b chugs ineluda morc fuel assemblies disdwgcd per fuel load bacause of the l q a r cycle, MI offload of the cofs each

outago, mtlne operation in wtagw with tha fual paols intattimi with OM Fuef Pool Cooling System in service, and E o u t i opemtion in o u h p with RHR in b a y hcrrt ramoval mode to assist the Fuel Pool Cool i i System during full offload.

My mvlm indicattd that tha normal practica of fully offloading the core each outage d t s in a condition mote severe than the basis for tho "Maxirnum N m d Heat Load" case addrrrsseul in FSAR 9.1.3 and more severe then the initid conditions assumed in the E A R Appendix 9A. Using both actual numbers lor the m m t outage (see Referem 8) and review of the opMahicmal changes, I estimate that the actual heat load is about 4 times that used in the FSAR analysis for these two situations. This increasa in heat load inctcafes the total heat load given in FSAR table 9.1-Za a d -2b by increasQ the M load dua to the most recently discharged fUel (the amount of heat due ,

to previously discharged wemblies i s low and i s relatively unaffected by the concerns Identified in this EDR.) These tables an, provided for information, only.

My review indicated that we cwmtly do not e x c d the 'Emcylency Heat L d eonditians identified in FSAR Table 9.1-Zc end -26 but am only a d l amount lower.

' FSAR Appendix 9A nnalyzea rrdiological eonssquences of the loss of fuel pool cooling during normal opemtion. T h o a p p r o x i d y four fold incrsrrss in hsat load due to full core of f ld would lead to an incmsa m evaporation rate with a cw154puBntial increase m the rdeass of 1-131 from the boiling pool. The EAR analyses provides zr pammtric study but the largost calculated dose is 0.046 REM (Thyroid Dose at the LPZ for 30 days), The NRC in their SER POP SSES (Refemme 9) identines ths limit t h y us8 as 1.5 REAII. Using a faetor of four increase in releases would lead to a totat dose of abut, 0.4 REM tPhidr is still a small fraction of the value ussd by NRC in the SER as a yardstick.

CmcUm - The changes in operation of ths fuel pool, principally poutinsly offldlng the core, are not analyzed in ths FSAR and do lead to an increase in radiological -. The magnitude of the ralsase is still small, howem?, and is still d l blow NRC's guidelines as identified in our SER. Under tbe current rules this does not appear to bs a conditian that signiflcaurtly caqtmisar plant safety.

The various modes of operation of the M p I s and the mvitieo ara not idantified in the FSAR. One analysis has btm submitted to ths NRC ( R e f e m 10) which pravldes us with some un- of the results we might get w h the analysis is compkt4. Using the valw identihl for the analysis in Safety Evaluation NL-86-W5 (sttackid to Reference lo), my raviw indicatedthat either the configuration 1 pool and 1 cavity or 2 fuel pools (at the appcqtirte t h e post shutdown, would probably result in a time to boil not significantly dlffemt from the 25 hours d in the Appendix 9A anal@s, and therehe the mode of aparatiari would not w e tha d y s i s substMtidly.

!&JI&&A - The changes in mode of o p t i o n of the f k l paols (other than tha cmplete core offload stmtagy d i d earlier) probably does not substantially affect the results of tha FSAR analyses. Tbem is a need to do tha appMpriate analyses of the nrrrsnt o p a r a t i i modes. %cb a review mi& uncow? additional failure modes not addtsssed in ths FSAR ot mi@ give different results b d on the modeling of movement of water betaem the pools. This anal*$ will also verify the adequacy of admMstrative and tests currently done to ensure ths safaty of the spent hel. Based on tha Informatian available at this time, I canchde that there is no basis for conccmcluding that this condition significantly GanpMmises safety.

The conditions that make up ths tvo items above do not haw cxmqmms that exceed current reportability standards, howver, the existence of t h e errors over a long period of t i with many people knowin# of the change$ that had taken place constitute a cmem larger than might exist lf a singis oversight had gone undiscovered. While ths provisions of 10CFR50.9[b) do not appear to apply because of tha lack of a "sipificant implication for public health d dety or coanmon defense and security", it may b prudent to F d y not@ the NRC of both the condition and our co t r ec t i~ action so that they cpn cdrm our jtillpnt of signlficame.

Fuel P- NRC in their SER (ahlch refers to Re@datory Guide 1.29 and 1.52) identlf'ies ths nand for the SGB to msst certain crlterla w a basis for approving the use d ~ - ~ s m i c Class I fuel pool coo@ systems. While no credit appears to havc bsen tahen in E A R Appendix 4A for the W i t s of filtration through SGTS, it is reasonable to assme thrxt it was NRC's intent to have SGTS available for mitigsrtion of releases from tb fusf p l . The SGTS is a~plrrently desipd far tha environment (using prduatii to rsduw humidity). The D R indicates that ths impact, of condmsatiwr buildup on the strudunl intsgrity of the ductwork has not hen properly analyzed. It is not clear 'prtrsther tho duct& vould or vould not fail, d if it failed, it is not clear that it would prevent SOrS from functimb# for thls event.

Ekedonthmeuncartainties,Icannot make a J u d p m t i n ~ d t o t b n impact onplant safety.

S~WQQ - The requlted &pis should be completed expeditiously (or an applicable e x i w and+ found) and the reportability of this itarn be re-assf~smi. It appears, from the evidence that I reviawed, that this matbr is reportable if it b wncluded that the condensate buildup prevents SGTS from performYng its function.

The issue in ths previous three questions involved h g u s in the design or operation of the plant or the @acy of an andysis specifidy done to rtssun aaceptability. The remaining two qurstim involve whether or not the phenomena associated with the fuel pool boiling even need to be considered as a hazard to other safety equipmnt during normal operation and accidents. PP&L rnay wll decide that these conditions are ones that thsy believe need to be addressed to meet Its standards, but the reportability question is whether or not the matter is wWn our licsnsing hie ms. The recard is not perfectly clear.

W e em find no evidence in the form of NRC guidance. puestlans to w duriq Beensing, or matters on our docLst that indieatsthat NRC considerd these phenomrma as pat of out licensing basis. Thare is discourse about the seismic capability of ths fuel p l cmling systems, about the methods and lusumptions of lnafysls of the &logid a m s m of the event, and about the design af the SGTS, but not about the impact of boil@ on the other safe shutdown or accident mitl@ion squfpnrsnt.

Historimlly, conditions that might devdop substsntially aftof the initiation of the design W s e m t s ym frsgusJltfy not considered patt of the licmsiog M s . This was because the focus was on fmmdcate actions wd diruct consequences, and the presumption that other matters could be haadled by the emergency staff (m control). Iwmdrqly with timo, NRC has started to address such matters ad- lcqsr term mukaup of cooling water, 30 day analysis of spray ponds, dc. N o ~ e the less most of the conseguanm of deign basis events outside the d e t y related arns are still nat m t h l y , EosmaIIy evaluated durw inltlal l l c e n s ~ activitiw.

There are elements in tha raoordthat do iafsa soms questions:

D k i i the llcenslag of SSES, the WW-2 plant was dealing with a caneern for the impact of fuel pool boiling cdansata fbodi i on ECCS equipment; after substantial time, they uppded thar fual pool cooling system (Refefdrance 13). Despite the fact that SSES was not yet l i , I f d on evidence that the WNP-2 issue was raised on SSES.

Out Rsackrx Buildiq WAC anatysis in the late 1980's (idsatid as the COlTAP d p l ~ ia the LDR) toot into account thn 'sensible W k m a fuel pwl at 212T but not the 'latent heata. It is my understanding that this was dona for extra canservatism. Note that the model used at the time apparsntly could not address the effects of moisture POW to the alr Prom the pool, so such conservatism (in the form of highwheat input) might ham, therefara, been appmptiats.

Repmtedly, earlier (1970's) heat load eateulations have baa purported to include from a bollfng pool. (I haM not h p c b d the prBnant documents

myself.

Conclualrm - While them are some Factors that raise doubt, the preponderance of the evidance says that the effect of the boiling envlrenment on both norm1 and post accident environments va not considered to be a Mpuirement for the I i h g of SSES. A coonpliwt@ factor that gives me substmial carrann is that the effect of intersst falls in tha a t ep ry of consspurmtid failures, that is, the is a direct result of evaluating the deterrninistlc progress of the event and not due to assumption of arbitrary failures. 1 therePore have reviewed gemally what tb c v might bs for loss of fuel pool c p o l i during normal operation.

For this situation, tha Raactar Bullding is sufficiently accessible so that EW car^ be aligned to provide the nsedsd makeup and initially the pool temperature and level can be determined by direct observatim. Wysis is required to varifir that sufficient imtmnmtation is available or a safe strategy is availabla to maintain mteF low1 is available without instmntatian alter the Go1 temperature lncreass ptscludss direct obserrntion. It appears that a mce&l str~tegy is possible without mJm ciwmp. Duting this event, RB HVAC Zone 3 (refueling flwr) should remala isolated frm tha rsmainder of the Reactor Building (or be capable of being isolated). Thi~ should pmvsnt the hot, moist envlronmant from W l y affecti i most ECCS equipment. Cdenwtion In tbs zm, however, will en* the drain systems, and might, therefore, be able to impad the ECCS systems due to flwdiryl or s&y evaporation and heating due t o hot water in the sump. NOM tbs lsss, giwn the tima involved, it appears feasible to tab damage contiof step to a c t o d a t e the impcts.

Conclusian - Whlle substantial lnatysis is required, and procedures mi@ need to be developed, I do not currently believe that this d t u t e s a matter that significantly compromises plant safety.

The introductory pamgn@j in the previous s e c t b discuss the imp& of this mMm on the lhs ln# b d s of the plant. Ths limiting event appglrs to be ths LOCAJLOOP bacause in thtseventbothfu$lpod~systems nrerwrolndfromse~thrwrghtoldshaddlng. Initial review intbtes that, practically, offsits power L rquind (for ths service water system) to retam Pus1 pool cooling. RHR is fully oocupied with cooli i the cores &or suppression pools on both unlts.

An a d d i t i d concern exists in regsrd to the effect of both direct radiation and radiation in the Mldina atmwphm on operatian of essential equipment. Specifically, if substantial amounts of radiation are ralsased to certain plant system or to the building atmwphare, damage c d r o l actions will be limited and access to the ESW to Laitlate makeup to the pool may by precluded or extmmely diicult. The safety signficance of this condition is therefore depadsat on the release of fission products during the event. Us of more realistic analyses may be appmpriats in handling matters like this that are outside thn bounds considered in orlginal l iming. The FSAR Chapter 15 provides such a "realisticu analysis. Recent GE worlt (being used in our Power Upmte apptieatirml indicates that the potential for damage from a LOCA (and LOCAILOOP) is rrmch lover than orighdly estimated (due to lower peak clad tsmpsratures) so the FSAR 'realistic" analysis m y be overly c-tlve. None the less, the informatian available does net provide a basis for cancludin~ wh.dw E W is sufficiently accessible or what access is available for damage control eflorts.

Cadusions and Remnmdathrw: . Wi ths exception of EDR 20MO item 91, the umcrms about the analysis and phenomena appear to bs valid.

2. Tbn matters idantifled in EDR 20020 items 9E, 9F. 9G and QH inmlve c h p s in the design and o p t i o n of the plant tbat ham not been properly mflseted in the PSAR. They must ba prqmrly analyzed (fncluding iOCFR50.59 safety eMhwtions w h not cutruntly avallabla or complete) and the FSAR uplatad. My review indicated that ths calculated iediological conssqusnws (FSAR Appendix 9N would be imeased but still wWn the limits identified by NRC in our SER. As such. the condition doss not appear to meat exWi criteria Por reportability.

3. Not Pithstrmdlng the information in 2 above, I bellevu that it is pFudant to offldally and formally inform NRC of Ws matter bacause it appears to to associated with a program weakness tht allowed to occur that we not ptoprly documented in the FSAR. 4. Given my cwent that Nuclew h#imviq hi dst4llmlned that the cgsideratiim QP hl pool boiling impacts on safety s y & m (othcr than SGTS and the b h l i n g f'loor enclosure) ls not patt of the desfen basis of the plant and my review indicating that it wes not part of the lkming basis of the plant, I do not find a baris for tho matter to be reporteble. Note that this mattar appears typical of rhat 0 t h han faced during design bask reconstitution efforts and that we will face in the fUurs an the DBD project.

5. Thu above (piqppb 4) d m not mean that ths matter may not be si#nificant b safety. k e d cm the damage occurria from the cansarmenng of the fuel pool boiling event or ths LOCAlLOOP ewnt and considering ths difficulties that the operators might, in fad, experience tryiq to c o p vlth thosa canseguerres, I believe the matter merits prompt attention. Those snrlyses may well uncover c c a s e ~ o c s that wiU be determined to be reportable so continual review of reportability i s reguired

6. Ths impact of heat and molstum (from the spent htel pool duriq loss of fuel pool ccmw on the SGTS should be evaluated immediately; if the SGTS is not det- ta be capeible of prforrming ik? functlans, 1 Miwe that this condition would bs reportable as an unanalyzd condition that significantly compromises plant safety.

7. I sugesst use of the voluntary LER approach to address the rmumicatlon r a m in paragraph 3 above. If you chess to follow this approach. I rocoramend that a discussion of the concum in regafd t o loss of fuel pool cooling durlrq the WA/LOOP be d i d and the general approach to addressin# tk4 matter be identiffed, ThaM Is Nfffcimt connection between the two matters that addressing only ans arould constitub an incomplete communicatim. NRA is ready to assist you in report preparation ff you wish.

8. During my review, the actual seismic design bash for the spent fuel coaling system was not clear. While it is clearly not Seismic Class I, the portlcm should be capable of maintaining its pressure boundary from at least during an earthquake and preferably Wing tho loading due to the suppression pool dynamics during a LOCA. (Nan-Seismic Catsgory I, W i t y GMup C). 9. The racord does not provide auy evidence that MRC Iw detstminsd or sgpsbd that this matter is e i t h repartable or not ~portabie. Wbila NRC has realmi infarmation Prom PP&L about this matter. they haw not to the best of my imodedga provided any feedback in regard to reporbbility, nor do I expeet tham to Wle we am still processing the Issue.

10.1 would riots that I have substantial sgmpathy for the EDR originators. 1 Wisw that thry have performed us r servica by idmtif'yiq the issues and mPintaining the prsssuru to emu& that the iuatw am addnrssed. I mc@a the fealiq that a matter of safety importarm n u t . as m matter of course. be "rfptabld, The repuimmts for repofbbiUty, homvw, are not written in that way. They are Intendad to emure that oen operates within the bo~mds authorid by onw licslus. I was unable to ht@ a condition that put us outsids our license that mst the reportablility ctitaria. Not w i t b t d l q this, 1 believe vs have csrkrin moral responsibilities to not only address these sbfuty Issuss but to fomally notify NRC of ow cancstn and our actions.

11. I recognize that you have recently re-directed your staff to attack the tschnical issues involved. I believe that was appropriate and that the urguncy should remain.

(All of the f o l l o m references wen, usad in my revlew even thou. not a11 ara cited In the above memorandum. Additional infopmation was loohd at durlng this revlew but did not provide specific Mormatian on the issues at hand different hom fht included in the referenrxs below.)

1. lOCFR50.72(b](ll(~)(A), (B), arrd (CJ

2. lOCFRN.72(b) (2) (i)

7. lOCFR50.9 (b)

8. PL1-72230, A. Dyne1 to T. C. Dalpiaz, dated 21 A@ 1992, 'UZ.RI05 Fuel Pool h a y Heat Evaluation. This rmmordum ldentifla the total U1 and U2 spent M hat load with a Ml core of'fload in tb Unlt 2 Pool. It also idantifleo DmnMsttativa l i i t s that, if observed will provide at least 25 hours to boiling ( E A R value) if tha fuel pool c o o l i i system is lost while it is grotriding tha only cookig to the spent fuel. Tha l i t aka assu?es that teguiml makeup Praon ths ESW is nat greater than 60 gpn (FSAR value).

9. NRC SER for SSES. 10. PLA-2120, H. W. Ksiser to NRC, dated 12 September 1986, 'Propod Amdment 41 To Lloense No. NPF-22, Supplemental InEormation.". Wudcs Safety Evaluation NL-86- 005. This submittal addrsssbs an Emsrgsncy Technical SpsGifimtion C h q Rsquest to t h the s d c d traln of RHR out of service dwiq rm outalpe with the cote off l d d aad the cavity and bath fuel pols c d d . The analysis shoved that vitb tke largsr water voluma amifable tha time to boll was increased even thou& the hsat load wm substmtially higher than FSAR Appendix 9A analysis used.

11. FsAR Section 9.1.3. Spsnt Fuel Pool Cooling and C h u p System, includlug assodated tables.

12. F U R Appendix 9A, Analysis For Non Seismic Spwt Fuel Pool Coaling S y & m

13. SERCH llctmhq inforrnatlon, dated April 1987, entitled Won-Cate#oty I Spent Fuel Pod Coourlg Systrrm (SPCS1' 14 ET-0149, htad 19 Mlrrck 1992, Susquehanua Steam Electric Station Spent Fuel Pool Baiting Isms. Docmeatsthe originat cancerns that -re later c l o d m EDR 20020.

15. EDR 20020, Rev 0, dated 16 April 1992, "Lass of Spent Fwl Pool Coo@ Event Design Discrepancies' 16. ET-0471. dated 22 June 1992, Dave Lochbaum and Don Prevatte to Joe Zola, entitled "flrppl~rmtntel fafdnnatian for EDR G20020 an Boiling Spent Fuel Pool"

17. Not numbeted, dated 26 June 1992, presumsd to be from Dave Lochhum and Don Pravatte, entitled "EDR G2WM Rcfersnces' .la. Report, dated 27 July 1992, by David A. Lochbaurn and Donald C. Prsvatte, entitled "Safety Consequences of a Boiling Spent Fusl Pool at the' Suspushanna Steam Electric Station"

19. PLi-72288, dated 1 saptimber 1992. Kevin W. EWnckmm to George T. Jones, "Review of Fuel Pool Cooling". Contains a study by Kevin and commlssioncd by George, to evaluate the safety cansequences of the boiling condition. The conclusim of the originators of EDR G20@0, Kevin l ihckmn, an4 myself m not substentially diirent in regard to the consequmms should boiling occur.

20. PL1-72367, dated 9 Saptsmher 1992, J. R. Miltenbetgar to k r g e T. Jom, 'Spent Fuel Pool Cooling". IdentlfPss NSAG concerns and actions k i u t e d with spent bl pool cooling. The reneetns iderrtifled closely parallel those identified in the EDR and in my own review. This reference also goes in to tba most detail about spent fbl pool operatins m&.

21. Umumbeeed, dated 9 October 1992, memo from David A. Lochbaum and D d C. Ptevatte to Geosgs T. Jones, "Reportability of Boiling Fuel Pool Carrezm*. This reference raised cwcerns about the h d h q of the subject matter, principally the amount of time involved, but also includes some tedutical discusions pertinent to tbt mattst. 22. ET-0784, dated 7 Octobw 1992, memo Prom J. E. Agnew to J. M. Ksnny. Ttansmits EDR G2OOZO. and the most recent rop0,rtabfUt.y and oper~bility assssmaat from Nuclear Tactmology.

23. Unnumbered, drtd 14 October 1992, memo from David A. Lochbarn and Dwald C. Pfivatte to Gmrp T. Jones, ~~ witb Scnsning, Repoftability and OpmMlity Evrluatim for HIR W W . Identifies spec& problems with Referem 22. 24. Draft d y s l s , dated 9 October 1992, by David G. Kostelnik, d t i e d "Loss of Fuel Pool Cooling Eva&, Evaluation for EDR G20[120". Thts .&ah paper identifies the operating modes currently in usa by SSES; I uoed it to pet some indspendsnt d r m a t i o n in regard to these opemtiqmodss.

Attachment 26

PP&L Memo from Glenn D. Miller to George T . Jones, "Evaluation of EDR 620020 - Spent Fuel Cooling Issue", October 21, 1992 (PLI-72711)

Note: In this memo, PP&L for the first time addresses the concerns in EDR G20020 individually. In doing so, PP&L agrees that seven of the nine concerns are valid, yet determines that the operability of the plant is not affected and it is not reportable.

October 21, 1992


SUSQUEHANNA STEAM ELECTRIC STATION EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-72711 F ILE A45-1A

The attached evaluation o f EDR 620020 i s provided i n response t o your memo PLI-72640. This evaluation was prepared by myself and a team o f engineers working on the act ion plan t o resolve the subject EDR.

I n the course o f reviewing t h i s issue i n de ta i l , we have concluded t h a t seven o f the n ine i d e n t i f i e d discrepancies are not v a l i d def ic ienc ies. This i s explained i n d e t a i l i n the evaluation. The two remaining discrepancies are v a l i d de f i c ienc ies but are not considered safety s i g n i f i c a n t nor reportable.

I bel ieve tha t the technical basis on several o f these issues has been c l a r i f i e d considerably i n the course o f the past week. Therefore, I suggest prov id ing t h i s evaluation t o Nuclear Regulatory A f f a i r s f o r reconsiderat ion o f the repo r tab i l i t y aspects.

We are cont inuing w i th the remaining act ion items as requested. A de ta i l ed engineering design repor t and j u s t i f i c a t i o n f o r i n te r im operation w i l l provide more d e t a i l than contained i n the attached evaluation. We are scheduled t o meet w i t h PORC on Monday October 26, 1992 a t 2:30 pm t o review t h i s issue. We are working w i th Systems Engineering, Operations and NRA-Compl i ance w i th respect t o po ten t ia l compensatory measures.

A x 2 - P . 4 Glenn D. M i l l e r

cc: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W. Simpson - A1-2 D. F. Roth - SSES H. G. Stanley - SSES J. M. Kenny - A2-4 J. S. Stefanko - A9-3 F.G. But ler -A6-3 J. R. Mil tenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records - A6-2 1

October 21, 1992 Page 1

Eva1 ua t ion o f EDR 620020

This document contains an evaluation o f the discrepancies documented i n EDR 620020, "Loss o f Spent Fuel Pool Cooling Event Design Discrepancies." Conclusions o f the author w i th respect t o r e p o r t a b i l i t y o f these concerns and o p e r a b i l i t y impact on SSES are also provided.

Desiqn Basis

The design bases f o r the Fuel Pool Cooling System are found i n FSAR sect ion 9.1. The por t ions o f the design basis relevant t o EDR 620020 are as fo l lows:

1. Maintain the fue l pool water temperature below 125F under "normal maximum heat loads" defined as 12.6 MBtufhr (equivalent t o a t yp i ca l fue l cycle discharge schedule which f i l l s the fue l pool, l a s t quarter core o f f l o a d a t 6.7 days a f t e r shutdown).

2. Maintain fue l pool water temperature a t o r below 125F dur ing the "emergency heat load" condi t ion o f 32.6 MBtufhr (equivalent t o a f u l l core o f f l o a d 10.5 days a f t e r a shutdown fo l lowing a t yp i ca l fue l cyc le discharge schedule which f i l l s the fue l pool) u t i l i z i n g the RHR system (w i th o r wi thout normal fue l pool cooling) f o r fue l pool cooling. This mode o f operation applies "during periods o f higher than MNHL generation i n the fue l pool, eg., s to r ing o f a f u l l core o f i r r ad ia ted fue l sho r t l y a f t e r shutdown". The RHR system i s used under these condi t ions t o ass i s t the FPCCS i n d iss ipa t ing the decay heat. Thus, any heat load i n excess o f 12.6 MBtufhr i s considered t o be w i th in the design basis f o r the RHR FPC ass i s t mode o f operation.

3. Redundant Seismic Category I ESW connections t o each pool are provided t o al low f o r makeup o f evaporative losses i n the event o f f a i l u r e o f the FPC system. The condi t ions are bounded by a fue l pool t ime-to-boi l analysis based on the same typ ica l fue l cycle discharge schedule as i n basis #1 except the time a f t e r shutdown i s 10.5 days instead o f 6.7 days r e s u l t i n g i n a heat load o f 9.8 MBtufhr. (This explains the d i f ference between the two d i f f e r e n t heat loads, ie., 12.6 MBtufhr f o r basis #1 and 9.8 MBtufhr f o r basis #3. This i s not a discrepancy.) The ESW makeup l i n e i s sized on the basis o f t h i s ca lcu la t ion (Reference FSAR sect ion 3.13).

4 . The cause o f the Loss o f Spent Fuel Pool Cooling event i s s ta ted t o be a seismic event.

5. A l l p ip ing and equipment shared w i th o r connecting t o the RHR i n t e r t i e loop are Seismic Category I and can be iso la ted from any p ip ing associated w i th the non-Seismic Category I fue l pool cool ing system.

Evaluation o f D isc re~anc ies Noted i n EDR 620020

EDR 620020 describes nine discrepancies r e l a t i n g t o the loss o f spent f ue l pool cool ing event. This discussion w i l l summarize each issue. The reader i s


re fe r red t o the complete t e x t o f the EDR.

General Statement

The in t roductory paragraph o f the EDR states: ". ..the design p rov is ion f o r the loss o f spent fue l pool cool ing event i s t o permit the fue l pool t o b o i l and maintain i t s water l eve l above the fue l through makeup from the ESW system. This design prov is ion i s necessary because the fue l pool cool ing system used f o r normal operation and the RHR fue l pool cool ing ass is t mode used f o r abnormal heat loads are no t designed t o s a t i s f y seismic category I and s ing le f a i l u r e c r i t e r i a . "

As s ta ted i n design basis #5 above the RHR fue l pool cool ing ass is t po r t i on o f the p ip ing i s designed t o seismic category I requirements. No c r e d i t however i s taken f o r t h i s mode o f operation i n the fue l pool b o i l i n g analysis i n the FSAR. Credi t i s taken f o r t h i s mode f o r emergency heat load s i tua t ions as defined by basis #2.

Discussion o f EDR Items A throuqh I

I n order t o discuss and evaluate each o f the nine discrepancies 1 i s t e d i n the EDR i t w i l l be more l og i ca l t o review them i n a d i f f e r e n t order. Items E & F both r e l a t e t o the t ime-to-boi l ca lcu la t ions and w i l l be reviewed f i r s t fol lowed by items G through I, which are re la ted t o the t ime-to-boi l concern. Items C & D invo lve operator act ion considerations and w i l l be discussed next. F i n a l l y items A & B r e l a t i n g t o the evaporation e f fec ts w i l l be discussed.


Item E: Analy t ica l Time-to-Boil

"The ana ly t i ca l 25 hour t ime-to-boi l f o r the spent fue l pool i s nonconservative f o r the maximum normal heat load i n the spent fue l pool."

As stated i n basis #1 the maximum normal heat load i s 12.6 MBtu/hr. As stated i n basis #2 the t ime-to-boi l analysis i s based on a heat load o f 9.8 MBtu/hr. These two design bases are i n f ac t consistent and are based on the same " t yp i ca l fue l discharge schedule" and re fue l ing outage scenario. The d i f ference i n the heat load i s due so le ly t o the time a f t e r shutdown assumed f o r purposes o f estab l ish ing the design basis.

Focusing on the t ime-to-boi l analysis, a time a f t e r shutdown value o f 10.5 days i s used. This i s the time a t which it i s assumed tha t re fue l i ng i s completed and the reactor cav i t y t o fue l pool gates are re ins ta l led . P r i o r t o t h a t po in t the addi t ional water stored i n the reactor cav i t y i s also avai lab le as a heat s ink and the RHR system i s ava i lab le f o r fue l pool cooling. For times greater than 10.5 days the appropriate heat load w i l l be even lower than the analyzed value o f 9.8 MBtu/hr. For the SSES Un i t 2 5RIO the time from reactor shutdown t o fue l pool gates i n s t a l l e d was 38 days. The decay heat i n the Un i t 2 pool a t t h a t time i s calculated t o be 5.65 MBtu/hr (Reference SEA-ME-405). The corresponding time- to -bo i l i s 45 hours.

The EDR goes on t o discuss other calculat ions which r e s u l t i n d i f f e r e n t heat loads using various assumptions. Calculat ion NFE-B-NA-053 was performed by Nuclear Fuels t o account f o r actual fue l discharge h i s to ry and fu tu re of f loads accounting f o r power uprate condit ions. The fue l pool heat load versus time curves obviously w i l l increase subsequent t o power uprate, however, these curves do not apply t o the ex i s t i ng design. As long as the calculated decay heat i s less than 9.8 MBtu/hr a t the po in t where the fue l pool gates are r e i n s t a l l e d the o r i g i n a l design basis t ime-to-boi l ca lcu la t ion i s s t i l l va l i d .

Calculat ion M-FPC-009 determined time-to-boil condit ions post power uprate. This ca lcu la t ion shows tha t the t ime-to-boi l f o r the design basis heat load o f 9.8 MBtu/hr i s s l i g h t l y greater than 25 hours.

I n conclusion, the design basis f o r the t ime-to-boi l condi t ion i s establ ished by the 9.8 MBtu/hr value used i n the o r i g i na l calculat ions. This design basis i s met by planning the outage so tha t the fue l pool i s not i so la ted from the reactor cav i t y o r the RHR system p r i o r t o a po in t i n time where the actual heat load i s 9.8 MBtu/hr o r less.

This discrepancy i s no t a v a l i d deficiency, i s therefore no t repor tab le and has no impact on the o p e r a b i l i t y o f the plant.


Item F: Time-to-Boil f o r Emeraencv Heat Load

"The ana ly t i ca l 25 hour t ime-to-boi l f o r the spent fue l pool does not account f o r the emergency heat load i n the spent fue l pool."

As discussed above, the t ime-to-boi l condit ions apply t o conf igurat ions where the spent f ue l pools are i so la ted from the reactor cav i t y ( ie. , non-refuel ing conf igurat ions). As i s co r rec t l y stated i n the EDR, current p rac t i ce i s t o f u l l y o f f l oad the core dur ing each re fue l ing outage. Speci f ic ca lcu la t ions are performed by Nuclear Fuels t o determine the a b i l i t y o f the FPC system t o remove the combined decay heat o f the cross-t ied re fue l ing pools. Tests are also conducted t o determine tha t the actual heat removal capab i l i t y exceeds the actual fue l pool heat loads dur ing the outage (Reference TP-235-011). Normally the reactor c a v i t y i s maintained flooded and cross-t ied t o the fue l pools. One loop o f Core Spray i s always operable i n t h i s conf igurat ion. One d i v i s i o n o f RHR i s maintained i n shutdown cool ing mode except f o r a b r i e f per iod required f o r the common RHR system outage window.

Design basis #2 s ta tes t h a t heat loads i n excess o f the MNHL are considered t o be emergency heat loads. The design o f the RHR system t o ass is t the FPC system dur ing emergency heat load condit ions assures t ha t fue l decay heat i s removed. No t ime-to-boi l ca lcu la t ion f o r t h i s conf igurat ion i s required since the RHR system w i l l be i n operation o r avai lable. A t any rate, such a ca l cu la t i on should consider the e f f e c t o f the addi t ional water inventory ava i lab le from the flooded reactor cav i ty , cask storage p i t and dryer and separator storage pool which are a l l cross-connected dur ing t h i s time. Makeup inventory i s also avai lab le from Core Spray and the RHR system i s normally in-service except f o r the common RHR system outage window.

I n conclusion, no t ime-to-boi l analysis i s required f o r the emergency heat load design basis. Single f a i l u res o f the RHR system are not required f o r t h i s design basis f o r the emergency heat load (Reference SRP 9.1.3).

This discrepancy i s no t a v a l i d deficiency, not reportable and has no impact on p lan t operabi 1 i t y .


Item G: Radioloqical Release Calculation for Boilinu Soent Fuel Pool

"The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates."

This discrepancy is directly related to the heat load assumed for the time-to- boil analysis. The evaporation rate used in the dose calculation is based on a heat load of 9.8 MBtu/hr which is the design basis heat load for the time-to-boil calculation. Heat loads in excess of 9.8 MBtu/hr obviously result in higher evaporation rates. Since the discussion under Item E above establishes that 9.8 MBtu/hr is the correct original design basis and still bounds current operation there is no discrepancy in the offsite dose calculation. It uses an evaporation rate consistent with the design basis heat load.

This discrepancy is not a valid deficiency, not reportable and has no impact on plant operabi 1 ity.


Item H: Nonconservative Activity Terms

"The radiological release analysis for a boiling spent fuel pool uses nonconservative activity terms. The original design calculation (200-0048) assumed 12 month operating cycles and 184 bundle equi 1 i bri um re1 oad sizes to determine the activity terms for failed fuel in the fuel pool. SSES currently has 18 month operating cycles with approximately 230 bundle reloads which will increase to approximately 254 bundles after power uprate. Since the calculation implied that most of the activity results from the most recent discharge batch, the effect of increasing the discharge size from 184 bundles assumed in the calc to 230 and 254 bundles would appear to be nonconservative with respect to the radiological release analysis. "

The original radiological release analysis as referenced above is conservative for the following reasons:

(1) the activity levels used as a source term are based on 1% failed fuel. All of the failed fuel rods are assumed to be in the offloaded batch of 184 fuel assemblies. Therefore, increased batch sizes will not increase the amount of the source term used in this analysis.

(2) the activity levels used for the iodine source term are based on saturation level inventories for a core operating at 3440 MWt for one thousand days. Therefore, the fuel cycle length will not affect the source term.

In conclusion, the offsite dose calculation remains valid.

This discrepancy is not a valid deficiency, not reportable and has no impact on plant operabi 1 ity.


Item I: Analysis for Max Time Prior to M a k e u ~

"The analysis for maximum time prior to makeup to a boiling spent fuel pool is based upon nonconservative assumptions. The original design calculation (175-14) determined the time using evaporation of the entire fuel pool water inventory. The maximum time should be based upon a minimum fuel pool water level which is sufficiently above the top of the fuel to provide the shielding required to allow corrective operator actions."

The purpose of the referenced calculation was to determine refueling floor atmosphere conditions under various operating modes. The evaporation rates and assumptions used in the cited portion of the calculation were used solely to determine if condensation could be expected under fuel pool boil ing conditions. The conclusion of the calculation regarding time to boil the pool dry is not relevant to any operator action. Operator actions are based on maintaining normal pool level and temperature conditions. In any case, the cited nonconservatism would have a minor effect on the calculated 19 days to boil the pool dry, a result which is not used elsewhere in the design.

This discrepancy is not a valid deficiency, is not reportable and has no impact on plant operabi 1 i ty.


I tem C: Manual ESW Valve Actions

"The manual valve manipulations required t o provide ESW makeup f low t o a b o i l i n g spent fue l pool may not be possible."

