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Omaha Public Power District -....

1623 '4 A R N E Y a OMAHA. NEeMASMA e5102 e TEL EPHONE S36 4000 AWE A CODE 402

October 20, 198*P Q

Mr. Robert A. Clark, Chief O 8TU. S. Nuclear Regulatory Commission

gCI2 6 Jggju -SrOffice of Nuclear Reactor Regulation C

Division of Licensing , M 8 a r m e , % ,,, . ' '*%Operatir.g Reactors Branch No. 3 g 'gWashington, D.C. 20555

g8sReference: Docket No. 50-285

Dear Mr. Clark:

The Commission's letter to Omaha Public Power District datedAugust 21, 1981, regarding pressurized thermal shock, requested certaininfonnation be provided for the Fort Calhoun Station within 60 days andadditional infonnation be provided within 150 days of your letter. TheDistrict's letter dated September 24, 1981 provided our reply to yourrequest for an acknowledgement within 30 days of your letter. Thisletter provides the District's responses to your 60-day request for theFort Calhoun Station, using the item numbers from your letter.

1) Enclosure (1) provides a " map" of the cylindrical portion ofthe Fort Calhoun Station reactor vessel showing the locationsof all inlet / outlet nozzles and all welds. The initial RTNDTvalue for each plate and weld is indicated within a rectangleon the figure. Please note that for the welds, the RTNDTvalues are based upon generic material properties as detailedon sheet (2) of Enclosure (1). The plant specific materialproperties for the Fort Calhoun Station welds will be deter-mined from archiveu m.iterial testing and the weld RTNDT valueswill be recalculated and provided as part of the District's150-day response. Sheet (2) of Enclosure (1) lists thelocations and provides the designations assigned to eachlocation on the figure. Sheet (2) also lists the weld chemistryused in the RTNDT shift pre:tictions, and the values of thepeak vessel fluene - the I.D. for the end of this year

(December 31,1981) 6M w four more calendar years ofcontinuedoperation(December 31,1985).

Sheet (3) of Enclosure (1) provides a " map" of adjusted RTNDTvalues for critical locations at the inner surface of the FortCalheen Station vessel predicted for the eno of this yaar(Cecember 31,1981). The method used to predict these valuesis summarized on Sheet (4).

(OY8110270181 811020DR ADOCK 05000285 &Pl'R

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Mr. Robert A. ClarkOctober 20, 1981 .

Page Two -,

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; 2) In response to your request for the rate of change of RTNQTfrom the current condition, Enclosure (2) provides a " map;

values at the inner surface of thei of the predicted RTNDT~ Fort Calhoun Statfori vessel after four more calendar years

of continued operation. t,

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3) The District does not consider it appropriate to define anvalue for continued operation for theupper limit RTN

reasonsdiscussNinitem4)below. |,

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4) The capability of a reactor vessel to withstand the effects| of pressurized thermal shock cannot be represented independent

of the pressure-temperature transient which might cause the,

is basedcondition of concern 'The concepp)of using RT'

NDTon the work of Pellini and Puzakt , who studied test ex-perience with essentially isothermal materials. This is

4

a valuable concept av useful design tool for materialswhich experience slowly varying temperature; however, thecurrent issue of reactor vessel thermal shock involves rapid'

'

cooldown ' transients which canst be adequately addressedby the simple fracture analysis technique used in Reference(1).

,

Research on the effects of thermal shock on reactor vessel ,

integrity indicate that the " phenomenon of warm pre-stress'

should _be considered in predictions of reactor vessel in-!

tegrity and that this phenomenon may form a key elementupon which to base assurance of vessel integrity..."(2);

: Adoption of a simple RTNDJ limit would n3t permit consider-! ation of the " key element of warm pre-s tress and is there-

fore an inappropriate criterion for continued operation..T

t

The pressurized thermal shock evaluation program presently| underway for the Fort Calhoun Station reactor vessel con-i siders in detail the effects of warm pre-stress, as well! as the values of the RTNDT at all critical locations aroundthrough the vesselthe vessel and the variation of RTNDTwall. The end result of this program will be the deter-

mination of the number of effective full power years ofplant operation during which vessel pressure boundary

; integrity can be assured for the events which result inthe most challenging pressure-temperature transients.

1

5) Item 5 of tne Commission's letter of August 21, 1981 rcquested information 01 cperator actions which are requiredto prevent pressurized thermal shock and to ensure vesselintegri ty. The District has. reviewed its emergency pro-

| cedures and has confirmed that appropriate infonnationcurrently exists to minimize repressurization followingthermal shock. The District has' alsc identified thoseemergency procedures which address the events during which

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. .34

Mr. Robert A. ClarkOctober 20, 1981Page Three

pressurized thennal shock is possible. Enclosure (3) providesthe relevant steps excerpted from procedures for the FortCall;un Station along w*th a brief explanation of each step.These explanations reveal how each procedure is structured tominimize repressurization following thermal shock to thereactor vessel.

