ML20050E007

Requests Meeting W/S&W on or About 820614 to Discuss & Resolve Any & All Open Items in Preparing Final Ser.Draft SER & Questions Encl

ML20050E007
Person / Time
Site:	River Bend
Issue date:	03/15/1982
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	William Cahill GULF STATES UTILITIES CO.
References
NUDOCS 8204120461
Download: ML20050E007 (44)

Text

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DISTRIBUTI0ft:

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01&E Mr. William Cahill, Jr.

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Senior Vice President en Inspector

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D River Bend Nuclear Group 2

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1 SHanauer 6-ed Culf States Utilities Company RMa ttson 4

P. O. Box 2951 3

HThompson Be umont, Texas 77704 a

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Attn: Mr. J. E. Booker R

k De r Mr. Cahill:

a gBra er Subject; Early Trans mittal of Mechanigpr$81neering Draft SER Evaluation /

9 Questions for the River Bend Station (Units 1 and 2)

The Draft SER for River Bend is scheduled for issuance in early April,1982.

However, the NRC mechanical engineering review staff de ires to convene a s

meeting with GSU, S&W and the NSSS in mid-Done,1982 to discuss and resolve any and all open items / issues in preparing for the final SER, which is scheduled for release on October 4,1982.

It has therefore been de ided c

to provide you with an advance copy of their draft SER evaluation and related questions, which are eaclosed, to ensure that you may have sufficient time to prepare for the meeting in June.

It is accordingly requested that arrangements be made to schedule the meeting, preferably at Stone & Webster, on/or about June 14, 1982. A three to five day meeting is envisioned and the meeting agenda structured as such. We expect that the meeting participants will be prepared to resolve the open items / issues, and to commit a response date for those areas which cannot be sat sfactorily resolved, so that we may accurately address then in the final SER(. The meeting agenda should be submitted not later than llay 15, 1982 to allow us to issue the fonnal meeting notice sufficiently in advance of the meeting.

If there are any questions pertaining to this request or on the enclosed evaluation / questions contact either R. Perch or John Stefano of my staff.

Your immediate attention to this request will be most appreciated and is urged.

Sincerely, I

8204120461 820315 PDR ADOCK 05000458

/7 Licensing Branch No. 2 E

PDR K(

Division of Licensing

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Nnc ronu m now uncu cuo OFFICIAL RECORD COPY usa m mi-m.w

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Mr. William J. Cahill, Jr.-

Senior Vice President River Bend Nuclear Group Gulf States Utili, ties Company

- Post Offic'e Box 2951 Beaumont, Texas 77704 ATTN:

Mr. J.E. Booker cc: Troy B. Conner, Jr., Esquire Conner and Wetterhahn 1747 P.ennsylvania Avenue, NW Washington, D.C.

20006 Mr. William J. Reed, Jr.

Director - Nuclear Licensing Gul f States Utilities Company Post Office' Box 2951 Beaumont, TX 77704 i

Stanley Plettm'an, Esquire Orgain, Bell and Tucker Beaumont. Savings Building Beaumont, TX 77701

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-1 William J. Guste, Jr., Esquire

-Attorney. General State of Louisiana P.O. Box 44005 State Capitol

_70804 Baton Rouge, LA

. Richard M. Troy, Jr., Esquire Assistant Attorney General in Charge

_ State of..Loui.siana Department of, Justice 234 Loyola Avenue New Orleans. LA 70112 A. Bill Beech

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Resident Inspector P.O. Box 105).

-St..Francisville, LA 70775 e

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EtiCLOSURE 1 F

3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS 3.2.1 Seismic Classification The staff has reviewed the material submitted in the FSAR for Section-3.2.1 and finds i: te be generally a:cep:acle.

A few itens recuiring clarifica:icn anc fur-her justification have been icentifiec dealing wi n the class.ification of certain systems associated with :ne Diesel Generators.

These open items have Oeer. transmitted to the

.apolicant. Upon resciution cf these open items, cur findings wiLL ce as follows:

- Structures, systems and components ~ (excluding' electrical features) t r. a are imocr: ant to safety anc :nat are recuirec to w ths and the ef'ects c' a saf'e shu:dewn ea-:he ake ard d

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.5re cee- : Lass " ee as se's- : Ca:egc y I j

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' ems anc have ceen ider. ifiec in an accectable manner in Ta'bles 3.2

.1 and on syste.m piping anc instrumentation m

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diagrams in the SAR.

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r. e r struc ures, syste s and ccmconents na may ce recuirec for cceration of the facility (excluding electrical features) need. net ce designed to seismic Catecgry I requirements.