In-p lant post-accident rad ia t i on leve ls are analyzed f o r SSES t o the requirements spec i f ied i n NUREG-0737. This document requires t ha t post-accident rad ia t i on l eve l s be determined f o r purposes o f v i t a l area access by p lan t operators t o perform short-term f i r s t p r i o r i t y actions. It spec i f ies t h a t rad ia t i on l eve l s be determined on the basis o f contained sources, and core damage source terms equivalent t o those used f o r lOCFRlOO calculat ions. These assumptions are c l e a r l y based on degraded core condit ions which are beyond the design basis LOCA. Airborne r a d i o a c t i v i t y sources from containment leakage are required t o be analyzed f o r environmental qua l i f i ca t i on o f equipment but not f o r personnel access.

A review o f FSAR chapter 18 shows tha t access t o the equipment necessary t o provide makeup t o the fue l pool from ESW i s r e s t r i c t e d f o r the approximately the f i r s t 30 hours fo l low ing the design basis event (Figure 18.1-9). This analysis i s based on a conservative source term equating t o 100% fue l damage r e s u l t i n g from core melt condi t ions as o r i g i n a l l y u t i l i z e d f o r o f f s i t e dose ca lcu la t ions used t o determine p lan t s i t i n g adequacy. These source terms were based on experiments invo lv ing heated i r rad ia ted uranium dioxide pe l l e t s .

An evaluat ion o f actual fue l thermal response dur ing design basis accidents r e s u l t s i n no predicted fue l f a i l u res (Reference PLI-72696). Thus, the source term r e s u l t i n g from the DBA LOCA would only be equivalent t o the r a d i o a c t i v i t y present i n the reactor coolant as a r e s u l t o f normal operations (a l lowing f o r fue l defects as permitted by Technical Speci f icat ions) . To bound the po ten t i a1 e f f e c t s o f a design basis accident, a r e a l i s t i c y e t conservative analysis using an assumed 1% fue l damage resu l t i ng from core degradation under LOCA condi t ions was performed (Reference EP-548) and concludes tha t access t o equipment necessary t o m i t i ga te the e f f ec t s o f a loss o f fue l pool cool ing fo l low ing a DBA LOCA i s assured.

I n conclusion, post-accident operator actions are v iab le f o r a l l po ten t i a l scenarios under consideration, f o r both the current design basis and those outside o f the current design basis.

This discrepancy i s no t a v a l i d deficiency, i s no t reportable and has no impact on p lan t operabi 1 i ty.


Item D: Instrumentation

"The instrumentation avai lab le t o the operator post-LOCA does not provide adequate i nd i ca t i on o f spent fue l pool temperature and l eve l t o al low proper response t o a loss o f fue l pool cool ing event."

The instrumentation avai lab le t o the operator i s not required t o be q u a l i f i e d since the design basis loss o f spent fue l pool cool ing i s not coincident w i t h the DBA LOCA condi t ions. This instrumentation i s powered from an un in te r rup t i b l e power supply and i t s ' associated 1E AC source.

The minimum water l eve l required per Tech Specs i s below the weir e levat ion. Since ESW makeup i s provided t o the pool the operators w i l l know t h a t when they see a r i s e i n skimmer surge tank leve l the fue l pool leve l i s a t l eas t as high as the weir. This provides a confirmation o f adequate pool l eve l wi thout requ i r i ng access t o the re fue l i ng f l oo r .

Furthermore, on the basis o f the discussion i n i tem C above, access t o the re fue l i ng f l o o r i s possible under a l l considered condit ions. Therefore, i t i s possible t o v e r i f y adequate fue l pool leve l v i sua l l y from the re fue l i ng f l o o r which i s accessible from several locat ions.

While the avai lab le instrumentation i s adequate f o r operator act ions and meets the regulatory requirements o f Reg Guide 1.13, improvements t o the instrumentation have been recommended i n the past and should be imp1 emented. This would enhance p lan t safety.

I n conclusion, the ex i s t i ng instrumentation i s adequate f o r performance o f requi red operator act ions f o r the current design basis and f o r scenarios not included i n the current design basis.

This discrepancy i s n o t a v a l i d deficiency, it i s no t repor tab le and has no impact on p lan t operab i l i t y .


I t em A: Reactor B u i l d i n o Desiqn Heat Loads

"Reactor b u i l d i n g design heat loads do n o t account f o r t h e b o i l i n g spent f u e l pool event."

The r e a c t o r b u i l d i n g temperature ana lys i s i s performed f o r DBA LOCA cond i t i ons . The presumption o f i t e m A i s t h a t t h e spent f u e l pool w i l l reach b o i l i n g c o n d i t i o n s p r i o r t o r e s t o r a t i o n o f f u e l pool c o o l i n g subsequent t o a LOCA. The e x i s t i n g temperature ana lys i s does account f o r t h e sens ib le heat l o a d f rom t h e f u e l pool a t 212F.

A l o s s o f spent f u e l pool c o o l i n g event can r e s u l t f rom several cond i t i ons . The design bas i s c o n d i t i o n i s a seismic event as analyzed i n t h e FSAR. The Fuel Pool Coo l ing system i s n o t designed f o r seismic loads. I n t h i s case, t h e Fuel Pool Coo l ing system i s assumed t o be l o s t . An eva lua t i on o f t h e p l a n t response shows t h a t several methods a r e a v a i l a b l e t o assure t h a t t h e spent f u e l remains cooled. These inc lude: (1) t h e RHR system can be used t o cool t h e f u e l pools w i t h a l t e r n a t e shutdown c o o l i n g o f t h e r e a c t o r us ing Core Spray and RHR f o r suppression pool coo l ing ; o r (2) a l l o w t h e f u e l pool t o b o i l w i t h makeup supp l i ed by ESW w i t h cons ide ra t i on o f e i t h e r SGTS opera t ing on Zone 111 o r p r o v i d i n g a vent pa th from Zone 111. The o f f s i t e dose ana lys i s takes no c r e d i t f o r SGTS. I f ava i l ab le , normal r e a c t o r b u i l d i n g v e n t i l a t i o n would be used t o p rov ide c o o l i n g and ven t i ng o f t h e Zone 111 atmosphere. Under any o f these scenar ios t r a n s p o r t o f mo is t a i r t o o the r p o r t i o n s o f t h e r e a c t o r b u i l d i n g would n o t occur. Th i s scenar io i s t h e design bas is f o r l o s s o f f u e l pool coo l i ng .

Other scenar ios n o t inc luded i n t h e design bas i s i nc lude LOCA and LOOP events, and combinat ions the reo f . The t ime frame f o r cons ide ra t i on o f ope ra to r a c t i o n s i s based on reasonable expecta t ions f o r t h e t ime- to -bo i l c o n d i t i o n . As s t a t e d p rev ious l y , f o r t h e c u r r e n t opera t ing p r a c t i c e , t h e f u e l pool heat l o a d p r i o r t o r e a c t o r r e s t a r t i s approximately 4.65 MBtu/hr. Time t o b o i l under t h i s c o n d i t i o n i s on t h e o rde r o f 55 hours. Note t h a t t h i s i s t h e s h o r t e s t p o s s i b l e t ime-to- b o i l f o r t h e c u r r e n t f u e l cyc le . With t h e pools c ross - t i ed t h e t ime- to -bo i l i s g r e a t e r than 100 hours.

For a LOCA scenario, t h e FPC system w i l l be l o s t i n i t i a l l y due t o t h e Aux Load Shed p rov i s ions . Although t h e Fuel Pool Cool ing system and o t h e r non-safety re1 a ted systems are n o t speci f i c a l l y analyzed f o r t h e e f f e c t s o f hydrodynamic loads i t i s expected t h a t they w i l l be ab le t o per form t h e i r normal f u n c t i o n s f o l l o w i n g a broad spectrum o f design bas i s events. C r e d i t f o r these systems i s n o t needed t o meet t h e design basis , however, p l a n t opera tors w i l l u t i l i z e any equipment a v a i l a b l e t o them du r ing emergency s i t u a t i o n s . Therefore, i n t h e course o f e v a l u a t i n g t h e e f f e c t s o f a DBA LOCA on t h e f u e l pool c o o l i n g system, we acknowledge t h e a v a i l a b i l i t y o f normal p l a n t systems i n responding t o t h e emergency.

Independent o f t h e LOCA cond i t ion , o f f s i t e power i s needed t o r e s t o r e normal c o o l i n g systems. The SSES I n d i v i d u a l P lan t Eva lua t ion considered l o s s o f o f f s i t e power (Reference IPE Appendix F) . The IPE conse rva t i ve l y est imated t h e inc idence o f LOOP t o be .04/year ( p l a n t re la ted ) , .008/year ( g r i d r e l a t e d ) , .00807/year


(severe weather re1 ated) , and .00066\year (extreme1 y severe weather re1 ated) . The p r o b a b i l i t y o f recovery from the LOOP w i th in spec i f ied times was also ca lcu la ted as fo l lows:

Time l h r l PlRecoverv w i t h in T h r s l 12.0 97.96% 24.0 99.53% 60.0 99.923% 75.0 99.953%

Thus, i t can be reasonably concluded tha t o f f s i t e power w i l l be avai lab le w i t h i n 24 hours fo l low ing the i n i t i a t i n g event.

The remaining fac to r i n addressing restorat ion o f fue l pool cool ing i s access t o the reactor bu i ld ing. This issue i s discussed under Item C above. Except f o r degraded core condi t ions access t o the reactor bu i ld ing i s feas ib le a f t e r the f i r s t twelve hours o f the i n i t i a t i n g event (Reference EP-548).

Notwithstanding the above basis, we have also considered the scenario where o f f s i t e power i s not ava i lab le and access t o the lower reac to r bu i l d i ng elevat ions i s r e s t r i c t e d (based on FSAR chapter 18 contained source terms). Under these condi t ions (representative o f a degraded core event) access t o the re fue l i ng f l o o r remains avai lable. Provision f o r fue l pool cool ing i s made through use o f the p lan t f i r e protect ion system. Venting o f Zone I11 v i a the f i l t e r e d exhaust system i s also possible f o r t h i s scenario. While access t o l eve l and temperature instruments would be questionable i t i s possible t o v e r i f y adequate pool l eve l v i sua l l y from the re fue l ing f l o o r which i s accessible a t several 1 ocations.

A Loss o f Fuel Pool Cooling Event Tree i s attached t o t h i s evaluat ion t o help guide the reader through these various postulated scenarios.

I n conclusion, f o r the design basis loss o f fue l pool cool ing the p lan t as cu r ren t l y designed and analyzed i s acceptable. For other scenarios not spec i f i ca l l y included i n the design basis we have reasonable assurance t h a t the e f f ec t s o f a loss o f fue l pool cool ing can be mi t igated without adverse consequences on the p lan t .

This discrepancy i s a v a l i d deficiency. It i s no t a sa fe ty s i g n i f i c a n t issue because we have establ ished reasonable assurance t h a t the e f f ec t s o f a l oss o f fue l pool cool ing can be m i ti gated without adverse consequences on the p lan t and pub l i c hea l th and sa fe ty and i s therefore not reportable. The evaluat ion above a lso shows t h a t t h i s concern does not impact p lan t operab i l i t y .

I n considerat ion o f t h i s concern, addi t ional analyses are warranted t o f u r t he r quant i f y the e f f ec t s o f evaporation and b o i l i n g condi t ions on the Zone 111 atmosphere and the po ten t ia l t ransport o f moist a i r t o other locat ions i n the reactor bu i l d i ng f o r condit ions outside o f the current design basis.


Item 8: I m ~ a c t o f ESW Makeu~ Water

"The impact o f the ESW makeup water t o the spent fue l pool on equipment i n the reactor bu i l d i ng has not been evaluated."

The analysis under i tem A above applies t o t h i s issue as wel l . This evaluat ion shows t h a t w i t h the current p lan t design and f o r ex i s t i ng design basis condi t ions the e f f ec t s o f a loss o f fue l pool cool ing are acceptable.

This discrepancy i s a v a l i d deficiency. As w i th i tem A i t i s no t a sa fe ty s i g n i f i c a n t issue and i s no t reportable. The evaluat ion above a lso shows t h a t t h i s concern does not impact p lan t operab i l i t y .

I n consideration o f t h i s concern, addi t ional analyses are warranted t o f u r t he r quant i f y the e f f ec t s o f evaporation and b o i l i n g condi t ions on the Zone 111 atmosphere and the po ten t ia l t ransport o f moist a i r t o other loca t ions i n the reactor bu i l d i ng f o r condit ions outside o f the current design basis.


En~ inee r i na Reoort on Loss o f S ~ e n t Fuel Pool Coolinq

A de ta i led repor t , SEA-ME-405, i s being prepared t o document t h i s evaluat ion i n f u r t he r d e t a i l . This repor t contains technical input from several engineering groups and w i l l provide a comprehensive set o f references on t h i s subject. The repor t w i l l be completed w i th in by October 28, 1992.


Enqineerinq R e ~ o r t on Loss o f S ~ e n t Fuel Pool Cooling

A de ta i led report , SEA-ME-405, i s being prepared t o document t h i s evaluat ion i n f u r t he r d e t a i l . This repor t contains technical input from several engineering groups and w i l l provide a comprehensive set o f references on t h i s subject. The repor t w i l l be completed by October 28, 1992.

Attachment 27

PP&L Memo from David A. Lochbaum and Donald C. Prevatte t o George T . Jones, "Evaluation o f EDR G20020 ReportabilitylOperability", October 26 1992 (PLI-72739)

October 26, 1992


SUSQUEHANNA STEAM ELECTRIC STATION EVALUATION OF EDR G2002O REPORTABILITYIOPERABILITY PLI-72739 FILE A45-1A

We have read Mr. Myers' memo of October 20, 1992 evaluating the reportability/operability of EDR G20020. We would like to thank both you and Mr. Myers for your considerations in performing these evaluations.

Our overall impression is that, in this document, we have received another confirmation of the technical validity of the concerns that we raised. However, there are still areas of disagreement as to reporting requirements to the NRC, and misunderstanding as to some of the facts associated with this EDR. The following paragraphs address these disagreements and misunderstandings.

First, on reportabil-ity. For an operating ~lant; 10CFR50.72 requires licensees to report in one hour any instance of the plant being (a) in an unanalyzed condition that significantly compromises plant safety; (b) in a condition that is outside the design basis of the plant; or (c) in a condition not covered by the plant's operating and emergency procedures. If any one of these criteria applies, the condition is reportable. We believe that all of these criteria are satisfied by the concerns described in EDR G20020 and other documents that have been communicated to you, to the EDR Group, and to others handling the evaluation of these concerns.

For the first condition, we have pointed out numerous areas where safety-related equipment in the reactor building is not analyzed for the conditions that would result from a boiling spent fuel pool, and where there is high potential that such analyses would show the conditions to be unacceptable, e.g., potential flooding of safety-related equipment, potential exceeding of EQ temperatures by large margins, potential pressurization of the reactor building, potential wetting of the charcoal in the SGTS, potential structural failures of ductwork due to condensation, etc..

These examples also illustrate areas where we satisfy the second condition of reportability. For every example, the probable consequences of a boiling spent fuel pool are outside their design bases.

The third condition of reportability is satisfied since we have no plant procedures which address how to cope with the conditions that would be generated by a boiling spent fuel pool, or even recognize that a boiling spent fuel pool would create these adverse conditions.

Legalistic literal interpretations of the CFR have been made that you must actually be in these conditions, that is, you must actually have an accidentin progress where these conditions exist, for them to be reportable. This is, as we have discussed, a ludicrous argument. The obvious intent of the NRC is that we should never get into these conditions in the first place. Therefore, the common sense interpretation is, any status that would create such a condition were an accident to occur. Using this interpretation, our status would be reportable.

Even if the above described provisions of 10CFR50.72 were not applicable, another provi.sion ic Paraqraph (h) (2) (iii] requires that reports shall be made within four hours of any condition that alone could have prevented the fulfillment of the safety function of structures or systems needed to (a) shut down the reactor and maintain safe shutdown, (b) remove residual heat, (c) control radioactive release, or (d) mitigate the accident. The concerns described in the EDR potentially could have prevented the fulfillment of all four of these.

Again, legalistic arguments have been raised with the fact that, at the time, it was not known for certain that any of these conditions would result. We maintain that the intent of this provision is clear; if it is reasonable to believe, based on our knowledge and experience with our plant and its analyses, that a condition could have prevented the fulfillment of a safety function, then it is reportable. There has been sufficient knowledge to have this reasonable belief for several months. It isn't necessary to cross every l1TIg and dot every "I" to have that reasonable belief. In addition, other independent formal reports have confirmed our concerns and reinforced this reasonable belief (Reference Mr. K. W. Brinkman's report, PLI-72288 of 1/9/92 and Mr. J. R. Miltenbergervs report, PLI-72367 of 9/9/92).

The following paragraphs address specific comments made in Mr. Myers8 memo:

1. First paragraph, we agree with Mr. Myers8 conclusion that these concerns "...should be formallv brought to the NRC1s attention...", and we would like to cite several very important reasons.

First, it is the law.

Second, informal reports are not well documented and may be incomplete.

Third, informal reports don't tend to get the requisite level of attention, either internally or externally.

Fourth, formal reports set into motion certain actions, commensurate with the safety significance of the concerns, both internally and externally, that are not necessarily set into motion by informal reports.

2. Page 3, Item 1, this section purports to discuss Items 9G and 9H from the EDR. It actually addresses neither.

Item 9G discusses the radiological release from a boiling fuel pool with respect to the increased heat load and the resultant shorter time to boil and increased boiling rate, both of which increase the radiological release. Mr. Myers' discussion does not touch on this point.

Item 9H discusses the increases in radiological release due to the increase in the number of bundles offloaded (the most significant contributor to the increase in releases) from 184 to 230 (a 25% increase). Mr. Myers defers discussion of this item to another section of his report.

He concludes that the existing analysis results are still bounding. Without considering these two factors, it is difficult to understand how this conclusion can be reached.

3. Page 3, Item 2, this section discusses Items 9E, 9F, 9G, and 9H from the EDR. In this discussion, Mr. Myers concludes that in the pool configuration currently used for refueling, Inat the appropriate time post shutdown", the time to boil would not be significantly different from the 25 hours from the FSAR (This is the same time

currently relied upon by the operators from procedure ON- 135-001.). Reference 8 cited by Mr. Myers shows that Ifthe appropriate time post shutdownN does not occur for 14 days. However, until 14 days have past, the time to boil is less than 25 hours (as low as approximately 8.8 hours at the beginning of this time). This is not reflected in the plant procedures.

4. Page 5, in Mr. Myers8 addressing the "Effect of Spent Fuel Pool Boiling on the SGTS", he states that "The SGTS is apparently designed for the environment (using preheating to reduce humidity)." In fact, SGTS requires this preheating to withstand the environment of a LOCA without fuel pool boiling (inlet conditions of 125OF, 100% relative humidity). It is specifically not designed to accommodate fuel pool boiling conditions (inlet of 180°F, 100% relative humidity) as described in Bechtel Calculation 175-17, Rev 4.

Additionally, Mr. Myers' review only addresses one of the concerns with the SGTS design, the structural integrity of the ductwork. Several other concerns have been communicated in documents and conversations subsequent to the original EDR. These include moisture carryover and/or condensation in the charcoal beds, fusing of the fire dampers in the ductwork (rated at 165O~), exceeding the EQ conditions in the SGTS room, and pressurizing the reactor building, among others, any one of which could incapacitate or degrade the system.

Even with the single concern addressed by Mr. Myers, he concludes that the condition is unanalyzed, and Ifit is not clear" if the SGTS would function. This alone, per lOCFR50.72(b)(ii)(A), makes the condition reportable.

5. Page 6, Mr. Myers states that "...the reportability question is whether or not the matter is within our licensing design basis." Although this is certainly one of the criteria from 10CFR50.72 that must be considered, there are others that do not appear to be considered in his report. There is also the question of the correctness of the original design criteria. If it is not correct, and as a result, unanalyzed conditions exist which have the potential to compromise the integrity or functionality of a safety feature or system in the plant, then that is also certainly reportable.

Mr. Myers also makes the point that historically the NRC did not consider all of the subsequent consequences of design basis events. This is an interesting historical footnote, but it is not pertinent. The consequences should have been considered, and the fact that they were not does not diminish their safety significance or the obligation to report them and correct them.

Mr. Myers cites three points which he concedes "...do raise some questions:"

The first is that WNP-2 (similar in design to SSES) did consider the effects the boiling spent fuel pool, came to the conclusion that the condition was unacceptable, and upgraded their fuel pool cooling system to preclude this condition. This would seem to lend credibility to the contention that the current design for SSES is inadequately analyzed and probably unacceptable.

The second point made is that our latest COTTAP analysis of the reactor building temperatures is conservative because it takes into account the sensible heat from the fuel pool at boiling conditions. Mr. Myers does not appear to understand the problem with the current analysis. The problem is that this analysis considers the sensible heat a; it does not consider the latent heat released during boiling which is many time greater than the sensible heat. Considering the sensible heat only, yields a total building heat load of approximately 5.2 million BTU/hr. Considering the latent heat adds approximately 20 million BTU/hr. Clearly, the results of the latest analysis are very non-conservative.

The third point made is that he understands that earlier calculations did include considerations of a boiling spent fuel pool. Indeed, this is true; the previously cited Bechtel Calculation 175-17, Rev 4 is an example, and based on the results of these calculations, it was concluded that the boiling spent fuel pool was not acceptable. However, this conclusion was somehow lost and was not integrated into the original design and licensing of the plant.

6. Page 7, first paragraph, Mr. Myers states that '... the preponderance of the evidence says that the effect of the boiling environment on both normal and post accident environments was not considered to be a requirement for the licensing of SSES."

On this point we strongly disagree. First of all, it was a requirement; 10CFR50.49 specifically requires that electrical equipment be environmentally qualified 'for the most severe design basis accidents." In this case, the design basis accident of LOCA/LOOP, or even LOCA without LOOP under the current operating procedures, mechanistically results in a boiling spent fuel pool which produces environmental conditions which are not analyzed and are likely to exceed the current EQ limits for the safety-related systems in the building. Secondly, the preponderance of evidence is not that it was not considered to be a requirement (We have an EQ program that's evidence of our understanding of the requirement.), but simply that it was overlooked except for the SGTS, and in that case, the effects were found to be unacceptable.

7. Page 7, in the second paragraph, Mr. Myers begins a discussion of the loss of fuel pool cooling during normal operation. In this and the ensuing two paragraphs he describes analyses that need to be performed and conditions that would be required to be maintained for this event which are not addressed by the current design or operating procedures. He concludes that "...given the time involved [until boiling begins after loss of cooling], it appears feasible to take damage control steps to accommodate the impacts."

We strongly maintain that this is not a valid approach to plant design or operations. Design shortcomings are required to be corrected when they are found, not when they are manifested in actual failures, and procedures to address anticipated accident conditions are also required to be developed ahead of time, not while the accident is in progress. It is not valid to say that we will develop these at the time of the event as a part of damage control.

Additionally, it should be pointed out that since our letter of October 9, 1992, between ten and twenty engineers have been working in the Allentown office and others at the site, continuously, late nights, and weekends for two weeks to revise the designs and the procedures to justify interim operation. The significance of this is twofold: First, if a JIO is required, this would appear to concede that the existing designs and procedures are inadequate. Second, if such massive and concentrated effort is required to explore

all of the ramifications of making these changes, then it is difficult to imagine how in an accident we could take the right actions or even know what the right actions are, with less people, in a shorter time frame, and under extreme pressure, as a part of "damage control".

The fact that we do not currently have these design features and procedures in place satisfies the conditions for reportability per 10CFR50.72, paragraphs (b) (ii) (A) and (b) (ii) (c) respectively.

8. Page 8, Conclusion 1, there appears to be no reason given why Item 91 from the EDR is not valid.

9. Page 8, Conclusion 4, Mr. Myers states that because the effects of fuel pool boiling were not considered as part of the original design basis or licensing basis, the concerns are not reportable. This appears to be convoluted logic. Their not being considered in the original design and licensing is, in and of itself, a reportable condition. The effects not being considered is the prime focus of the EDR - the reason for concern in the first place. If the original design and licensing bases were not adequate, the fact that it was not recognized until today does not make them adequate today.

Conclusion 5 appears to agree with this in principle, but not to the point of saying these concerns are reportable. It is at least gratifying to see that Mr. Myers shares our concern for the potential difficulties of the operators and the need for prompt attention.

Mr. Myers makes the point in the last sentence of Conclusion 4 that these concerns are typical of those uncovered in DBD efforts and implies that DBD concerns are exempt from being reportable.

On this point the regulatory guidance is very clear; NUREG-1397, 2/91, An Assessment of Design Control Practices and Design Reconstitution Programs in the Nuclear Power Industry, states in Section 3.8, Operability and Reportability, "Once the determination has been made that the facility has been or is operating outside its design bases or that systems, structures, and components may be incapable of performing their specified function(s), the requirements for reportability as specified in 10CFR50.72 and 10CFR50.73 become operative and the time clock starts for any affected action

statements as defined in the facility's technical specifications." Additionally, in Section 4.3.2 Reportability, the NUREG states, "The reporting requirements specified in 10CFR50.9, 50.72, and 50.73 apply equally to discrepancies discovered during DDR [DBD] programs. Therefore, there is no regulatory basis to treat discrepancies discovered during the conduct of a DDR program differently than any other reportable item."

10. Conclusion 6 states that the impact of heat and moisture on SGTS should be evaluated immediately, and that, if the conditions are unanalyzed, that is reportable. The fact is that, as described in Item 4 above, one of the concerns was analyzed by Bechtel and found to be unacceptable for SGTS. This should be reportable as Mr. Myers says. The other conditions of concern, as described in Item 4 above, are unanalyzed, and therefore they also should be reportable as Mr. Myers says.

11. Conclusion 7, we agree that an LER should be produced.

12. Mr. Myersg comments in Conclusion 10 are appreciated, and we believe they are sincere. Unfortunately, they will not be read by the engineers in the tren-hes of the Nuclear Departmen.;, of if they are, they will not be believed. To them, this is a test case. Their approach to the EDR System in the future will be governed in large measure by how this issue has been handled, and up until our meeting October 9, 1992, they have seen this issue being brushed off. They will believe what they see, not what they are told.

Again, we appreciate the attention of yourself, Mr. Myers, and all of the others who have been engaged in addressing these concerns. We are gratified that the approach of Mr. Myers appears to have taken has been more common sensical and less legalistic than others have taken. However, our basic concerns and positions reflected in our letter of October 9, 1992 and subsequent conversations still remain. As always, we remain at your service.

\ x % a David A. Lochbaum

\A\?& Donald C. Prevatte

cc: C. A. Myers G. D. Miller R. R. Sgarra J. M. Kenny J. E. Agnew D. F. McGann G. J. Kuczynski H. G. Stanley J. R. Miltenberger H. W. Keiser R. G. Byram W. R. Corcoran

J. S. Kemper

R. L. Doty A. F. Iorfida A. R. Sabol W. R. Licht J. S. Stefanko F. G. Butler J. A. Zola M. R. Mjaatvedt C. A. B0schett.i T. J. Sweeney G. D. Gogates M. J. Manski J. D. Richardson

A2-4 A6-3 A2-4 A2-4 A6-3 SSES S&A-4 SSES SSES A6-1 TW-16 A6-1 21 Broadleak Circle Windson, CT 06095 115 Polecat Road Glen Mills, PA 19342 A9-3 SSES A2-5 A6-1 A9-3 A6-3 A6-3 A6-3 SSES SSES SSES Enercon Enercon


PP&L Memo f r o m D a v i d A . Lochbaum and D o n a l d C . P r e v a t t e t o George T. Jones, "Response t o E v a l u a t i o n o f EDR G2002OU, Oc tobe r 28, 1992 (PL I -72751 )

Note: I n t h i s memo, t h e a u t h o r s responded t o PP&L1s e v a l u a t i o n o f t h e i n d i v i d u a l conce rns i n EDR G20020 ( A t t a c h m e n t 26) . The a u t h o r s ag reed w i t h t h e t e c h n i c a l j u s t i f i c a t i o n p r e p a r e d b y PP&L showing t h a t two ( 2 ) o f t h e n i n e conce rns were n o t v a l i d d i s c r e p a n c i e s . The a u t h o r s a l s o ag reed w i t h PP&L t h a t t h e r e m a i n i n g seven conce rns were v a l i d d i s c r e p a n c i e s , b u t s t r o n g l y d i s a g r e e d t h a t t h e s e v a l i d d i s c r e p a n c i e s d i d n o t a f f e c t t h e o p e r a b i l i t y o f t h e p l a n t and were n o t r e p o r t a b l e .

October 28, 1992


SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO EVALUATION OF EDR G20020 PLI-72751 FILE A45-1A

We have reviewed Mr. Glenn Miller's most recent evaluation of EDR 620020 dated October 21, 1992 (PLI-72711). While we concur with the technical justification provided for two of the concerns expressed in EDR G20020, we continue to disagree with the operability and reportability assessments for the majority of the concerns. A detailed discussion in response to Mr. Miller's evaluation is attached.

In his memo, Mr. Miller suggested that Nuclear Regulatory Affairs reconsider the reportability aspects of EDR G20020 in light of the clarified information. Since the initial evaluation dated October 6, 1992 for EDR G20020 concluded that none of the concerns was valid and Mr. Miller's evaluation indicates that two of the concerns are valid, we suggest that EDR G20020 be formally revised to reflect the latest PPhL position on these nine concerns.

Mr. Miller also indicated that a justification for interim operation would soon be completed. In Mr. Miller's evaluation, none of the nine concerns affect the operability of the plant and a justification for interim operation would not be required. We agree that the justification for interim operation should be completed, but because current operation of SSES is adversely affected by the concerns identified in EDR G20020.

While we are pleased to have the concerns we raised in EDR G20020 addressed individually, our basic concerns and positions reflected in our letter of October 9, 1992 and subsequent discussions still remain.

As always, we remain at your service.

Attachment - Response to Evaluation of EDR G20020 (PLI-72711) cc: C. A. Myers A2-4

H. G. Stanley SSES G. D. Miller A6-3 J. E. Agnew A6-3 M. R. Mjaatvedt A6-3 Nuclear Records A6-2

M. W. Simpson A1-2 J. S. Stefanko A9-3 J. R. Miltenberger A6-1 D. F. Roth SSES J. M. Kenny A2-4 F. G. Butler A6-3

Response to Evaluation of EDR 620020 (PLI-72711)

1. The nine concerns identified in EDR G20020 were listed in order of decreasing nuclear safety significance. Mr. Miller's evaluation rearranges the nine concerns and addresses them in essentially order of increasing safety significance. We have ordered our comments to match the order in EDR G20020.

2. We concede that EDR 620020 incorrectly stated that the RHR fuel pool cooling assist mode was not seismically designed. The fact that the fuel pool cooling mode of RHR is seismically designed does not materially change any of the concerns expressed in EDR G20020.

3. Mr. Miller's response to EDR G20020 Item A on reactor building heat loads is inadequate for many reasons. Mr. Miller states that the only scenario in the SSES design basis for loss of fuel pool cooling is a seismic event. However, the SSES design basis also includes loss of offsite power, LOCA, and failures of non-safety related components which can each result in loss of fuel pool cooling. Additionally, Mr. Kevin Brinckman in his report dated September 1, 1992 (PLI-72288) indicated that the hydrodynamic loads resulting from a LOCA may result in loss of fuel pool cooling. And finally, Reg Guide 1.13 states that the spent fuel pool shall be designed to maintain adequate cooling of the fuel under all normal operating and postulated accident conditions. Therefore, the SSES design basis implicitly covers failure modes for fuel pool cooling other than the seismic event.

Mr. Miller contends that the loss of fuel pool cooling event coupled with a LOCA or LOOP is outside the SSES design basis. However, SSES FSAR Chapter 6.2 reports that the LOCA scenario used for containment functional design is postulated to occur simultaneously with a LOOP and a safe shutdown earthquake. The calculated reactor building heat loads are inputs to the EQ program for safety related components located in the reactor building. Since operation of these safety related components is assumed for core and containment cooling in the containment functional design analyses, it is necessary that the reactor building heat load calculations consider a loss of fuel pool cooling. The latent heat load from a single spent fuel pool is approximately four (4) times greater than the current total calculated reactor building heat load and could result in calculated reactor building room temperatures exceeding EQ values for safety related components.

PP&L1s implementation in 1988 of procedures to manually initiate shedding of all non-Class 1E loads in the reactor building 24 hours after a MCA to control reactor building room temperatures should have been an opportunity to properly


address the consequences of a boiling spent fuel pool. The 10CFR50.59 safety evaluation for this activity should have covered the effects of loss of fuel pool cooling, particularly considering the change was made to prevent excessive reactor building room temperatures and loss of fuel pool cooling will adversely affect these same room temperatures.

Mr. Miller states that one SSES response to a loss of fuel pool cooling is the RHR system. The RHR fuel pool cooling assist mode has adequate decay heat removal capacity to handle the load, but it is a non-safety related system which has never been used at SSES and may never have even been successfully pre-operationally tested. Furthermore, use of the RHR fuel pool cooling assist mode is described in the SSES FSAR and SER only to supplement fuel pool cooling for the emergency heat load case (full core offload) . And finally, as Mr. Brinckman states in his report dated September 1, 1992 (PLI-72288), use of RHR in the fuel pool cooling assist mode following a LOCA is an unanalyzed condition which may compromise core and containment cooling.

Mr. Miller states that the second SSES response to a loss of fuel pool cooling is to t'allow the fuel pool to boil with makeup supplied by ESW with consideration of either SGTS operating on Zone I11 or providing a vent path from Zone 111. Per Mr. Dave Pai, the SGTS will not operate if the fuel pool boils because the fire dampers isolate at inlet temperatures above 165'F and the calculated inlet temperature resulting from a boiling spent fuel pool is =180eF. The normal reactor building ventilation relied upon by Mr. Miller to cool and vent Zone I11 is a non-safety related function which cannot be relied upon in this manner. Additionally, the current design of the reactor building W A C system and the standby gas treatment system for the LOCA scenario do not permit the alignment proposed by Mr. Miller. To utilize such an alignment would require extensive analyses to determine its feasibility and the design modifications necessary to accomplish this operation.

Mr. Miller reports that the SSES IPE determined a very low probability of LOOPS lasting over 24 hours. Such information would support a justification for interim operation, but cannot be used to eliminate a design requirement. In addition, Mr. Miller's contention that fuel pool cooling would be restored prior to the pool boiling after a LOCA is inconsistent with the assumed LOOP duration specified throughout the SSES FSAR. For example, the design basis for the ultimate heat sink water inventory provides makeup to both boiling spent fuel pools for the 30 days period following the


SSES design basis LOCA/LOOP, not just 24 hours.

Mr. Miller states that "although the fuel pool cooling system and other non-safety related systems are not specifically analyzed for the effects of hydrodynamic loads it is expected that they will be able to perform their normal functions following a broad spectrum of design basis eventsv1 and that "credit for these systems is not needed to meet the design basis1'. The design response to any postulated design basis accident must not rely on non-safety related equipment. Additionally, "it is expected8' as Mr. Miller states is insufficient rigor for design analyses of nuclear safety functions.

