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TOPICS REQUIRING FURTHER REVIEWRELATED TO THE SEISMIC UPGRADE OF THE YANKEE PLANT

H. J. RUSSELL

JULY 1988

Prepared for theUnited Sta'.es Nuclear Regulatory Comission

Washington, DC 20555Under DOE Contract No. DE-AC07-761001570

FIN No. A6808

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SUMMARY

The Yankee Atomic Electric Company, licensee for the Yankee NuclearPower Station, has recently indicated an interest in performing theseismic upgrade of the Yankee Plant in a fashion which will minimize theamount of review work required of the NRC staff. This document identifiesall the criteria, methodology, and issue resolution strategies whichrequire staff reviews. It also identifies the available alternativeswhich do not require such reviews.

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CONTENTS !

SUMMARY ............................................................ 11

1. INTRODUCTION .................................................... 1

2. TOPICS REQUIRING NRC STAFF REVIEW, AND ALTERNATIVES ............. 1

2.1 MS/ FW An al y s i s (4. 3.13, 2. 2. 5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

2.2 Response Ratio Method (4.4.1.1, 2.2.2) ..................... 2

2.3 Bracketing of Piping Models (4.4.1.1, 2.2.2) ............... 2 ''2.4 Nozzle Anchor Flexibility using WRC-297 (4.4.1.1, 2.2.2) ... 3

2.5 Strain Criteria Application Exceeding 1%(4.4.1.1,2.2.2)... 3

2.6 SRSS Combination of SI and SAMfor Pipe Supports (4.4.2. 2.3.1) ........................... 4

; 2.7 Gang Hanger Out of-Scope Piping Loads (4.4.2, 3.2.1) ....... 4

2.8 U-Bolts Used as Axial Restraints (4.4.2, 3.2.2) . . . . . . . . . . . . 5

2.9 Equipment Nozzle Analysis Using WRC-297 (4.5.1, 3.2.2) . . . . . 5

2.10 Review of Time Histories Used inReactor Sliding Issue Resolution ........................... 5

2.11 SRSS Combination of VC and RSS SAMs ........................ 6 ;

i3. CONCLUSION ...................................................... 7

4. REFERENCES ...................................................... 7'

TABLES

1. LIST OF ACRONYMS ................................................ 8

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TOPICS RE0VIRING FURTHER REVIEW l

RELATED TO THE SEISMIC UPGRADE OF THE YANKEE PLAST

1. INTRODUCTION j

In July of 1987, the Nuclear Regulatory Commission (NRC) staff issueda Safety Evaluation Report (SER) concerning the seismic upgrade of theYankee Nuclear power Station (Yankee). This SER is identified inReference 1. In March of 1988, the NRC staff issued a document clarifyingtheir position on several issues discussed in the July 1987 SER (Reference2). These documents benchmark the progress made in completing the latest ;

review cycle pertaining to the seismic upgrade of Yankee equipment and|piping. Criteria and methodology acceptable to the staff for implementingi

the upgrade are identified. Of interest here are the criteria and I

methodology, the application of which require case-by-case review by thestaff before acceptance. Also of interest are resolution strategies forunresolved issues identified in the July 1987 SrR which require staffreview. All topics in these documents which involve staff review arediscussed in the following section.

ITwo meetings i' ave occurred between the licensee and the staff since j

the July 19?? SER was issued concerning the upgrade. The first meeting |(held in Bethesda, MD December 21,1987) concerned the integrity of tne i

reactor support under seismic loading. The second (held in Walnut Creek, !CA, May 2-4, 1988) dealt primarily with the method used by the licensee in '

calculating support loads for nonseismic piping attached to gang hangers 1

in the seismic scope. Any potential staff review work related to thesemeetings is also discussed in the following section. 1,

2. TOPICS REQUIRING NRC STAFF REVIEW, AND ALTERNATIVES

In each of the following subsections, a topic which involves thepotential for future NRC staff review is identified. Specific areasrequiring review are identified, as well as alternatives which do notrequire review. Two section numbers are provided in parentheses in the

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title of each subsection below. The first is the section number in theJuly 1987 SER where the topic is discussed. The second is the sectionnumber in Attachment 3 to the SER, where a much more detailed discussionof the issue is typically provided. References to other pertinent.

|documents are explicitly made in the text.