The structures, systems and components not repuirec-:c be cesigned c seismi: Categcry I include these scrtions of Category I systems such as ven ' lines, crair.

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Lines, fill lines and test lines on the downstream side of isolation valves and those cortions of the systems which are

'not required.to perform'a safety function.

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Tne staff concludes that the structures, systems anc ccmpenents "important to safety that are within :ne scope of

he Mechanical Engineering Eranen have been procerly classified as~ seisnic Ca:escry I itens and nee: :he requirements of General Design criteria 2,

" Design Sases for Protection Against Natural Phenomena" and 10 CFR Part 100, Accencix A,

" Seismic' and Geologic Siting Criteria for Nuc, lear Power Plants."

This conclusion is based en the aonlicant having met the requirements of General Design Criterion 2, and 10 CFR Par: 100, Accencix A,

cy having crecerty class 4fied iheir-structures, systems.and cc cenents

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c sa"e y as seis.': :ste;c ; ; ;; ems 'n acccrdance witn :he ;csi:Icns of Regula cry GJ.ide 1.29,

" Seismic Design Classification" and.by our conclusion.that

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the icentifiec SSC are :ne plan: fea:ures recessary :c a s's u r e (1 I : n e integrity c'

he reac:ce ccctan cressure boundary, (2) the cacabil.ity ~tc shutccwn the r e a 'c t e r anc maintain it.in a safe shutdown condi 'icn, and (3) the

-capability to prevent and mitigate the consecuences of accidents which coulc result in. potential offsite ev osures comparable to the guidel'ine exposures of 10 CFR Part 100.

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3.2.2

-System Quality Group Classification T he st_af.f has.rev-iewed the mate' rial, presented in FSAR Section 3.'2.2 and find it generally acceptable.

The applican: has taken exception to certain sections of Regulatory Guice 1.26.

fWhile scme of :nese excecticns are

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acceptable, others reovire further clarificatioa and justification before they can ce accepted.

These exceptions have been accressed as questions transmittec to the applicant.

-Upon. resolution of these open issues, our fincings will be as follows:

Pressure-retaining components of fluid systems import. ant to safe:y such as pressure vessels, heat exenangers,.s:crace tanks, cumes, cicing and valves have been classified Guality Grcae 4,

5, C,

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manner in Table 3.2-3 and en syste: picing-and instrumenta ion diagrams in the SAR.

T h e s e c o m p o r.e n.,t s have.been constructed m ss 1

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to cuality standards ccamensura:e

'th the i:ccrtance of tne safety functicn to ce performec. Tne revie-of'Gwali:y Grcu:

A and B (ASME Section.III, Class"1 and'2) r e a c': c r ecolan pressure boundary components is discussed in Section 5.2.1.1 of the SER.

Other' Quality Group 5 comec$ents of s y s t e r.s

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identified.in Position C.1.a thrcugh1C.1.e cf Regula:ory Guide 1.26 are constructed to ASME Section III, Class 2.

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Components in systems identified in' Position C.2.a through C.2.d of Regulatory Guide 1.26 are constructed to Quality Group.C standards, ASME..Section_III,,,, class 3..

components in, systems identified in Position C.3 of Regulatory Guide 1.26 are constructed to Quality Group D standards such as ASME Section VIII and ANSI B31.1.

.The staff concludes that pressure-retaining comecnents of fluid systems important to safety have been properly classified as Quality Group A,

B, C,

or D items and meet the reovirements of General Design Criterion 1,

" Quality Stancards and Records".

This conclusion is based on the

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applicant having met the requirements of General Design Criterion i by having croperty :Lassified these pressure-retaining components;important to safety Quality Grcuo A,

S, C,

- : in at:crdar.ce with t r. e cositicn Of Reguia::ry G;4ce 1.2e, ";.aLi:y Grcus Classifications anc S:ancaros", anc by our conclusion t, hat the.identi.fied pressure-retaining c' c c c o e e n t s / a' rF't h'd re' 4. d ie s s a Fy * '( 1 i 'Yo' ?ii' ej e n' t -o r ::i'i-fi g a fe '

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within the reactor coolant pr' essure bouncary, (2) to cernit 4

snw:cown of tne reactor anc maintain i; in a safe shu ccan condition, and (3) to contain radioactive, materials.

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3.6.2 Determination of Rupture Locations and Dynamic Effects Associated with-the Postulatec Rupture of

. Piping.

The staff has reviewec the material suomitted in the FSAR fer Section 3.6.2 and fines that it in general covers all topics _re: vicing discuss 4:n.

Mc.ever, s e v e r a 's areas-neec f u r t '. e - clarification anc ;us-fication a r.c i ne:nsistencies are present in a few' areas.