Mr. Miller states that a loss of fuel pool cooling could be handled by allowing the pool to boil, providing makeup from the ESW system, and operating Zone I11 ventilation to avoid adverse consequences. We assume that Mr. Miller is conceding that SSES could not have endured a boiling spent fuel pool with the design features and procedures currently in place without significant adverse consequences.

We agree with Mr. Miller that EDR G20020 Item A is a valid deficiency. We strongly disagree with Mr. Miller's contention that this valid deficiency has no safety significance and is not reportable. If the fuel pool boils, the existing reactor building heat load calculations do not account for the latent heat load.

4. Mr. Miller's response to EDR G20020 Item B on the impact of ESW makeup water is inadequate for the reasons as given in Comment 3 and because his response does not address all of the potential impacts.

The ESW flow supplied to the fuel pool is controlled by manually positioning a throttle valve. If the ESW flow rate to the fuel pool exactly matches the boil-off rate, then the level in the fuel pool will be maintained constant. If the ESW flow rate is lower or higher, then the fuel pool level will drop or rise accordingly. The most probable outcome will be for more ESW flow than is required to be supplied to the fuel pool. Under this scenario, both the moist air from fuel pool boil-off and the water from fuel pool run-off must be considered. The adverse nuclear safety consequences include pressurization of the refueling bay and/or secondary containment, flooding, component failure due to humidity and condensation, and HVAC ductwork failures due to either flow blockage from condensed vapor or collapse from the added water weight. As stated in Comment 3, the SSES ultimate heat sink


design analysis covers ESW makeup to both spent fuel pools throughout the 30 day period following the SSES design basis MCA/MOP. However, this analysis is incomplete and invalid because the consequences of the 5 million gallons of water delivered to the spent fuel pools on systems and components in the reactor building is not taken into account.

We agree with Mr. Miller that EDR G20020 Item B is a valid deficiency. We strongly disagree with Mr. Miller's contention that this valid deficiency has no safety significance and is not reportable. If ESW is supplied to a boiling fuel pool, the consequences are virtually unanalyzed. For example, the boil-off might result in pressurizing the refueling bay and challenging secondary containment integrity. As Mr. Miller states, "additional analyses are warrantedto further quantify the effects of evaporation and boiling conditions on the Zone I11 atmosphere and the potential transport of moist air to other locations in the reactor building." However, these analyses are needed to support SSES operation within its existing design basis.

5. Mr. Miller's response to EDR G20020 Item C on manual ESW valve actions is inadequate because it is based on predictions of no more than 1% fuel failures. Mr. Miller's logic that recent evaluations of actual fuel thermal response during design basis accidents indicate no fuel failures occur would support tearing down secondary containment and removing the standby gas treatment system if it were justified.

Mr. Miller states that the radiation levels determined at SSES in response to NUREG-0737 requirements are "clearly based on degraded core conditions which are beyond the design basis MCA." Nevertheless, the requirements in NUREG-0737 were imposed by the NRC following the TMI accident and are clearly within the SSES licensing basis.

Mr. Miller claims that "airborne radioactivity sources from containment leakage are required to be analyzed for environmental qualification of equipment but not for personnel access." This claim is preposterous, and we believe that the SSES Operations staff does not support this position. If airborne radiation sources are considered, and they would be realistically considered by HP in determining if the post-MCA reactor building could be entered, severe core damage is not required forthe reactor building to be rendered inaccessible.

Therefore EDR G20020 Item C is a valid deficiency because the makeup supply to a boiling spent fuel pool may not be available post-MCA due to inaccessibility of the ESW manual

Response to Evaluation of EDR G2OO2O (PLI-72711)

valves.

6. Mr. Miller's response to EDR G20020 Item D on fuel pool instrumentation is inadequate because it is incomplete. Reg Guide 1.97 requires that instrumentation required for initiating and monitoring safety functions be qualified. The fuel pool instrumentation may be powered from Class 1E sources, but it has not been established that this instrumentation will function in the environment in which it will be exposed post-LOCA. In addition, Mr. Brinckman in his letter dated September 1, 1992 (PLI-72288) reported that "the fuel pool trouble alarm in the control room cannot be counted on for reliable indication." The readouts of fuel pool level and temperature are at local panels in the reactor building and would also be inaccessible post-LOCA. Mr. James Miltenberger in his letter dated September 9, 1992 (PLI-72367) indicated that the fuel pool level instrumentation needs to be upgraded to provide reliable control room indication.

Mr. Miller also states that the Ifinstrumentation available to the operator is not required to be qualified since the design basis loss of spent fuel pool cooling is not coincident with the DBA MCA conditions. However, Reg Guide 1.13 requires adequate cooling of the spent fuel pool to be available for all normal operating and postulated accident conditions. Since a fuel pool cooling failure can occur due to the hydrodynamic loads or environmental conditions associatedwith the design basis IDCA, it should have been considered as part of the design basis. In addition, if the existing unqualified instrumentation provides a false indication of increasing fuel pool temperature or dropping fuel pool water level post-LOCA, personnel may be unnecessarily exposed to radiation as they enter the reactor building and refueling floor area to respond to the perceived threat.

EDR G20020 Item D is a valid deficiency because the fuel pool instrumentation may not be adequate to provide the operator with sufficient information to implement and monitor required safety measures.

7. Mr. Miller's response to EDR G20020 Item E on the analytical time to boil for the maximum normal heat load case is inadequate because it does not address the probtem reported. EDR G20020 Item E did not dippute the 9.79~10 BTU/hr heat load value versus the 12.6~10 BTU/hr value, but rather that the maximum normal heat load upon which the time to boil calculation was based was rendered non-conservative by changes in fuel types and operating cycle lengths. EDR GO0005 was written in 1990 on the subject of outdated FSAR Chapter 9


data. Item E was included in EDR G20020 to identify another consequence of the problem originally reported in EDR G00005. It is our understanding that EDR GO0005 has not yet been resolved over two years after it was initiated.

EDR G20020 Item E is a valid deficiency because the basis for the time to boil calculation reported in the FSAR is invalid, but as we clearly stated in our memo dated June 22, 1992 to Joe Zola (ET-0471) it does not affect the present operation of S S E S because the exis8ing decay heat loads in the fuel pool are less than 9.79~10 BTU/hr. This deficiency represents a potential safety concern bfcause the maximum normal heat load at S S E S may exceed 12.6~10 BTU/hr when the spent fuel pool is filled to capacity.

8 . Mr. Miller's response to EDR G20020 Item F on the time to boil for the emergency heat load case is inadequate for many reasons. Mr. Miller states that "no time-to-boil calculation for this configuration [emeraencv heat load case] is required since the RHR system will be in operation or available." The time to boil calculation for the maximum normal heat load case is provided in FSAR Appendix 9A even though the fuel pool cooling system is initially operating because its failure must be considered. An equivalent calculation must be provided for the emergency heat load case because the RHR fuel pool cooling assist mode could fail. As Mr. Miller points out, this calculation should consider the additional water inventory available. But this calculation must also account for operational events such as putting the fuel pool gates in to isolate the reactor cavity from the spent fuel pool volume.

Mr. Miller states that "single failures of the RHR system are not required for this design basis for the emergency heat load1@ and references S R P 9.1.3. S R P 9.1.3 indeed supports the position that a single active failure need not be considered for the emergency heat load case. However, since the emergency heat load case is defined as a full core offload in S R P 9.1.3 and S S E S FSAR Chapter 9 and refueling operations described in the S S E S FSAR do not entail full core offloads, then PP&L8s routine use of full core offloads conflicts with the FSAR and increases the probability of "newtq and "unanalyzed" events with consequences potentially more severe than the analyzed event.

It is our understanding that the RHR fuel pool cooling assist mode has not been used at S S E S and was not even successfully pre-operationally tested.


EDR G20020 Item F is a valid deficiency because current S S E S refueling operations routinely place the "emergency heat load8' in the spent fuel pool without a corresponding time to boil analysis.

9. Mr. Miller8 s response to EDR G20020 Item G on the radiological release calculation is inadequate for the reasons stated in Comment 7.

EDR G20020 Item G is a valid deficiency because the basis for the time to boil calculation reported in the FSAR is invalid, but as we clearly stated in our memo dated June 22, 1992 to Joe Zola (ET-0471) it does not affect the present operation of S S E S because the exis2ing decay heat loads in the fuel pool are less than 9.79~10 BTU/hr.

10. Mr. Miller's response to EDR G20020 Item H on nonconservative activity terms in the radiological release calculation is adequate. Based on the technical justification providd by Mr. Miller, we agree that EDR G20020 Item H is not a valid deficiency.

11. Mr. Miller's response to EDR G20020 Item I on maximum time prior to makeup is adequate. We agree that EDR G20020 Item I is not a valid deficiency, but we recommend that calculation 175-14 be either revised to clarify its purpose and usage or deleted.

12. Mr. Miller focused much of his evaluation on the ability of S S E S to withstand a loss of fuel pool cooling during a refueling outage. EDR G20020 did not emphasize this aspect to the same degree, although many of the problems are just as pertinent under this condition. In fact, the current S S E S practice of performing full core offloads each refueling outage places the station in a very vulnerable (and unanalyzed) condition. At the point when the common RHR system outage is entered on the unit in refueling, the operating unit's fuel pool cooling system is handling the entire heat load from the cross-tied fuel pools. A design basis LOCA/LOOP at this time subjects the station to a loss of fuel pool cooling at a time when one unit's RHR system is totally unavailable and the remaining unit's RHR system is dedicated to core and containment cooling functions. If realistic events such as the single failure of one RHR on the operating (LOCA) unit and/or installation of the fuel pool gates to the reactor cavity on the unit in refueling are considered, the potential consequences can be quite severe. In any case, this condition is routinely entered by S S E S without the necessary analyses to support it.

Attachment 29

PP&L Memo from Glenn D. Miller to George T . Jones, "Evaluation of EDR 620020 - Spent Fuel Pool Cooling Issue", October 29, 1992 (PLI-72763)

1 z. . ' . . . : b ..: . - ., ....-.&$..- i : * , 1 October 29, 1992

George T. Jones A6-2 I I

SUSQUEHANNA STEAM ELECTRIC STATION EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-72763 FILE A45-1A

Attached, please f i n d a copy o f NE-092-002, Rev. 0 "Loss o f Fuel Pool Cooling Event Evaluation". This evaluation i s provided i n response t o your memo PLI-72640. This repor t i s intended t o supplement my previous evaluat ion o f EDR 620020 transmit ted i n PLI-72711.

This document evaluates the various accident scenarios i d e n t i f i e d by EDR 620020 f o r which a loss o f fue l pool cool ing can be expected. I n addi t ion, t h i s document provides per t inent design basis informat ion w i th regard t o the Fuel Pool Cooling system and recovery from a loss o f Fuel Pool Cooling event.

This repo r t concludes t h a t recovery from a l l design basis accidents i s possible wi thout compromising safety re la ted equipment, whi le assuring t h a t the spent fue l pool w i l l remain s u f f i c i e n t l y cooled. Therefore, i t can be concluded t h a t the p lan t i s operable and continued safe operation o f the p lan t i s assured w i th regard t o the concerns ra ised i n EDR 620020.

This repo r t a lso recommends a number o f immediate and short term act ions (see Section 5.0) which w i l l enhance p lan t safety and reduce the r i s k t o the p lan t environment resu l t i ng from the events studied. Implementation o f these act ions i s required t o support the conclusions reached i n t h i s evaluation.

I f you have any questions regarding t h i s evaluation, please contact me a t your convenience.

A s . 4

Glenn D. M i l l e r

cc: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W. Simpson - A1-2 D. F. Roth - SSES H. G. Stanley - SSES J. M. Kenny - A2-4 J. S. Stefanko - A9-3 F. G. Bu t le r - A6-3 J. R. Mil tenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records - A6-2 D. A. Lochbaum - Enercon


PP&L E n g i n e e r i n g Repor t , "Loss o f F u e l Poo l C o o l i n g Even t E v a l u a t i o n f o r EDR #G20020", Oc tobe r 29, 1992 (NE-92-002 Rev. 0 )

Note: T h i s r e p o r t documents t h e e x t e n s i v e e n g i n e e r i n g e f f o r t u n d e r t a k e n b y PP&L t o e x p l o r e t h e l o s s o f f u e l p o o l c o o l i n g e v e n t . I t recommends s u b s t a n t i a l m o d i f i c a t i o n s and p r o c e d u r e changes i n o r d e r f o r t h e u n i t s t o h a n d l e a l o s s o f f u e l p o o l c o o l i n g under a l l d e s i g n c o n d i t i o n s . T h i s r e p o r t c o n c l u d e s t h a t t h e o p e r a b i l i t y o f t h e p l a n t i s n o t a f f e c t e d , b u t seems t o base t h i s c o n c l u s i o n o n t h e c o n d i t i o n s xEbx a l l t h e recommended m o d i f i c a t i o n s and p r o c e d u r e changes a r e imp lemented , n o t as t h e p l a n t e x i s t s c u r r e n t l y .

HE-092-002 R e v . 0

LO88 OF F W L POOL COOLING

EVENT EVALUATIOM

FOR

EDR $ 020020

10-29-92

P r e p a r e d by:

David G. Kostelnik

R e v i e w e d by:

1c;.s-.; L

Michael H. Crowthers

A p p r o v e d by:

LO88 OF FUEL POOL COOLING EVENT EVALUATION

W-T

I ,

This document evaluates .the various accident scenarios identified by EDR 620020 for which a loss of fuel pool cooling is currently not discussed in the FSAR or other design calculations. In addition to evaluating these events, this document provides pertinent design basis information with regard to the Fuel Pool Cooling system and recovery from a loss of fuel pool cooling event.

For each event a design basis and a realistic evaluation is performed in order to assess the impact of the environment that results from a boiling spent fuel pool on safety-related equipment in Zones I & 11. These evaluations also describe how the plant would recover from a loss of spent fuel pool cooling. The report also identifies what actions are required by the operators in order to assure recovery without compromising the operability of safety-related equipment in Zones I & II.

This report concludes that for all DBA LOCA events it is possible to recover without compromisinq the operability of safety related equipment required to mitigate the events, and assure that the spent fuel will remain sufficiently cooled. Therefore, it can be concluded that the plant is operable with regard to the concerns raised in EDR G20020. This report draws no conclusions with regard to the reportability of this EDR since this was outside the scope of this evaluation.


The author wishes to recognize the numerous individuals without whose efforts this report would not be possible. These individuals are commended for their efforts in providing timely quality information under adverse conditions.

Phil Brady Prakash Dave Kevin Browning Ron Vazquies Mark Saxon Chuck Dvorscak Dave Matchick Terry Mckay A1 Derkacs Juan Caj igas John Akus Don Kohn

Len West Greg Gogates Dale Roth Gerry Meartz Frank Gruscavage Tim Sweeney John Winders Tony Roscioli Fred Curry John Schleicher Tony Borger Glenn Miller


TABLE OF CONTENTB

.............................................. 1.0 purpose

.......................................... 2.0 Assumptions

3.0 Background ........................................... 3.1 Fuel Pool System Design ......................... ....................... 3.2 Fuel Pool Instrumentation 3.3 W A C ............................................ 3.4 DBA Acccident Dose .............................. 3.5 RHR Fuel Pool Cooling Assist .................... 3.6 Seismic and Hydrodynamic Design ................. 3.7 Flooding Effects ................................ 3.8 Recovery from Loss of Fuel Pool Cooling .........

4.0 Event Evaluations .................................... Event #1: Units 1&2 @loo% Power; ..................... WCA; Pools Separated

Event #2: Units 1&2 @loo% Power; LQCA/LOOP; Pools Separated ................

Event #3: Units 1&2 @loo% Power; ..................... LOOP; Pools Separated

Event #4: One Unit Operating. One Unit in a Refueling Outage; WCA/LOOP ...............

5.0 Conclusions & Recommendations ........................ 5.1 Conclusions ..................................... 5.2 Recommendations .................................

6.0 References ........................................... TABLES & F m

Table 1 : Unit 2 5RI0 Heat Loads and Time to Boil

Figure 1: Unit 2 Time to Boil with Pools Isolated

Figure 2: Time to Boil with Pools Connected

Figure 3: Unit 1 Time to Boil with Pools Isolated

Figure 4: SGTS Duct Routing

Figure 5: Loss of Fuel Pool Cooling Event Tree


ATTACHMENTS:

Attachment 1: Engineering Assessment of Fuel Pool Cooling Piping EDR-G20020

I I

Attachment 2: EP-548, dated 10-20-92, Radiological Evaluation in Support of EDR G20020

Attachment 3: Evaluation of Zone I11 Venting

Attachment 4: Time to Make-up for Fuel Pool

Attachment 5: PLI-72764, dated 10-29-92, Revised Evaluation of EDR G20020 -Spent Fuel Pool Cooling Issue

Attachment 6: PLI-72696, dated 10-20-92, Expected Number of Fuel Failures During the DBA LQCA

Attachment 7: ET-0870, 10-27-92, Use of UHS with RHR in Fuel Pool Assist Mode

Attachment 8 : ET-0871, dated 10-28-92, Drainage of Condensation from 818' Elevation During Fuel Pool Boiling

Attachment 9: ET-0750, dated 10-20-92, SGTS Fire Dampers

LOSS OF FUEL POOL COOLING EVENT EVALUATION

1.0 PURPOSE

The purpose of this- evaluation is to examine the various scenarios and configurations under which fuel pool cooling could be lost and to determinefthe resulting impact on the plant. For each scenario, an evaluation is performed under both design basis and current realistic operability assumptions. Particular emphasis is placed on how cooling to the fuel pool can be restored.

For each scenario, the following assumptions are made concerning the design basis and realistic evaluations:

Design Basis:

- DBA accident event scenarios. - Zone I, 11, & I11 HVAC operate/isolate as currently designed.

- Use of only safety-related or otherwise reliable equipment to recover from the event.

- Use of existing procedures as of 9-23-92. Current in-plant dose evaluations in the FSAR for contained sources.

. Extended loss of offsite power (224 hours). Time to boil is 25 hours. (Ref. 14)

Systems not hydrodynamic /seismic designed are unavailable.

Hydrodynamic loads act as originally analyzed affecting the non-LOCA unit with the same loading as the LOCA unit.

Realistic:

. Both units reactor buildings will remain accessible after 12 hours post-accident.

- The probability that offsite power can be restored in 24 hours is 99.53% per the SSES IPE.

- Plant procedures are changed to direct the use of all available means of cooling the pool prior to


allowing the fuel pool to boil as the means of cooling .

- The probability of a MCA/LOOP concurrent with both loops of RHR out of service in an outage is sufficiently small so asjnot to be considered.

Time to boil values are as identified in Table 1, and Figures 1-3.

Non-MCA unit will experience small hydrodynamic loads from the LOCA unit which are not expected to disable non-safety systems/equipment.

Fuel pool temperature, level, and skimmer surge tank level instrumentation will be available.

- Non-safety-related systems are assumed to be available post accident.

With a LOCA signal present, HVAC Zone I11 can be isolated from the recirculation plenum and operation of normal Zone I11 ventilation can be restored.

This section provides pertinent design information regarding the SSES design relative to the issues identified in EDR G20020. This information is used herein serving as input for the event evaluations presented in Section 4.0 of this report. The issues are evaluated from both a design basis viewpoint and a realistic viewpoint.

3.1 FUEL POOL SYSTEM DESIGN

The Fuel Pool Cooling system is designed to maintain the fuel pool water temperature below 125'F at a nnMaximum Normal Heat Load (MNHL) It of 12.6 MBtu/hr. The MNHL is defined as equivalent to a typical fuel cycle discharge schedule which fills the fuel pool with the last quarter core offload at 6.7 days after shutdown (Ref. 14). For the %mergency Heat Load (EHL)In condition of 32.6 MBtu/hr, the fuel pool water temperature can be maintained below 125°F by the Fuel Pool Cooling system in combination with the RHR system or by the RHR system alone. The emergency heat load is defined as a full core offload 10.5 days after a shutdown following a typical fuel cycle discharge -schedule which fills the fuel pool (Ref. 14) . The original design relies upon fuel pool boiling with makeup supplied by ESW as the nnsafety-relatednn backup to a loss


of Fuel Pool Cooling. The FSAR assumes a seismic event causes the loss of Fuel Pool Cooling (the system is not seismically designed). It is further postulated in the FSAR that the earthquake occurs during a refueling outage, 10.5 days after shutdown, at a time when the heat load is 9.8 MBtu/hr. In this situation, the pools are isolated and assumed to boil in 25 hours with make-up provided by ESW. The resulting radiological release was well within offsite dose limits specified by 10CFR100. The impact of a boiling pool on reactor building (Zone I & 11) equipment was not evaluated.

Note that the current heat load is much lower and thus the time to boil is longer than the original design assumptions as delineated in Table 1 and Figures 1 and 3.

3.2 FUEL POOL INSTRUMENTATION (Ref. 17)

Note that Unit 1 is discussed here. Unit 2 is identical.

The skimmer surge tank level (LT-15312), fuel pool level (LE- 15332), and fuel pool temperature (TE-15333) instrumentation strings would be utilized in a potential loss of pool cooling event. These instruments are powered from UPS units. UPS units maintain power to the instrument strings on loss of power until the diesels pick up the load. Therefore, in a LOOP scenario, these instruments will remain powered.

Skimmer surge tank level is alarmed and indicated in panel 1C206 and alarmed on panel OC211. 1C206 is located at elevation 749' and OC211 is located on elevation 818'. Several other alarms also exist in these panels (pump low flow, pump low pressure etc.). When one of the local panel alarms trip, group alarms in control room panels 1C651 and OC653 respectively also alarm. Thus the operators must go to the local panels to determine which alarm has actually come in.

The fuel pool level and temperature instrument strings are similarly configured. These, however, only alarm locally on the OC211 with a group alarm on control room panel 1C653. Fuel pool temperature indication is additionally provided on the local OC211 panel (818' elevation).

Fuel pool level can be determined visually on elevation 818. The acceptable minimum level would be above the bottom of the weir. The bottom of the weir is above the level minimally required to maintain 22 feet of water above the top of the fuel bundles per technical specifications. When on the 749 elevation, fuel pool level can be determined even though fuel pool level is not indicated on the 749 elevation. When makeup is being provided to the fuel pool (via ESW, etc), the


operator will know when the fuel pool level is above the weir when a rise in skimmer surge tank level occurs. This will indicate that the fuel pool level is at least above the weir and thus above the technical specification minimum. If the skimmer surge tank were to become full, then it will be necessary to drain a portion so that the level rise can be seen. This can be done by opening the 153068A valve located on elevation 779.

While the instrumentation is deemed adequate for determining fuel pool level (and meets the requirements of Regulatory Guide 1.13 as documented in the FSAR), it could be significantly improved to make mitigating the loss of fuel pool cooling easier for the operators. This could be achieved by providing safety related fuel pool level and temperature indication directly in the control room and on the 749 elevation on panel 1C206.

Having assured level and temperature indication on elevation 749 will preclude the need to possibly access the 818 elevation. With temperature indication in the control room, the operators will be able to track the pool temperature. This will allow the operators to determine when the pool might begin to boil. Similarly, with control room fuel pool level indication, the operators will be able to determine when actions to begin pool makeup will be necessary. These judgements will be able to be made from the control room instead of requiring entry of the reactor building to assess these parameters.

3.3 HVAC

The design of Susquehanna is such that the effects of a boiling pool coincident with a DBA LOCA are not considered in the design of the reactor building HVAC systems. Due to the common refueling floor, any moisture generated through evaporation and/or boiling of the fuel pool will be spread to the Unit 1 and/or Unit 2 secondary containment in a LOCA and /or LOOP. Excessive moisture in the ECCS Pump rooms of secondary containment could cause the room coolers to condense the moisture and remove less sensible heat in the room. This could result in excessive room temperatures potentially overheating pump motors, as well as, other equipment. The Emergency Switchgear Cooling System may also experience similar affects from the moisture.

It is worth noting that a similar condition was previously evaluated by PP&L as part of our response to NUREG-0803 concerning a SDV pipe break. While this event was not combined with a DBA LOCA to produce large radiation doses and had a lower heat input to the reactor building, it did produce a high moisture content and did result in


temperatures which exceeded EQ limits. A justification for interim operation was prepared (Ref. 22) and transmitted to the NRC demonstrating operability.

This mixinq of the ventilation zones raises an additional concern. With cooling to the f ~ e l pool during a LOCA and/or LOOP moisture will be spread throughout secondary containment with reactor building recirculation in operation. Calculations (ref. 13) indicate that the evaporation rate for both pools ranges from 60 gallons per hour at a pool temperature of 125'F with cooling in operation to 480 gallons per hour at a pool temperature of 210'F (no pool cooling). It is possible that high humidity conditions will be experienced in the ECCS pump rooms and switchgear rooms even without a loss of fuel pool cooling. Boiling the pool may greatly increase the molsture content and heat load in the reactor building. This may have an adverse affect on the environmental conditions experienced by safety-related equipment in Zones I and 11.

The environmental impact on Zone I and I1 can be averted by implementing actions to isolate these zones from the high humidity environment on elevation 818. This can be accom~lished by implementing the actions for Zone 111 Venting described in Attachment 3.

3.4 DBA ACCIDENT DOSE

DESIGN BASIS

Another important aspect of the original design basis relative to EDR# G20020 is postulation of fuel failure and its impact on equipment andaccess to the reactor buildings post accident. The current design accessibility evaluation for SSES is contained in FSAR Chapter 18. It uses methodology for determining "off site" dose to establish in- plant dose levels.

This approach is specified by NUREG-0737 (Ref. 19) which states that radiation levels should be determined on the basis of contained sources, and core damage source terms equivalent to those used for 10CFR100. The NUREG further states that airborne radioactivity from leakage of systems outside containment is not required to be analyzed for personnel access. While the NUREG is silent with regard to leakage from other containment penetrations ( e . hatches, airlocks, flanges, etc. which are lOCFR5O Appendix J "Type Bnl tested) it does recognize that intermittent access to the reactor building is necessary and requires that shielding for personnel access need only be considered. It can be inferred from this point that airborne leakage from "Type B"


penetrations need not be evaluated with regard to personnel access. This is reinforced by the fact that FSAR Chapter 18 only evaluates contained source radiation as stated in Section 18.1.20.2.1 and as shown on radiation zone maps. Also, note that radiation zones are based on a determination of the highest contact readi ,Par contaminated systems in that zone. This is very conse%ative.

FSAR Chapter 18 implies that the reactor building may be inaccessible due to airborne radiation. Based on the NUREG requirements, it appears this was a conservative statement made on the basis that access to the reactor building for operator actions was not required for first priority vital actions. This is further supported by the absence of a specific airborne accessibility calculation. Therefore, from a design basis perspective, the accessibility of the reactor building for operator actions required to recover from a loss of fuel pool cooling will be affected by contained sources only.

It should be noted that Attachment 2 evaluates the airborne radioactivity that would be present on the refueling floor for design basis conditions and reports large doses would be present. It should be further noted that the evaluation uses a highly conservative model and dose not restrict leakage to only that resulting from "Type B1' penetrations. The desiqn basis evaluation was done for information purposes only in order to compare the realistic evaluation (with airborne) to design basis evaluation which includes airborne radiation. Attachment 2 is not intended to. replace the NRC approved design basis for personnel access presented in FSAR Chapter 18.

Thus, per Chapter 18, access to .the refueling floor is possible post-accident. The radiation impact from contained sources is less than. 15 mR/hr which is well within the legal limit of 5 R for the duration of the accident specified in GDC 19. Access to the remaining 'areas of the reactor building will be severely hampered for many days following an accident with severe core damage due to the conservative source term assumption of 100% fuel damage.

Realistic

Access to the M C A unit reactor building and the refueling floor will be possible based on the fact that no fuel failures are expected from realistic LOCA Peak Cladding Temperatures. An evaluation (using NRC approved methodologies and computer codes) of actual fuel thermal response during design basis accidents results in no predicted fuel damage. Thus, the source term resulting from the DBA LQCA would only be equivalent to the radioactivity present in the reactor water through normal operation as

LOB8 OF FUEL POOL COOLING EVENT EVALUATION

permitted by Technical Specifications (Reference Tech Spec 3/4.4.5).

In order to perform a realistic, yet conservative in-plant accessibility evaluation for a DBA LQCA, an analysis was performed assuming 1% fuel cladding damage and consideration of airborne radiation (Ref. 18) . The results of this evaluation indicate that at 12 hours post-accident an airborne exposure of 1.4 R/hr would be expected on the refueling floor, while the reactor building will have an exposure of 47 mR/hr. Gross airborne activity concentrations tend to rise to a maximum at around 12 hours post-accident and holds constant until about 24 hours post-accident. The contained source exposure experienced by operators manipulating valves for the RHR Fuel Pool Cooling Assist mode would be 2-3 R/hr. Therefore, doses will be within the legal limits (Ref. 16).

3.5 RHR FUEL POOL COOLING ASSIST

A key element of the realistic evaluation of this document is taking credit for the use of RHR as a substitute for normal Fuel Pool Cooling. Original design of the RHR Fuel Pool Assist mode was intended to supplement Fuel Pool Cooling during refueling operations rather than replace it as the sole source of cooling, although it is fully capable of functioning by itself. This explains why numerous manual operator actions required to place RHR in the Fuel Pool Assist mode. It also explains why the piping connections to RIIR prevent use of both Fuel Pool Cooling Assist mode and shutdown cooling simultaneously.

The Ultimate Heat Sink (UHS) analyses performed to date do not account for the additional heat load resulting from the use of RHR FPC Assist coincident with a DBA WCA. These analyses can accommodate the additional heat load as long as two loops of spray arrays are available to dissipate the heat. These analyses will need to be updated formally to account for such capability in the future. This will, however, require additional operator actions to obtain optimum spray efficiencies. See Attachment 7.

Notwithstandinq the preceding, the RHR Fuel Pool Cooling Assist mode exists as a viable option for use with a reactor at power. There are numerous drawbacks to its use but they can be overcome with prompt operator action and clear procedures that provide for adequate make-up and control of the RHR pumps.


3.6 SEISMIC AND HYDRODYNAMIC EFFECTS ON FUEL POOL COOLING

The Fuel Pool Coolinq and support systems such as service water are not specifically designed to withstand the loads that result from an OBE or SSE. The supports consist mostly of spring and rigid rod type components. Given the magnitude of the reactor buildings seismic response at the elevations where the Fuel Pool Cooling System is located, there is a high probability that continued operation of the system will be compromised. The Fuel Pool Cooling System and Service Water System piping are not specifically designed to withstand the hydrodynamic loading that result from a LOCA. Given the magnitude of the hydrodynamic loads at the elevation of the reactor building where FPC is located, there is a high confidence that the system will remain operational. There is a reasonable probability that the support systems required for the fuel pool cooling system (service water) will be operational as well even though it has piping at lower elevations where hydrodynamic loads are greater. (Ref. 15)

3.7 FLOODING EFFECTS

One of the issues raised in the EDR concerns flooding effects from the water generated from a boiling spent fuel pool. It is suggested that as much as 60 gpm could accumulate on the refueling floor as a result of condensation. The 60 gpm value is equivalent to the design make-up from ESW for two boiling spent fuel pools. An assessment of where this flow would go is made in Attachment 8 (Ref. 20) . This assessment concludes that no significant water will leak down the stairwells, and that it will drain to the reactor building sump.

It is also reasonable to assume that a siynificant portion of that water will remain on the 818 elevation in areas such as the dryer/separator pits and wash down areas. In PP&L0s response to NUREG-0803 (Ref. 21), a flooding evaluation of the reactor building was performed for a SDV pipe break. This pipe break evaluated a leak of 550 gpm which tapered off to 43 gpm (for a single reactor building) long term and concludedthat safety-related equipment could remain functional until the water could be removed via reactor building sump pumps.

Based on these evaluations, it is concluded that flooding effects from a fuel pool boiling event are not significant and would occur over a sufficiently long period of time that operator actions can easily be performed to remove the accumulated water.


3.8 RECOVERY FROM LOSS OF FUEL POOL COOLING

As a result of the evaporation/boiling caused by a loss of fuel pool cooling, it will be, necessary to take numerous corrective actions. These actions are identified in Figure 5. While the following event evaluations do not proceed down each branch of the event tree, they do identify all of the actions listed on Figure 5. The actlons are discussed herein relative to plant scenarios rather than loss of Fuel Pool Cooling scenarios discussed in the event tree of Figure 5.

Besides providing makeup or cooling to the pool, one of the key elements to recovery is to provide isolation of HVAC Zone I11 from the recirculation plenum and provide a means of relieving pressure buildup from a boiling pool. This will require access to Zone I11 and power to the Zone I11 ventilation equipment.

NOTE: The evaluation which follows identifies the use of venting HVAC Zone 111. For the purposes of this evaluation, zone 111 Venting is defined as follows:

Isolating the connections between Zone I11 and the recirculation plenum and providing Zone I11 ventilation by running filtered exhaust only or unfiltered exhaust and supply if radiation levels permit.

Consequently, many of the assumptions in the realistic evaluations regarding accessibility and power restoration will be directed to the isolation of Zone 111.

Therefore, in this evaluation, success is considered to occur when the pool is prevented from boiling or by permitting boiling and preventing the environmental consequences of a boiling fuel pool on equipment in secondary containment.

This success criteria can be expanded to the following:

1. Do not allow pool boiling/evaporation to adversely affect proper function of safety-related equipment in the reactor buildings..

2. Pursue all available means for maintaining adequate pool cooling before using pool boiling and make-up.

3. Provide filtered exhaust to the Zone I11 environment.

4. Keep operator actions required in the reactor building at a minimum.

LOB8 OF FUEL POOL COOLING EVENT EVALUATION

4.0 EVENT EVALUATIONS

This section evaluates how a loss of fuel pool cooling could occur as the direct result of various Design Basis Accident scenarios. It also evaluates the different spent fuel pool configurations possible for each event.

Each event is divided ' into subsections which evaluate different aspects of the event. The information provided for each subsection is based on the following ground rules:

- LOCA UNIT: Informationis provided to describe how equipment pertinent to this issue will be affected by the event in the LOCA Unit. This information is presented from a neutral position with regard to design basis versus realistic.

WON-WCA UNIT:

Information is provided to describe how equipment pertinent to this issue will be affected in the Non-LOCA as a result of the event. This information is presented from a neutral position with regard to design basis versus realistic.

CONCERNS:

This section raises issues specific to the event being analyzed that will affect the evaluation of the event. These issues are factors that can affect both the realistic and design basis evaluations.

- Design Basis/Current Procedures Evaluation: This section evaluates the event under the design basis assumptions listed in Section 2.0 and other pertinent design basis information presented in Section 3. It also evaluates the event based on plant procedures in effect prior to the corrective actions identified in Section 5.2.

This section evaluates the event under the realistic assumptions listed in Section 2.0 and other pertinent realistic information presented in Section 3. It also evaluates the events

LOSS OF FUEL POOL COOLING EWNT EVALUATION

under the assumption that the corrective actions listed in Section 5.2 can be taken credit for.