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2.1 MS\FW Analysis (4.3.13. 2.2.5)

The analysis of the main steam (MS) and feedwater (FW) piping outsidethe Vapor Container (VC) is a special case. The steel framing supportingthis piping is comparable in mass and stiffness to that of the piping.The licensee has proposed special analytical techniques to analyze theMS/FW piping and supports. Approval was given to the proposal inReference 1, with the understanding that the analyses to be done will besubject to a detailed review to determine final acceptance by the NRCstaff. The only acceptable alternative to the use of special analyticaltechniques requiring case-by case review is for the licensee to modify thesteel frame to the point where decoupling the piping from the frame ispossible. This is not considered a feasible alternative.

2.2 Resoonse Ratio Method (4.4.1.1. 2.2.2)

During the course of evaluating piping inside the Vapor Container, thelicensee regenerated spectra for branch piping connected to the reactorcoolant loop. This regeneration made use of a refinement of the reactorsupport portion of the structural model which substantially reduced andbroadened the spectra generated. Rather than reanalyzing the pipingalready analyzed, the licensee developed a response ratio method for

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calculating stresses for the new spectra based on the original stresses.This method was reviewed as a result of an audit question, and foundacceptable on a case by-case basis. The alternative to the response ratiomethod is a reanalysis using the new spectra.

2.3 Bracketino of Pioina Models (4.4.1.1. 2.2.2)

In most cases, piping models extended from anchor to anchor. Two

exceptions to this oce.urred. In these cases, bracketing criteria were

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used. With bra:keting criteria, the coundary between piping analysismodels was not located at an anchor. Interaction effects between themodels were included by extending each model past the arbitrary boundary asufficient distance to include such effects. The amount of extension wasper the engineering judgement of the analyst. Confirmatory analysi,sestablished the acceptability of these two applications. The -

acceptability of future applications of bracketing criteria will need tobe established on a case-by-case basis because of the engineeringjudgement exercised in their application. There are two alternatives tothe use of bracketing criteria: (1) create a large model with all pipingbetween anchors included, if possible; and (2) install an anchor at aconvenient intermediate location, with piping models terminating at theanchor.

2.4 Nozzle Anchor Flexibility Usino WRC-297 (4.4.1.1. 2.2.21

The licensee proposed defining nozzle terminations of the piping asanchors in the models in most cases. This is industry practice acceptableto the NRC staff. In a limited number of cases, the use of WeldingResearch Council Bulletin No. 297 (WRC-297) was proposed for defining morerealistic boundary conditions. This was judged a reasonable approach, butit lacked sufficient basis for blanket acceptance. Therefore, the use ofWRC 297 in defining nozzle flexibility was accepted with the restrictionthat every application of it be subject to a case-by-case review by thestaff. The alternative is to use anchors at the terminatien points.

2.5 Strain Criteria Acolications Exceedino 1% (4.4.1.1. 2.2.2)

For NP.C spectrum loading, the licensee proposed functional capabilitycriteria based on the SEP Guidelines. As a first check, stresses due tooccasional loads are compared to allowable stresses of 1.8Sh forequivaleni Class 1 piping, and 2.4Sh for equivalent Class 2 and 3piping. If these stresses are exceeded, strain criteria were proposed.With these criteria, a strain is calculated based on the ASME Plastic,Simplified Elastic methodology. This strain is compared to a limitingstrain of 1% for carbon steel and 2% for stainless steel, in addition, a;

wrinkling check is performed for large bore, Sch 40 piping; a maximum3

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stress based on low cycle fatigue limits is imposed; and factoredincreases on flanged joint, nozzle, and support loads are imposed in thearea of strains in excess of 1% (this being applicable only to stainlesssteel piping due to the 1% limit for carbon steel). The clearance checkperformed in the plant will use displacements increased by a factor of

a 3.33 in areas of strain criteria application. Applications of thesecriteria where strains exceed 1% are subject to a case-by case review bythe NRC staff. The alternative to a case by-case review is to reconfigurethe supports to reduce strain levels in the pipe to below 1%.