Additional justification and clarification are required for the limits used in the criteria for the no break :cce, the pipe break criteria, the pipe thrust coef.ficients, j et impingement analyses and pipe uhip restraint design.

In acdition, a very large amount of information will nct be available until cocolet' ion of~tne new t'

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Our r e v i e., cann:t ce :

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r tne app.L'icant.

U en resoluti:n of these cpen items, our f i, n d i n g.,w.i l l. b e -a.s..fo.4.l.ows; 2

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_consicered in the plant design, ano therefore are acce able-and meet the recuir,ements of General Design Criterion I.

i This_ conclusion is_basec en the following:

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The proposed pipe rupture locations have been adequately assumed and the cesign of piping restrain *: and measures to deal with the ' subsequent. dynamic effects of pipe whip and jet inpingenent provide acecua:e protection Oc the integrity and functionality of safety-related s ructures, systems, and cc cenerts.

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2.

Tae crevisiens for protection agaims: cynamic effects associated with pipe ruptures cf the reac:cr coolant pressure bouncary inside containment and the resulting discharging fluid provide adecuate assurance ina: cesign basis less-cf-coolant accidents will not be aggravatec by secuential failures of safety-related piping, and emergen.cy core cooling system perfernance will net ce degracec by these dynamic e'fects.

3.

7. e. p r c ;.c s e c p " c ' r g. a r. c restra r.: arrar.ger.ent anc age.;cacie cesign considerations for high-anc moderateenergy fluid s'ystens i n s'1 d e a n c.. c u r s ;-de. f c a n : a '. n r. e n t,. irplu. ding' he c

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assurance tha* the structures, systems anc cen;cnents t

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• are ii c.cse p ic x. ; ;, Ic :~e postula*ed pipe rupture will be prote,cted.

The cesign will be C a nature tc

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m-maintained in a safe shutdown condition i n the event of a postulated rupture of a high or moderate energy pi:ing system inside or outside of' containment.

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3.9-MECHANICAt SYSTEMS ~AND COMPOSITES

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3.9.1 Special Topics for Mechanical Ccmponents The' staff has reviewed the material sucmitted in the

• FSAR for Section 3.9.1 and find it generally acceptable.

Mcwever, more infccmation is recuired on the verification of certain computer cregrams.

There are a(sc cuestions regarcing :ne ccmcLe:eness of the design transients anc the numoer of cycles.

Adcitional clarification is neeced regarding tests performed on a limitec number cf components.

These c :: e n items have been transmittec to the applicant.

Upon resolution of these open items, our fincings will be as follows:

S The staff Concludes tha-the cesign transients anc resst:ing leads and load ccmbinations witn a :repriate s:ecif ed cesign and service limits ':-

e: a-icat :c ::n.e-s is a c c e::: a b l e and meets the relevant recuirements of General, Design Criteria 1, 2, 14,,15, 10 CFR Part 50, Appendix B,

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a. c 10 CFR Part 1-0 0, A::::e n d i x A.

This concLusi:n' is tase:

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.The acclicant has met the relevant recuirements of

. General Design Criteria 14 and 15 by'de=cnstrating that' the design transients anc resulting loads and combinations with appropriate specified design and j

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.i service limits which the applicant has used for designing Ccde Class 1 anc C L components anc supports, and reactor internals prcvide a ccmplete basis for design of the reactor coctant pressure boundary for all concitions and events expec e over :ne service Lifetime

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2.

The applicant has met the relevan; recuiremen:s of General Design Criteria 2 and 10 CFR Pa. : 100, Accencix A by inclucing seismic events in cesign transients.hich serve as design Dasis Ic withstand :ne effects cf natural phenomena.

3.

The acclican: has me the relevan; re uirements of 10 CTE Ca--

50, A cendix E,

an Se erst :ss gn Cr :e s

.cy naving sueni :eo 'rf rma: 1:n : a: :e cnstra:es :ne acclicabil'ity and. validity cf the design m e t h o c's and c'e m'd o : 4'r' c r e g r a n s u's'e d ' f o r t h e' 'c e s i g n a c' a n a ly s i's of

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structures, and non-Coce strue: ares.i:nin the cresen; s:ste-of-:r.e-ar: Lir.1:s a r. c 0;

.a.ing cesign :ce rc, measures which are acceptable to a s s u,r e the cuality of the computer Oregrams.

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s 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment The staff has reviewec-the material for Section 3.9.2 provided in the FSAR and find it generally acceptable.

Hesever,.t h e number of OBE stress cycles,censidered in the NSSS scope is not in ccmpliance with Standard Review. Alan.