The evaluations for each event contained in this section address all of the conditions identified in Figure 5 except for a seismic event. This case is omitted from this evaluation, since recovery from this event was not a concern in the EDR and it is already evaluated in Chapters 9 & 9A of the FSAR. It is also bounded by the following event evaluations.


NOTE: Power operation with the pools crosstied is bounded by the pools separated condition evaluated for this event as well as for events # 2 , 3 , & 4. Crosstied pools would result in a longer time to boil (compare Figures 1 and 2). Based on this, crosstied pools would represent a less severe condition to recover from than a pools separated configuration, and therefore is bounded by the evaluation presented here.

LOCA UNIT:

- Long term RPV cooling maintained by 1 loop of Core Spray and 1 loop of RHR in suppression pool cooling, with water leaking through the break to the suppression pool.

- One loop of RHR would be available for Fuel Pool Cooling (FPC) Assist mode.

- The availability of the normal Fuel Pool Cooling System will depend upon the extent hydrodynamic loads affect the piping and components in the fuel pool cooling and service water systems.

- The availability of ECCS keepfill to support RHR in the fuel pool cooling assist mode is dependent upon the extent the system is affected by hydrodynamic loads.

- LOCA load shed will also cause the fuel pool cooling s stem and its support systems to be lost, however, t f: ese systems can be restored.

NON-LOCA UNIT:

- All ECCS available. - May need the 88B18 loop of RHR for Shutdown Cooling

(SDC) as early as 6 hours into the event. However, Alternate Shutdown Cooling could be used which would eliminate the need to swap back and forth between FPC and Shutdown Cooling (SDC) . Other options include use of the condenser as the heat sink, or maintaining the reactor at pressure using HPCI/RCIC.

- Status of Fuel Pool Cooling will depend on the magnitude of the hydrodynamic loads from LOCA unit.


- Status of ECCS Keepfill will depend on the magnitude of the hydrodynamic loads from the M C A unit.

- RHR would be available for Fuel Pool Cooling Assist since Alternate Shutdown Caoling could be used to cool the vessel. Additionally, the "An loop of RHR would have to be used for FPC Assist in order to allow for suppression pool cooling via the "B" loop. This is due to the piping configuration which requires the discharge to the fuel pool through the "Aq1 loop only.

CONCERNB :

1. For design basis purposes, access to some areas of the reactor building could be limited for several days post-accident. The refuelin? floor will remain accessible at all times post-accident per chapter 18 of the FSAR.

From a realistic standpoint, access to the M C A unit reactor building and the refueling floor will be possible based on the amount of fuel failure expected from realistic LQCA analysis.

2. Loss of the keepfill system requires the use of an alternate system to fill the Fuel Pool to allow use of RHR Fuel Pool Assist mode (ESW could fulfill this function) .

3. The skimmer surge tank is marginally sized for the use of RHR at the flow rate prescribed by procedures. Potential is real for a loss of suction to the RHR pumps in this mode of operation. This in fact, happened at SSES during the test program where flows greater than 2000 gpm could not be attained without running the skimmer surge tank dry. (Ref. 4)

4. The time to boil is a key factor since it may allow for sufficient time to take actions necessary to place RHR into service for fuel pool cooling.

EVALUATION:

Design Basis/Current Procedures:

Under the strict design basis assumptions identified in Section 2 and the concerns identified above, it may not be possible to prevent the M C A Unit's Spent Fuel Pool from boiling. This is primarily because the normal fuel pool cooling system is assumed to be lost and the RHR Fuel Pool Cooling Assist mode procedure


requires the use of non safety-related equipment, such as ECCS keepfill. Additionally, numerous manual operator actions are required in potentially radioactive areas of the plant. The procedure for loss of fuel pool cooling, reference 10 would most likely lead the operator toward bailing given the uncertainty of establishing RHR as a viable cooling source.

Accessibility to the valves required to operate RHR in Fuel Pool Cooling Assist will not be possible for the duration of the accident due to the radiation that may be present from contained sources during an accident which results in severe core damage, as described in FSAR Chapter 18. Access to the ESW valves required for fuel pool make-up will also be greatly hampered due to radiation; although access to Unit $2 valves could be available within 30 hours. The Unit #1 valves remain inaccessible throughout the accident. Radiological doses, based on contained sources only, would not preclude actions to perform HVAC Zone I11 venting, thereby preventinq the moisture from reaching safety related equipment in Zones I and 11. Should significant airborne radiation be present, Zone I11 venting would not be possible.

The non-LOCA unit has some of the same concerns as the LQCA unit, however, there would not be an accessibility problem associated with the RHR or ESW valves.

Therefore, assuming all the above occurs during a DBA LQCA, the pools qre expected to boil. However, it appears possible that molsture could be prevented from reaching the safety related equipment in Zones I and 11. The existence of airborne radiation complicates the actions necessary for recovery during a loss of pool cooling event.

Realistic/Operability:

Access to the LQCA unit's reactor building and the refueling floor will be possible based on the fact that no fuel cladding failures are expected from realistic LOCA analysis. Conservatively assuming 1% cladding damage and primary containment leakage occurs at 1% per day, an airborne of 1.4 R/hr would be expected on the refueling floor at 12 hours post accident, while the reactor building will have an exposure of 47 mR/hr in a typical 20' X 20' x 20' room. The contained source exposure experienced by operators manipulating valves for the RHR Fuel Pool Cooling Assist mode would be 2-3 R/hr (Ref. 16). Therefore, any doses will be within the legal limits.


Operators must be made aware of the need to provide pool cooling by other than boiling if at all possible. Additionally, the operators will need to perform Zone I11 venting. Procedures need to be changed to reflect the need to place RHR into fuel pool cooling as early as possible and to identify ESW, Fire Protection and RHRSW as safety-related sources to fill the pool. Time to boil, thus time to take these actions are in excess of 60 hours for the upcoming operating cycle as shown in Figures 1-3. Time to makeup is estimated in Attachment 4.

The non-LOCA unit should be able to cool its pool via the normal Fuel Pool Cooling system, since hydrodynamic loads would be significantly less than the analytical model indicates. Also, the operators would have sufficient time to put RHR in Fuel Pool Cooling Assist mode if necessary. The operators should be made aware of the need to use Alternate Shutdown Cooling to permit the use of RHR in Fuel Pool Cooling Assist.

If the pools can be connected, time to boil will be extended. Additionally, it will allow use of the LOCA units pool cooling systems to cool both pools. This will alleviate the problem of swapping the NON-LOCA units RHR system between shutdown cooling and fuel pool cooling.

Therefore, it appears that the plant could prevent the pools from boiling, however, accessibility (Ref. 16) and the time it will take for the pools to boil (Table 1) are key factors in this judgement.

Note that should it not be possible to provide cooling to the pools, it would be possible to allow the pool to boil and vent zone 111.


EVENT 2: UN T &2 100 0

LOCA\NON-LOCZi UNIT:

- The situation will be more complicated than event #1 due to the LOOP. The LOOP will cause the HVAC recirculation system to mix all three (3) zones. For the duration of the LOOP, the following will not be available:

- ECCS Keepfill - Fuel Pool Cooling System - An ability to crosstie the pools Once power is restored, these will become available subject any possible LOCA impacts as discussed in event #1.

NO~-LOCA Unit:

The situation is made worse by the fact that normal Fuel Pool Cooling will be lost with the LOOP. This will necessitate use of RHR Fuel Pool Cooling Assist mode. Reactor building accessibility may also be hampered due to the three zone mixing of the ventilation system as a result of the LOOP.

ADDITIONAL LOOP RELATED CONCERNS:

1. The three zone mixing of the ventilation system could create a radioactive environment in the non-LOCA unit.

2. Assumption for length of time offsite power is lost is critical, especially for the non-LOCA unit where this will impact when RHR is utilized for shutdown cooling vs FPC Assist mode.

EVALUATION:

Design Basis/Current Procedures

The evaluation of the LOCA unit remains the same as event #l. It appears that the non-LOCA unit's pool may eventually boil due to accessibility concerns with regard to operation of ESW and RHR valves.

Therefore, the potential exists that the pools may


boil and have an adverse effect on the emvironmental conditions experienced by safety related equipment.

Realistic/Operability

The evaluation for the MCA unit remains the same as event #l. The non-MCA unit will lose normal Fuel Pool Cooling until power is restored (within 24 hrs). Since calculated time to boil exceeds the LOOP duration, operators can be expected to adequately restore normal Fuel Pool Cooling.

If normal Fuel Pool Cooling cannot be restored, operators must use Alternate Shutdown Cooling, to enable use of RHR FPC Assist. If this mode of RHR is not available, operators can establish a makeup source and allow the pool to boil. Initiation of Zone I11 venting will be necessary to prevent building pressurization and to provide a path for removal of moisture buildup on the refueling floor. Due to the three zone mixing during the LOOP, the non-LOCA unit must be isolated from the recirculation plenum to prevent the spread of moisture to its reactor building.

Therefore, it appears that the plant would adequately prevent pool boiling. However, the M O P has the potential to further complicate the plant response to this event for the non-MCA unit.


EVENT #3: BOTH UNITS @loo% POWER: LOOP: POOLS SEPARATED

UNIT 1L2:

Response will be the same as the non-MCA Unit in event #2, except that there will be no accessibility concerns. Operators will be required to use Alternate SDC and FPC for long term cooling.

ADDITIONAL CONCERNS:

None.

EVALUATION:

Design Basis/Current Procedures:

Accessibility is not an issue in this case; however, expected duration of the MOP may influence the operators decisions with regard to pool boiling. Safety related systems are available to provide pool cooling via ESW makeup and RHR Fuel Pool Cooling Assist. However, if the duration of the M O P is long operators may choose to go to cold shutdown and allow the pool to boil, as permitted by current procedures. This may be preferred rather than swapping the "B" loop of RHR between SDC and FPC operation, or using Alternate Shutdown Cooling.

Therefore, it is possible to prevent the pools from boiling during a LOOP without relying on non-safety systems.

As noted above, prevention of pool boiling can be achieved.


EVENT #4: ONE UNIT OPERATING, OTHER UNIT IN A REFUELING OUTAGE; MCA/LOOP

Note: A IX)CA/LOOP for the refueling scenario is the bounding event since it potentially disables more equipment than other accident events. Therefore, LOCA/LX)OP is the only event being evaluated.

As documented in Table 1, the U2-5R10 evolution bounded the FSAR Appendix 9A design basis loss of fuel pool cooling event. During an outage, the fuel pools are at times crosstied to each other and to the RX well, and at times isolated.

For a brief period during the U2-5RI0 outage, both loops of RHR were out of service for maintenance. During this period however, one loop of Core Spray remained in service providing a safety-related source of make-up. Additionally, the non- outage unit's FPC system and RHR system would be available to provide cooling to the crosstied pools. Furthermore, the probability of a LOCA/LOOP, concurrent with both loops of RHR out of service in an outage, is sufficiently small so as not to be considered.

During outages, the outage unit reactor building is normally isolated from the recirculation plenum, effectively isolating the outage unit from a potentlal LOCA environment. This allows accessibility to the outage unit reactor building should anything happen in the non-outage unit or on the refuel floor. With the pools crosstied, both pools can be cooled by either units pool cooling systems (via a Fuel Pool Cooling system or an RHR Fuel Pool Cooling mode). Since time to boil is relatively long and the outage units reactor building is accessible, the actions necessary to mitigate the loss of pool coolin9 event during a LOCA\MOP will be more readily achievable in an outage situation than during non- outage situations. Therefore, the previous event evaluations present more challenging situations than does an outage event.


5.1 Conclusions

Upon review of the preceding event evaluations, it is concluded that the most limiting event with regard to loss of fuel pool cooling is a LOCA/LOOP with both units at 100% power (event # 2 ) . This event results in the least amount of disabled equipment and provides the greatest challenqe to performin operator actions. As noted in Figure 5, it is possible ?n all cases to provide adequate pool cooling and protect safety related equipment in Zones I and I1 from the environment produced from a boiling pool.

The following conclusions result from this evaluation.

DESIGN BASIS

Based on the preceding evaluations, it appears that it would not be possible to prevent one or both of the spent fuel pools from boiling if one makes the desiyn basis assumptions identified in the introduction to this document. The principal assumptions that result in this conclusion are:

- Use of only safety-related equipment to recover from the event.

- Use of the referenced procedures. - Significant in-plant accident doses. - Extended loss of offsite power (224 hours). - Current time to boil projections (525 hours). - HVAC operation in recirculation mode with normal HVAC

unavailable.

Zone I11 venting, as described herein, can be accomplished to prevent the boiling pool from adversely affecting safety- related equipment in Zones I & 11. It should be noted that if significant airborne radition is present it will be necessary to evaluate Zone I11 venting against mixing Zones and using SGTS.

REALISTIC

The precedinq evaluations indicate that with extensive operator actions, actual time to boil, realistic accident source term and credit for non-safety related equipment and power, it appears that fuel pool boil can be avoided and HVAC


Zone 111 venting can be achieved under certain key assumptions:

- Both reactor buildings will remain accessible for a DBA LOCA to enable performance of operator actions.

- Offsite power can be restored within 24 hours. - Plant procedures are changed as indicated below.

5.2 Recommendations - The following enhancements to the SSES plant procedures and design are deemed necessary to assure mitigation of the events evaluated herein. These actions are categorized into "Immediate" and "Short Term" actions, based in part on the relative timeframe in which they can be accomplished. "Short termt1 actions generally represent modifications which will require some time to design and install. For each Itshort term1I action noted below, interim measures have been identified to enable time for implementation of a permanent design modification. Additional evaluations will be performed to determine the need for additional actions.

Immediate Actions:

1. Procedure changes to ON-135(235)-001, "Loss of Fuel Pool Cooling/Coolant Inventory" and OP-149 (249) -003, W H R Operation in Fuel Pool Coolinq Mode". These procedure changes are necessary to identify all available options and important parameters to be considered during mitigation of a loss of fuel pool cooling resulting from SSES design basis events (i. e., MCA/MOP) . The following changes should be considered as a minimum:

ON-135 (235) -001;

1. Identify the potential need to isolate Zone 111, as described herein, with reference to the appropriate implementing instructions.

2. Identify ESW and fire water as a potential source of fuel pool makeup.

3. Place greater emphasis on the importance of providing cooling to the pools rather than allowing the pools to boil.

4. General recognition of this evaluation. 5. Identify the potential need to periodically drain

the skimmer surge tank to provide a means of gross fuel pool level indication.

OP-149 (249) -003;

1. Identify fire water as a potential source of fuel


pool makeup. 2. Identify accessibility and timing concerns

discussed herein. 3. Prerequisites may not exist in the scenarios

discussed herein. 4 . General recoqnition of this evaluation. 5. Identify actions necessary to provide optimum UHS

spray efficiencies (for operation with two loops of spray arrays).

2. Procedure change to EP-PS-102 "Technical Support Coordinator Emeraencv Plan Position S~ecific Procedure". A new section will -be added to this position specific procedure which provides guidance for mitigation of the loss of fuel pool cooling during design basis events.

3. Operator training will be necessary to provide a complete understanding of the potential consequences of a loss of fuel pool cooling during design basis events. This training should include identification of all procedural changes noted above.

Short Term Actions:

1. Replace existing SGTS fusible links with links rated for 212'F or above. The current links are rated at 160'F - 165°F. As discussed in the SGTS evaluation herein, it is possible that this temperature could be reached as a result of a boiling fuel pool condition. In this case, Zone I11 is isolated and SGTS is in operation. It would be desirable to operate SGTS as long as possible, therefore it is recommended that these links be changed out to assure system operability. These links exist to assure 325°F air does not traverse a fire wall. The SGTS evaluation is contained in Attachment 3 and the fusible link evaluation is provided in Attachment 9. This action is not deemed to be "Immediate", because Zone I11 venting is an acceptable alternate to use of SGTS.

2. Install fuel pool level and temperature indication on the 749 elevation and in the control room. Refer to Section 3.2 above, for a detailed discussion on the fuel pool instrumentation. This is not considered to be an I1ImmediateB8 action, because of the ability to indirectly determine fuel pool level via the skimmer surge tank indication on the 749 elevation (refer to Section 3.2 above).

3. Install remote actuation capability for ESW valves 153500 & 153501. As discussed in Section 2.0 above, these valves are located in high radiation zones within the reactor building. To preclude the need for operator access to these zones, remote operating capability is recommended


for these valves. This is not considered an "Immediate" action, because alternate means of fuel pool makeup will be available (i. e. , fire hose) .

LOSS OF FUEL POOL COOLING EVENT EV?aLUATION

6 .0 REFERENCES

P&IDs M-151: M-2151: Residual Heat Removal.

p&IDs M-153: M-2153: Fuel pool cooling and Clean-up.

P&ID M-154: Fuel Pool Filter Demineralizer.

PLI-72367, dated 9-9-92, Letter from J.R. Miltenberger to G.T. Jones concerning Spent Fuel Pool cooling.

PLI-72288, dated 9-1-92, Letter from K.W. Brinckman to G.T. Jones transmitting an Independent Review of the Fuel Pool Cooling Issues identified in EDR# G20020.

EDR # G20020: ebLoss of Spent Fuel Cooling Event Design Discrepancies".

FSAR Chapter 9.1: Fuel Storage and Handling

FSAR Chapter 18.1.20: Plant Shielding in Response to NUREG-0737.

OP-249-003, Rev. 8: RHR Operation in Fuel Pool Cooling Mode, Unit #2. (typical of both units)

10. ON-235-001, Rev.11: LOSS of Fuel poll Cooling/Coolant Inventory, Unit #2. (typical of both units)

11. OP-235-001, Rev.12: Fuel Pool Cooling and Cleanup System, Unit #2. (typical of both units)

12. GO-200-010, Rev. 6: ECCS/Decay Heat Removal in Condition 4, 5, or Defueled. (typical of both units)

13. M-FPC-010, Rev. 0: Calculation of Spent Fuel Pool Time To Boil and Evaporation Rate.

14. FSAR Appendix 9A

15. Fuel Pool Cooling Piping Assessment.

16 EDR #G20020 Radiological Evaluation.

17. Fuel Pool Instrumentation References:

SKIMMER SURGE TANK

a. J153 b. El67 SHT 4 C. E325 SHT 10 d. E323 SHT 8

MSS OF FUEL POOL COOLING EVENT EVALUATION

FUEL POOL

e. FF10451 SHT 4201 f. El67 SHT 6 a. J453 SHT 3 6 . J~AC-142 SHT 1 i. E331 SHT 11 j. E371 SHT 19 k. E25 SHT 4

GENERAL REFERENCES:

1. SEIS INSTRUMENT INDEX m. P&ID M153

18. PLI-72696

19. NUREG 0737

20. ET-0871

21. PLA-1132 (NUREG 0803 SUBMITTAL)

22. PLI-20806 (NUREG 0803 JIO)

23. ET-0870

24. Calculations E-AAA-718 AND E-AAA-719


TABLE 1

UNIT 2 5RI0 HEAT MAD8

AND

TIME TO BOIL


TABLE 1: U2 5RI0 HEAT LOADS AND TIME TO BOIL

TIME HEAT LOAD TIME TO BOIL (DAYS) (MBTU/HR) (HOURS)

110 - - 125

0 - 9/14/92 U2 = .75 >340 340 * Pools are isolated. No fuel is offloaded. U1 = .8 >340 340

18 - 9/30/92 * U1 FPC cooling both 2 0 pools. U2 defueled. Pools crosstied and connected to RX well.

38 - 10/19/92 U2 = 5.65 * RPV refueled. 1/3 of U2 core left in U2 Ul = .8 pool. Pools are isolated.

57 - 11/7/92 * u2 S/U. Pools are isolated.

NOTES :

1. Time in days from U2-5RI0 outage schedule 8/28/92.

2. Heat load based on measured values to the extent available at the time of this evaluation and on calculated projections. Measured values from TP-235-011 and calculated values from NFE-B-NA-048, rev. 3 which provided the heat load of 1/3 of the core as a function of time. See M-FPC-010.

3. Time to boil from M-FPC-010 assuming initial pool temperatures of 125 and 110 F. Values identified should be viewed as the minimum time to boil. Due to conservative nature of the calculation methodology, actual time to boil will be in excess of the values provided.


NE-092-002 Rev. 0

FIGURE 1

UNIT 2 TIME TO BOIL

POOLS ISOLATED

U2 FUELPOOL TIME TO BOIL

1 I I I I 0 , I

4.18€1(16 3.79E46 350E*06 3mE*06 2.75E 46 2Jl9E46 4.6SE106 HEAT LOAD BNMR


FIGURE 2

TIME TO BOIL

POOLS CONNECTED

TIME TO BOIL FUEL POOLS CONNECTED

A

A/

100 -

60 -

0 1 4.59E+06 4.30E + 06 3.80E +06

HEAT LOAD BTUlHR


NE-092-002 REV. 0

FIGURE 3

UNIT 1 TIME TO BOIL

POOLS ISOLATED

TIME TO BOIL - HOURS


NE-092-002 Rev. 0

FIGURE 4

8GT8 DUCT ROUTING


FIGURE 5

I088 OF FUEL POOL COOLING

EVENT TREE

Figure 5a - Loss of Fuel Pool Cooling Due to Seismic Event

Use RHR FFC assist Wfih Alternate SDC M available.

otherwise:.

AIIOW to boil with ESW makw. e w SGTS shldavn is rewired,

then:.

LOSS oi Fuel Pod Coding. I Use RHR FFC assist w l h Alternate SDC ll available.

otherwise:.

8 Zlil vmed.

Figure 5c - Loss of Fuel Pool Cooling Due to Other Causes

Loss of Fuel Pod Coding. a Restart RB HVAC. a

Restore FPC with mm4 systems.

Use AHR FPC assist wlth Allemate SDC II avdlable.

othanvise:.

\ AIIOW to boil with ESW makew 6 2111 vected. \


ATTACHMENT 1

ENGINEERING ASSE88MENT OF

FUEL POOL COOLING PIPING

EDR-020020

ENGINEERING ASSESSMENT OF FUEL POOL

COOLING PIPING EDR-620020

Reviewed by: Date: I o / z o / ~ ~

Approved by: Date: , 0 / 6 / / ? ~

Rev. 1 10/20/92

Background

A pre l iminary engineering assessment has been made o f the SSES Fuel Pool Cooling p ip ing t o judge i t s a b i l i t y t o withstand earthquake and hydrodynamic loadings. St ructura l i h t e g r i t y and system o p e r a b i l i t y were considered. This assessment i s based upon a review o f approximately 25 FPC fab r i ca t i on isometrics and associated pipe support drawings. Also reviewed were approximately 15 other fabr ica t ion isometrics and the associated pipe support drawings f o r the service water and condensate systems.

The conclusions reached as t o the ope rab i l i t y o f the FPC and i n te r fac ing p ip ing have been based, t o a la rge degree, on engineering judgement. This assessment was performed i n order t o provide a pre l iminary assessment o f EDR 620020.

. Current Design

A l l o f the Fuel Pool Cooling p ip ing which comprises the primary system f low path was reviewed. This includes 6" , En, and 10" l i n e s extending from the skimmer surge tank, through the heat exchangers and pumps and back t o the spent f ue l pool; reference PEID M-153. Almost a l l o f the subject p i p ing i s ASME Class 3, however, a review o f the FPC Design Report, DR-114, shows t h a t only a small por t ion o f the system has been dynamically analyzed, i .e. a sect ion o f p ipe coming from the surge tank and the pipe d i r e c t l y adjacent t o the fue l pool. The ma jo r i t y o f the FPC p ip ing has not been dynamically analyzed, i .e. seismic and hydrodynamic loads not analyzed.

The Un i t I isometrics reviewed contained approximately 1000 1 inear f ee t o f FPC la rge bore piping, t y p i c a l l y o f 6" , 8", o r 10" diameter. There are also a number o f small bore branch l i n e s connected t o the main runs which were not i n d i v i d u a l l y reviewed. The Un i t I 1 FPC p ip ing i s s i m i l a r i n quant i ty, s ize and layout.

The FPC non-seismic p ip ing i s supported using spr ing can supports, r i g i d rod supports and a t some l i m i t e d locations, r i g i d s t ru ts . The design o f these supports included deadweight and thermal expansion loads only. No evidence o f the consideration o f dynamic loads f o r "two-over-one" concerns has been found. Also noted i n the review was the lack o f many ana l y t i ca l anchor po in ts which t y p i c a l l y i so la te branch connections and thereby r e s u l t i n smaller, more manageable analy t ica l boundaries. There i s a t l e a s t one branch l i n e which cross-connects w i t h Un i t I1 p ip ing t h a t i s not anchored o r a n a l y t i c a l l y iso la ted.

The service water p ip ing supplying the fue l pool heat exchangers was also reviewed, reference PEID M-110. This p ip ing i s supported using p r i m a r i l y spr ing can hangers and r i g i d rod hangers. The pipe ranges i n s ize from 4" t o 14" diameter. I n addi t ion, the condensate l i n e s supplying fue l pool

Page 2 o f 5

pump seal ing water were reviewed , reference P&ID M-108. These l i n e s are a lso non-seismic p ip ing s im i l a r t o FPC and SW. This p ip ing cons is ts o f mainly small bore diameter (1" diameter) p ip ing. This p ip ing was p r i m a r i l y f i e l d supported, therefore, no hanger d e t a i l s o r support conf igurat ions ex i s t .

Seismic Loadings

It i s the judgement o f the reviewer t h a t there i s a high r i s k t o system o p e r a b i l i t y i f the FPC and in te r fac ing p ip ing experiences seismic (OBEISSE) loadings i n i t s current design conf igurat ion. Areas o f primary concern include pipe support adequacy, pipe stresses, equipment 1 oadings and system spat ia l in te rac t ion .

As was stated e a r l i e r , the ex i s t i ng pipe supports are designed f o r deadweight and thermal loads. A1 though there are design margins inherent i n these supports i t i s l i k e l y t h a t the capacity o f some o f these supports may no t be s u f f i c i e n t t o accommodate seismic loads. A hanger f a i l u r e would impose higher loads on adjacent supports which i n t u rn could f a i l . This "unzipping" e f f e c t would obviously create major system challenges. I n addi t ion, some o f the supports reviewed consisted o f one way deadweight supports, e.g. the pipe res t i ng on a spring can load p la te . This type o f hanger conf igurat ion could al low the pipe t o s l i d e o f f dur ing a s i g n i f i c a n t seismic event since la rge side-to-side displacements may be present.

Pipe stresses produced by la rge dynamic displacements could be large. High pipe stresses i n themselves do not necessari ly lead t o pressure boundary f a i l u r e . I n fact , recent t es t s conducted by EPRI and GE Nuclear demonstrated t h a t p ip ing systems are very res i s tan t t o gross f a i l u r e , even when exposed t o dynamic loads f a r i n excess o f any c red ib le SSE. I n addi t ion, the actual performance o f p ip ing i n past earthquake ind icates t h a t i n e r t i a l stresses i n pipes are not cred ib le r o o t causes o f p ipe f a i l u r e ( r e f . EPRI-NP-5617). However, la rge stresses can adversely a f f e c t system funct ional capab i l i t y , i .e. the a b i l i t y o f the system t o d e l i v e r ra ted f low. The r e s u l t would be an i n t a c t pressure boundary but a t reduced flow.

Another area o f seismic vu lne rab i l i t y o f the FPC system i s equipment loadings. This includes nozzle allowables and hold down anchorage capaci t ies o f the heat exchangers and pumps. The seismic adequacy o f system valves which must be operable fo l lowing the event i s another issue tha t would need t o be addressed, however, i t i s judged t h a t the manual values i n the FPC system would remain operable.

Spat ia l i n te rac t i on o f the FPC and in te r fac ing p ip ing would need t o be considered. This concern i s associated w i th the close prox imi ty o f other p lan t systems and components. A p lan t walkdown would be required t o i d e n t i f y po ten t i a l areas o f pipe damage due t o predicted dynamic displacements o f the system.

Page 3 o f 5

. Hydrodynamic Loadings

Fuel Pool Cooling

It i s the judgement o f the reviewer t ha t there i s a low r i s k t o system o p e r a b i l i t y i f the FPC p ip ing experiences hydrodynamic loadings i n i t s current design conf igurat ion. This judgement i s based on the f o l l owing.

Hydrodynamic events due t o a LOCA are most impacting f o r those p ip ing systems located ins ide containment, attached t o containment o r located a t lower elevat ions i n the reactor bu i ld ing. The FPC p ip ing i s located i n the reac to r bu i l d i ng a t r e l a t i v e l y high p lan t elevations (above 764') where the effect o f hydrodynamic events i s much less. A review was made o f the FPC p ip ing which was analyzed for both seismic and hydrodynamic loads a t these elevations. It was found tha t the load con t r ibu t ion due t o hydrodynamic events resu l ted i n approximately a 25% maximum increase over p ipe support deadweight loads. I t i s l i k e l y t ha t the FPC supports have enough reserve capacity t o accomnodate a 25% increase i n loads. The reduced e f f e c t due t o hydrodynamic loads on supports implies a reduced e f f e c t on pipe s t ress and equipment.

It i s the judgement o f the reviewer tha t there i s a moderate r i s k t o system o p e r a b i l i t y f o r i n te r fac ing systems such as service water and condensate. This judgement i s based on the fol lowing.

These systems are located a t lower elevations o f the reactor bu i l d i ng where the e f fec t o f hydrodynamic loads are higher. A review was made o f s i m i l a r p i p ing located low i n the reactor bu i ld ing which was dynamically analyzed t o determine the load con t r ibu t ion due t o hydrodynamic loads. The load increases were only s l i g h t l y la rger than found a t upper reactor bu i l d i ng elevat ions. It i s judged tha t the supports have s u f f i c i e n t design margin t o accommodate the increases due t o hydrodynamic loads. As w i th the support loads, i t i s ant ic ipated tha t pipe stresses and equipment loadings would be s l i g h t l y l a rge r a t these lower elevations but not o f a magnitude which would a f f e c t system operab i l i t y .

These i n te r fac ing systems are judged t o be s l i g h t l y more vulnerable t o hydrodynamic loads due t o the increased magnitude as described above, the quant i t y o f p ipe and the uncertaint ies associated w i th f i e l d supported p ip ing, i .e. no hanger d e t a i l s o r support conf igurat ions avai lable.

Conclusions

Based upon the prel iminary review performed on the Fuel Pool Cooling Pip ing and i n te r fac ing systems as described above, i t i s concluded t h a t there i s a h igh r i s k t o system ope rab i l i t y when considering the po ten t ia l e f f e c t s o f seismic loadings (OBE/SSE). It i s judged tha t pipe support modi f icat ions would be requi red t o ensure system qua1 i f i c a t i o n .

Page 4 o f 5

The e f fec ts o f hydrodynamic loads on the FPC p ip ing and i n t e r f a c i n g systems i s much less impacting than those associated w i th seismic events. The r i s k t o the ope rab i l i t y o f these systems due t o hydrodynamic loads i s judged t o be low t o moderate. A f i e l d walkdown o f the subject p ip ing could i d e n t i f y areas o f dynamic f r a g i l i t y which are not evident on design drawings.

The judgements presented here are made i n consideration o f the l a rge amount of p ip ing/p ipe hangers involved, thefvar ious support types i n s t a l l e d and the m u l t i p l e challenges t o system ope rab i l i t y inherent i n non-seismic designed systems when subjected t o dynamic loadings. The de ta i l ed reviewslanal y s i s and wal kdowns necessary t o ensure operabi 1 i t y o f the FPC system, inc lud ing a1 1 i n te r fac ing systems, would requi re considerable resources t o complete.

Page 5 o f 5


NE-092-002 Rev. 0

ATTACHMENT 2

RADIOLOGICAL EVALUATION

IN SUPPORT OF EDR# 020020

(EP-548)

M E M O R A N D U M

TO: G . D. Miller, A6-3 DATE: October 21, 1992

FROM: D. A. Matchick, A6-2m COPIES: C . J . Kalter, A9-3 K. E. Shank, A9-3

JOB :

FILE :

NUMBER: EP - 548

REPLY: NO

SUBJECT: Radiological Evaluation in Support of EDR G20020

As per your request, radiological engineering evaluations of in- containment and off-site doses have been performed in support of the resolution of EDR G20020. The attachments to this memorandum summarize the results of those evaluations. Documentation packages containing the computer code runs and detailed results of these analyses will be transferred to you under separate cover.

MILMEM. WPF

Radiological Evaluation

I. Airborne Doses

The TACT5 computer code was run to provide realistic estimates of radioactivity concentrations that might be expected on the Refuelling Floor (818') and a representative room (20'x20'~20'). The TACT5 code is validated and verified for design work under EPM-QA-104. The Chapter 15 DBA-LOCA model as obtained from PP&L calculation FX-C-DAM-014 was used as the base model, with a realistic estimate of 1% cladding failure as the source term. The 1% fuel cladding failure parameter was provided by Nuclear Fuels. A TACT5 run was conducted for a 100% cladding failure, and the activity concentrations were linearly scaled back to 1% cladding failure. Activity concentrations were then input into the MICROSHIELD 3.0 computer program to compute gamma exposure dose rates in the subject areas. The MICROSHIELD 3.0 computer program is validated and verified for design use under EPM-QA-104.

From an earlier analysis, SE-B-NA-074, it is known that gross airborne activity concentrations tend to rise to a maximum at around 12 hours post-accident, and hold constant until about 24 hours post accident in the reactor building. This is because the activity flow out of the drywell into the reactor building competes with radioactive decay processes to hold the gross activity concentration rather constant over this period. As such, the total activity concentrations of iodines, noble gases, and cesium at 12 hours were used in the MICROSHIELD analyses for airborne dose rates.

For the Refueling Floor, the exposure dose rate at 12 hours post-accident for the above scenario with 1% cladding failure was found to be 1.411 R/hr. This dose considers gammma only. Protective clothing to shield the beta component will be required.

For a representative 2O'x20'x20' room in the reactor building, the exposure dose rate for the above scenario with 1% cladding failure was found to be 46.8 mR/hr. This dose considers gamma only. Protective clothing to shield the beta component will be required.

The TACT5 code was subsequently used to determine doses on the refueling floor from a design basis (conservative) source term. This source term consists of 100% of core noble gases and 25% of core iodines being released into primary containment instantaneously at the start of the accident. Gamma exposure doses on the refueling floor from the design basis containment leakage rate of 1.0 % per day and a more realistic .25% per day were calculated for various times post accident. The gamma exposure doses at the center of the refueling floor as a function of time are:

Time Post-accident Dose @ 1% / day Dose @ .25% / day

4 days 30 days 45 days 60 days

11. Contained Sources

It is postulated that a valve in the RHR Pump room will be required to be manually manipulated. The approximate dose rate at the valve location at 12 hours post accident was determined. In this case, the source is radioactivity in suppression pool water from 1% failed fuel cladding which is contained in RHR piping . The analysis consisted of using the MICROSHIELD 3.0 computer program to estimate the exposure dose rate from two sections of piping, a long section below the area the operator would be working in, and a shorter section of pipe where the valve would have to be manually opened.