2.6 SRSS Combination of SI and SAM for Pioe Suocorts (4.4.1.3. 2.3.11

A limited application of square root of the sum of the squares (SRSS)combination of seismic inertia and anchor motion loads was proposed forevaluation of supports under NRC spectrum loading. This was approved,with the understanding that all such applications were subject tocase by case reviews. The alternative is absolute summation.-

2,7 Gana Hanaer Out-of-Scone Pioina loads (4.4.2. 3.2.1)

Use of in-scope piping loads to estimate the loads associated with| out-of-scope piping on gang hangers is not acceptable. This issue wasi discussed at the April 1987 meeting. Four alternatives acceptable to the

NRC staff were identified to resolve it: (1) The licensee could performstatic analyses to calculate the non-safety related piping loads. Such

analyses would include use of peak spectral accelerations with a factor of

| 1.5 as recommended in SRP section 3.7 (Reference 3). (2)Thelicenseej could perform confirmatory analysis of a few representative samples to2 show acceptability of the practice. (3) The licensee could provide a

similarity argument between the safety related and non-safety relatedpiping configurations to justify using the loads for one to estimate loads

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for the other. (4) The licensee could perform standard response spectrumanalysis of the non safety related piping and combine the resulting loadsusing absolute summation. Options (2) and (3) would reouire review by theNRC staff for acceptability. Options (1) and (4) are alternatives that.

I would not require review..

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All current gang hanger loads ha* e been generated by the licensee, and Ireviewed and approved by the ataff during the May 2 4, 1988 meeting !(Reference 4). The particular methodology used would need review in the . |future if it were to be useo with piping other than that involved in theapplication already approved. If such a need arises in the future, ;options (1) and (4), described above, could be used without further review |

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2.8 U-Bolts used as Axial Restraints (4.4.2. 3.2JL),

Use of bullets as axial restraints without a testing basis for theallowable loads used is not useptable. Several alternatives acceptable I

to the NRC staff were developed to resolve this issue: (1) The licenseecould use available test results to validate the allowable loads used; (2) |The licensee could perform the necessary testing t.a validate the allowableloads used; and (3) The licensee could replace the subject supports with lconventional supports. If one of the alternatives that involve validatingthe allowable loads is chosen, review by the NRC staff will be required. lThe remaining alternative would not require review.

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2.9 Eouiement Nozzle Analysis Usino WRC-297 (4.5.1. 2.1.2) i

The methods contained in Welding Research Council (WRC) Bulletin 107were proposed for calculating stresses in the vicinity of nozzles. Insome cases, the methods of WRC bulletin 297 were proposed as an

|alternative. In either case, the geometrical parameters of the nozzleunder evaluation would be checked to ensure they fall in the range ofapplicability of the Bulletin being applied. These methods areacceptable, with the understanding that all applications of WRC Bulletin297 are subject to a case-by-case review by the NRC staff. If WRC

Bulletin 107 is used, no staff review is required.

2.10 Review of Time Histories Used in Reactor Slidina Issue Resolution

During the December 21, 1987 meeting, the licensee presented theresults of an analysis intended to demonstrate that the reactor vessel

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will not siide during an NRC spectrum earthquake. The analysis involvedtime histories generated at the attachment points of the reactor vesselsupports to the Reactor Support Structure (RSS) by a nonlinear analysis ofthe RSS. The analysis which generated these time histories was notreviewed previously by the staff, and a need for such a review wasidentified during the meeting. A review by the Structural EngineeringBranch of the NRC has been planned, but has not yet been completed.Although the licensee has withdrawn in original proposal to use floorspectra generated uring the new time histories in piping analysis, thesetime histories still support the licensee's conclusion that the reactorvessel will not slide under an NRC spectrum earthquake. A review i still

needed. As an alternative, the licensee can reanalyze using timehistories previously reviewed and approved by the staff. No furtherreview would be required, provided that the reactor vessel modal used inthe analysis has been previously reviewed and approved by the staff. Thismay not be possible, be:tuse the lack of sliding predicted by the analysisresulted primarily from reductions in excess margin obtained in the newtime histories.

2.11 SRSS Combination of VC and RSS SAMs

Audits performed during the May 2 4 meeting (Reference 4) establishedthat seismic anchor motions of the Vapor Container (VC) relative to theReactor Support Structure (RSS) used in analyzing piping in the VC weregenerated using the SRSS combination methodology. Justification for thisis required, which implies further staff review. As an alternative,absolute sua combination can be used in place of SRSS combination withoutfurther review.

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3. CONCLUSION

Although many of the approaches requiring staff review have viablealternatives which do not require such review, some may not. The

following are likely to require further review:

1. MS/FW analysis - Modifications to simplify the analytical requirementsare probably prohibitive.

2. Use of unreviewed time history in reactor sliding issue resolution -Under the assumption that the licensee would not have generated thenew time histories if the reactor vessel could be shown not to slideusing existing, previously reviewed time histories, a review will beneeded of the analysis which generated the new time histories.