Tnis is an open item.

Accitional information-regarding the.

Oreoperational test program is required.

In carticular, more information is recuired on the locations of measurements and visual inspections, the limits for. steady state and transien vibration and the limits a'nd tolerances for thermal expansion.

Comcliance with NUREG-0619, "BWR Feedwater No::le and Control' Rod Drive Return Line Nozzle

ac'i g", 's ecuirec.

These c en items have ceen

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r. e a c c l.i c a c t. anc.ucen Ineir resclu:icn, :ne-staff's finding will be as follows:

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ri, ::m;:ne-::,.anc e:w':.en-and meets the gelevant recuirements of' General Design Criteria 1,

2, 4,

14 and 15.

This conclusion is casec cn the following:

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The' applicant has met the relevant requirements of General Design criteria 14 and 15 wi a respect to the design and testing of':he reactor coolant pressure bcundary ic assure t h a.: there is a lew pr:bacility of racidly crocagating failure and of gross ru :ure and to assure tha: cesig-cen:.:::ns are no: exceeced : rin:

normal cpera: 1cn inclucing anticicated c:erational occurrences by having an acceptable vidration, thermal excansicn, and dynamic effects tes pr: gram which will s

ce conductec during startu; anc initial operation en specified high-and moderate-energy picing, anc all associated systems, restraints and succorts.

The tests provice acequa:e assuranc'e that tre ci:ing ar.d ciping restrain;s of tre system have been cesigaec c witns:and eiera:":naL Oyna- : e'fe: s cae ::. a '., e :L:s re5, tri[ps, and c:ner ::erating medes asse::a:ed wi:n :ne design basis flow conditions.

In acdition, the tests

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provide assurance : ". a t ace;uate c L e a r a r. c e s and ' ee ncvemen: Of snu::ers ex s: :r.nres: ainec :r.e ma; movemen; of piping and su;; orts. curing normal system heatu: and coolcown operations.

The planned tests will develo: loacs similar to these exceri'encec curing reactor operation.

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The applicant has met the relevant requirements of s.t.

General Design' Criteria 2 with respect to demonstratinc

' design adequacy of all Category I systems, components, eouipmenttand their supports to aithstand earthcuakes by

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meeting the regulatory positions c,#

Regulatory Guides

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systems analysis precedure and criteria.

The scope c' review of th'et seismic system analysis included the s

.t seismic analysis methods of all Category I systems,

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components, ecuipment and their suppcrts.

It i ncluded review of procedures for modeling, inclusion of torsional effects, seismic analysis of Catescry I piping s y s t e rr s, seismic analysis of multiply-succorted

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ecuioment and components with c i s/t i n c t inputs, u

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i factors and d e t e r fi ria j i o n of c c m c c s i't e ca.4. ping.

The

's review has included design criteria and procedures for s

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A criteria and,s ei.smi c analysis procedures' for reactor internals.anc Category I'ouried piping cuisice'

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i. 'T h e system analyses are performed by the applicant on an elasticbasis.

Mocal response spectrum multidegree of freedom and time history methods form the bases for the analyses of all major Category I systems, components, ecuicment and their~sucports.

When the medal response

- scec rum methec is used, sever.ing res;cese carameters are ccmbined by the scuare root of the sum of the scuares rule.

However, the absclute sum cf the modal r e s co'n s e s a r e used for modes with closely scace5 frecuencies.

The scuare root cf the sum cf the scuares

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of the maximum codirectional responses is used in accounting for. three components of the earthquake miotion

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for both the tine nistory anc resconse scectrum.nethces.

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FLocr scectra incuts te be used for cesign and test ver 'ications ci

s. stems, cc ec eets, ec.i.c-e t a r. c their sucperts.are gener-atec frcm the ti.me histcry method,'taking into ' account variation of parameters by v...~,.. ~ ~

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A vertical se sric sys:e--dynamic anatysis nas ceen emptcyec for att systems, anc components, ecuipment.'and'th'eir supocrts wnere analyses show.significant structural amplification in the s

vertical' direction.

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The applicant has met the relevant requirements of General Design Criteria 1 and 4 with resoect to the i

reactor internals being designed snd tested to' quality standard commensurate with the importance of the safety functions being performed and being accrocriately Orc:ec:ed agains: dynamic e'fects by eeting tne regulatory positions of Regulatory Guioe 1.20 fer the

onduct of preoperational vibration tests anc by,having a creoperatienct vibration program planned for the reactor internals which crovides an acceptable basis fcr verifying the desig.n adequacy of these internals under test loading conditions comparable to those that will be excerienced during opera-ion.