The exposure dose rate from the long section of pipe was 471 mR/hr at the point the operator was projected to be working. The dose rate to the operator from the short section was estimated to be 629 mR/hr. The total of these doses is 1.1 R/hr. To allow for additionalunanalyzed source contributions, it is recommended that the estimated dose be increased to 3 R/hr.

The suppression pool activity decays away continuously, therefore the dose rates from the suppression pool activity decrease with time. Calculations with other piping models indicate that the exposure,dose rate will be approximately 3 times higher at 2 hours post accident ( estimated exposure dose rate = 9 R/hr @ 2 hours), and at 24 hours, the dose rate is approximately a factor of 1.5 lower ( estimated exposure dose rate = 2 R/hr @ 24 hours)

111. Offsite Consequences Of Reactor Building Purge

The radiological dome consequences of purging Zone I11 of the reactor building under design basis LOCA accident conditions for the offsite Low Population Zone and Control Room were evaluated. It was assumed that all activity contained in the Zone I11 would be exhausted to the environment instantaneously, and the additional dose consequences were calculated using a spreadsheet program. The results were then tabulated. Note that the doses shown assume a purge with no charcoal filtration. If filtration can be used the thyroid doses should be multiplied by a factor of (1.0 - f) where f is the filter fractional efficiency. As in the above analyses, the TACT5 computer code was run to provide activity inventories in the reactor building as a function of time post-LOCA. The FSAR Chapter 15 DBA-LOCA model was used for this evaluation.

Two cases were analyzed. In Case 1, The drywell is assumed to leak at its design leakage rate of l%/day.In Case 2, The drywell leakage rate is 0.25%/day which is an actual calculated average leakage rate for Post-LOCA conditions. Off site and control room doses for purging were calculated assuming that all of the activity in ZONE I11 of the reactor building at the time purging begins is released instantaneously to the environment. The doses shown below comprise an additional contribution to be added to the current doses calculated for the Chapter 15 design basis LOCA.

TABLE 1. Additional DBA LOCA Doses Due To Purge Of Reactor Building ZONE III--- Drywell Leakage = 1% / day

ADDITIONAL DBA LOCA DOSE DUE TO ZONE I11 PURGE

Time At Which Purge occurs

I

60 day I 2.13 I 0.0003 0.380 I neg . I 0.030 11

4 day

30 day

45 day &-

Dose AT LPZ (REM)

489.2

38.5

9.06

Control Room Doses (REM )

THYROID THYROID WHOLE BODY

0.092

0.005

0.001

WHOLEBODY

87.2

6.86

1.62

BETA SKIN

0.020

0.0005

neg .

0.991

0.064

0.038

TABLE 2. Additional DBA LOCA Doses Due To Purge Of Reactor Building ZONE 111---Drywell Leakage = 0.25% / day

ADDITIONAL DBA LOCA DOSE DUE TO ZONE I11 PURGE

Dose AT LPZ Control Room Doses (REM) (REM)

I I

Time At Which Purge THYROID WHOLE BODY THYROID WHOLEBODY BETA SKIN occur8

24 hr 276.4 0.112 49.3 0.020 0.826

4 day 192.5 0.034 34.3 0.006 0.305

30 day 17.6 0.0022 3.14 0.0002 0.024

45 day 4.53 0. 0005 0.807 neg . 0.016

60 day 1.16 0.0001 0.208 neg . 0.014


NE-092-002 Rev. o

ATTACHMENT 3

ZONE I11 VENTING EVALUATION


For the purposes of this evaluation, this process will be defined as Zone I11 Venting as following:

Isolating the connections between Zone 111 and the recirculation plenum and providing Zone I11 ventilation by running filtered exhaustTonly or unfiltered exhaust and supply if radiation levels permit.

This section examines the various aspects of how to provide Zone I11 venting and ensure that it is a viable option that will not compromise other plant systems. The areas evaluated are:

- Zone I11 Venting - Fuel Pool Cooling Instrument Air Supply - 480 VAC Circuit Breaker Changeout For Zone I11 Venting - SGTS Operability - Assessment of Emergency Ventilation Options for Zone

I11 Venting during Fuel Pool Boil Scenarios

- EQ Equipment Located in HVAC Zone I11

FUEL POOL COOLING

ZONE I11 VENTING

1) OPTION I - Isolate Zone I11 from the Recirculation Plenum and vent

1.0 Isolate Zone I11 from the Recirculation Plenum by:

1.1 Blanking off the Recirculation Supply Plenum (closing balancing damper is not adequate)

1.2 Mechanically lock close the following Back Draft Dampers to isolate the Recirculation Plenum:

1.3 Jumper from PDY-07554A4 Terminal 9 to PDY- 07554A Terminal 6

2.0 Block Open Exterior doors on Elevation 818 or open hatches.

2) OPTION I1 - Isolate Zone I11 from Recirculation Plenum and start 1V2178, 1V212B and 1V213B





2.0 Open the Dampers

3.0 Jumper out LOCA contact in Isolation Stop Logic for 1V212B and 1V217B (E-192 shl)

4.0 Start Zone I11 HVAC. If HVAC does not start due to Loss of Off-site Power provide Class 1E power to 1V212B, 1V213B and 1V217B

5.0 Provide Class 1E power to Zone I11 W A C Fans

5.1.1 Install and connect cable from 1B227 to 1B261-024 (1-3/C #2)

5.1.2 Change wiring in 1B227 and 1B261-24 cubicles to remove O/L and contactor contacts from power circuit

5.1.2 Block open all circuit breakers on 1B261 except 1B261-024

5.1.3 Close the 1B227 supply circuit breaker and 1B261-024 circuit breaker

5.2 1V212B and 1V213B

5.2.1 Install and connect cable from 1B220-034 to 1B280-023 ( 3-1/C # 500Kcmil)

5.2.2 Block open all circuit breakers on 1B280 except for the 1V212B and 1V213B circuit breakers

5.2.3 Close the 1B220 and 1B280-024 supply circuit breakers

6.0 Start Zone I11 HVAC Train B

3) OPTION I11 - Isolate Zone I11 fromthe Recirculation Plenum and start 1V217B Filtered Exhaust Fan





2.0 Open the following Dampers to 1V217B:

3.0 Jumper out IX)CA contact in Isolation Stop Logic for 1V212B and 1V217B (E-192 shl)

4.0 Jumper out the following contacts in 1V217B control:

5.0 Start 1V217B. If HVAC does not start due to Loss of Off-site Power provide Class 1E power to lV217B

6.0 Provide Class 1E power

6.1.1 Install and connect cable from 1B227 to 1B261-024 (1-3/C $2)

6.1.2 Change wiring in 1B227 and 1B261-24 cubicles to remove O/L and contactor contacts from power circuit

6.1.3 Block open all circuit breakers on 1B261 except 1B261-024

6.1.4 Start 1V217B

FUEL POOL COOLING

INSTRUMENT AIR SUPPLY

Instrument Air is required to hold open the following Dampers for Zone I11 HVAC 'B' Fans:

1V212B 1V217B

Instrument Air also supplies the instruments that actuate FSL-17512B which is used in the control of 1V217B.

, The Instrument Air Compressors and Controls are power from Diesel Generator Backed supplies. In the event of Loss of Of f-site Power, the Instrument Air Compressors and Control have electrical power.

The Above Zone I11 Dampers are powered from Diesel Generator backed supplies . SCENARIO

The above listed Zone I11 Dampers for the *B9 Fans can be opened by using normal procedures to open the Dampers.

MCA/MOP with no Seismic Event

The control logics for the following Dampers have LOCA actuation contacts in the circuits which close the Dampers and prevent their opening for a MCA. These contacts MUST be defeated so the Dampers can be opened by normal procedures.

The other listed Dampers open when their respective Fans start.

Seismic Event

The Instrument Air System is considered lost for a seismic event since the tubing in the system is not designed with seismic support.

For this event the above listed Dampers must be mechanically locked open in preparation for starting the Zone I11 1V212BI 1V213B and 1V217B Fans.

EDR GZOOZO 480VAC CIRCUIT BREARER CHANGEOUT

FOR' ZONE I11 VENTING

In order to provide Class 1E power to the Zone I11 HVAC Fan 1V217B, an f'Isolation System" must be install between the Class 1E equipment and the Non Class 1E equipment. To accomplish this the circuit breakers, terminating the new cable between 18227 and 1B261 for the Class 1E power supply to 1V217B, must be replaced with:

Ratinq

TYPE HFB-TM 60,70,80,90 or lOOA


SGTS OPERABILITY

In a postulated seismic event that presumably causes the loss of the fuel pool cooling system, Zone I11 will be isolated (normal Zone I11 HVAC shutdown, Zones I and I1 isolated from recirc plenum) and the SGTS will be used to maintain Zone I11 pressure (overpressure control). No radiation source terms will exist in Zone 111. The only possible source term that might exist would be from leaking fuel in the fuel pool. Currently five ( 5 ) leaking fuel rods are maintained in the pool. This is within that assumed in Appendix 9A. Thus the analysis contained in FSAR Appendix 9A is bounding (this will

- need to be confirmed each cycle). Note that the Appendix 9A offsite dose analysis takes no credit for operation of SGTS. Even without SGTS, it still concludes that offsite doses are acceptably low.

In the scenario evaluated here, assuming no pool cooling is established, it is possible that the SGTS system will be operating with incoming Zone I11 air conditions created by the high temperature pools. This could result in elevated Zone I11 air temperatures and high humidity.

Carried to extremes, these conditions could result in condensation collecting in the duct work and SGTS filter train. It is expected that it would take a significant amount of time for this to occur. This is in part due to the large time to boil that exists during the current operating cycle (See Figures 1 and 2). It is deemed probable that pool cooling would be reestablished prior to this condensation becoming threatening to the SGTS train components or ductwork.

If during this scenario SGTS is rendered inoperable (as identified above this is not expected to occur or, at worst, it is not expected to occur for an extended period of time) pressure control of Zone I11 can be achieved by Zone I11 venting . Note that calculation 175-017 revision 3 determined that with 10500 CFM at 180eF, 100% relative humidity, the heaters (OElOlAa and OElOlB) needed to be rated for 90KW. In this scenario the flow rate through SGTS will be limited to the air inleakage to zone 111. This flow rate (on the order of 2000 CFM) will be significantly lower than the 10500 CFM evaluated in the calculation. (10500 CFM is the maximum system flow rate). Thus for the scenario evaluated here, the heat required of the air stream to drop the relative humidity to acceptable levels will be less than at the conditions evaluated in the calculation. The calculation thus bounds the conditions expected in this scenario. The installation of the 90 KW heaters is verified by review of drawing E242 sht 3 revision 15.


It is expected that it will take a long time for the ductwork to fill with condensation. This is due to the extended time to boil and to the relatively warm air conditions that will exist on the outside of the ductinq. Condensation is felt to more readily occur where the relatively cool dilution air is introduce to the system. As seen from Figure 4, the dilution air is introduced just upstream of the filter train. The airflow is expected to draw this condensation into the train rather than allowinq it to drain back down into the ducting. Thus condensation wlll sooner collect in the SGTS train than in the duct work.

Condensation collection in the SGTS train has been previously evaluated in NCR 89-0427. The operability assessment of the NCR identifies that with significant amounts of water in the train that "water will not significantly affect the airflow in the SGTS plenum nor will it cause the heaters or filters to fail".

The impact of water in the ducting has not been quantitatively evaluated. It is assumed that the ducting will not be structurally capable of maintaining integrity should it fill with water. Review of the duct runs reveals that water condensing in the duct will collect at the low points, possibly creating a loop seal. More likely, however, it is felt that the duct will structurally fail prior to the creation of a loop seal.

Therefore, it is concluded that the SGTS train will not be adversely affected. If it is affected, it is not expected to be affected for an extended period of time. Even if it is affected, the alternate acceptable method of maintaining Zone I11 pressure is Zone I11 venting.


Assessment of Emeraencv Ventilation O~tions for Zone I11 Ventina durina Fuel Pool Boil Scenarios

purpose

The purpose of this evaluation is to provide a qualitative assessment of three proposed options for ventilating Zone I11 during fuel pool boiling scenarios. All three options require that Zone I11 be isolated from the recirculation plenum and involve the following scenarios:

I) Pool boiling with a vent path and no fans in operation;

11) Pool boiling with the normal HVAC supply, exhaust and filtered exhaust fans in service;

111) Pool Boiling with only the filtered exhaust fan running.

Assum~tions

1) Zone I11 is completely isolated. This includes complete isolation from the secondary containment common recirc plenum and the Standby Gas Treatment System. The required isolation lineups are achieved by manually closing various dampers and/or the installation of blank flanges.

2) Electrical power is available to the Zone I11 fan motors and their associated control schemes.

3) For the purposes of damper operation and flow control, it is assumed that either: a) the instrument air system is functional, or b) balancing dampers can be adjusted manually.

4) Additional assumptions are stated where applicable.

References

1) PP&L Calculation M-FPC-009 Rev. 0 2) Combustion Engineering Steam Tables 3) Bechtel Calculation 175-19 Rev. 2 4) Bechtel Calculation 175-17 Rev. 3 5j Bechtel Specification $856-A-7 6) 1985 A S H ~ E Fundamentals Handbook 7) PP&L P&IDs M-175 Rev.25 and M-2175 Rev. 15 81 PP&L S~ecification M-1070 Rev. 1 9 i ~s?iii fable 6.2-17 10j ASHRAE Psychrometric Chart #3 - High Temperature 11) PP&L Drawing FF108570 Sht. 6401 Rev. 4 12) PP&L Drawing FF108880 Sht. 1301 Rev. 1

Conclusions

The following conclusions are based on a qualitative analysis which assumes the worst case rate of pool boiling as determined in Reference 1. Hence these conclusions also apply to scenarios which involve a more realistic rate of pool boiling.


1) If no fans are available, and Zone I11 needs to be vented (Option I) : a) An opening, 2.2 ft2 in size, between the outside and

Zone I11 will provide an adequate exhaust path for the worst case fuel pool rate of boiling. This area corresponds to a hatchway with a diameter of 1.7 ft.

b) Under these conditions, the maximum Zone I11 ambient temperature is 180°F with a relative humidity of 100%. (See sheets 7 - 8 of this discussion)

2) With respect to returning the normal HVAC equipment to service (Option 11): a) It is likely that portions of the normal W A C system

will be available to re-establish normal Zone 111 ventilation during fuel pool boiling scenarios. Although certain system interlocks must be defeated (i.e. LOCA Fan Trips), it is anticipated that the equipment which normally supplies ventilation cooling for Zone I11 will be available to provide this function in a post accident environment.

b) In this case, the steady state environmental conditions on the refueling floor would consist of 100% relative humidity with a temperature approximately 6°F above the outside supply air temperature. (See sheets 9 - 11 of this discussion)

c) Just after normal W A C is restored, these conditions could result in the condensation of moisture inside the exhaust ductwork. In this event, itis likely that the condensate will drain to the lower elevations, and possibly out the exhaust registers.

3) If a complete complement of normal system components is not available, or it is desired to only operate a Zone I11 filtered exhaust train (Option 111): a) The existing plant equipment is capable of supporting

this mode of operation for some period of time. As with Option 11, certain system interlocks must be defeated.

b) At some point in time, the effectiveness of the charcoal filters will degrade as a result of the high moisture content.

During normal plant operations, Zone 111 ventilation is provided by separate Unit 1 and Unit 2 air supply and exhaust systems. The design for each of these inde~endent systems provides for 100% redundancy for each unit during normal operation: each sub-system is provided with two 100% capacity supply, exhaust and filtered exhaust fans. Currently, the Zone I11 supply fans (xV212A,B) are balanced to provide 92050 CFM for Unit 1, and 85,350 CFM for Unit 2. The exhaust fans (xV213A,B) are balanced for 86,350 CFM (Unit 1) and 80500 CFM (Unit 2). The filtered exhaust fan (xV217A,B) flows are 6600 CFM and 5550 CFM for Unit 1 and 2 respectively.


In the event that Zone I11 venting is revired after an accident, actions can be taken to physically isolate the Zone I11 air volume from the rest of the reactor building air spaces. Zone I11 can also be isolated from the common recirculation fan plenum and the Standby Gas Treatment System.

In this configuration, it is possible to start a filtered exhaust fan to take advantage of the filtering capability of a filtered exhaust train. Also, if available, it is possible to re-establish ventilation to this volume (Zone 111) with those systems which normally serve this function.

Although the function of the normal supply and exhaust systems are not related to safety, there is a hiqh probability that portions of these systems and their associated components will be available to aid in post accident ventilation cooling, including seismic scenarios. The following rationale supports this argument:

1) The performance of these major plant components is closely monitored as part of the predictive maintenance program. Since failure of a major HVAC component could jeopardize plant reliability, this equipment is maintained to high standards. While these components are not related to safety, they maintain the secondary containment pressure within the Tech Spec operating envelope.

2) Although portions of the associated duct work are not seismically qualified, significant portions of the QQNon-Qu duct work are qualified as IQClass A" which is supported to withstand the safe shutdown earthquake. The remaining Wlass Bw duct work is not seismically designed. However, as with many other non-seismic plant components, design specifications require that it be capable of withstanding modest seismic accelerations of .05g (horizontal) and .03g (vertical) without exceeding code allowable stresses.

3) There are four complete sub-systems, comprised of a total of twelve fans, which provide the normal Zone I11 HVAC function. Considering this equipment complement, it is likely that at least one set of components will be available to establish supply and exhaust ventilation flow.

E c r u i ~ m e n t E v a l u a t i o n

The following discussions are intended to provide a general qualitative assessment of the post accident availability and operation of the normal Zone I11 HVAC system components. Option I is not addressed in these evaluations since equipment operation is not required to support this mode.

LOSS OF FUEL POOL COOLIN(3 EVENT EVALUATION

Fans As previously stated, there are a total of twelve fans which could be utilized to establish a post accident ventilation line-up . Ovtion I Prior to're-establishing Zone I11 HVAC flow, the assumed conditions are 180'F at a relative humidity of 100% (reference Bechtel Calc. 175-17 Rev. 3). These conditions would have no effect on the supply fans since they would be handling outside air which would be at, or near their normal operating conditions. While the exhaust fans would handle air at the accident conditions, there would be little or no effect on their performance. As a result of the additional moisture content, the fans would tend to operate further back on their performance curves, at a slightly decreased flow. At elevated temperatures, the lower density of the air would more than offset the effects of the higher moisture content. Further, these high temperatures would only be experienced until a sufficient volume of the Zone I11 air is "changed- outf1 with fresh outside air.

If one unit's normal lineup is re-established at an assumed flow of 80,000 CFM, the total time to completely changeout the entire Zone I11 volume would be:

Time = Total Zone 111 Volume / 80,000 = 2.668.400 / 80.000 = 33 minutes'

The effect, if any, of exhaust fan operation at the higher temperatures for this relatively brief period would be at worst, accelerated component wear. Once normal operating temperatures are restored, fan operation would be essentially the same as during normal plant operation.

Per the attached calculations, the steady state environmental conditions after re-establishing a normal complement of Zone I11 fans would be a relative humidity of 100% with a temperature approximately 6°F above the outside supply air temperature.

O~tion I11 The steady state conditions which would be expected during operation of a filtered exhaust train would be the same, or near the conditions described for the venting process (Option I). The filtered exhaust fans would handle air at the accident conditions, but there would be little or no effect on their performance. As a result of the additional moisture content, the fans would tend to operate further back on their performance curves, at a slightly decreased flow. At elevated temperatures, the lower density of the air would more than offset the effects of the higher moisture content.


The effects of exhaust fan operation at the higher temperatures would be at worst, accelerated component wear.

The capacity of the Zone I11 filtered exhaust fans is 6500 CF'M. Per the attached calculations, it is seen that this capacity is less than the worst case volumetric pool boiling rate. Although the calculation implies that the remaining vapor would exit through the 2.2 fta opening, it is more realistic to expect that the direction of flow would actually be into the Zone I11 volume for the following reasons: 1) In this Wolumetric balance1', no credit is taken for

condensation which would occur on the wall and ceiling surfaces. These surfaces will in fact, provide cooling since they would ultimately transport heat to the environment.

2) More realistic estimates of the pool boiling rate are within the volumetric capacity of the filtered exhaust fans .

Fan Motors & Power SUDD~Y Eaui~ment

Per the discussions above, fan flow and dP would be essentially the same as during normal operation. The fan motors are located in the "no-zone" spaces on elevations 779' and 799', which are not connected to Zone I11 and hence, would not be subjected to the high temperature and humidity levels induced by fuel pool boilinq. If the normal supply and exhaust fans are running (Option 11) , these Itno-zone" areas will be supplied with ventilation cooling, as during normal operation. If only a filtered exhaust fan is in operation, the ambient operating temperature may be slightly elevated, but nonetheless acceptable.

Ductwork/Damuers

As previously stated, portions of the Zone I11 normal supply and exhaust ductwork are not seismically qualified. However, significant portions of the "Non-Q" duct work are qualified as I1Class A" which is supported to withstand the safe shutdown earthquake, and the remaining "Class B" duct work is also designed to withstand small seismic loadings. It is therefore reasonable to assume that significant portions of duct required to support operation of the Zone I11 fans will be intact, or only require minor repair.

The design, fabrication, and installation of all of the associated ductwork is governed via PP&L Specification M-1070 which provides assurance of high quality workmanship and design configuration. Further, all ductwork and balancing dampers are qualified for a continuous operating temperature of 150'F at a relative humidity of 100%.

In the event that automatic flow control cannot be re- established, the system's ductwork contains multiple flow


balance dampers which could be used to manually adjust flow.

For the purposes of this evaluation, it is assumed that fan interlocks are operative such that duct overpressurization or underpressurization will not occur. Nonetheless, all affected duct is rated to withstand static. pressures (at least 6IsWG), and duct failure is highly unlikely.

If normal W A C is restored (Option 11), temperatures in Zone 111 will return to conditions near normal operation. With a relative humidity of loo%, it is possible that condensation may occur in the exhaust duct work as the temperature decreases from 180 'F to it's long term steady state value. During this period, the condensate would tend to drain from the higher elevations down to the lower elevations and possibly exit via exhaust reyisters. Accumulation which could result in duct failure is possible, but even in this event, a flow path for Zone I11 air would still exist. Once the Zone I11 temperature stabilizes, the majority of condensation will occur on the 818' elevation near the fuel pool (i.e. near the origin of evaporation) . It is therefore concluded that condensation in the exhaust ductwork does not compromise the viability of Zone I11 venting. It is recommended that the configuration of the Zone I11 ductwork be reviewed/surveyed to corroborate this determination.

Filter Exhaust Trains

Operation of a filter exhaust train under the accident conditions will be feasible for some period of time. The 180°F assumed inlet temperature is below the 190'F high charcoal temperature alarm setpoint, and well below the 450'F fan shut-off permissive. Under normal operating conditions, it is recommended to maintain the relative humidity at the filter inlet to less than 70% to assure superior charcoal effectiveness. However, during performance testing of these filters, they are subjected to relative humidity of 95%. Studies have shown that operation with an inlet of 100% RH is acceptable, but filter effectiveness will degrade after some period of time. The actual point in time at which performance will be affected depends on the actual humidity as well as the initial condition of the charcoal media.

Summary

Based on the discussion above, the attached calculations, and evaluations performed to date, it is concluded that the various proposed options to vent Zone I11 during fuel pool boiling scenarios are feasible.

ASHRAE PSYCHROMETRIC CHART NO. 3

ASHRAE PSYCHROMETRIC CHART NO. 3

EQ Equipment Located i n HVAC Zone I11

A l i s t o f EQ equipment located on e levat ion 779 and above was generated from SEIS. This l i s t was reviewed by Kevin BrowingfMark Mjaatvedt t o determine which equipment was required t o operate dur ing the per iod when h igh humidity condi t ions from fue l pool b o i l i n g might occur. Items i d e n t i f i e d are as fo l lows:

Plant I D Coma Tvoe Old Binder New Binder FT-07557 Tavis PC8 EQDF-29 EQAR-016 PDT-O7554Al,A2,A3,Bl ,B2,B3 II I I

n

HDM-07545A, B NH-90 EQDF-31B EQAR-074 PDDM-07554A, B NH-90 EQDF-31B EQAR-074 PDSL-07544A,B Dwyer dp EQPL-JOl EQAR-087 OY201A,B WHSE Motor EQDF-26 EQAR-057

The binder on each o f these items was reviewed t o determine i f items were qua l i f i ed f tes ted f o r 100% humidity condi t ions dur ing normal and/or accident condit ions.

Tavis PC8 t ransmi t ters

These t ransmi t ters were not subjected t o 100% humidity i n e i t h e r normal o r accident tests . These devices are potted which may al low an evaluat ion t o be performed tha t el iminates humidity as a q u a l i f i c a t i o n concern. The o l d binder SCEW sheets s ta te t ha t these t ransmi t ters are q u a l i f i e d f o r 90% humidity.

NH-90 Actuators

Actuators were exposed t o steam and spray dur ing f i r s t 12 hours o f accident tes t ing . The e l e c t i c a l compartment was not sealed and the environment was allowed entry. Low humidity was maintained dur ing thermal aging. Ant ic ipate t h a t 100% humidity dur ing normal and a t l eas t the ea r l y po r t i on o f the accident can be j u s t i f i e d . The o l d binder SCEW sheet.s s ta te t h a t these actuators are q u a l i f i e d f o r 100% humidity.

Dwyer Pressure Switches

The t e s t repor t s ta tes t h a t switches are q u a l i f i e d f o r 100% humidity dur ing both normal and accident condit ions. A cursory review o f the t e s t ind icates t h a t 100% humidity condi t ions were applied only dur ing the accident tes t . Ant ic ipate t ha t 100% humidity dur ing normal and accident condi t ions can be j u s t i f i e d . The o l d binder SCEW sheets s ta te t h a t these switches are qua1 i f i e d f o r 100% humidity.

Westinghouse Fan Motors

Aged motorettes were exposed t o 48 hours i n a condensing atmosphere whi le de-energized. While s t i l l i n the chamber and v i s i b l y wet the motorettes were tested t o 600 VAC f o r 10 minutes both phase t o phase and phase t o ground. Ant ic ipate t h a t 100% humidity dur ing normal and accident condi t ions can be j u s t i f i e d . The o ld binder SCEW sheets s ta te t h a t these motors are q u a l i f i e d f o r 100% humidity.

Figure C - Loss of Fuel Pool Cooling Due to Other Causes

Loss of Fuel Pool Coding. rl

Use RHR FPC &st with Aitemae SDC Il aviilable,

dhsmisa:.

Allow to bail with ESW Mk 6 l l vmed.


ATTACHMENT 4

TIME TO MAKE-UP

FOR FUEL W O L


A88UMPTIONS:

Fuel ~ o o l level is at the normal level of 81711*1. lJ653 sht 65) Evaporation rate prior to boiling will be based on 170'F pool temperature. 170°F,is the approximate average of the pool temperature between the initial assumed temperature of 110°F and the final o r boiling temperature - C ...,.om V L L I U r . The minimum acceptable level is the bottom of the fuel pool weir which is at 816' 6 3/4" ( M29-9). Before pool level reaches this level, makeup needs to be provided. Note that the minimum acceptable level per technical specifications is:

779'4" Bottom of fuel pool (C1932 sht 5) + 14'8" Height of fuel bundles (XN-NF-304,998) + 22' Tech Spec margin (3/4.9.9)

816 ' Thus use of 816' 6 3/4" provides considerable margin.

No fuel pool level indication is available. If it is available, then actual conditions should be used versus this estimate.

Total volume and thus mass that will be evaporated prior to reaching the minimum acceptable pool level:

Fuel pool surface area = 1350 FT2 Depth = 817'1'- 816'= 111=.9167 FT

Volume

Using (conservatively) a water density at 110'F = 61.8 (Crane), the mass loss will be:

1237.4 FT3 (61.8 lbm/FT3) = 76477.5 Lbm

For the current operating cycle 11/92 to 9/93:

Once boiling occurs;

Q = heat load H = Latent heat of Vaporization

Mass loss = Q/H


Q for Unit 2 = 5 MBTU/HR (M-FPC-010) Q for Unit 1 = -8 MBTU/HR

UNIT 2 M = 4849 LBM/HR

UNIT 1 M = 776 LBM/HR

Looking at T (time) less than time to boil, thus only considering evaporation losses;

T = 76477.5 lbm/ M (evap loss in lbm/hr)

From page 13 of M-FPC-010:

M = 797 lbm/hr

Thus

T= 76477.5 (lbm)/ 797 (lbm/hr) = 95.9 hours

From M-FPC-010;

Table 1: Time to boil = >340 hrs

Table 2: Time to Boil = 65 hrs on 11/10/92 Time to boil = 94 hrs on 1/8/93

Conclusions:

1. Unit 1 - Makeup to the pool should be initiated prior to 90 hrs after the loss of fuel pool cooling.

2. Unit 2 Makeup to the pool should be initiated prior the time to boil as identified in M-FPC-010 Figure 9 or 90 hrs whichever is sooner.

3. Significant time is available to provide makeup to the fuel pools.


ATTACHMENT 5

REVISED EVALUATION OF EDR 020020

SPENT FUEL POOL COOLING ISSUE

(PLI-72764)

October 29, 1992


SUSQIJEHANNA STEM ELECTRIC STATION REVISED EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-72764 FILE A45-1A

The attached rev ised evaluation o f EDR 620020 i s provided f o r your information. The rev i s i on accounts f o r a change t o the engineering repo r t reference, corrects a typographical e r r o r on page 10, expands the event t r ee t o three pages and provides several graphs used as p a r t o f the PORC presentat ion on Monday October 26, 1992. The conclusions remain unchanged.

& ~ 4 Glenn 0. M i l l e r

cc: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W . Sirnpson - A1-2 D. F. Roth - SSES H. G. Stanley - SSES J. M. Kenny - A2-4 J. S. Stefanko - A9-3 F. G. Bu t le r - A6-3 J. R. Mil tenberger - A6-1 D. C . Prevatte - A6-3 Nuclear Records - A6-2 D. A. Lochbaum - Enercon

October 21, 1992


SUSQUEHANNA STEAN ELECTRIC STATION EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-72711 FILE A45-1A

The attached evaluat ion o f EDR 620020 i s provided i n response t o your memo PLI-72640. This evaluat ion was prepared by myself and a team o f engineers working on the ac t ion plan t o resolve the subject EDR.

I n the course o f reviewing t h i s issue i n de ta i l , we have concluded t h a t seven o f the n ine i den t i f i ed discrepancies are not v a l i d def ic ienc ies. This i s explained i n d e t a i l i n the evaluation. The two remaining discrepancies are v a l i d de f i c ienc ies but are not considered safety s i g n i f i c a n t nor reportable.

I bel ieve t h a t the technical basis on several o f these issues has been c l a r i f i e d considerably i n the course o f the past week. Therefore, I suggest prov id ing t h i s evaluat ion t o Nuclear Regulatory A f f a i r s f o r reconsiderat ion o f the repo r tab i l i t y aspects.

We are cont inuing w i th the remaining act ion items as requested. A de ta i l ed engineering design repor t and j u s t i f i c a t i o n f o r i n te r im operation w i l l provide more d e t a i l than contained i n the attached evaluation. We are scheduled t o meet w i t h PORC on Monday October 26, 1992 a t 2:30 pm t o review t h i s issue. We are working w i t h Systems Engineering, Operations and NRA-Compliance w i t h respect t o po ten t i a l compensatory measures.

4 e - D . 4 Glenn D. M i l l e r

cc: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W. Simpson - A1-2 D. F. Roth - SSES H. G. Stanley - SSES J. M. Kenny - A2-4 J. S. Stefanko - A9-3 F. G. Bu t le r - A6-3 J. R. Mil tenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records - A6-2 D. A. Lochbaum - Enercon

October 21, 1992 Page 1 Rev. 1.

Evaluation o f EDR 620020

This document contains an evaluation o f the discrepancies documented i n EDR 620020, "Loss o f Spent Fuel Pool Ccoling Event Design Discrepancies." Conclusions o f the author w i t h respect t o r e p o r t a b i l i t y o f these concerns and o p e r a b i l i t y impact on SSES are a1 so provided.

Desian Basis

The design bases f o r the Fuel Pool Cooling System are found i n FSAR sect ion 9.1. The por t ions o f the design basis relevant t o EDR 620020 are as fo l lows:

1. Maintain the fue l pool water temperature below 125F under "normal maximum heat loads" defined as 12.6 MBtu/hr (equivalent t o a t yp i ca l f ue l cycle discharge schedule which f i l l s the fue l pool, l a s t quarter core o f f l o a d a t 6.7 days a f t e r shutdown).

2. Maintain fue l pool water temperature a t o r below 125F dur ing the "emergency heat load" condi t ion o f 32.6 MBtu/hr (equivalent t o a f u l l core o f f l oad 10.5 days a f t e r a shutdown fo l lowing a t yp i ca l f ue l cycle discharge schedule which f i l l s the fue l pool) u t i l i z i n g the RHR system (w i th o r wi thout normal fue l pool cool ing) f o r f ue l pool cooling. This mode o f operation applies "during periods o f higher than MNHL generation i n the f ue l pool, eg., s to r ing o f a f u l l core o f i r r a d i a t e d f u e l sho r t l y a f t e r shutdown". The RHR system i s used under these condi t ions t o ass is t the FPCCS i n d iss ipa t ing the decay heat. Thus, any heat load i n excess o f 12.6 MBtu/hr i s considered t o be w i th in the design basis f o r the RHR FPC ass i s t mode o f operation.

3. Redundant Seismic Category I ESW connections t o each pool are provided t o al low f o r makeup o f evaporative losses i n the event o f f a i l u r e of ' the FPC system. The condit ions are bounded by a fue l pool t ime-to-boi l analysis based on the same t yp i ca l fue l cycle discharge schedule as i n basis X 1 except the time a f t e r shutdown i s 10.5 days instead o f 6.7 days r e s u l t i n g i n a heat load o f 9.8 MBtu/hr. (This explains the d i f fe rence between the two d i f f e r e n t heat loads, ie. , 12.6 MBtu/hr f o r basis X I and 9.8 MBtu/hr f o r basis X3. This i s not a discrepancy.) The ESW makeup l i n e i s sized on the basis o f t h i s ca lcu la t ion (Reference FSAR sect ion 3.13).

4 . The cause o f the Loss o f Spent Fuel Pool Cooling event i s s ta ted t o be a seismic event.

5. A l l p ip ing and equipment shared w i th o r connecting t o the RHR i n t e r t i e loop are Seismic Category I and can be i so la ted from any p ip ing associated w i th the non-Seismic Category I fue l pool cool ing system.

Eva1 uat ion o f Di screoancies Noted i n EDR 620020

EDR 620020 describes nine discrepancies r e l a t i n g t o the loss o f spent f u e l pool cool ing event. This discussion w i l l summarize each issue. The reader i s

October 21, 1992 Page 2 Rev. 1


General Statement . .