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4. REFERENCES

1. Letter from M. B. Fairtile (NRC) to G. Papanic (YAEC) datedJuly 17, 1987, Subject: NUBEG 0825, Section 4.11 Seismic DesignConsiderations (TAC No. 51807).

2. Letter from M. B. Fairtile (NRC) to G. Papanic (YAEC) datedMarch 21, 1988.

3. NUREG 0800, "Standard Review Plan for the Review of Safety Analysis :Reports for Nuclear Power Plants, LWR Edition," July,1981.

4. Letter from C. F. Obenchain (INEL) to P. Y. Chen (NRC) datedJune 8, 1988, Subject: Trip Report for the May 2-4, 1988 MeetingConcerning the Seismic Upgrade of the Yankee Plant (A6808) -,

Oben 71-88. (Included in the document package which contains thisj docE* ant)

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Table 1. List of Acronyms

ACRONYM DESCRIPTION

FW FEED WATER PIPING SYSTEM,

INEL IDAHO NATIONAL ENGINEERING L/BORATORYtis MAIN STEAM PIPING SYSTEM

NRC NUCLEAR REGULATORY COMMISSIONRSS REACTOR SUPPORT STRUCTURESAM SEISMIC ANCHOR MOTION,

SER SAFETY EVALUATION REPORTSI SEISMIC INERTIASRSS SQUARE ROOT OF THE SUM 0F THE SQUARES

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VC VAPOR CONTAINERWRC WELDING RESEARCH COUNCILYAEC YANKEE ATOMIC ELECTRIC COMPANY

YANKEE YANKEE NUCLEAR GENERATING STATION

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cc: R. G. Rahl 8MA.sa .

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p Wh40AM File A6808E G s G Idahog

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|NTEROFFICE CORRESPONDENCE.

Date: December 28, 1987

To: R. C. Guenzler

Frem: M.J. Russell */7|

subject MEETING SUMMARY: YANKEE R0WE REACTOR SU/ PORT INTEGRITY - MJR-13-87 |

Refs: (a) M. 8. Fairtile (NRC) ltr to G. Papanic (Yankee Atomic |Electric Company), NUREG-0825, Section 4.11 Seismic Design I

Consideraticns (TAC No. 51807), July 16, 1987(b) Meeting summary fru. T. Cheng to YAEC, Meeting Sumary on .

Systematic Evaluation Program (SEP) Seismic Review (TopicIII-6), July 25, 1986

A meeting was held in Bethesda, MD on December 21, 1987 to discuss theresults of a licensee evaluation of the potential for the Yankee Rowereactor vessel to slide under NRC spectrum loadings. This was identifiedas an unresolved issue in the last $afety Evaluation Report covering theseismic upgrade of the Yankee plant a). Other issues related to thetupgrade were also discussed: (1) the performance of case-by-case reviews '

soecified in Reference (a); and (2) use of spectra generated using therasults of the reactor vessel sliding evaluation in the evaluation ofpiping and supports mounted on the reactor support structure (RSS). The |meeting was attended by representatives of the NRC staff (EMEB), |representatives of the Yankee Atomic Elsetric Company (the licensee forYankee), representatives of the licensee's consultant (Cygna Corp.), andmyself. A list of attendees is presented in ths first attachment. A copy |

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of the licensee's meeting handouts is included in the second attachment. '

In evaluating the central issue, the licensee regenerated free field groundmotion time histories for the NRC ground spectrum, performed a nonlinearanalysis of the RSS to obtain time histories for the reactor support area,and used the resulting time histories in a linear analysis of the reactorvessel, neutron shield support tank and associated supports. The analysisresults were used to calculate a coefficient of friction needed to keep thereactor frvm slidirg. This coefficient was shown to be less than thosedefined for the steel-concrete interface under question in both the PCI andACI standatds.

Since the first analysis of the evaluation, that of the RSS, is beyond myarea of experi!se, no judgement as to its general validity will be madehere. This includes the proposal to generate response spectra for theanalysis of piping supported from the RSS using the results of the RSSanalysis. Based on my attendance at many of the meetings concerningbuilding analysis conducted during the last review round. I questioned the

"Providing research and development services to the govemment" -

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R. C. GuenzlerDecember 28, 1987

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MJR-13-87 iPage 3 of 3 |.