The combination of tests, credictive analysis, and post-test inspection provide acecuate assura ce :na,

9e reactor internals will, curing : heir service lifetime, withstano.the flow-induced vibrations of reactor opert: ion without loss of

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s:*actural integr':y.

The ihtegrity cf the reac cr int.ernais in s e r '.1 c e is essen:ial to assure ne proper.

p o s.1 t t e n i n g,o f.....' reactor fuel as'semblies ano unimpaired operation of the control rod assemblies to permit safe reactor operation and' shutdown.

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The applicant has met the relevant requirements of General Design criteria 2 and 4 with rescect to the design of systems and components imper: ant to safety te withstand the effects Of earthquakes and the aporceriate combinations of the effects of normal and postulated acciden: conditions.ith the effects cf the sa e d

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prc.4 ces an s: cec: acte casis fer confir' ming tne structural design acequacy of the reac;ce internals and unbroken ciping Lcces to withstand :ne ecmbinec cynamic loacs Of costula ed Lcss of coclant accidents (LOCA) and the SSE.

The analysis provides adecuate assurance that -he ecmcinec stresses anc strains 'n the ecmpenents c'

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. T.aterials of constructicn, anc nat ne resulting

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..vy cf ne rea: ce internals.~ i '. L 0-c' sect-the "es: cr internals geometry to One extent t r. a t core ccoling may s

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b e 'i n c a *. r e d ; 'The methcds use3'for cimbereat a n'a l y s i s '

have been found to be compatitte with those used for the systems analysis.

The croposed ccmcinations cf component and system analyses are, :nerefore, acceptable.

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d assurance of structural integrity of the reactor internals under LOCA concitions for the most adverse postulated loading event provides added confidence tha:

the design will.ithstanc a spectrum of lesser pipe breaks and seismic Leading events.

5.

The acclicant has me the relevant recuirements of General Design Criterion 1 with respect :c systems and components being designed and testec to cuality standarcs c0mmepsurate.ith the imocrtar.ce of the s a f e.: y functions te be performed by the proposed pregram to

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correlate the test measurements with the analysis results.

The program c o r. s t i t u t e s an acce; abLe casis for demonstrating the ccecatibility of the resu':s from es:s are ara'>ses, ne ec siste :,- ce ween na: e.a:ical modeLs used for.cifferent.Loacings, and : n, e vatici y of the interpretation of the test and analysis re'sults.

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-l 3.9.3 ASME Code Class 1,

2 and 3 Components, component Supports, anc Core Support Structures The staff has reviewec the information in the FSAR for Section 3.9.3 anc find it ic ce generally a c c e ; : s t '. e except i-t'e areaI :' fer.c:icnal :aca:

s ii arc d e s i g r.

li-':s.

Compliance with r4UREG-0500 must ce sh0wn.

n aediticq, further clarificatio,n is recuirec on certain aspec*s of sup;cr: cesign.

Tnese c;en issues nave cee- :ransmi :ec to Tne ac Licant and ucen resclution, the staf*'s 'incing shalL ce as fellows:

1 The ap licant net the recui'rements of 10 CFR Part 50, 550.55a ar.d General Design Criteria 1,

2, anc 4 with res:ect ::

5e cesig-and service L:a cencinat i ens a.c ass:c a:e: s ress ar

e'c-a-

-.--- s see: ' e: f:-

ASME Coce Class 1,

2, anc 3 ce=cenents y insuring that

. systems and, components i.mportant to safety are cesignec tc cua'.

y stancarcs.conmensara:e.i:P :nei-i n::r ance a

to safety and tha: these sys* ems :an ac: mecca:e :ne e.' ' e c': s of ncem.at Oceration as.setL as 60s ut'a:ec even s'

such as Loss-of-coctant accidents anc the dynamic effects resulting from earthcuakes. The s ecified cesign a r.d service combinations of loadings as at: Lied to ASME l

i

• /

_.e.

+

a s

Cod'e Class 1,

2, and 3 pressure re:aining components in systems designed to meet seismic Category I standards are such as to provide assurance that in :ne event of an earthquake affecting the site or other service loadings due to ocstulated events er system o; era-ing transients, the resulting ccmbined stresses imcesed en system c c i c e n e r.: s. lL no e4: eec allo-acle s: es: anc strain limits for the materials cf construction.

Limiting the stresses under such loading coccinations crovices a conservative basis for :ne design cf system ccmcenents c w :nstand the mes acverse ccmcinatien cf loacing events witnout loss of structural integrity.

2.