The in t roductory paragraph o f the EDR states: ". . .the design prov is ion f o r t he loss o f spent fue l pool cool ing event i s t o permit the fue l pool t o b o i l and maintain i t s water l eve l above the fue l through makeup from the ESW system. This design p rov is ion i s necessary because the fue l pool cool ing system used f o r normal operation and the RHR fue l pool cool ing ass i s t mode used f o r abnormal heat loads are no t designed t o s a t i s f y seismic category I and s ing le f a i l u r e c r i t e r i a . "

As s ta ted i n design basis #5 above the RHR fue l pool cool ing ass i s t po r t i on o f the p ip ing i s designed t o seismic category I requirements. No c r e d i t however i s taken f o r t h i s mode o f operation i n the fue l pool b o i l i n g analysis i n the FSAR. Credi t i s taken f o r t h i s mode f o r emergency heat load s i tua t ions as def ined by basis #2.

I n order t o discuss and evaluate each o f the nine discrepancies l i s t e d i n the EDR i t w i l l be more l o g i c a l t o review them i n a d i f f e r e n t order. Items E & F both r e l a t e t o the t ime-to-boi l ca lcu la t ions and w i l l be reviewed f i r s t fo l lowed by items G through I, which are re la ted t o the t ime-to-boi l concern. Items C & D invo lve operator act ion considerations and w i l l be discussed next. F i n a l l y items A & B r e l a t i n g t o the evaporation e f fec ts w i l l be discussed.


"The ana l y t i ca l 25 hour t ime-to-boi l f o r the spent fue l pool i s nonconservative for the maximum normal heat load i n the,spent fue l pool."

As s ta ted i n basis #1 the maximum normal heat load i s 12.6 MBtu/hr. As stated i n basis #2 the t ime-to-boi l analysis i s based on a heat load o f 9.8 MBtu/hr. These two design bases are i n f ac t consistent and are based on the same " t yp i ca l fuel discharge schedulen and re fue l ing outage scenario. The d i f fe rence i n the heat load i s due so le ly t o the time a f t e r shutdown assumed f o r purposes o f estab l ish ing the design basis.

Focusing on the t ime-to-boi l analysis, a time a f t e r shutdown value o f 10.5 days i s used. This i s the time a t which i t i s assumed t h a t re fue l i ng i s completed and the reactor c a v i t y t o fue l pool gates are re ins ta l led . P r i o r t o t h a t p o i n t the addi t ional water stored i n the reactor cav i ty i s also avai lab le as a heat s ink and the RHR system i s ava i lab le f o r fue l pool cooling. For times g rea te r than 10.5 days the appropriate heat load w i l l be even lower than the analyzed value o f 9.8 MBtu/hr. For the SSES Un i t 2 5RI0 the time from reactor shutdown t o f u e l pool gates i n s t a l l e d was 38 days. The decay heat i n the Un i t 2 pool a t t h a t t ime i s ca lcu la ted t o be 5.65 MBtu/hr (Reference NE-092-002). The corresponding time- (A to -bo i l i s 45 hours.

The EDR goes on t o discuss other ca lcu la t ions which r e s u l t i n d i f f e r e n t heat loads using various assumptions. Calculat ion NFE-B-NA-053 was performed by Nuclear Fuels t o account f o r actual fue l discharge h i s t o r y and fu tu re o f f loads accounting f o r power uprate condit ions. The fue l pool heat load versus time curves obviously w i l l increase subsequent t o power uprate, however, these curves do not apply t o the ex i s t i ng design. As long as the ca lcu la ted decay heat i s less than 9.8 MBtu/hr a t the po in t where the fue l pool gates are r e i n s t a l l e d the o r i g i n a l design basis t ime-to-boi l ca lcu la t ion i s s t i l l va l id .

Calcu la t ion M-FPC-009 determined t ime-to-boi l condi t ions post power uprate. This ca lcu la t ion shows t h a t the t ime-to-boi l f o r the design basis heat load o f 9.8 MBtu/hr i s s l i g h t l y greater than 25 hours.

I n conclusion, the design basis f o r the t ime-to-boi l condi t ion i s establ ished by the 9.8 MBtu/hr value used i n the o r i g i na l ca lcu la t ions. This design basis i s met by planning the outage so tha t the fue l pool i s not i so la ted from the reac to r cav i t y o r the RHR system p r i o r t o a po in t i n time where the actual heat load i s 9.8 MBtu/hr o r less.



"The ana l y t i ca l 25 hour t ime-to-boi l f o r the spent fue l pool does not account f o r the emergency heat load i n the spent fuo1,pool."

As discussed above, the t ime-to-boi l condit ions apply t o conf igurat ions where the spent fue l pools are i so la ted from the reactor cav i t y (ie., non-refuel ing conf igurat ions). As i s co r rec t l y stated i n the EDR, current p rac t i ce i s t o f u l l y o f f l o a d the core dur ing each re fue l ing outage. Spec i f i c ca lcu la t ions are performed by Nuclear Fuels t o determine the a b i l i t y o f the FPC system t o remove the combined decay heat o f the cross-t ied re fue l i ng pools. Tests are also conducted t o determine t h a t the actual heat removal capab i l i t y exceeds the actual f ue l pool heat loads dur ing the outage (Reference TP-235-011). Normally the reactor c a v i t y i s maintained flooded and cross-t ied t o the fue l pools. One loop o f Core Spray i s always operable i n t h i s conf igurat ion. One d i v i s i o n o f RHR i s maintained i n shutdown cool ing mode except f o r a b r i e f per iod required f o r the common RHR system outage window.

Design basis #2 states t h a t heat loads i n excess o f the MNHL are considered t o be emergency heat loads. The design o f the RHR system t o ass i s t the FPC system dur ing emergency heat load condit ions assures t h a t fue l decay heat i s removed. No t ime-to-boi l ca lcu la t ion f o r t h i s conf igurat ion i s required since the RHR system w i l l be i n operation o r avai lable. A t any rate, such a ca l cu la t i on should consider the e f f e c t o f the addi t ional water inventory ava i lab le from the f looded reactor cav i ty , cask storage p i t and dryer and separator storage pool which are a l l cross-connected dur ing t h i s time. Makeup inventory i s also ava i lab le from Core Spray and the RHR system i s normally in-service except f o r the common RHR system outage window.

I n conclusion, no t ime-to-boi l analysis i s required f o r the emergency heat load design basis. Single f a i l u res o f the RHR system are not requi red f o r t h i s design basis f o r the emergency heat load (Reference SRP 9.1.3).

This discrepancy i s no t a v a l i d deficiency, not reportable and has no impact on p lan t operabi 1 i ty.


"The rad io log ica l release analysis f o r a b o i l i n g spent f ue l pool uses nonconservative evaporation rates." , .,

This discrepancy i s d i r e c t l y re la ted t o the heat load assumed f o r the time-to- b o i l analysis. The evaporation r a t e used i n the dose ca l cu la t i on i s based on a heat load o f 9.8 MBtufhr which i s the design basis heat load f o r the t ime-to-boi l ca lcu la t ion. Heat loads i n excess o f 9.8 MBtu/hr obviously r e s u l t i n higher evaporation rates. Since the discussion under Item E above establ ishes t h a t 9.8 MBtu/hr i s the cor rec t o r i g i na l design basis and s t i l l bounds current operation there i s no discrepancy i n the o f f s i t e dose ca lcu la t ion. It uses an evaporation r a t e consistent w i t h the design basis heat load.

This discrepancy i s not a v a l i d deficiency, n o t repor tab le and has no impact on p l a n t operabi 1 i ty .


"The rad io log ica l release analysis f o r a b o i l i n g spent fue l pool uses nonconservative a c t i v i t y terms. The o r j g i na l design ca lcu la t ion (200-0048) assumed 12 month operating cycles and 184 bundle equ i l ib r ium reload sizes t o determine the a c t i v i t y terms f o r f a i l e d fue l i n the fue l pool. SSES cu r ren t l y has 18 month operating cycles w i t h approximately 230 bundle reloads which w i l l increase t o approximately 254 bundles a f t e r power uprate. Since the ca l cu la t i on impl ied t h a t most o f the a c t i v i t y resu l ts from the most recent discharge batch, the e f f e c t o f increasing the discharge s ize from 184 bundles assumed i n the ca lc t o 230 and 254 bundles would appear t o be nonconservative w i t h respect t o the rad io log ica l release analysis."

The o r i g i n a l rad io log ica l release analysis as referenced above i s conservative f o r the fo l low ing reasons:

(I) the ac ' t i v i t y l eve l s used as a source term are based on 1% f a i l e d f ue l . A l l o f the f a i l e d fue l rods are assumed t o be i n the of f loaded batch o f 184 f ue l assemblies. Therefore, increased batch sizes w i 11 not increase the amount o f the source term used i n t h i s analysis.

(2) the a c t i v i t y leve ls used f o r the iodine source term are based on sa tu ra t ion l eve l inventor ies f o r a core operating a t 3440 MWt f o r one thousand days. Therefore, the fue l cycle length w i l l not a f f e c t the source term.

I n conclusion, the o f f s i t e dose ca lcu la t ion remains va l id .

This discrepancy i s no t a v a l i d deficiency, no t reportable and has no impact on p lan t operabi 1 i ty.

October 21, 1992 Page 7 Rev. 1.

Jtem I: Analvsis f o r Max Time P r i o r t o Makeuo

"The analysis f o r maximum time p r i o r t o makeup t o a b o i l i n g spent f ue l pool i s based upon nonconservative assumptions. R e o r i g i n a l design ca l cu la t i on (175-14) determined the time using evaporation o f the e n t i r e fue l pool water inventory. The maximum time should be based upon a minimum fue l pool water l e v e l which i s s u f f i c i e n t l y above the top o f the fue l t o provide the sh ie ld ing required t o a l low cor rec t i ve operator actions. "

The purpose o f the referenced ca lcu la t ion was t o determine r e f u e l i ng f l o o r atmosphere condi t ions under various operating modes. The evaporation ra tes and assumptions used i n the c i t e d por t ion o f the ca lcu la t ion were used s o l e l y t o determine i f condensation could be expected under fue l pool b o i l i n g condi t ions. The conclusion o f the ca lcu la t ion regarding time t o b o i l the pool d r y i s not re levant t o any operator action. Operator act ions are based on maintaining normal pool l e v e l and temperature condit ions. I n any case, the c i t e d nonconservatism would have a minor e f f e c t on the calculated I9 days t o b o i l the pool dry, a r e s u l t which i s not used elsewhere i n the design.

This discrepancy i s no t a v a l i d deficiency, i s no t reportable and has no impact on p lan t operabi 1 i ty .


"The manual valve manipulations required t o provide ESW makeup f low t o a b o i l i n g spent f ue l pool may not be possible." , -

In-plant post-accident rad ia t i on l eve l s are analyzed f o r SSES t o the requirements specif ied i n NUREG-0737. This document requires t h a t post-accident r a d i a t i o n l eve l s be determined f o r purposes o f v i t a l area access by p lan t operators t o perform short-term f i r s t p r i o r i t y actions. It speci f ies t h a t rad ia t i on l e v e l s be determined on the basis o f contained sources, and core damage source terms equivalent t o those used f o r lOCFRlOO calculat ions. These assumptions are c l e a r l y based on degraded core condit ions which are beyond the design basis LOCA. Airborne r a d i o a c t i v i t y sources from containment 1 eakage are required t o be analyzed fo r environmental q u a l i f i c a t i o n o f equipment but no t f o r personnel access.

A review o f FSAR chapter 18 shows tha t access t o the equipment necessary t o provide makeup t o the fue l pool from ESW i s r e s t r i c t e d f o r the approximately t he f i r s t 30 hours fo l low ing the design basis event (Figure 18.1-9). This analys is i s based on a conservative source term equating t o 100% fue l damage r e s u l t i n g from core mel t condi t ions as o r i g i n a l l y u t i l i z e d f o r o f f s i t e dose ca lcu la t ions used t o determine p lan t s i t i n g adequacy. These source terms were based on experiments invo lv ing heated i r r ad ia ted uranium d iox ide pe l l e t s .

An evaluat ion o f actual fue l thermal response dur ing design basis accidents r e s u l t s i n no predicted fue l f a i l u r e s (Reference PLI-72696). Thus, the source term r e s u l t i n g from the DBA LOCA would only be equivalent t o the r a d i o a c t i v i t y present i n the reactor coolant as a r e s u l t o f normal operations (a l lowing f o r fue l defects as permit ted by Technical Speci f icat ions). To bound the po ten t i a l e f f ec t s o f a design basis accident, a r e a l i s t i c ye t conservative analysis using an assumed 1% fue l damage resu l t i ng from core degradation under LOCA condi t ions was performed (Reference EP-548) and concludes tha t access t o equipment necessary t o m i t i ga te the e f f ec t s o f a loss o f fue l pool cool ing fo l low ing a DBA LOCA i s assured.

I n conclusion, post-accident operator actions are v iab le f o r a l l po ten t i a l scenarios under consideration, f o r both the current design basis and those outside o f the current design basis.

This discrepancy i s no t a v a l i d deficiency, i s no t reportable and has no impact on p l a n t operabi 1 i ty .


I tem D: Instrumentation

"The instrumentation avai lab le t o the operator post-LOCA does not provide adequate i nd i ca t i on o f spent fue l pool .temperature and l eve l t o a1 low proper response t o a loss o f fue l pool cool ing event."

The instrumentation avai lab le t o the operator i s not required t o be q u a l i f i e d since the design basis loss o f spent fue l pool cool ing i s not coincident w i t h the DBA LOCA condit ions. This instrumentation i s powered from an un in te r rup t i b l e power supply and i t s ' associated 1E AC source.

The minimum water l eve l required per Tech Specs i s below the wei r e levat ion. Since ESW makeup i s provided t o the pool the operators w i l l know t h a t when they see a r i s e i n skimmer surge tank l eve l the fue l pool l eve l i s a t l e a s t as h igh as the weir. This provides a confirmation o f adequate pool l e v e l wi thout requ i r i ng access t o the re fue l ing f loor .

Furthermore, on the basis o f the discussion i n i tem C above, access t o t he re fue l i ng f l o o r i s possible under a l l considered condit ions. Therefore, i t i s possible t o v e r i f y adequate fue l pool leve l v i s u a l l y from the r e f u e l i n g f l o o r which i s accessible from several locat ions.

While the avai lab le instrumentation i s adequate f o r operator act ions and meets the regulatory requirements o f Reg Guide 1.13, improvements t o t he instrumentation have been recommended i n the past and should be implemented. This would enhance p lan t safety.

I n conclusion, the ex i s t i ng instrumentation i s adequate f o r performance o f requi red operator actions f o r the current design basis and f o r scenarios no t included i n the current design basis.

This discrepancy i s no t a v a l l d deficiency, it i s not repor tab le and has no impact on p lan t operab i l i t y .


I tem A: Reactor Bui ld ina Desian Heat Loads

"Reactor bu i l d i ng design heat loads do not account f o r the b o i l i n g spent f u e l pool event." , .,

The reactor bu i l d i ng temperature analysis i s performed f o r DBA LOCA condi t ions. The presumption o f i tem A i s tha t the spent fue l pool w i l l reach b o i l i n g condi t ions p r i o r t o res to ra t ion o f fue l pool cool ing subsequent t o a LOCA. The ex i s t i ng temperature analysis does account f o r the sensible heat load from the fue l pool a t 212F.

A loss o f spent f ue l pool cool ing event can r e s u l t from several condi t ions. The design basis condi t ion i s a seismic event as analyzed i n the FSAR. The Fuel Pool Cooling system i s not designed f o r seismic loads. I n t h i s case, the Fuel Pool Cooling system i s assumed t o be l o s t . An evaluation o f the p lan t response shows t h a t several methods are avai lab le t o assure t h a t the spent fue l remains cooled. These include: (1) the RHR system can be used t o cool the fue l pools w i t h a l te rna te shutdown cool ing o f the reactor using Core Spray and RHR f o r suppression pool cooling; o r (2) a l low the fue l pool t o b o i l w i t h makeup supplied by ESW w i th consideration o f e i t he r SGTS operating on Zone I11 o r prov id ing a vent path from Zone 111. The o f f s i t e dose analysis takes no c r e d i t f o r SGTS. I f avai lable, normal reactor bu i ld ing ven t i l a t i on would be used t o provide cool ing and venting o f the Zone I11 atmosphere. Under any o f these scenarios t ranspor t o f moist a i r t o other port ions o f the reactor bu i l d i ng would no t occur. This scenario i s the design basis f o r loss o f fue l pool cool ing.

Other scenarios not included i n the design basis include LOCA and LOOP events, and combinations thereof. The time frame f o r consideration o f operator act ions i s based on reasonable expectations f o r the t ime-to-boi l condi t ion. As s ta ted previously, f o r the current operating practice, the fue l pool heat load p r i o r t o reactor r e s t a r t i s approximately 5.65 MBtu/hr. Time t o b o i l under t h i s cond i t ion i s on the order o f 45 hours. Note t ha t t h i s i s the shortest possible t ime-to- b o i l f o r the current fue l cycle. With the pools cross-t ied the t ime-to-boi l i s

Im greater than 100 hours.

For a LOCA scenario, the FPC system w i l l be l o s t i n i t i a l l y due t o the Aux Load Shed provis ions. Although the Fuel Pool Cooling system and other non-safety re la ted systems are not s p e c i f i c a l l y analyzed f o r the e f f ec t s o f hydrodynamic loads i t i s expected tha t they w i l l be able t o perform t h e i r normal funct ions fo l low ing a broad spectrum o f design basis events. Credi t f o r these systems i s not needed t o meet the design basis, however, p lant operators w i l l u t i l i z e any equipment ava i l able t o them during emergency s i tuat ions. Therefore, i n t he course o f evaluat ing the e f f ec t s o f a DBA LOCA on the fue l pool cool ing system, we acknowledge the a v a i l a b i l i t y o f normal p lan t systems i n responding t o the emergency.

Independent o f the LOCA condit ion, o f f s i t e power i s needed t o restore normal cool ing systems. The SSES Ind iv idual Plant Evaluation considered loss o f o f f s i t e power (Reference IPE Appendix F). The IPE conservatively estimated the incidence o f LOOP t o be .04/year (p lant re lated), .008/year ( g r i d re la ted) , .00807/year


(severe weather related), and .00066\year (extreme1 y severe weather re1 ated) . The p r o b a b i l i t y o f recovery from the LOOP w i th in spec i f ied times was also calculated as fol lows:

, ., Time l h r ) PlRecoverv w i t h in T hrs)

12.0 97.96% 24.0 99.53%

Thus, i t can be reasonably concluded tha t o f f s i t e power w i l l be ava i lab le w i t h i n 24 hours fo l low ing the i n i t i a t i n g event.

The remaining fac to r i n addressing res to ra t ion o f fue l pool cool ing i s access t o the reactor bu i ld ing. This issue i s discussed under Item C above. Except f o r degraded core condi t ions access t o the reactor bu i l d i ng i s feas ib le a f t e r the f i r s t twelve hours o f the i n i t i a t i n g event (Reference EP-548).

Notwithstanding the above basis, we have also considered the scenario where o f f s i t e power i s not ava i lab le and access t o the lower reac to r bu i l d i ng elevat ions i s r e s t r i c t e d (based on FSAR chapter 18 contained source terms). Under these condi t ions (representative o f a degraded core event) access t o the re fue l i ng f l o o r remains avai lable. Provision f o r fue l pool coo l ing i s made through use o f the p lan t f i r e protect ion system. Venting o f Zone 111 v i a the f i l t e r e d exhaust system i s also possible f o r t h i s scenario. While access t o l eve l and temperature instruments would be questionable i t i s possible t o v e r i f y adequate pool l eve l v i sua l l y from the re fue l ing f l o o r which i s accessible a t several locat ions.

A Loss o f Fuel Pool Cooling Event Tree i s attached t o t h i s evaluat ion t o he lp guide the reader through these various postulated scenarios.

I n conclusion, f o r the design basis loss o f fue l pool cool ing the p lan t as cu r ren t l y designed and analyzed i s acceptable. For other scenarios no t s p e c i f i c a l l y included i n the design basis we have reasonable assurance t h a t t he e f f ec t s o f a loss o f fue l pool cool ing can be mi t igated without adverse consequences on the p lant .

This discrepancy i s a v a l i d deficiency. It i s no t a sa fe ty s i g n i f i c a n t issue because we have establ ished reasonable assurance t h a t t he e f f e c t s o f a l oss o f f ue l pool cool ing can be m i t i ga ted without adverse consequences on the p l a n t and pub l i c hea l th and sa fe ty and i s therefore not reportable. The evaluat ion above a lso shows t h a t t h i s concern does not impact p lan t operab i l i t y .

I n considerat ion o f t h i s concern, addi t ional analyses are warranted t o f u r t h e r quant i f y the e f f ec t s o f evaporation and b o i l i n g condi t ions on the Zone 111 atmosphere and the po ten t ia l t ransport o f moist a i r t o other loca t ions i n t he reactor bu i l d i ng f o r condit ions outside o f the current design basis.


"The impact o f the ESW makeup water t o the spent fue l pool on equipment i n the reactor bu i l d i ng has not been evaluatedr%

The analysis under i tem A above applies t o t h i s issue as we l l . This eva luat ion shows t h a t w i t h the current p lan t design and f o r e x i s t i n g design basis condi t ions the ef fects o f a l oss o f fue l pool cool ing are acceptable.

This discrepancy i s a v a l i d deficiency. As w i t h i tem A i t i s n o t a sa fe t y s i g n i f i c a n t issue and i s no t reportable. The evaluat ion above a lso shows t h a t t h i s concern does no t impact p lan t operabi l i ty .

I n considerat ion o f t h i s concern, addi t ional analyses are warranted t o f u r t h e r quant i f y the e f f ec t s o f evaporation and b o i l i n g condi t ions on the Zone I11 atmosphere and the po ten t ia l t ransport o f moist a i r t o other loca t ions i n the reactor bu i l d i ng f o r condit ions outside o f the current design basis.


Enaineerina R e ~ o r t on Loss o f S ~ e n t Fuel Pool Coolinq

A de ta i l ed report , NE-092-002, i s being prepared t o document t h i s evaluat ion i n )A fur ther d e t a i l . This repor t contains tocbnical inpu t from several engineering groups and w i l l provide a comprehensive set o f references on t h i s subject . The repor t w i l l be completed by October 28, 1992.

Design Basis Decay Heat Last ~ a c h Offload to Single Fuel Pool

- -- A

14-

13

h

12 L

f 11 3

5 10 15 20 25 30 35 40 Days After Shutdown

10 z w

5 9 Q,

* 8 > a $ 7 n

6

5

4

\

l'ii

-

. - ------------------

I U2 5RIO

Actual Decay Heat U2 Cycle 5 113 Core, Pool Isolated

Days After Shutdown

d 8 Decay Heat L-- 7

7 A// -30 f , i=

6 - --- c--- -25 5 //

4 20 5 10 15 20 25 30 35 40 45

Figure A - Loss of Fuel Pool Cooling Due to Seismic Event

#low to boil with ESW n~&ecp. 0 I Use RHR FPC assis with

Ulenrte SDC ll audlable. ahemise:. I

M88 OF FUEL POOL COOLING EVENT EVALUATION

NE-092-002 Rev. 0

ATTACHMENT 6

EXPECTED NUMBER OF FUEL FAILURES

DURING A DBA M C A

October 20, 1992

G. D. M i l l e r A6-3

SUSOUEHANNA STEAM ELECTRIC STATION EXP~CTED NUMBER OF FUEL FAILURES DURING THE DBA LOCA CCN 741087 FILE A7-8

Attachment I provides an evaluat ion o f the expected number o f fue l rod f a i l u r e s dur ing the design basis LOCA based on the power uprate SAFERIGESTR analysis resu l ts . This information i s provided i n support o f the assessments i n response t o EDR 620020.

A. J. ~ d c i v o l i Senior Pro ject Engineer-Nucl ear Nuclear Fuels Engineering

Attachment

cc: J. M. Ku l i ck A9-3 C. R. Lehmann A9-3 J. S. Stefanko A9-3 D. A. Matchick A6- 1 M. R. Mjaatvedt A6-3. NR F i l e A6-2

ATTACHMENT I

FUEL PERFORMANCE DURING A LOCA

During a LOCA there are two fue l rod f a i l u r e mechanisms tha t have the po ten t ia l t o cause damage t o the fue l rods. F i r s t , cladding rupture can occur fo l low ing fue l rod bal looning when the ,cladding temperature r i s e s and the fue l rod in te rna l pressure exceeds the external coolant pressure. Current LOCA analysis methodologies include empirical models t o ca lcu la te the onset o f bal looning and rupture. Second, i f the fue l rod cladding reaches h igh temperatures, the cladding becomes embr i t t led by steam ox idat ion o f the z i rca loy cladding and can fragment upon in t roduct ion o f the emergency core cool ing water. This thermal shock phenomenon i s precluded f o r events which remain below the 2200°F and 17% peak l oca l ox idat ion c r i t e r i a i n 10CFR50.46 (Ref. NUREG-1230). The best estimate and 1 icensing basis LOCA analysis resu l t s (Ref. NEDC-32071P) f o r Susquehanna using the GE SAFERIGESTR methodology were reviewed t o determine the extent o f fue l damage t h a t i s expected dur ing the design basis LOCA.

The SAFERIGESTR methodology (Ref. NEDC-23785P, Vol 11) f o r ca l cu la t i ng fue l rod bal looning and rupture consists o f a cladding hoop s t ress versus temperature curve below which cladding per fo ra t ion i s assumed not t o occur. This methodology was used f o r the Susquehanna power uprate p ro jec t and i s conservative f o r current condi t ions i n t ha t the analysis assumed h igher LHGR and MAPLHGR values than the current p lan t operating l i m i t s . No bal looning was predicted f o r the best estimate DBA LOCA case and no fue l rod f a i l u r e s are predicted f o r the l i cens ing basis DBA LOCA case (Ref. Telephone c a l l w i t h D. Pappone (GE) on lO/l9/92). Therefore, fue l f a i l u r e fo l low ing bal looning i s not expected t o occur f o r the design basis LOCA:

Figure 1 shows the thermal shock f a i l u r e curve and associated experimental data (Ref. NUREG-1230, Section 6.14). For Susquehanna, the peak l i cens ing basis cladding temperature i s 1510°F (1094'K) and peak 1 icensing basis l o c a l ox idat ion i s less than 0.25%. These resu l t s are wel l below the f a i l u r e boundary i n Figure 1. Therefore, fue l f a i l u r e as a r e s u l t o f ox ida t ion embritt lement and subsequent thermal shock i s not expected t o occur f o r the design basis LOCA.

I n summary, the po ten t i a1 f ue l f a i l u r e mechanisms and SAFER/GESTR LOCA analysis resu l t s have been reviewed. No fue l rod f a i l u r e s are expected t o occur f o r the design basis LOCA.

References

1. NUREG-1230, "Compendium o f ECCS Research For Real i s t i c LOCA Analysis," December 1988.

2. NEDC-32071P, "Susquehanna Steam E l e c t r i c S ta t ion Uni ts 1 and 2, SAFERIGESTR-LOCA Loss-of-Cool ant Accident Analysis," May 1992.

FIGURE 1

Thermal Shock Failure Map for Zircaloy Cladding

Source: NUREG-1230, "Compendium of ECCS Research for Realistic LOCA Analysis," December 1988.


NE-092-002 Rev. o

ATTACHMENT 7

USE OF UHS WITH RHR IN

FUEL POOL COOLING ASSIST MODE

(ET-0870)

M E M O R A N D U M

TO: D. K o s t e l n i k A6-3

FROM: J. Ca j igas A6-3

JOB: SSES

l :

NUMBER: ET-0870

DATE: October 27,1992

COPIES: M. M jaa tvedt A6-3 Nuc. Records F i l e A6-2 Eng. Tech. F i l e A6-3

FILE: P8B-6 REPLY: NO

SUBJECT: USE OF UHS WITH RHR I N FUEL POOL COOLING ASSIST MODE

The RHR system can be a l i gned i n t h e f u e l pool c o o l i n g a s s i s t mode i n t h e event o f a l o s s o f f u e l pool c o o l i n g cond i t i on . Fo l lowing a DBA, use o f t h e RHR system i n f u e l pool c o o l i n g mode r e q u i r e s two u l t i m a t e heat s i n k (UHS) spray d i v i s i o n s t o be ava i l ab le . Loop "A" would support t h e " A " RHR system loops, i n f u e l pool c o o l i n g mode, f o r bo th f u e l pools w h i l e l oop "I?" supports suppression pool/shutdown c o o l i n g a t each u n i t . Since t h e design bas i s s i n g l e f a i l u r e o f t h e UHS s a f e t y ana lys i s i s f a i l u r e o f one loop o f sprays, t h i s al ignment i s beyond the design bas i s o f t h e spray pond.

The expected maximum normal f u e l pool c o o l i n g heat l o a d on t h e UHS i s 13 m i l l i o n BTU/HR per pool o r 26 m i l l i o n BTU/HR per RHRSW loop ( o r per UHS spray d i v i s i o n ) . UHS thermal analyses performed t o da te seem t o i n d i c a t e t h a t t h i s RHR f u e l pool c o o l i n g heat l o a d and DBA shutdown heat loads can be d i s s i p a t e d by two spray pond loops w i thou t exceeding t h e system design temperature. Th is scenar io, however, has never been q u a n t i f i e d by an ana lys is . Fur ther , t h e c u r r e n t RHRSW opera t i ng procedure w i l l need r e v i s i o n t o p rov ide s u f f i c i e n t guidance on how t o o b t a i n t h e optimum spray e f f i c i e n c i e s requ i red f o r t h i s al ignment.

4

Per t h e above d iscuss ion , i t can be concluded t h a t t h e UHS design should support t h e RHR f u e l pool c o o l i n g mode prov ided t h i s al ignment i s j u s t i f i e d w i t h UHS thermal analyses and t h e RHRSW opera t i ng procedures are r e v i s e d t o p rov ide proper spray ar rays a1 ignment i n s t r u c t i o n s .


NE-092-002 Rev. 0

ATTACHMENT 8

DRAINAGE OF CONDENSATION FROM THE

818' ELEVATION DURING FUEL POOL BOILING

(ET-087 1)

Page 1 of 2

M E M O R A N D U M

TO: D.G. Kostelik DATE: 10/28/92

FROM: K . G . Browning 7 7 ,

JOB: EDR G20020 NUMBER: ET-0871 COPIES: ET Memo File

FILE: REPLY: None Required

SUBJECT: Drainage of Condensation from the 818' Elevation During Fuel Pool Boiling

This memo is written in follow-up to our discussion regarding the possibility of condensate flowing down the reactor building stairwells during fuel pool boiling scenarios.

-- - - - - -

1 Problem During accident scenarios where the fuel pools are allowed to boil, condensation will occur on the wall and ceiling surfaces of the refueling floor elevation. Since the stairway access doors to the 818' elevation (stairwells 101, 102, 201, & 202) are not water tight, it is possible for this condensation to eventually drain down the stairwells to the basement of the reactor building.

Investiqation To address this concern. a review of the 818' Floor Plan Drawings (c-243 thru C-i55), along with the Area Drainage Drawings (P-25-7 thru P-34-7) was performed. The results are summarized below.

.

Results The high point elevation for the refueling floor is 818'1". In certain places, such as the steam dryerlseparator pit and washdown areas, the high point is several inches higher since these areas are curbed. However, in general, area perimeters are defined by the high point elevation.

From this high point perimeters, the floor is sloped 2" to drains which are located at an elevation of 817'1lV. There are over 40 drains on the 818' flooring which are generally located at the centers of these perimeters. Each of these drains has a diameter of 4 " which provides for ample collection pathways for the fuel pool condensate. Since there is such a large cumulative collection area, flow of the condensate would be essentially unrestricted.

There are several hatchways on the 818' elevation, including the train & truck bay hatches which are potential flow paths. However, flow through these plugs would be minimal, if any,

EPM-lOlB, Rev. 1

Drainage of Condensation from 818' During Fuel Pool Boiling scenarios

for the following reasons: 1) The edges around hatches are high points (818'lW), from

which the floor slopes away to drains. As a result, there would be minimal accumulation around the hatch seal.

2) Some of these hatches, such as the New Fuel Vault Hatch, are specifically designed to be water tight.

3) For those hatches which are not water tight, the size, weight, and fit of the hatch plugs would act to adequately seal off this flow path.

Similarly, flow through the stairwell doorways (if any), would be inconsequential for the following reasons:

1) The perimeters near doorways are high points (818'ln), from which the floor slopes away to drains. As a result, flow would tend to drain away from these points of access.

2) There are airlocks between the stairwells and the 818' air space. As a result, there are several barriers to prevent whatever water "seepagew might occur. Also, at least one of these doorways is curbed; the Unit 2 access point.

While there may be a small amounts of accumulation (i.e. puddles) near the doorways, there would be no motive force to drive significant quantities of condensate through the door seals. Therefore, the flow of condensation down the stairwells of the reactor buildings is not a significant issue, and does not compromise the viability of venting Zone 111 (i.e. allowing the fuel pool to boil).

References The following references were used in the resolution of this issue:

1) Architectural Drawing: A-17 2) 818' Floor Plan Layout Drawings: C-240 thru C-255 3) Drainage Piping Drawings: P-25-7 thru P-34-7

EPM-1019, Rev. 1


NE-092-002 Rev. o

ATTACHMENT 9

SGTS FIRE DAMPERS

(ET-0750)

ENGINEERING TECHNOLOGY MEMORANDUM

TO: M . R. Mjaa tved t A6-3 DATE: October 2 0 , 1992

FROM: D . J. Kohn A6-3 COPIEB: G . D. Miller A6-3 J . E . Agnew A6-3 J . E . Schleicher A1-2 S . E . Dav i s S S E S

JOB : NUMBER: ET-0750

FILE: A20-1, E D R G 2 0 0 2 0 REPLY: NO

BUBJECT: EDR G20020

...................................................................

T h i s c o n f i r m s o u r d i s c u s s i o n concern ing the SGTS f ire dampers be tween the Control S t r u c t u r e and the Unit 1 Reac to r B u i l d i n g (P- l i ne w a l l ) .

S p e c i f i c a t i o n 323c i n d i c a t e s t h a t a l l f u s i b l e l i n k s t h a t o p e r a t e fire dampers a r e r a t e d a t 1 6 0 - l 6 5 O F . The h i g h e s t t e m p e r a t u r e shou ld be k e p t a t l e a s t 3 0 ' ~ be low the f u s i b l e l i n k r a t i n g o r 130°F . I f t h i s t empera tu re i s t o o low, then the f u s i b l e l i n k s c o u l d be r e p l a c e d w i t h l inks r a t e d a t a h i g h e r t e m p e r a t u r e . L i n k s r a t e d a t 2 1 Z ° F and 286OF a r e r e a d i l y a v a i l a b l e . These p r o v i d e for a maximum o p e r a t i n g t empera tu re o f abou t 190°F or 250°F . B o t h would be a c c e p t a b l e since the f u s i b l e l inks would o p e r a t e and close the fire dampers before a l l o w i n g 325OF a i r t o pas s t h rough the f ire w a l l . B a s i c a l l y the accep tance c r i t e r i a for a fir& w a l l i s t o k e e p the c o l d s i d e o f the f ire w a l l be low 3 2 5 O ~ .