Specifically, detailed information with regard to HPS1 throt-'

tling/tennination criteria and feedwater throttling criteriai' are provided in Enclosure (3). HPSI throttling and termi-

nation criteria are containea in both the Loss of Coolant andL

the Steam Line Break emergency procedures. Feedwater throt-tling criteria are necessary only for the Steam Line Breakprocedure. Both the HPSI and the feedwater operating criteriaare included to minimize the total RCS cooldown and subsequent

: repressurization while still ensuring that the core remains,

cooled.'

While the existing procedures are adequate, the District willcontinue to evaluate this guidance as a near-term effort.Where it is warranted, future changes will be made to the Fort

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Calhoun Station emergency procedures. The results of thenear-term effort will be provided in the District's 150-dayresponse.

A two-phase program comprising the near-term effort has also| been initiated. Phase 1 consists of a generic effort being; performed by Combustion Engineering, Inc. (C-E) for the C-E ,'

Phase 2 is a District effort to review theOwners Group.plant'.; emergency procedures where such changes are desirable.

During Phase 1, C-E will review existing procedures in lightof the C-E Owners Group generic guidelines currently underdevelopment and information generated in related analytical|

|

work. Where it is warranted, revisions to existine procedureswill be proposed and evaluated. Evaluation will includeengineering analyses, review of proposed changes by the' FortCalhoun Station staff, and validation of the altered procedureon a simulator. During Phase 2, detailed procedural changes

Iwill be made to the Fort Calhoun Station emergency proceduresi

This willfor any procedure changes initiated in Phase 1.also include retraining of the operating staff as necessary.

Since elys,|

/ fYh '"

W. C. J nesDivisio ManagerProduc on Operations

.WCJ/KJM/TLP:jmm

Enclosures

cc: LeBoeuf, Lamb, Leiby & MacRae'

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REFERENC.ES

i

W. S. Pellini and P. P. Puzak, " Fracture Analysis Program Procedures.

for the Fracture Safe Engineering Design of Stee' Ctructure", NRCReport 5920, March, 1963.

2. F. J. Laus, R. A. Braz, Jr., and J. R. Hawthorne, " Significanceof Warm Prestress to Crack Initiation During Thermal Shock," NRC/NUREG Report 8165, September 1977.

F

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UNITED STATES OF' AMERICANUCLEAR REGULATORY COMMISSION

4

In the Matter of )))

Omaha Public Power District ) Docket No. 50-285*

(Fort Calhoun Station, )Unit No.1) - )

i

AFF2 DAVIT:

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. . . . . be!ng duly sworn, hereby deposes and says that he is DivisionManager - Production Operations of Omaha Public Power District; that he-

is duly authorized to sign and file with the Nuclear Regulatory Com-mission the attached response to the Commission's letter dated August 21'

1981 regarding thermal shccking of the reactor pressure vessel; that heis familiar with the content thereof; and that the matters set forththerein are true and correct to the best of his knowledge, information-

;

anci belief.

W. C. Jo sDivision HanagerProdt!ction Operations

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STATE OF NEBRASKA). . . ) ss .

COUNTY OF DOUGLAS)

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|Subsciibed and sworn to before me, a Notary Public in and for the Stateof Nebraska on this yo f4 day of October, 1981.

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^' " W"nW? g*|{y'j'D,?|z& Nota y Public # /-M. My comm En vay a.1984

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. Enclosure 1Sheet 2*-

Designation on FigureFORT CALHOUN

Longitudinal Seam Welds -

Upper Shell Course 1-410 A, B and C

Intermediate Shell Course 2-410 A, B and C

Lower Shell Course 3-410 A, B and C

Girth Seam Welds -Upper to Intermediate Shell 8-410

Intermediate to Lower Shell 9-410

Shell Course Plates -Upper D-4801-1, 2 and 3

Intermediate D-4802-1, 2 and 3

Lower D-4812-1, 2 and 3

Assumptions

Residual Chemistry - as measured or:

Copper 0.35% Upper Bound

Phosphorus 0.012% Upper Bound,

Nickel (as measured)

Initial RTNDT - as measured or:-20 F (21 upper bound for submergad arc welds)Branch technical position MTEB 5-2 for plates ,

Peak Vessel 10 Fluence18 2

5.36 EFPY 7.04 x 10 n/cm (12/31/81)

8.56 EFPY 1.12 x 1019n/cm2 (12/31/85)

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Enclosure 1Sheet 4 . .

TRANSITION TEMPERATURE SHIFT METHODOLOGY

*

FOR ALL PLATE MATERIALS -

REGULATORY GUIDE 1.99 (REV. 1): ,,5-

ART NDT = [(40 + 1000 (Cu.08) + 5000 (P .008) )[ (h)

FOR ALL WELD MATERIALS -

REGULATORY GUIDE 1.99 (REV. 1) IF-

NICKEL CONTENT-IS MORE THAN 0.30%-

MODIFIED REGULATORY GUIDE l'.99 IF-

NICKEL CONTENT LESS THAN 0.30%:

ART NDT = (90 + 600 (Cu .24))[(y h ),,,

.

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Enclosure (2) .