The acclicant h a s -m e t the recuirements of 10 CFR Part 50, 550.55a and Gen'eral Design Criteria 1,

2, anc 4 a :h resce;; :c

-*e criter's usec 'er ces'gn anc i

.1 r, s ; a.. a t i e n. c f-A S M E Coce'C-lass 1,

2, anc 3 cvercressure relief devices ~ by insuring that safety anc relief valves a n d ' t n e i r - i n s t a t l'a t'i o n s' a r e c"e s ig n'e ct O c' s t a n'd a r d s whi'cS

~

' ~' '

s3 :r-e s.r2 e

,e" ie-I's-

---

're5, l'c

• 3-t h, e y, c an atc0mmodate.the e.f f.e c t s c f. c i s en a r g e.. d u e t o.

n r al ccera-icn as -ell as ccstulatec even s sucn as loss-of-coolant accidents and the dyn,amic effec:s resulting from the safe shutccwn eartnquake.

The relevant recuirements of General Oes gn Criteria 11 anc 9

+

m~-~--'

• /

15 are also met with respect to assuring that the reactor coolant pressure boundary design limits for normal operation including anticipated operational occurrences are not exceeded.

The criteria used by the applicant in the design and installation of ASME Class 1,

2, and 3 safety and relief valves crovide adecuate assurance that, under cischarging conditions, the esulti g stresses will not ex:ee: a l '. : a t '. e stress anc strain limits fo'r the materials of construction.

Limiting the stresses under the loading ccmcinations associated with the actuation of these =ressure relief devices provides a conservative basis for the design and installation of the devices to withstand these loads with6ut loss of structural integrity or impairment of the ever:ressure Orctection funct ier.

3.

The'acclica'nthas met' the rehuire'ments'of TO'C.:h~

Part 50, 550.55a and General Design Criteria 1,

2,- and 4 V :.-,.

With,......

... c:,a...the design and service teac.

..c.o m e : n a t i o n s respect to a r. : asse: tate: stress a c.ce':rmat r.

.i.-its s e:'fie:

for ASME Code class 1, 2,'a'nd 3'componen l. supports'by ensuring that component supports important to safety are designed:to qua.lity standards commensurate with their i

incortance to safety, and that these succorts can

(

accommocate the effects of. normal-oceration as ell as b

L

e postulated events such as loss-of-coolant accidents and the cynamic e tec s resulting from the safe shutdown e,arthquake.

.The combination of loacings (inclucing system operating transients) considerec for each component succcr: within a system, including :.h e cesigna:icn Of :ne accrecriate service stress limi for each loading concinat i ce, has me

ne cesitiens and criteria of Regula: cry Guides 1.124 snc 1.130 anc are in accordance with NUREG-04c4 and NUREG-0609.

The spec fied cesign anc service loacing ccmcinations usec for ne cesign of ASME Coce Class 1,

2, anc 3 componen:

supports in systems classified as seismic ' Category I c Ovide assarance that in the even cf an earthcuake or other service teacings cOe :: costulated events er

::em ccera-i g transients, -he res;L:ir.; ccme, ec s reSses incesec en system compCner. s W'll not 'exceec a'l l o w a p l 'e stress and strain limits for the materials of.

• '- ' c c r-s t r'u c t'i'c'n.

1. i m. i t iM g ' t h e s Yesses' uNder"soch (cading

.c' 3:': s c:

01- : a :: :e s:

.s cas

Or :"s design of.suppor.

ccmponents to withstanc the mes:

acserse ccmcina-icn of Loacing events wi:ncut less cf structural integrity.

b Class CS componen evaluation findings are ecvered in SRP Section 3.9.5 in connection with reac:cr internals.

O s*

~

• ?

'3.9.4

' Control Rod Drive Systems The staff has reviewec the information in the FSAR for Section 3.9.4 and' find it generally a c c e p t a b'l e.

If the crctetype information is r.c: available from the Kuo Sheng I in - timely manner, the applicant may have te

..a m e alter.a:e cia.s.

.:s : cree *r r, a s ceen ransni ted te ne acclicant.

bocn receiving acceptacle cretotyce information, the staff's findings will be as fellcws:

Tne sta'f concludes that the design of the cen:rcl rec drive system is acceptable anc meets the recuirements of General Design Criteria 1,

2, 14, 26, 27, and 29, and 10 CFR Part 50,

$50.55a.

This cenclusien is basec en the fellcwing:

1 The a'oplican has me: i n 'e r e c u t'r e m e n t of GDC'1 and 10 CFR Part 50, 650.5'a, with respect c designing 5

x -

..... ~........

. c'Jal*. !y stanCards femconenTs inoCr:an!'!c s'a ety IC s

c rer.s.r5:e.;;- : ". e ' cc*:ince c-

e sa e:f '.n:: i:r :

to De performed.