The f o l l o w i n g dampers would r e q u i r e m o d i f i c a t i o n :

FPD-3-2 7-8-lsc FPD-3 -2 7-8-3SC FPD-3 -29-8-1 FPD-3-29-8-2 FPD-3-29-8-3 FPD-3-29-8-4

I f I can p rov ide a d d i t i o n a l i n f o r m a t i o n , p l e a s e c o n t a c t me.

n

Attachment 31

PP&L Memo from Glenn D. Miller t o George T . Jones, "Revised Evaluation o f EDR 620020 - Spent Fuel Pool Cooling Issue", October 29, 1992 ( P L I - 7 2 7 6 4 )

October 29, 1992


SUSQUEHANNA STEAM ELECTRIC STATION REVISED EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-72764 F ILE A45-1A

The attached revised evaluation o f EDR 620020 i s provided f o r your information. The rev i s i on accounts f o r a change t o the engineering repor t reference, corrects a typographical e r ro r on page 10, expands the event t r ee t o three pages and provides several graphs used as pa r t o f the PORC presentat ion on Monday October 26, 1992. The conclusions remain unchanged.

Glenn D. M i l l e r

cc: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W. Simpson - A1-2 D. F. Roth - SSES H. G. Stanley - SSES J. M. Kenny - A2-4 J. S. Stefanko - A9-3 F. G. But ler - A6-3 J. R. Mil tenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records - A6-2 D. A. Lochbaum - Enercon

October 21, 1992


SUSQUEHANNA STEAM ELECTRIC STATION EVALUATION OF EDR 620020 - SPENT FUEL POOL COOLING ISSUE PLI-7 711 1

The attached evaluat ion of EDR 620020 i s provided i n response t o your memo PLI-72640. This evaluation was prepared by myself and a team o f engineers working on the act ion plan t o resolve the subject EDR.

I n the course o f reviewing t h i s issue i n de ta i l , we have concluded t h a t seven o f the n ine i d e n t i f i e d discrepancies are not v a l i d def ic ienc ies. This i s explained i n d e t a i l i n the evaluation. The two remaining discrepancies are v a l i d de f i c ienc ies but are not considered safety s i g n i f i c a n t nor reportable.

I be1 ieve t h a t the technical basis on several o f these issues has been c l a r i f i e d considerably i n the course o f the past week. Therefore, I suggest prov id ing t h i s evaluat ion t o Nuclear Regulatory A f f a i r s f o r reconsiderat ion o f the repo r tab i l i t y aspects.

We are cont inuing w i th the remaining act ion items as requested. A de ta i l ed engineering design repor t and j u s t i f i c a t i o n f o r in te r im operation w i l l provide more d e t a i l than contained i n the attached evaluation. We are scheduled t o meet w i t h PORC on Monday October 26, 1992 a t 2:30 pm t o review t h i s issue. We are working w i t h Systems Engineering, Operations and NRA-Compliance w i t h respect t o po ten t i a1 compensatory measures.

A D . 4 Glenn D. M i l l e r

cc: G. J. Kuczynski - SSES J. E. Agnew - A6-3 C. A. Myers - A2-4 M. R. Mjaatvedt - A6-3 M. W. Simpson - A1-2 D. F . Roth - SSES H. G. Stanley - SSES J. M. Kenny - A2-4 J. S. Stefanko - A9-3 F. G. Bu t le r - A6-3 J. R. Mil tenberger - A6-1 D. C. Prevatte - A6-3 Nuclear Records - A6-2 D. A. Lochbaum - Enercon


This document contains an evaluation o f the discrepancies documented i n EDR 620020, "Loss o f Spent Fuel Pool Cooling Event Design Discrepancies." Conclusions o f the author w i th respect t o r e p o r t a b i l i t y o f these concerns and ope rab i l i t y impact on SSES are also provided.

Desian Basis

The design bases f o r the Fuel Pool Cooling System are found i n FSAR sect ion 9.1. The por t ions o f the design basis relevant t o EDR 620020 are as fol lows:

Maintain the fue l pool water temperature below 125F under "normal maximum heat loads" defined as 12.6 MBtu/hr (equivalent t o a t yp i ca l fue l cycle discharge schedule which f i l l s the fue l pool, l a s t quarter core o f f l o a d a t 6.7 days a f t e r shutdown).

Maintain fue l pool water temperature a t o r below 125F dur ing the "emergency heat load" condi t ion o f 32.6 MBtu/hr (equivalent t o a f u l l core o f f l oad 10.5 days a f t e r a shutdown fo l lowing a t yp i ca l f ue l cycle discharge schedule which f i l l s the fue l pool) u t i l i z i n g the RHR system (w i th o r wi thout normal fue l pool cooling) f o r fue l pool cool ing. This mode o f operation appl ies "during periods o f higher than MNHL generation i n the fue l pool, eg., s to r ing o f a f u l l core o f i r r ad ia ted fue l sho r t l y a f t e r shutdown". The RHR system i s used under these condi t ions t o ass is t the FPCCS i n d iss ipa t ing the decay heat. Thus, any heat load i n excess o f 12.6 MBtu/hr i s considered t o be w i th in the design basis f o r the RHR FPC ass i s t mode o f operation.

Redundant Seismic Category I ESW connections t o each pool are provided t o al low f o r makeup o f evaporative losses i n the event o f f a i l u r e o f the FPC system. The condit ions are bounded by a fue l pool t ime-to-boi l analysis based on the same typ ica l fue l cycle discharge schedule as i n basis #1 except the time a f t e r shutdown i s 10.5 days instead o f 6.7 days resu l t i ng i n a heat load o f 9.8 MBtu/hr. (This explains the d i f ference between the two d i f f e r e n t heat loads, ie . , 12.6 MBtu/hr f o r basis #1 and 9.8 MBtu/hr f o r basis 83. This i s not a discrepancy.) The ESW makeup l i n e i s sized on the basis o f t h i s ca lcu la t ion (Reference FSAR sect ion 3.13).

The cause o f the Loss o f Spent Fuel Pool Cooling event i s s ta ted t o be a seismic event.

A l l p ip ing and equipment shared w i th o r connecting t o the RHR i n t e r t i e loop are Seismic Category I and can be iso la ted from any p ip ing associated w i th the non-Seismic Category I fuel pool cool ing system.

Evaluation o f Discreaancies Noted i n EDR 620020

EDR 620020 describes nine discrepancies r e l a t i n g t o the loss o f spent fue l pool cool ing event. This discussion w i l l summarize each issue. The reader i s



General Statement

The in t roductory paragraph o f "the EDR states: "...the design p rov is ion f o r the loss o f spent fue l pool cool ing event i s t o permit the fue l pool t o b o i l and maintain i t s water l eve l above the fue l through makeup from the ESW system. This design prov is ion i s necessary because the fue l pool cool ing system used f o r normal operation and the RHR fue l pool cool ing ass is t mode used f o r abnormal heat loads are n o t designed t o s a t i s f y seismic category I and s ing le f a i l u r e c r i t e r i a . "

As s ta ted i n design basis #5 above the RHR fue l pool cool ing ass i s t po r t i on o f the p ip ing i s designed t o seismic category I requirements. No c r e d i t however i s taken f o r t h i s mode o f operation i n the fue l pool b o i l i n g analysis i n the FSAR. Credi t i s taken f o r t h i s mode f o r emergency heat load s i tua t ions as def ined by basis #2.

I n order t o discuss and evaluate each o f the nine discrepancies l i s t e d i n the EDR i t w i l l be more l o g i c a l t o review them i n a d i f f e r e n t order. Items E & F both re1 ate t o the t ime-to-boi l ca lcu la t ions and w i l l be reviewed f i r s t fo l lowed by items G through I, which are re la ted t o the t ime-to-boi l concern. Items C & D invo lve operator act ion considerations and w i l l be discussed next. F i n a l l y items A & B r e l a t i n g t o the evaporation e f fec ts w i l l be discussed.


"The ana ly t i ca l 25 hour t ime-to-boi l f o r the spent fue l pool i s nonconservative f o r the maximum normal heat load i n the spent fue l pool ." As stated i n basis #1 the maximum normal heat load i s 12.6 MBtu/hr. As s ta ted i n basis #2 the t ime-to-boi l analysis i s based on a heat load o f 9.8 MBtu/hr. These two design bases are i n f ac t consistent and are based on the same " t yp i ca l fue l discharge schedule" and re fue l ing outage scenario. The d i f fe rence i n the heat load i s due so le ly t o the time a f t e r shutdown assumed f o r purposes o f estab l ish ing the design basis.

Focusing on the t ime-to-boi l analysis, a time a f t e r shutdown value o f 10.5 days i s used. This i s the time a t which i t i s assumed t h a t re fue l ing i s completed and the reactor cav i t y t o fue l pool gates are re ins ta l led . P r i o r t o t h a t po in t the addi t ional water stored i n the reactor cav i ty i s also avai lab le as a heat s ink and the RHR system i s ava i lab le f o r fue l pool cooling. For times greater than 10.5 days the appropriate heat load w i l l be even lower than the analyzed value o f 9.8 MBtu/hr. For the SSES Un i t 2 5RI0 the time from reactor shutdown t o fue l pool gates i n s t a l l e d was 38 days. The decay heat i n the Un i t 2 pool a t t h a t time i s ca lcu la ted t o be 5.65 MBtu/hr (Reference NE-092-002). The corresponding time- (A to -bo i l i s 45 hours.

The EDR goes on t o discuss other ca lcu la t ions which r e s u l t i n d i f f e r e n t heat loads using various assumptions. Calculat ion NFE-B-NA-053 was performed by Nuclear Fuels t o account f o r actual fue l discharge h i s to ry and fu tu re of f loads accounting f o r power uprate conditions. The fue l pool heat load versus time curves obviously w i l l increase subsequent t o power uprate, however, these curves do not apply t o the ex is t ing design. As long as the calculated decay heat i s l ess than 9.8 MBtu/hr a t the po in t where the fue l pool gates are r e i n s t a l l e d the o r i g i n a l design basis t ime-to-boi l ca lcu la t ion i s s t i l l va l id .

Calculat ion M-FPC-009 determined t ime-to-boi l condit ions post power uprate. This ca l cu la t i on shows t h a t the t ime-to-boi l f o r the design basis heat load o f 9.8 MBtu/hr i s s l i g h t l y greater than 25 hours.

I n conclusion, the design basis f o r the t ime-to-boi l condi t ion i s establ ished by the 9.8 MBtu/hr value used i n the o r i g i na l calculat ions. This design basis i s met by planning the outage so tha t the fue l pool i s not i so la ted from the reactor cav i t y o r the RHR system p r i o r t o a po in t i n time where the actual heat load i s 9.8 MBtu/hr o r less.



I tem F: Time-to-Boil f o r Emeraencv Heat Load

"The ana ly t i ca l 25 hour t ime-to-boi l f o r the spent fue l pool does not account f o r the emergency heat load i n the spent fue1,pool."

As discussed above, the t ime-to-boi l condit ions apply t o conf igurat ions where the spent fue l pools are i so la ted from the reactor cav i t y ( ie. , non-refuel ing conf igurat ions). As i s co r rec t l y stated i n the EDR, current p rac t i ce i s t o f u l l y o f f l oad the core dur ing each re fue l ing outage. Speci f ic ca lcu la t ions are performed by Nuclear Fuels t o determine the a b i l i t y o f the FPC system t o remove the combined decay heat o f the cross-t ied re fue l i ng pools. Tests are also conducted t o determine tha t the actual heat removal capab i l i t y exceeds the actual fue l pool heat loads during the outage (Reference TP-235-011). Normally the reactor c a v i t y i s maintained flooded and cross-t ied t o the fue l pools. One loop o f Core Spray i s always operable i n t h i s conf igurat ion. One d i v i s i o n o f RHR i s maintained i n shutdown cool ing mode except f o r a b r i e f per iod required f o r the common RHR system outage window.

Design basis #2 states t ha t heat loads i n excess o f the MNHL are considered t o be emergency heat loads. The design o f the RHR system t o ass is t the FPC system dur ing emergency heat load condit ions assures t ha t fue l decay heat i s removed. No t ime-to-boi l ca lcu la t ion f o r t h i s conf igurat ion i s required since the RHR system w i l l be i n operation o r avai lable. A t any rate, such a ca lcu la t ion should consider the e f f e c t o f the addi t ional water inventory ava i lab le from the flooded reactor cav i ty , cask storage p i t and dryer and separator storage pool which are a l l cross-connected dur ing t h i s time. Makeup inventory i s a1 so avai lab le from Core Spray and the RHR system i s normally in-service except f o r the common RHR system outage window.

I n conclusion, no t ime-to-boi l analysis i s required f o r the emergency heat load design basis. Single f a i l u r e s o f the RHR system are not required f o r t h i s design basis f o r the emergency heat load (Reference SRP 9.1.3).

This discrepancy i s not a v a l i d deficiency, not reportable and has no impact on p lan t operabi 1 i ty.


Item G: Radioloqical Release Calculation for Boilinq Spent Fuel Pool

"The radiological release analysis for a boiling spent fuel pool uses nonconservative evaporation rates."

This discrepancy is directly related to the heat load assumed for the time-to- boil analysis. The evaporation rate used in the dose calculation is based on a heat load of 9.8 MBtu/hr which is the design basis heat load for the time-to-boil calculation. Heat loads in excess of 9.8 MBtu/hr obviously result in higher evaporation rates. Since the discussion under Item E above establishes that 9.8 MBtu/hr is the correct original design basis and still bounds current operation there is no discrepancy in the offsite dose calculation. It uses an evaporation rate consistent with the design basis heat load.

ThSs discrepancy i s not a valid deficiency, not reportable and has no impact on plant operability.


"The rad io log ica l release analysis f o r a b o i l i n g spent fue l pool uses nonconservative a c t i v i t y terms. The o r i g i na l design ca l cu la t i on (200-0048) assumed 12 month operating cycles and 184 bundle equi l ibr ium reload sizes t o determine the a c t i v i t y terms f o r f a i l e d fue l i n the fue l pool. SSES cur ren t l y has 18 month operating cycles w i th approximately 230 bundle reloads which w i l l increase t o approximately 254 bundles a f t e r power uprate. Since the ca lcu la t ion impl ied t h a t most o f the a c t i v i t y resu l ts from the most recent discharge batch, the e f f e c t o f increasing the discharge size from 184 bundles assumed i n the ca lc t o 230 and 254 bundles would appear t o be nonconservative w i t h respect t o the radio1 ogical re1 ease analysis."

The o r i g i n a l rad io log ica l release analysis as referenced above i s conservative f o r the fo l low ing reasons:

(1) the a c t i v i t y l eve l s used as a source term are based on 1% f a i l e d f ue l . A l l o f the f a i l e d fue l rods are assumed t o be i n the of f loaded batch o f 184 f ue l assemblies. Therefore, increased batch sizes w i l l not increase the amount o f the source term used i n t h i s analysis.

(2) the a c t i v i t y l eve l s used f o r the iodine source term are based on saturat ion l eve l inventor ies f o r a core operating a t 3440 MWt f o r one thousand days. Therefore, the fue l cycle length w i l l not a f f e c t the source term.

I n conclusion, the o f f s i t e dose ca lcu la t ion remains va l i d .

This discrepancy i s no t a v a l i d deficiency, not reportable and has no impact on p lan t ope rab i l i t y .


Item I: Analvsis f o r Max Time Pr io r t o Makeuo

"The analysis f o r maximum time p r i o r t o makeup t o a b o i l i n g spent f ue l pool i s based upon nonconservative assumptions. The or ig ina l design ca lcu la t ion (175-14) determined the time using evaporation o f the e n t i r e fue l pool water inventory. The maximum time should be based upon a minimum fue l pool water l eve l which i s s u f f i c i e n t l y above the top o f the fue l t o provide the sh ie ld ing required t o al low cor rec t i ve operator actions."

The purpose o f the referenced ca lcu la t ion was t o determine re fue l i ng f l o o r atmosphere condi t ions under various operating modes. The evaporation ra tes and assumptions used i n the c i t e d por t ion o f the ca lcu la t ion were used so le l y t o determine i f condensation could be expected under fue l pool b o i l i n g condit ions. The conclusion o f the ca lcu la t ion regarding time t o b o i l the pool d r y i s not re levant t o any operator action. Operator actions are based on maintaining normal pool l eve l and temperature condit ions. I n any case, the c i t e d nonconservatism would have a minor e f f ec t on the calculated 19 days t o b o i l the pool dry, a r e s u l t which i s not used elsewhere i n the design.

This discrepancy i s not a v a l i d deficiency, i s not reportable and has no impact on p lan t operab i l i t y .


Item C: Manual ESW Valve Actions

"The manual valve manipulations required t o provide ESW makeup f low t o a b o i l i n g spent fue l pool may not be possible."

In-plant post-accident rad ia t ion l eve l s are analyzed f o r SSES t o the requirements spec i f ied i n NUREG-0737. This document requires t ha t post-accident rad ia t i on l eve l s be determined f o r purposes o f v i t a l area access by p lan t operators t o perform short-term f i r s t p r i o r i t y actions. It speci f ies t ha t rad ia t i on l eve l s be determined on the basis o f contained sources, and core damage source terms equivalent t o those used f o r lOCFRlOO calculat ions. These assumptions are c l e a r l y based on degraded core condit ions which are beyond the design basis LOCA. Airborne r a d i o a c t i v i t y sources from containment leakage are required t o be analyzed f o r environmental qua1 i f i c a t i o n o f equipment but not f o r personnel access.

A review o f FSAR chapter 18 shows tha t access t o the equipment necessary t o provide makeup t o the fue l pool from ESW i s res t r i c ted f o r the approximately the f i r s t 30 hours fo l lowing the design basis event (Figure 18.1-9). This analysis i s based on a conservative source term equating t o 100% fue l damage r e s u l t i n g from core me1 t condi t ions as o r i g i n a l l y u t i l i z e d f o r o f f s i t e dose ca lcu la t ions used t o determine p lan t s i t i n g adequacy. These source terms were based on experiments invo lv ing heated i r r ad ia ted uranium dioxide pe l l e t s .

An evaluat ion o f actual fue l thermal response dur ing design basis accidents r e s u l t s i n no predicted fue l f a i l u res (Reference PLI-72696). Thus, the source term r e s u l t i n g from the DBA LOCA would only be equivalent t o the r a d i o a c t i v i t y present i n the reactor coolant as a r e s u l t o f normal operations (a l lowing f o r fue l defects as permitted by Technical Speci f icat ions). To bound the po ten t ia l e f f ec t s o f a design basis accident, a r e a l i s t i c y e t conservative analysis using an assumed 1% fue l damage resu l t i ng from core degradation under LOCA condi t ions was performed (Reference EP-548) and concludes tha t access t o equipment necessary t o m i t i ga te the e f f ec t s o f a loss o f fue l pool cool ing fo l low ing a DBA LOCA i s assured.

I n conclusion, post-accident operator actions are v iab le f o r a l l po ten t ia l scenarios under consideration, f o r both the current design basis and those outside o f the current design basis.

This discrepancy i s no t a v a l i d deficiency, i s not reportable and has no impact on p lan t operabi 1 i ty .


Item D: Instrumentation

"The instrumentation available to the operator post-LOCA does not provide adequate indication of spent fuel pool tzmperature and level to allow proper response to a loss of fuel pool cooling event."

The instrumentation available to the operator is not required to be qualified since the design basis loss of spent fuel pool cooling is not coincident with the DBA LOCA conditions. This instrumentation is powered from an uninterruptible power supply and its' associated 1E AC source.

The minimum water level required per Tech Specs is below the weir elevation. Since ESW makeup is provided to the pool the operators will know that when they see a rise in skimmer surge tank level the fuel pool level is at least as high as the weir. This provides a confirmation of adequate pool level without requiring access to the refueling floor.

Furthermore, on the basis of the discussion in item C above, access to the refueling floor is possible under all considered conditions. Therefore, it is possible to verify adequate fuel pool level visually from the refueling floor which is accessible from several locations.

While the available instrumentation is adequate for operator actions and meets the regulatory requirements of Reg Guide 1.13, improvements to the instrumentation have been recommended in the past and should be implemented. This would enhance plant safety.

In conclusion, the existing instrumentation is adequate for performance of required operator actions for the current design basis and for scenarios not included in the current design basis.

This discrepancy is not a valid deficiency, it is not reportable and has no impact on plant operabi 1 i ty.


Item A: Reactor Bui ld ing Design Heat Loads

"Reactor bu i l d i ng design heat loads do not account f o r the b o i l i n g spent fue l pool event."

The reactor bu i l d i ng temperature analysis i s performed f o r DBA LOCA condit ions. The presumption o f i tem A i s tha t the spent fue l pool w i l l reach b o i l i n g condi t ions p r i o r t o res to ra t ion o f fue l pool cool ing subsequent t o a LOCA. The ex i s t i ng temperature analysis does account f o r the sensible heat load from the fue l pool a t 212F.

A loss o f spent fue l pool cool ing event can r e s u l t from several condi t ions. The design basis condi t ion i s a seismic event as analyzed i n the FSAR. The Fuel Pool Cooling system i s not designed f o r seismic loads. I n t h i s case, the Fuel Pool Cooling system i s assumed t o be l o s t . An evaluation o f the p lan t response shows tha t several methods are avai lable t o assure t ha t the spent fue l remains cooled. These include: (1) the RHR system can be used t o cool the fue l pools w i th a l te rna te shutdown cool ing o f the reactor using Core Spray and RHR f o r suppression pool cooling; o r (2) a l low the fue l pool t o b o i l w i t h makeup supplied by ESW w i th consideration o f e i t he r SGTS operating on Zone 111 o r prov id ing a vent path from Zone 111. The o f f s i t e dose analysis takes no c r e d i t f o r SGTS. I f avai lable, normal reactor bu i ld ing ven t i l a t i on would be used t o provide cool ing and venting o f the Zone 111 atmosphere. Under any o f these scenarios t ranspor t o f moist a i r t o other port ions o f the reactor bu i l d i ng would not occur. This scenario i s the design basis f o r loss o f fue l pool cool ing.

Other scenarios not included i n the design basis include LOCA and LOOP events, and combinations thereof. The time frame f o r consideration o f operator act ions i s based on reasonable expectations f o r the t ime-to-boi l condi t ion. As s ta ted previously, f o r the current operating pract ice, the fue l pool heat load p r i o r t o reactor r e s t a r t i s approximately 5.65 MBtu/hr. Time t o b o i l under t h i s condi t ion i s on the order o f 45 hours. Note tha t t h i s i s the shortest possible time-to- b o i l f o r the current fue l cycle. With the pools cross-t ied the t ime-to-boi l i s

Im greater than 100 hours.

For a LOCA scenario, the FPC system w i l l be l o s t i n i t i a l l y due t o the Aux Load Shed provis ions. Although the Fuel Pool Cooling system and other non-safety re la ted systems are not s p e c i f i c a l l y analyzed f o r the e f f ec t s o f hydrodynamic loads i t i s expected tha t they w i l l be able t o perform t h e i r normal funct ions fo l low ing a broad spectrum o f design basis events. Credi t f o r these systems i s not needed t o meet the design basis, however, p lan t operators w i l l u t i l i z e any equipment ava i lab le t o them during emergency s i tuat ions. Therefore, i n the course o f evaluating the e f fec ts o f a DBA LOCA on the fue l pool cool ing system, we acknowledge the a v a i l a b i l i t y o f normal p lant systems i n responding t o the emergency.

Independent o f the LOCA condit ion, o f f s i t e power i s needed t o restore normal cool ing systems. The SSES Indiv idual Plant Evaluation considered loss o f o f f s i t e power (Reference I P E Appendix F). The IPE conservatively estimated the incidence o f LOOP t o be .04/year (p lant re1 ated), .008/year ( g r i d re la ted) , .00807/year


(severe weather re1 ated) , and .00066\year (extreme1 y severe weather re1 ated) . The p r o b a b i l i t y o f recovery from the LOOP w i th in spec i f ied times was also calculated as fo l lows:

Time f h r l P(Recoverv w i t h in T h r s l 12.0 97.96% 24.0 99.53% 60.0 99.923%

Thus, it can be reasonably concluded tha t o f f s i t e power w i l l be avai lab le w i t h i n 24 hours fo l low ing the i n i t i a t i n g event.

The remaining fac to r i n addressing res to ra t ion o f fue l pool cool ing i s access t o the reactor bu i ld ing . This issue i s discussed under Item C above. Except f o r degraded core condit ions access t o the reactor bu i ld ing i s feas ib le a f t e r the f i r s t twelve hours o f the i n i t i a t i n g event (Reference EP-548).

Notwithstanding the above basis, we have also considered the scenario where o f f s i t e power i s not ava i lab le and access t o the lower reactor bu i l d i ng elevat ions i s r e s t r i c t e d (based on FSAR chapter 18 contained source terms). Under these condi t ions (representative o f a degraded core event) access t o the re fue l i ng f l o o r remains avai lable. Provision f o r fue l pool cool ing i s made through use o f the p lan t f i r e protect ion system. Venting o f Zone 111 v ia the f i l t e r e d exhaust system i s also possible f o r t h i s scenario. While access t o l eve l and temperature instruments would be questionable i t i s possible t o v e r i f y adequate pool l eve l v i s u a l l y from the re fue l ing f l o o r which i s accessible a t several locat ions.

A Loss o f Fuel Pool Cooling Event Tree i s attached t o t h i s evaluat ion t o help guide the reader through these various postulated scenarios.

I n conclusion, f o r the design basis loss o f fue l pool cool ing the p lan t as cu r ren t l y designed and analyzed i s acceptable. For other scenarios not s p e c i f i c a l l y included i n the design basis we have reasonable assurance t h a t the e f f ec t s o f a loss o f fue l pool cool ing can be mi t igated without adverse consequences on the p lan t .

This discrepancy i s a v a l i d deficiency. It i s not a safety s i g n i f i c a n t issue because we have establ ished reasonable assurance t h a t the e f f ec t s o f a l oss o f fue l pool cool i ng can be mi t igated without adverse consequences on the p lan t and pub l i c heal th and safety and i s therefore no t reportable. The evaluat ion above a lso shows t h a t t h i s concern does no t impact p lan t operab i l i t y .

I n consideration o f t h i s concern, addi t ional analyses are warranted t o f u r t he r quant i f y the e f f ec t s o f evaporation and b o i l i n g condit ions on the Zone 111 atmosphere and the po ten t ia l t ransport o f moist a i r t o other locat ions i n the reactor bu i l d i ng f o r condit ions outside o f the current design basis.


"The impact o f the ESW makeup water t o the spent fue l pool on equipment i n the reactor bu i l d i ng has not been evaluated.",

The analysis under i tem A above applies t o t h i s issue as wel l . This evaluat ion shows tha t w i t h the current p lan t design and f o r ex i s t i ng design basis condi t ions the e f f ec t s o f a loss o f fue l pool cool ing are acceptable.

This discrepancy i s a v a l i d deficiency. As w i th i tem A i t i s no t a sa fe ty s i g n i f i c a n t issue and i s no t reportable. The evaluat ion above a lso shows tha t t h i s concern does no t impact p lan t operab i l i t y .

I n consideration o f t h i s concern, addi t ional analyses are warranted t o f u r t h e r quant i f y the e f f ec t s o f evaporation and b o i l i n g condi t ions on the Zone I11 atmosphere and the po ten t ia l t ransport o f moist a i r t o other loca t ions i n the reactor bu i l d i ng f o r condi t ions outside o f the current design basis.


Enaineerinq Reoort on Loss of Soent Fuel Pool Coolinq

A detailed report, NE-092-002, is being prepared to document this evaluation in llil further detail . This report contains technical input from several engineering groups and will provide a comprehensive set of references on this subject. The report will be completed by October 28, 1992.

Design Basis Decay Heat Last ~a'tch Offload to Single Fuel Pool

13 Max Normal Heat Load

12 \ -.

11 Time-to-Boil Calc

10

9

8

7 -----------------: -

6 - U2 5RIO - * A

5

4 5 10 15 20 25 30 35 40

Days After Shutdown

Figure A - Loss of Fuel Pool Cooling Due to Seismic Event

Loss of Fuel Pwl Coding. r i

I Use RHR FPC assist with Anemate SDC if available.

otherwise:. I Use RHR FPC assist with Ailemate SDC if available.

otherwise:.

$.

Allow to boil with ESW rnakw.

1 if SGTS shtdown is rewired,

then:.

1 vect 2111.

h ZIII vetted.

Restore FPC wiih mrmal -\

Figure C - Loss of Fuel Pool Cooling Due to Other Causes

Use RHR FPC assist wilh Ailemate SDC if available,

othemise:. r i


PP&L Memo f r o m D a v i d A. Lochbaum and Dona ld C . P r e v a t t e t o George T. Jones, " P o s i t i o n on EDR 620020 and P lanned A c t i o n s " , November 2, 1992 (PLI -72783)

Note: T h i s memo p r o v i d e d PP&L w i t h an u p d a t e as t o t h e a u t h o r s ' i n t e n t i o n s w i t h r e s p e c t t o t h e d e c l a r a t i o n made on Oc tobe r 9, 1992 ( A t t a c h m e n t 19) o n a r e p o r t t o t h e NRC on November 2, 1992. The a u t h o r s d e f e r r e d a r e p o r t t o t h e NRC p e n d i n g PP&L 's p r o m i s e d f o r m a l r e p o r t o f t h e i r own on t h i s m a t t e r .

November 2, 1992


SUSQUEHANNA STEAM ELECTRIC STATION POSITION ON EDR G20020 AND PLANNED ACTIONS PLI-72783 FILE A45-1A

--

In a meeting with you on October 9, 1992, we submitted a letter again expressing our deep concern reyarding the boiling spent fuel pool safety issues we raised in EDR G20020. After months of unsuccessful attempts to have these issues properly addressed per PP&L procedures and federal regulations, we formally requested of you that the screening, reportability, and operability processes required by procedures be completed, the attendant documentation be provided to us, and that these concerns be formally communicated to the NRC in accordance with the Code of Federal Regulations by November 2, 1992. We stated that should these actions not be taken by that date, or should they be technically inadequate or incomplete, it was our intent to report these concerns to the NRC, also as required by federal regulations.

Since that meeting, significant efforts have been mounted by PP&L to address these concerns, significant progress has been made, and we have been assured by you that a formal report will be made to the NRC. The latest manifestations of these efforts are a revision to Mr. Miller's October 21 evaluation which, in essence, still maintains that our concerns are of no safety significance and not reportable, and a report, NE- 092-002, Rev 0, Loss of Fuel Pool Cooling Event Evaluation for EDR # G20020, 10-29-92, which is substantially more balanced, more complete, and better documented than any previous assessments, but with which we still have major disagreement on a number of basic technical points.

Therefore, in recognition of this continuing progress and ycur assurance, we will not make our report to the NRC on November 2. However, our fundamental positions with respect to the major technical concerns and the obligation for prompt reporting have not changed. We believe that today, within the NRC mandated design basis requirements, most if not all of the safety-related systems and functions in the reactor buildings are "In an unanalyzed condition that significantly compromises plant safety, in a condition that is outside the design basis of the plant, and in a condition not covered by the plant's operating and emergency procedures".

We know that the above cited report is the primary product of the efforts of the last three weeks, and we expect that in generating your report to the NRC you will give it heavy

consideration. We urge you to review it very carefully, considering the following comments.

Although the report provides a wealth of information which could be used as justifications for interim operation, we strongly disagree with its conclusion that ''...it is possible in all cases [emphasis added] to provide adequate pool cooling and protect safety related equipment in Zones I and I1 from the environment produced from a boiling pooln, as well as many of the positions the report takes to support these conclusions. Although we agree it would be possible in some cases, it would not be possible for the design basis case of LOCA/LOOP and many lesser cases. Additionally, the report's conclusions are based on expectations for future analyses, future modifications, and future procedure changes, not the plant as it exists today. The evaluations of operability and reportability are required by law to be based on the plant today, not our expectations for the future.

We believe that the report's conclusions are incorrect for the following reasons:

1. They place heavy reliance on non-safety-related equipment.

2. They place heavy reliance on modifications to the plant which have not yet been implemented or even designed.

3. They place heavy reliance on procedure changes which have not yet been made.

4. They place heavyreliance on analyses which have not yet been performed.

5. They place heavy reliance on operator and EOF personnel training which has not yet been accomplished or even developed.

6. They place heavy reliance on operator actions during a LOCA when there is already heavy reliance on operator actions and monitoring. These additional actions must be performed under extremely adverse environmental conditions in the reactor building.

7. The conclusions are based on assessments of operator accessibility to the reactor building which in turn are based on assumptions of core damage which are unreviewed by the NRC and are substantially less than the assumptions required by NUREG-0737 and our licensing basis reflected in Chapter 18 of the FSAR. Additionally, the accessibility position taken in the report with respect to airborne radiation is inconsistent with NUREG-0737; 10CFR50, Appendix J;

actual Appendix J test results for SSES; the design of other plant system systems (e.g. secondary containment and SGTS); and common sense. For NRC mandated DBA conditions, the reactor building is inaccessible for days into the accident.

8. They rely on probability arguments, arguments which may be acceptable for an IPE or a JIO, but which are not acceptable substitutes for NRC mandated design bases, unless they are reviewed and approved by the NRC. These have not been.

9. In numerous areas, the report's conclusions are not consistent with the facts presented, e.g., the report concludes that Zone I11 venting is acceptable; the supporting documentation, by contrast, shows that the lOCFRlOO and 10CFR50, Appendix A, Criterion 19 allowables for offsite and control room doses respectively are exceeded.

Our more detailed specific comments on the report are provided on the attached sheets.

We would like to express our appreciation to you and to all who have worked hard in addressing these concerns. We have come a long way; we still have some distance to go; and like you, we are committed to going that distance. As always, we are at your service.

cc: C. A. Myers Miller Sgarra Kenny Agnew McGann Kuczynski Stanley Miltenberger Keiser Byram Corcoran

Kemper

Doty Iorf ida Sabol Licht Stef anko Butler Zola Mjaatvedt Boschetti Sweeney Gogates Manski Richardson

A2-4 A6-3 A2-4 A2-4 A6-3 SSES S&A-4 SSES SSES A6-1 TW-16 A6-1 21 Broadleaf Circle Windsor, CT 06095 115 Polecat Road Glen Mills, PA 19342 A9-3 SSES A2-5 A6-1 A9-3 A6-3 A6-3 A6-3 SSES SSES SSES Enercon Enercon

COMMENTS OF DAVID A. LOCHBAUM AND DONALD C. PREVATTE ON NE- 092-002, REV 0, LOSS OF FUEL POOL COOLING EVENT EVALUATION FOR EDR G20020, 10-29-92

Abstract, first sentence, EDR G20020 also expressed concerns with the existing FSAR discussion for the seismic event of loss of fuel pool cooling (i.e., effect on reactor building heat loads and effects from boil-off and spillover) . Page 3, the assumption of 25 hours to boil is improper given that EDR G20020 challenges the basis for that time and EDR GO0005 questions the accuracy of FSAR Chapter 9 and FSAR Appendix 9A.