~~F5itfE3[HbUN~'~'. PRESSURE VESSEL VIEW

- -

-ADJtGTED NOT F 'AT 8.SG EFpy 0F PLANT NUMBER 02

12 -31- 8 5-

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C. 3 4tO A_C**y 180 - g _ _ q _ __ _ _ _ _.___________________a ___.________________

D 4812 2_ _ _ _ _ . _ _ _ _ _ _

L1f11 .

.

220 --

.

15 0 3609'O 180 ,

260 --

.

RZIMUTHAL LOCRTION!

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Enclosure (3)''

FORT CALHOUN PLANT PROCEDURE

,

Loss of Coolant AccidentEP-5

/ Excerpt from Loss of Coolant Accident Procedure:

Start auxiliary feedwater pump (s) and feed steam generators via emer-

gency feedwater nozzles as requirea to restore and maintain level.

Bases for Action:

ains step ensures that RCS heat removal is controlled by maintaining

a normal steam generator level, so as to minimize an RCS cooldown dur-

ing the event.

f

/ Excerpt from Loss of Coolant Accident Procedure:.

Dqring the INJECTION PHASE, ensure balanced HPSI flow to each loop (HPSI

maximum design f' w rate is 400 gpm/ pump). When HPSI loop valves are-i

throttled, the primary system pressure may be reduced. Observe system,

pressure during valve repositioning in relation to maintaining system|

| pressure above saturation pressure. Continued operation of the HPSIf

| system should not present an overpressurization problem as long as the

| RCS temperature is greater than 250 F.

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Bases for Action:

. .

operation of the HPSI syst m ensures that fluid is being de-Continued

livered to the primary system, thus maintaining heat removal via fluid

flowing out the break. This step alerts the operator's attention to

the possibility of overpressurization.,

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FORT CALHOUW PLANT PROCEDURE

Uncontrolled Heat ExtractionEP-6

/ Excerpt from Uncontrolled Heat Extraction Procedure:

Secure the main feedwater pumps.

Bases for Action:

Securing the main feedwater pumps cuts off feedwater flow to the steam

generators. This action is required because excessive feedwater to

the steam generators could result in an uncontrolled cooldown rate.

This procedural action minimizes a severe RCS heat extraction transient.,

/ Excerpt from uncontrolled Heat Extraction Procedure:; ,

Secure all feedwater flow to the affected steam generator.

|i

Bases for Action:

I A break in the steam line results in excess steam flow on the secondary1

fside. This results in an increase in the affected steam generator's heat

f removal rate, and an RCS temperature reduction. Isolating feedwater toi

the affected steam generator minimizes a further uncontrolled plant

|cooldown. The unaffected steam generator can then be used to maintain

an orderly RCS cooldown.

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' / Excerpt from Uncontrolled Heat Extraction Procedure:

Verify that auxiliary feedwater flow is being delivered to the unaffected

steam generator and maintain normal level (greater than 80% level if the

auxiliary feedwater nozzles are being used).

Bases for Action:

If a normal level is maintained in the unaffected steam generator by

verifying an adequate auxiliary feedwater supply, the unaffected steam

generator will continue to be available for controlled Reactor Coolant

System heat removal purposes. This minimizes RCS cooldown.

/ Excerpt from Uncontrolled Heat Extraction Procedure:,

Establish normal water level in the pressurizer. Secure charging pumps

as needed to prevent a water solid RCS after verifying that the HPSI

pumps are operating (if they are needed) and capable of delivering flow0to maintain RCS pressure at a point to assure 50 F subcooling in the RCS.

1

Bases for Action:!

!

! A reduction in RCS temperature, due to the steam line break, will cause|

an RCS in/entory loss (shrinkage). This inventory loss will be made up

by the safety injection system (HPSI pumps) and/or the charging pumps.

The charging pump 3 are secured to prevent the RCS from going solid

while the lost inventory is being replaced. This, along with tha re-

quirement that 50 F of subcooling is maintained, prevents an RCS over-0;

|pressurization condition.

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FORT CALHOUN PLANT PROCEDURE

:

Reset of Engineered SafeguardsEP-35

- / Excerpt from Reset of Engineered Safeguards Procedure:

HPSI cannot be shut down unless the following conditions are met:

The HPSI system has ' een in operation for at least 20 minutes and untila

all RCS hot and cold leg temperatures are at least 50 F below the satur-

ation temperature for the existing RCS pressure. The available instru-

mentation includes incore thermocouples, RCS hot and .old leg tempera-

0tures and pressurizer pressure. If 50 F of subcooling cannot be main-

tained after HPSI has been shutdown, the HPSI shall be re-initiated.

Pressurizer level indication alone should not be relied on in deter-

mining RCS heat removal capability. Charging pump / heater control may

be manual..

Bases for Action:

Once the above conditions are established, the operator is prompted to

stop the HPSI system to prevent the RCS from going solid and overpres-

surizing. Termination of HPSI flow, according to the criteria of above

action step, cannot take place until after the HPSI systems heat removal

i

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function has been achieved, and RCS pressure control has been achieved.

Pressurizer level may not necessarily indicate that pressure control has

been achieved, since the presence of RCS voiding would not be detectable

by pressurizer level alone. An indicated pressurizer level coexisting

with RCS subcooling_will ensure that RCS heat removal capabilities con-

tinue to exist.

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