The-design-procecures anc criteria used for the control roc drive system are in conformance with the requirements of appropriate ANSI and ASME codes.

e m

O W

. _ ~.a o _

j

~ ~ -... __

9 2.

The apel.icant has met-the recuirements of GDC 2, 14, and 26 with respec..to designing the, control r'od. drive system to withst'and effects of earthquakes and anticipated normal-operation occurrences with adequate margins to assure its reactivity control functicn anc with extremely lo. crecability Of leakage or gross ru;;ure cf reac:cr cc Lan: pressure beundary.

The s: ecd fied design t r a r. s i e n t s,

cesign and service loadings, ccmbination of loads, and Limiting the stresses and defermations under such loading coccinations ard in conformance with the recuirements cf acercoriate ANSI and ASME Codes and a c c e p t.a b L e regulatory positiens specified in SRP Secticn 3 '. 9. 3.

3.

The a::Lican: has ne: the receirements of.G00 27 anc 29 with' e5:ec:

0 c e s i g n i r. ; :ne :en:r:.

cr;.e syster' s

to.asscre its. capability 1: 1 con:rclling reacti'vity and coc.li.ng the

... :. 3 s

,.c o.r e. -wit..h reactor v a,pp r o p-r i a t e m.a rc.i n, -in-n.

.. z.%.,-

a y.,.

..,.9. c

~.

0 0 n : v n c t i cr. witn e i, n e r. tre emergency cere,cecting system or the reactor prctection system.

Tne operability assurance' pro' gram' i s. e c c e :: a c t 'e with res:ect :c meeting system design recuirements in observed performance as to wear, functioning times, latching, and overcoming a stuck red.

T I

l mY

o

• i 3.9.5 Reactor Pressure Vessel Internals The staff has reviewec Ib.e naterial in,the~FSAR for

~

Section 3.9.5.

This naterial is generally acceptable except that additional clarification and justifica:icn are recuired

1. or the stress, cef r.aticn anc buc' Ling L'-i: 3 given.

In d s

en issue has teen transmi ec t r. e a;;L :ar:.

U: n res;>u::cn of :ne Open issue, t r. e craff's fincings.;LL Ce as f;llows*

T*e s;aff Concludes ha; ; r. e cesign Of "eac:Or ir. e nals is acce::able and meets the recairements.cf General Design Criteria 1,

2, 4,

anc 10 and 10 CFR Part 50, 550.55a.. This conctusien is basec on the f;LLowing:

. =.......

e.--.

...=.=.....

.a.

10 C'R Par-50, s50.55a with res:ect :c cesigning the reactor i n,t e r n a.L s. t o.. q u.a l.i t.y. s t a n d a r d s. c o m e n..s.u.r a t.e with..

he i.CCr*ance c' the safety furcticri :: Ce Oe"# Cr ec.

The cesign prccecures anc criteria use: for :he,reac;;r i n t 'e r n a L s 'a r e in confor ance Jit'r the ecuir'e er:s c4 Subsection NG of the ASME Coce,Section II.

e

_._-a

~.

=----------:_.a.

.ww==...

=...

= _.,

+ ~. - _. _

4 2.

The applicant has met the requirements of GDC 2, 4,

and

~

10 with respect to designing co=conents imcortan: to

s a f e t y; t o w i t h s t a,n.d '. th e ',e f f e c t s o f earthquake-and.the-s effects of normal operation, maintenance, testing, and postulated loss-of-coolan accidents with sufficient margic to assure that cacabili

y :c cerfcrm its safety functions is maintaired and :ne specified acce::able fuel design limits are not exceedec.

The specified cesign transients, design and service loadings, and combination of loadings as applied to *he design of the reactor internals structures,and components provide reasenable assurance thai in the event of an earthcuake or of a system transient during normal plant 0;eration, the resul ir.; deflectiens and asscciated st esses '

cse: c r.

tr.ese str.c:.res a r. d ccmconer.:s would not exceed allowable stresses anc deformation L i m i t s. f o r, t h e., m a.t e. r i a..l s o. f ~c o n s t. r u c.: i o n.. L i m. i t i n e the m

.c

s :..

s y

- - m ~~ - ~.

y. * ' ' v.

stresses a: Seform.ations ur.cea sucr leacing,ccmbi a: ices or0 vices an acceptable basis for the cesign of t h e s,e c

+

structures'and combonents'te w i t h s't a n'd the mest 'acverse loading e' vent s whi c h have been postulated to occur curing service lifetime without loss of structural integrity or incairment.cf function.