Page 5, 1st sentence, even though the FSAR assumes a seismic event causes the loss of fuel pool cooling, it is only one of many loss mechanisms.

Page 5, 2nd sentence, although the FSAR postulates the loss during a refueling outage, it can occur at any time.

Page 5, second paragraph, the current heat load is not lower than 9.8 BTU/hr until well beyond 10.5 days because SSES performs full core offloads each refueling outage.

Section 3.2, 2nd paragraph, although the instruments are UPS powered, they are not 1E qualified, not seismically qualified , and not environmentally qualified for LOCA or for the boiling spent fuel pool. Therefore, the instruments would not necessarily remain operational.

Section 3.2, 3rd paragraph, the skimmer surge tank level is not a reliable indicator of fuel pool level. Page 6, is the procedure to drain the skimmer surge tank described here a part of current plant procedures and operator training?

Section 3.3, this section confirms our concern in EDR G20020. Although it refers to actions that can be taken, these are not part of the present plant procedures.

Section 3.4, second paragraph, although it is true that NUREG-0737 does not requlre consideration of airborne sources due to leakage of systems outside containment, this is aimed at the ECCS systems which operate post-LOCA and form an extension of the containment boundary. It is not necessary to consider their leakage because they are normally filled and pressurized, and therefore they would not be expected to leak. Additionally, the leakage involved would be liquid leakage and would be a relatively small contribution to the total airborne dose. However, NUREG-0737 does not exclude the leakage through containment isolation valves which are closed post-LOCA.

The leakage of these the Type B penetration leakage constitutes the leakage of concern which is consistent with the requirements of 10CFR50, Appendix J. It is a real and significant valve as our tests show, and it may not be igqored. Additionally, PP&L response in FSAR Chapter 18 to the NUREG requirement recognizes this leakage, it just doesn't account for it.

Page 8, first full paragraph, FSAR Chapter 18 does not im~lv that the reactor building will be inaccessible for several days following the LOCA - it plainly states this. On page 8, second paragraph, from a "design basis perspective " the Appendix J leakage is required to be considered, not just contained sources. Appendix J leakage also doesn't include leakage from the water filled ECCS systems that operate post-LOCA.

Page 8, third paragraph, the model uses the design basis leakage is not just from Type B penetrations; it is also from Type C penetrations which is consistent with the NUREG 0 7 3 7 requirements.

Page 8, fourth paragraph, access to the refueling floor is not possible post-accident for the design basis accident due to airborne radiation levels in the hundreds of Rfhour per PP&L analyses in the files.

Page 9, first paragraph, first sentence, the assumption of 1% fuel cladding damage is not conservative; this is the allowable fuel cladding damage for normal operation. This assumption basically allows for no damage as a result of the accident. This is non-conservative.

Second, the RHR system operating in the fuel pool assist mode is unanalyzed in the first place, and is particularly unanalyzed while at the same time responding to a LOCA. Additionally, the effects of this mode of operation in a LOCA on the RHRSW system, the ESW system and the spray pond are unanalyzed.

Third, the exposure to the operator in manipulating the ESW valves for fuel pool makeup is unanalyzed.

10. Section 3 . 5 , this section confirms our point, the RHR fuel pool cooling assist mode is unanalyzed for post-LOCA response .

11. Section 3 . 6 , this paragraph is true. However, for design basis for the safety function of fuel pool cooling, this is not adequate.

12. Section 3.7, there are several problems with this section.

First, the makeup flow to the pool is through a manual throttling valve which is not calibrated and for which there are not throttling instructions in the procedures. The procedures only require that the valve be opened. Unless otherwise stated, that means fully opened. At the full open position, the valve would likely flow considerably more than 30 gpm; the number that has been used for previous analyses has been 60 gpm/pool. At this flow rate, over 30 days, this is 5.2 million gallons of water - more than the sumps can hold. Second, for the design basis accident condition, the water cannot be purged out of the building for several reasons: The pumps are non-1E powered, non-seismically qualified, non-environmentally qualified, etc.; the same is true of the radwaste processing system; the radwaste systems are not designed to handle this volume; the radwaste systems are not designed to handle DBA LOCA contaminated water; and finally, this is an unreviewed safety question.

Third, this does not address all of the other flooding effects, e-g., possible structural failure of HVAC ductwork, blockage of HVAC ductwork, spillage from HVAC ductwork on safety-related components, etc.

Section 3.8 addresses recovery. It is predicated on access to the areas, significant operator actions in a potentially hostile environment for the DBA LOCA, use of non-safety related systems, modifications not yet made or even designed, and unanalyzed, nonexistent procedures. This may be possible, but is clearly outside many of the regulatory and design basis requirements.

14. Section 3.8 takes credit for operating the filtered exhaust of the normal HVAC system. It is unlikely that this system would provide adequate filtering function for pool boiling conditions due to wetting of the charcoal. This is an unanalyzed condition.

15. Page 13, the report does not consider the seismic event. This is in conflict with 10CFR50, Appendix A , Criterion 2 and FSAR Section 6.2 which describes the worst case scenario for containment functional design as LOCA/LOOP with concurrent SSE. The report's analysis nonconservative and inconsistent in that all consequences of SSE (e.g., loss of fuel pool cooling) are not considered.

16. Page 14, Event #1, this description requires both loops of RHR to be available. It does not address single failure in one of the loops.

17. Page 14, Event #1, LOCA unit, the last statement is not necessarily true. Current design basis calculations for

the reactor building temperature require the fuel pool cooling system to be out of service along with the other non-1E powered loads in the reactor building. The service water system is also non-safety related and cannot be taken credit for in the design basis.

Page 14, non-LOCA unit, RHR in the fuel pool cooling mode as well as RHRSW, ESW and the spray pool are unanalyzed for the condition described . Use of the condenser as a heat sink is outside the design basis since it is non- safety related and requires non-1E power.

Page 15, concern #1, the last statement is not necessarily true for design basis requirements, even without a LOOP.

Page 15, concern #2, ESW could not fulfill the purpose of makeup to the pool, due to inaccessibility of the valves.

Page 15, concern 8 3 , this concern states that for flows greater than 2000 gpm the skimmer surge tank runs dry. Operating procedures require a flow rate of 5000-6000 gpm for the RHR fuel pool cooling assist mode. ~t would appear that this is more than just a concern, it would appear to be a major discrepancy in the stated plan,

Page 16, third paragraph, if actions are not taken immediately to isolate Zone 111, for the DBA, then all zones become inaccessible, even the non-LOCA unit zone, due to airborne contamination. It is unreasonable to expect that the operator can take the necessary actions to isolate Zone I11 in time before the contamination is spread to all zones with the current design.

Page 16, last paragraph, this paragraph discusses exposures being within legal limits for the case that does not even represent NUREG 0737 requirements.

However, it does not appear not to consider the concept of ALARA.

Page 19, next to last paragraph, this states that if RHR is not available for fuel pool cooling, boiling with makeup is allowed if Zone I11 is vented. No. At this point in the accident, Zone I11 will already be highly contaminated from the accident since up until this time, the three zones were cross connected. Therefore, it will be impossible to gain access, and even if access were possible, the release of the unfiltered airborne radiation from Zone I11 would likely exceed the lOCFRlOO limits.

Event #3, page 20, fuel pool boiling may be permitted by current procedures, but it is an unanalyzed condition.

Although it may not rely on non-safety related systems for the non-boil scenario in this case as pointed out, it does involve an analyzed, untested mode of RHR, RHRSW, ESW and spray pond operation.

26. Event #4, page 21, this description dismisses the LOCA/ LOOP for the outaqe situation as "sufficiently small so as not to be considered.' This is a probability argument, appropriate for an IPE or JIO, but not appropriate for the design basis except with the approval of the NRC.

27. Section 5.1, page 22, we do not agree that "... it is possible in all cases to provide adequate pool cooling and protect safety related equipment in Zones I and I1 from the environment produced from a boiling spent fuel pool. "

28. Section 5.1, Design Basis, it is concluded that the fuel pools will boil for the design basis case. For today's plant, this would appear to create a reportable condition in light of short term action #1 on page 24 which shows the SGTS to fail if the pool boils.

29. Section 5.1, Design Basis, last paragraph, this concedes that I t . . . if significant airborne radiation is present it will be necessary to evaluate Zone I11 venting against mixing zones and usinq SGTS." If the airborne is high, neither option is available. If venting is chosen, the operator gets too much exposure and the lOCFRlOO limits are exceeded. If mixing is chosen, the reactor building safety systems plus SGTS are not qualified and are likely to fail. For these design basis conditions, we do not meet the design basis requirements.

30. Page 22, for the "realisticn case, even if these non- design basis conditions exist, it states that plant procedures must be changed to have a workable system. This would appear to meet the test, of reportabllity (required, not voluntary) under 10CRF50.72, paragraph (b) (ii) (c) , by being ''In a condition not [currently] covered by the plant's operating and emergency procedures."

31. Figure 1, the time-to-boil is shown for initial fuel pool temperatures of 1 0 0 ~ ~ and llo°F, yet the initial design temperature for the fuel pool is 125OF which would represent shorter times-to-boil.

32. Figure 2, same as Comment 27 above.

33. Figure 9, for the LOOP or LOCA/LOOP case, what prevents condensate from collecting in the ductwork at the low point and causing collapse or blockage?

34. The report does not address the fact that for isolation of Zone I11 for which great credit is taken, the airborne concentration in Zone I (11) will be significantly higher due to less dilution. Therefore, for the current SGTS flowrate the offsite dose will be significantly higher than shown by the current analyses. The airborne exposure to the operator in Zone I (11) will also be significantly higher. These are also currently unanalyzed conditions.

35. Attachment 2, Airborne Dose, Section I, 1% cladding failure is a non-reviewed safety question with respect to the assumptions required by NUREG-0737.

36. Attachment 2, Section 11, Contained Sources, this section does not calculate the doses for the NUREG-0737 required case release from the core of 100% noble gases, 50% halogen inventory, and 1% other constituents.

37. Attachment 2, Section 111, Offsite Consequences of Reactor Building Purge, this section states that the doses shown in the tables must be added to the FSAR, Chapter 15 doses. For the design basis case in Table 1, even without adding these doses to the Chapter 15 doses they exceed the lOCFRlOO and 10CFR50, Appendix A, Criterion 19 allowables for offsite and control room doses respectively. Therefore, they don't support the position taken by the report that Zone I11 venting is allowable.

38. Attachment 3, Zone I11 Ventinq, the actions required here include blanking off supply pienums since just-closing the dampers will not be adequate. These actions would require substantial manpower and time under adverse conditions even if all the materials and tools were pre- staged. This entire analysis is based on design features and procedures that are not in place and are unanalyzed.

Attachment 3, Instrument Air Supply, Seismic Event, this section only considers the effects of the seismic event on the system tubing. The rest of the system would also be susceptible. In addition, it is susceptible to failures from other causes since it is non-safety- related.

Attachment 3, SGTS Operability, this section is based on conjecture and unanalyzed scenarios, and it even contradicts itself in two places; first, it says the condensation will be drawn into the filter train (this would incapacitate the filters) and second, it says the ductwork would likely fail structurally, both of which would constitute failure of SGTS.

41. Attachment 3, Assessment of Emergency Ventilation options

for Zone I11 Venting during Fuel Pool Boil Scenarios, this entire section is based on designs and procedures that are currently unanalyzed and don't exist, e.g., none of the fan loads, if they are to be powered from the emergency diesel generators, have been analyzed with respect to the diesel generator available capacity.

42. Attachment 3, EQ Equipment Located in Zone 111, this section does not show that all devices required to qualified are indeed qualified with respect to humidity Additionally, the temperature and radiation effects are not addressed at all.

43. Attachment I, Fuel Performance During a LOCA, this analysis is an important consideration for a JIO, but it is not in accordance with the "minimum" requirements of NUREG-0737.

44. Attachment 7, Use of UHS with RHR in Fuel Pool Coolinq Assist Mode, this attachment concludes that (1) this is an unanalyzed condition, and (2) it will only work if single failure is not considered, both unacceptable for a design basis.

45. Attachment 8, Drainage of Condensation from the 818' Elevation During Fuel Pool Boiling, while this analysis appears to be very good with respect to conditions on the refueling floor, it does not address the questions regarding what happens to the water that goes down the drains, e.g., does it create a flooding hazard for safety-related equipment in the building.

46. Attachment 9, SGTS Fire Dampers, per this section, the existing fire damper fusible links are unacceptable and design modifications are required. This alone would constitute a reportable item.

Attachment 33

PP&L Memo f rom D a v i d G . K o s t e l n i k and Mark R . M j a a t v e d t t o George T . Jones, "Comments on P L I - 7 2 7 8 3 R e g a r d i n g EDR G20020", November 1 1 , 1992 ( P L I - 7 2 8 5 7 )

November 11, 1992

George T. Jones ~ 6 - 2

SUSQUESANNA BTEAM ELECTRIC STATION COMMENTS ON PLI-72783 REGARDING EDR 020020 ~%740181/003 PLI- 72857

FILE: A45-lA

This letter provides you with our assessment of the comments containe* in th" subject letter regarding report NE-092-002, Rev. 0, "Loss of Fuel Pool Cooling Event Evaluation for EDR G20020". In this letter, Messrs. Lochbaum and Prevatte identify forty-six (46) comments related to specific sections of the report. They also provide general comments in the cover letter regarding the intent of the report and its conclusions.

It is not our intent to provide a resolution to each comment listed in the subject letter. We believe that this would be counter productive and simply result in additional letters from Messrs. Lochbaum and Prevatte restating their position. We do, however, feel compelled to provide some comment on their letter in order to assure that the report is reviewed by others in an objective fashion.

Upon reviewing the statements made in the cover letter, it is apparent to us that the thrust of this letter is to promote the views of Messrs. Lochbaum and Prevatte by casting doubt as to the soundness of the evaluations presented in NE-092- 002. This is evidenced by the statements which indicate that the realistic assessments made in the report cannot be applied to design basis, and that the conclusions drawn in the report are incorrect because of the reliance on operator actions and non-safety-related equipment.

The report does not state that realistic assumptions will be applied to design basis, and further indicates that, if design basis assumptions are made, then plant operability may be questionable. The intent of the realistic evaluation of each event was to demonstrate that, under the type of realistic assumptions permitted by the NRC for operability evaluations, the plant is operable for a loss of Fuel Pool Cooling event. The report recognizes the need to evaluate the issue further so that appropriate changes can be made to the plant design and design basis.

After reviewing the comments attached to the cover letter we believe that they can be grouped into the following categories:

1. Twenty-six (26) comments make statements to the effect that it is improper to use realistic assumptions in place of design basis assumptions/requirements. These are predominantly made with regard to the realistic discussions of the report. We believe it is appropriate to make such assumptions. Examples are comments number 10, 11, and 19.

2. Three (3) comments represent statements taken out of context. A prime example of this is comment number - 26, wh'ich reads:

"Event #4, page 21, this description dismisses the LOCA/LOOP for the outage situation as llsufficiently small so as not to be considered.@# This is a probability argument, appropriate for an IPE or JIO, but not appropriate for the design basis except with the approval of the NRC.I1

The actual sentence, which is being made from a I#JIO1* viewpoint, reads as follows:

"Furthermore, the probability of a LOCA/LOOP, concurrent with both loops of RHR out of service in an outage, is sufficiently small so as not to be c~nsidered.~~

3. Eleven (11) comments represent statements made by Messrs. Lochbaum and Prevatte to promote their views on a specific area of the report. Examples are comments 3 and 4.

4. Five (5) comments represent areas where Messrs. Lochbaum and Prevatte apparently misunderstood the report. Examples are comments 15 and 21.

5. There is one (1) comment which is made that identifies a true inaccuracy in the report. This is comment number 1 which appropriately points out that the abstract does not fully identify all of the concerns raised by the EDR. This will be corrected if and when a revision is made to NE-092-002.

As previously stated, we are not planning to respond to PLI- 72783 on a comment by comment basis. We do not believe that this will be productive given the nature of the comments. We simply wish to express our opinion regarding these comments for your consideration when evaluating the report in light of

the comments made by Messrs. Lochbaum and Prevatte. If you wish to discuss this letter or the report, please feel free to contact Dave Kostelnik at x7788 or Mark Mjaatvedt at x7795.

David G. Kostelnik

cc: G. J. Kuczynski C. A. Myers M. W. Simpson H. 2. Stanley J. S. Stefanko J. R. Miltenberger J. E. Agnew G. D. Miller M. R. Mjatvedt D. G. Kostelnik M. H. Crowthers D. F. ROth J. M. Kenny R. R. Sgarro F. G. Butler

SSES A2-4 A1-2 SSES A9-3 A6-1 A6-3 A6-3 A6-3 A6-3 A6-3 A6-3 A2-4 A6-3 A6-3

afk,At?@ aatvedt

D. C. Prevatte A6-3

Nuclear Records A6-2

Attachment 34

PP&L Letter from H. G. Stanley to the U.S. Nuclear Regulatory Commission, "Licensee Event Report 92-016-OO", November 17, 1992 (PLAS-546)

Note: PP&L's formal report t o the NRC o n the concerns expressed in EDR 620020.

U.S. Nualeat Regulatory Conraiamion Document Control DML wamhinqton, W 2055S

SUSQUEHUHA STEM ELECTRIC STATION LICENSEE EVENT REPORT 92-016-00 FILE R41-2

-546

Dockat No. 50-387 Liasnos No. NPP-14

Attachmd is Limnmas w a n t Rope* 91-016-00. Although it warn datarmhod thnt thim condition is bot regortnblr, t h i s voluntary roport 5s baing eubmittad t o provide the C-hsion with i n f o r ~ a t i o n about thm Btation'a spent Fuel Storago Poola.

H.G. Stanley Suporintendant of Plant - Suaquehanna

oc: Hr. T. T. Martin ~ a g l o ~ l A d r l n i & t t 4 t O r , Region I 0.8. ~ h u r Rqulrtory Comminmion 475 Irllendrle Raad King of erUrrda, PA 19406

Mr. Q. S. Barber Sr. Roaidmt 1n.protor U.S. Huclmar Regulatory Commfsmion P.O. Box 35 Borwlok, PA 18603-0035

On April 16, 1997, Enginewm worm performing 8valuetiona as par t of M e fu turo upratod licmnmmd power project. Coneoram wmro ra1m.d th&t We mxhtinq analyaim for tha s tn t i onBs tw B p n t ?uml Storage Poolm did not r o f l a c t the currant fuml domign and omrat ion of the plant. Additional conoarnu cmtmrod around the

I a b i l i t y to re-emtab1i.h fuol Pool Cooling (?PC) and Fuel Pool Weup i f Purl Pool Cooling i m lo*. Thm concatnm anr. dommmtod on an Enginoaring Discrepancy Report and mubnmqumtly avaluatod. Tho won t w.6 detmrninoa not to M rq tor tablo par lOCFR!30.71, 50.73. R u l dmmign and plant opmxational chanqu m a d m i n amaoaiation with spant fuel .toraga wmrm not reflmctod i n t h e 8 ta t ion88 Final Safaty Analyuia Report (FSAR) but warm datMinmd to bm b u d a d By the mximting 4smign h i m . me corumquonaes of the loam of m a 1 Pool Coolinq were &terninad to ba mathfactory f o r tho domiqn basi8 lomm of PPC. Tho long t m r m oiiwtm of tho lomm of FPC for evmntm involving L a a m of Cooling Accident8 (LOCX'w) end C o r m o l Offmite Powmr ( W P 4 a ) arm beyond the domign b a d 8 amlynim. Additional anaLyur arm pluulmd t o f u r t h m r quantify thr e f f w t a of tho bayond design bud. mcanarion.

The ralwnnt deaigh bases raquiromants for tho s~.quohanna 8tem Elootric Station*# (BBES) Purl Pool Cooling S y m t a m (FPCS) (BIIB C0DS:m) -0 fwnd i n 8UbmCtion 9.1.3 of thq Station1= Final arfmty Anrlysir Report (FSAB). Thoy ho1ud.r

1. Maintain the fUol pool v a t m hmporatuto blow 125.F under %orma1 maxlaua heat loadsw (10111L) whioh is datineid as 12.6 nBtu/hr.

2. ? l a h t a i n fue l pool wator tompmrature a t or b low 129.1 during thm "emergency haat loadw condition equivalent to a f u l l core offlond 10.5 days a f tor a ahutdoun following a

'YP ical f u r l cyela tiisahat e which f i l l s the fuel pool. W s im accompliahod by u t 1 l i r i n g t h e R e m i d u a l Reat Romoval (m) (EXIS CODEICE) myatam (with or without normal fuel pool cooling) for fue l pwl cooling. Thim m o b OS oporation appliea "during porioda of higher than m L ganeration i n the fuel pool, og., storing of a f u l l aorm of frradiated rum1 ehortly a f t e r shutdownH. Thm RSiR aystam i s used under those oonditiono t o aaairnt tho FPcS i n dismipating the decay heat.

3. Rdundsnt Sdsrrio Category I Enorgurcy Smrvicr Water ((~m) (MIS C0DE:aI) wnmction8 to each pew1 are providad t o allow for -up of evaporative l o u r s in +hr eveht of f a i lu re of tho PPC swtm.

I 4. Tho design bnmim muam of loes of Fuml Pool Cooling i e a aoioa.ic *vent.

Thm Station's deeign reeponro t o a coraplrtm lws of Fuel Pool coolinq duo t o a 68 . i r i a w a n t i* t o al lw tha -1 P o o l to boi l w i t h inventory nrlcmup providad from a Safoty rolated mourca, the Eaorqonoy Sorvior water (EsPO) Bystor Thm analymir for t h i s wont I8 d0ouolent.d i n Appondlx 9A i thm P W .

During a typiaal refunling outage, a11 fue l aauahliorn are removed from tho raactor and plramd in UI. Spent -1 Pool. This provides fo r groatar control of the corm raams.rbly by lsmmning tho chcrnco of mimlwding a fun1 amammbly. Conplmto core offlod/reloadm a l w prwidm greater control of Shutdown t(rrqin and allow f o r flexibility while parforainq myatam maintonuace. Thim however, a1110 p h c u a grantor heat rmaval burdon on tho FPe8 than originally plnnnmd when approximately o m quartat of a core warn t o b. offlaaded and tho remalninq fu.1 5huffl.d am

modad. m i c a 1 outaga practicm is t o maintain ona loop of RKR i n lJIutdorn coolinq u n t i l fka fallowing t w o conditions a ro ut: A l l rum1 i m remvad and M a &cay h u t load ef the furl i8 less than tba capabi l i ty of tho rum1 Pool Cooling Symtm. A t t h i n tima with t ha r r fua l q8t.s ruovod, the RX Cavity and me1 pool arm cromathd. This provides r mubatantfa1 inarum* i n water voluma thereby dolaying t h e o m e t of boil ing ahould a11 cooling bm lomt.

On April I6, 1992, Enginearm (contractor, non-1icurm.d) vere ovaluatiwu am part of tha futurm upratad liconmod

power pmrf Omi"9 pro act [Power Upfatm Projoet). Tha Enpimum quamtloned t h e adaquaey of tho eximting analymim f o r tho Btationrm two Spent Puml Storaga Pool8. Additional conaomr, with r..prct t o th. a b i l i t y t o re-smtablinh Purl Po: ! Cooling and Fuel Pool makaup following pomtulatad accident c- 3i t ion8 ware 81.0 rained. The concurrim ueto documantad oh an ~ i ~ m r i n q Dimorepamy Roport (EDR) which wa8 than aubjretod a mcroaning procorns. Several evalwtionm haw k e n parfarnod 3 ortab1i.h tha nafoty mignificanco of tha isbsumm rais. . Am a r-lt. reviowlng, o v r l w t i n g aad dimpcrsitionlnq t ,- EDR f q u i r a d mvara l month8 to aoqle~te. Tha mR idantitlam nl:w mpocif5.o aonaornm. The c o n e a m faaum on thur Mtae rrcin areas:

1) Furl danign and plant o p e r c l t i o ~ l ahango# a r e not rmfloctod i n tho F8AR analyuim. Bub8rgumnt analy8ia ham shown that a r r a n t praatica i m bounded by tho domign h m i s a ~ l y d m .

I a) Tho long-term mtfwtm of incr-od evaporation rntsm that a boi l ing Spent Bus1 Pool could ormate.

3) Tba Bu.qruhanna demiqn baaim analyaom do not oonmidaf t h a t a lo88 of Fuol Pool Cooling ovant could b8 caumd by other than a niamic mnt. The EDR 1d.ntit imm aonaarno ammooiatd with r WCA or m / U I O P typm avoat and t h m a b i l i t y t o provldo h u l Pool Cooling following 8uah aventm. It should k nakd t ha t wnmid.rat5.m of this .p rc f f i c scnn8rio wam bryor\d tha dosign hsim of th. Fuel Pool.

The t u a l pool kriling and radiological raleaaa a n a l y s u i n tha FSAR arm d&p.ndmnt on fu.1 damign -tars and o p u a t i n g practicrm. ma PBAII anrlysim is ba8.6 on t ha original gum1 d u l g n (8 x 8 fual , 1 2 month fue l cyclo, 1/4 oat. rmloadm)

. . . . . ..

UCCNIil EVENT REPORT ILUI) TEXT CONTINUATION

whrrmam t h e ourront fuel dmmiqn i n 9 x 9 fu r l , 18 month f u e l a p l a , 1/3 core rmloadm. The F M S aamum*~~ rofualinq i m nccomp1imh.d by mhuffling fuml within thm reactor corm sfluream current 88118 praatiom i m t o fu l l y off1m.d the corm and then rrload. Xmviou of t l u m 8 concsmm ha8 .ham that current p lan t conditionm remain within tha bound8 of t h e o r ig ina l FSAR mrlymim.

Two concarnm ware raisod r agad ing ths long-tam effec t# impomad on the reactor building environment by tha increamm i n evaporation from the fua l poola mubmrqumt to loma of m o l Pool cooling. Th. i n c t e a ~ d waporation i m p o m e m additional long-term heat loads on tha reactor building and the uatmr mamm remalting from condmnmation of th is moistura could accurnulatr within the r.acr+or building etructura. _

. - Although a lorn of -1 ~oo$&mling w a n t aould r u u l t from mavaral conditionm, t h e d t m a b a m i s condition is a reimmic want am onslyrrd i n t h e PSAR. T&:Fur1 Bool Cooling mymtem is not damignad f o r maimio loada. I n thh aam, the Erual Pool cooling mymtmm i m asmuned to bm daapad and unrvailmble f o r w a l i n g t h e Fuel Pool. The design bnals p lant rampons6 i m analyard for the radiologioal oonmaquencsr of tblm avant, t h a t i m , allowing the f u a l pool t o bo i l witb mawup supplied by ESW. Ik.ul t ing o f f s i t e domu arm oa1culat.d t o be w a l l within required l i a i t a and adaquafa naltaup to maintain the fua l coverod wikh water i m a86urrd. Evaluation of tha e f fec ta of incrmnal evaporation and wndanuation on tha reactor building wrm kyona the o r ig ina l damfqn basis conaidarm4 for the Purl Pool Cooling mystam. Operation of BGTB i m a n t i e i p a t d undar thoma oonditionm, hwavar, offmito do- analydm f o r -1 Pool boiling t a k u no crmdit f o r M T S .

Other eourariom beyond the dmaign bas r l o r of Fuel Pool Cooling evmt miah aould cauwm a ah--tera loam of Pus1 Pool Cooling inoluda pomtulatul LOEL and LOOP evant8. Although they azr clmarly beyond tho clrrrmnt damign bad8 f o r the Fual Pool Cooling myatem, avaluatianm arm ongoing i n odor t o deternine tha nmd f o r any mub..prunt action..

I Thm cmuwm of th. f a i l u r e t o modify tlm analysim of l l w m of Spent Fuml Pool Cooling as dacumu1t.d i n thm FEAR am: 1). Bumquohanna'r raload mrlymw did not adequrtaly sddrw impacts on t h m ma1 Bool Cooling design analymim, i n part beoauw of lack

LICENIEE M l Y T REPORT ILEII) TEXT CONllNUATDN

¶ -uaqdmaStean-frie- m r w ~ * . u e M r r W ~ k ~ W 1 R

of adequate involvmwnt of nymtmn demign enginaerm i n reviewing the analysim of mloadm. 2) Raviaw of chmgem i n oporatlng modes, f u l l core offload 6s a nomul practicm, did not identify tho ne.d Ca revime tlu FBAR analymlm.

The con-nn invalving come enopc trm a LOCA or MC?~/rxloP m n t havm ariman due t o dam 'r qn bad8 review8 a680dat.d w i t h the s t a t ion fa Powu Uprata Projwt . The conaernm wmre not reviawe8 am par t of the original plant danign b e ~ u n r thmy w a r e ammidared t o bm beyond tha deuign baaic to r the -1 Pool Cooling myatsa.

The evmnt was determined net t o &a reportable undar the requiromentm o t the Cod0 of Federal xmqulatioru, mpt.r 10 Part8 50.72, 50.73. The operational chunqea made i n ammoolation with S p a t Fuel starage -re not reflected i n the Station's Fa*R, but w e e Swnd t o bo w i t h i n t h e demign b m i m for the plant. Uming input. t h a t rmfl.et t he ourrent demign and aprration of t h ~ plant, analymis of the pertinent desigh liliunuing rmquiruruatm of the r u m 1 Pool Coollng Byitom have &own that: t he Station laas not operated outside thm lieenming barnis. more apacifirrally, current operation i s bounded by the FSAB analysib r u u l t e for both dway h u t load and offmita do-. The concern. associated w l t h a LOCA or LQCAJLOOP mvmnt ~ 0 n C u r C ~ t w i t h a lonu of Fuel Pool Cooling have b a a detenuin.d to l ie beyond tho demlgn and liamnmihg basrs.

It waa a lso detemhmd tha t boiling of tth .el pool warn conaidormd only bmb.oaWa of a pre-liccmming docketed drclmion t o reclamsity thm F u m l Pool cooling Bystam am N O ~ - B @ ~ ~ I B I O category I. In order for thl6 t o b. approvd, fue l pool boiling waa rmqulred t o ha ammummd t o r a seismic avent, a d the result ing offmitm dome caloulated. Similar conditionm were not required t o b. applied to t he me1 Pool Cooling r y s t u undu LOOP o r XU% ecenariw.

The currant evaluation of repor tabi l i ty and operability CUIUid~at ionl for -0 matte8 waa OOaplmbd on 10/21/92. Thm evaluation concluded t ha t the demign bad. lor8 of fue l pool cooling, am currently deaignod and analyxd, i m aoceptable. him i m r contlnual procemm t h a t w i l l ba revisi ted am portlnurt informmtion b m c o m m available.

Although the evmt w a m detuminsd n o t t o be rmportable, Ulla rmport is h i n g mubittad to thm Caaolission for

purpomu. Lamsonm laarned in arrmociation w l t h evont could k u u f u l t o the rmt of the hdumtry.

11-23-1992 10:01 P. 08/08

UCENsLf EVENT REPORT f L W TEXT WIMTlNUATION

Pottinont mctionm ol thm FSMt will bm rovimd to account for slhanqu that have been made to tho dmmiqn of the rwl, fuol cycle operation, and the oporationrl modes of tho apmnt fuel pools.

Procodural and proae88 changu asmoaiatod witb fuol reload design I rill D. made to enmure tha inpact to otbor system is addnued I for futura reloads.

A review will b8 completed to identity any additional aroao whore changrm auociatad with iu8l reload doaign might impact the design of other iy8tem.u not currantly addra68.d.

Additional analy.08 are plannod to further quantity the effect. of evaporation and boiling conditions on th8 refuollng floor atmo8ph.m and thm potential tranmport of moimt air to othmr loutlonm in th. reactor building for conbitfan8 bmyond the current h u l Po01 Cooling ayetam design bamim.

Proc~uroa as weL1 as operatar training will bo dovelopad or aoditiad to ptwidm kttmr guidanccr to tho oporatorm in monitering tho swnt fuel pool, reemtabliohinq W o - u p to the pool, and ronrtabliahing fuel pool cooling should it be lort. #odificatione, including iaprnvrd instmumentation for monitorinq 02 tho fuol pool f r m tho main control room arr a100 undmr evaluation.


PP&L S a f e t y E v a l u a t i o n Summary, "P rocedu re EO-IP-055". 1988 (SER No. 88-127)

Note : T h i s summary o f t h e 10CFR50.59 s a f e t y e v a l u a t i o n p r e p a r e d b y PP&L i n s u p p o r t o f t h e i r 1988 i m p l e m e n t a t i o n o f a p r o c e d u r e t o m a n u a l l y shed t h e non-1E l o a d s i n t h e r e a c t o r b u i l d i n g a f t e r a LOCA w i t h o u t a LOOP t o p r e v e n t r e a c t o r b u i l d i n g room t e m p e r a t u r e s f r o m e x c e e d i n g v a l u e s used f o r E Q pu rposes shows how a "new f a i l u r e mode" f o r t h e l o s s o f f u e l p o o l c o o l i n g e v e n t d e s c r i b e d i n FSAR Append ix 9A was n o t p r o p e r l y addressed. I f PP&L 's c o n t e n t i o n t h a t t h e o n l y f a i l u r e mode i n t h e i r l i c e n s i n g b a s i s i s f r o m a s e i s m i c e v e n t , t h e n t h i s a c t i o n i n 1988 c o n s t i t u t e d an u n r e v i e w e d s a f e t y q u e s t i o n .

SER NO.: 88-127

CROSS REFERENCE: Procedure EO-IP-055

DESCRIPTION OF CHANGE:

Implementation of procedure EP-IP-055 which crates a selective load shed of all non-essential electrical loads within the Reactor Building.

SUMMARY:

I. No. FSAR Section 15.6.5 discussed loss of coolant accidents coupled with severe natural environmental conditions. The proposed action represents a less severe condition than discussed in this FSAR Section. The loads shed by EP-IP-055 do not affect any system required for safe shutdown.

11. No. The loads to be shed by the proposed action are non-essential electrical loads in the reactor building for the accident unit only. The load shed does not affect any system required for safe shutdown as described in FSAR Section 7.4.

111. No. The proposed action does not affect the existing Technical Specifications nor does it require the need for additional Technical Specifications. The margin of safety is not reduced.

Documents

Thomas T. Martin I - Amazon S3 · due to the seismic event indicated by PP&L to be within the licensing basis, clearly do not meet the design or licensing basis requirements if the