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|and methodology. The portion of the SER discussing strain criteria stated i

that strain criteria would be used only if allowable stresses wereexceeded. The portion discussing square-root-of-the-sum-of-the-squarescombination of seismic inertia and anchor motion loads for pipe support i

analysis stated that such application would be limited. The discussion ia i

the SER of the third case, the response ratio method, indicates ouri

understanding at the conclusion of the review that all applications of this '

method had already been identified and reviewed on a case-by-case basis.i'uture applications of this methodology were not expected. Considering theadditional cost and effort involved in case-by-case reviews, across theboard application of criteria and methodology requiring such reviews is notsv.tsable. ;

At the end of the meeting, the licensee had agreed to submit to the staffthe following: (1) a document defining the differences between the previous

iand current reactor vessel sliding analyses and providing justification for |

the differences, with the document specifically to include the methodologyused in obtaining the 10% damping response spectra from the 5% spectrapreviously reviewed and approved, the correlation coefficient between the

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horizontal and vertical time histories used in the analyses, and a plot ofdamping versus frequency with the RSS natural frequencies identified on it; 1

j and (2) a document clarifying the licensee's application of methodology {requiring hase-by-case-review.

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Attachments:As Stated (2)

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Attachment 1December 28, 1987MJR-13-87 :

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LIST OF ATTENDEES

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NAME AFFILIATION

P. Chen NRC/NRR/ EMESJ. Haseltine Yankee Atomic Electric Co. (YAEC)/ Project ManagerB. Holmgren YAEC/ Lead Mechanical EngineerP. Kuo* NRC/NRR/EMEBV. Nuses NRC/NRR/ Project ManagerG. Philley YAEC/ ConsultantM. Russell INELT. Wang Cygna/ Technical ConsultantR. Wessman* NRC/NRR/EMEB

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N. Williems Cygna/ Project Manager

* Attended only a portion of the meeting.

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Attachment 2,

December 28, 1987MJR-13-87Page 1

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MEETING HANDOUTS l-

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SUMMARYEVALUATION |

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REACTORPRESSUREVESSEL/NEUTR0!!SHIELDTAliK-

SEISMICRES?0liSIAllDARSGENERATI0li

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YANKEEATOMICELECTRICCOMPA!iY

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ISSUISMARY:.

Perfor. a stability analysis of ths Reacter Pressurs Visssi

(RPV)/ Neutron Shield Tank (NST) to deconstrate that the R?V/NSTWillnotupliftorslideunderthecombineddeadandNRCseistic,

spectra 10 ads.

CEnsratsthea:plifiedresponssspectra(ARS)atsignificantReactorSupportStructure(RSS)locationsandattheMainCoolantLoop (MCW nozziss.

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1.3 Potential For Uplift / Sliding j

Becauseofthetangentialrestraint,theRPVcannot-

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TheslidingoftheRSRcanonlybepreventedbythe

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21

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withGapElements

u"

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.

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Figure 4.3VerticalForceofSpringElement8 RepresentingUpperGap

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27

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RPV STABILITT ANALYSIS In00EL 209 L!wgami

Figure 4.5VerticalReactionatNode14RepresentingLowerGa;

29

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Figure 4.7RequiredFrictionCoefficienttoPreventRPVSliding

31

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PROGRAM INSPECCTGN4 (N(mCT S(RylCES

. -

Figure 5.1 llorizontal ARS at RSS Elev. 1079'-8" (Node 32).

33

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PROGRAM INSPEC l

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CONCLUSION:-

.

TheR?V/NSTwillneitherupliftnorslideunderthecombineddeadandh10seismicspectraloads. Therefore,thelinearRPVanalysisand the ARS generated using the time-histories generated by thisanalysisarevalidandtheARSareacceptableforuseinthepipinganalysis.

.

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Cross Connect Main Coolant

icop Safety Valve Discharge |

I4op 4 Safety Valve Discharge, X X

Charging Relief, Pressurizer,Spray, Aux. Pres. SprayPressuriser Sample and Vent X X X Note 1 |Hot Leg, Cold Leg, and X X

Pressurizer Drains, Drain

Hesder

Bleed Flov X

EX Vents and EX Cross Connect | X X |MC Pump Vents, Loop Bypass '-

Vents, Loop Safety Valves )SG Drains - Loop 1 X Note 1

Loop 2

Icop 3

140P 4 X Note 1SG Vents | | X | | )SG WRL Reference and | X X Note 1Variable Legs X X Note 1SG Differential X X Note 1Pressure

Pressurizer WRL and NRL | | X | |Pressurizer Pressure | | X | |SG NRL | X | X | |Reactor and Pressurizer | 1 X | X | Note 1Head Vent j

i

*.

6067R/26.190

.

A-Y9 -

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