4 8

e e*

L,--

./

3.9.6 Inservice Testing of Pumps and Valves The staff has reviewed the i n f o i,- m a t i o n in the FSAR for Section 3.9.6'and find'it incomplete.

The specifics of the inservice testing program for pumps anc valves are lacking.

The criteria upon which the program will be casec needs clarifica:1cn and anclifica:icn.

These c en items have ceen

• arsnit ed :: the a:elicant.

U:en rescl.:icn Of thes* 0;en issues, the staff's' findings will be as follows:

The staff cencludes that the acclican:'s cum:s anc valves test crogram is acceptacle and meets :ne requirements of 10 CFR Part 50, Appendix A,

General Design Criteria 37, 40, 43, 46, 54 and S50.55a(g).

This conclusion is based on the acplicant having provided a tes: crogram to ensure that safe y-relate: ;cm:s anc vaives w ll :e in a s:ste cf operational reaciness ro'perferm necessary'sa'fe:'y' functions t h r o u g h'c u t the' life of t.he plant.

This program includes

' 4 b a s e l'i n e c r e's e r'd i c e 't'e t t 'i'n gC s n d p e r i c c f e i hre r v i c e~ "t e s $ 'n g. ' '

' ' ' ~

. : ;. a.- ; ;.. es f:r :::-

. r. :. :. a. :es c-

.e 7

c componen:s.in.the. oper.ating. state-and for. visual insoection-for leaks anc ether signs of cis:ress.

Applicant nas also formulated his inservice test program to, include all safety-reta:ed Code Class 1,

2, and 3 cumps and valves and :c include these pumps and valves wnich are no: Coce Class 1,

2,'and 3 but are considered to be safety related.

e

L.

4 RIVER BEND QUESTIONS

'Section 3.2.1 Table'3.2-1 It is the staff's ocsition that certain systems i e.c c c : a n net identified in R e g u '. a t e r y Guide ".26 shoutc be e

Ouality Group C,

or its ecuivalent.

Among tnese classified s

systems are:

diesel fuel oil storage and transfer system, diesel engine cooling water system, ciesel engine Lubrication system,. diesel engine starting system, and diesel engine combustion air intake and exhaus: system.

Justify the absence of a cuality group classification of, certiens of these s y s't e m s listed belew:

Diesel ~ Genera:cr Cocling Water, System

.essi Genera:cr Star:ing Syste Oies'el. Gen 6ra:cr Lucrication System Diesel Generator Combustion Air Intake and Exhaust System a:. q.,.. :. =...... c;

. ve..

... ; ;. s..

~.

3 6 C : ' ; *. 3.2.2 t

P a g e ' 3. 2 -6',

S e c t i'o n 3 ' 2 '. 2. 2. 3 -

Excections to Regulatory Guide 1.26; Sections'C.1.e and C.'2.e needs further clarification and justifica:icn. e

..x.

6 e

g i

9

e

- - = -

^

r*

._ L_

l

• ?

i h orovide a list of

=

all lines to which these exceptions are intenced to apply.

1 i

Secticn 3.6.2.1.5.2.1A.2.a 1

r l

The design stress anc faticue limits for Class I pipinc l

l i- :ne ccata" mer: ;e e:Ps icn areas are c; in :ce:Lian;e

.ith Standarc Review Flan 3.6.2 anc STP MES 3-1.

If :ne maximum stress range cf Ecuatien (10) exceeds 2.4 Sm, cc:h i

Ecua t i on s -(12) and (13) must be less than 2.4 Sm.

In all L

cases the cumulative usage fac:ce us: be less than 0.1.

1 t

k

.vaC; O.

l 9

q).,.ww. :. 3 l

Page 3.6A-16

,,s.,

.s -

Section 3.6.2.1.5.2.i*.2.c Tne maxamum s ress range as ca':ula:ec by the sum cf E c u a t i o'n ( c ), and (10) ~should consider.sust.ained ' Loads,

)

occasional loads anc thermal expansion.

Have occasional Lcads been included as per STP MES 3-1?

6

I.

k < -... *

. ~ _..

Page 3.6A-17 Section 3.6.2.1.5.2.1A.2.g

P r o v i.d e a s s ur a'n c e that-thef.100% volumetric' inservice examination of all pipe welds will be conducted during each inspection i nterval as defined in IWA-2400, ASME Code,Section XI.

<1

.... -. ~

2-N Page 3.6A-24 Section 3.6.2.1.7A.2 P '. e.a s e : Larify :nis caragra:n.

List ary acditienal-

'.. Criteria used.

, *.. +. ~.

. v. ~ * : s.

h

.s

..P a g e g A : 2 3..

-2 e...

.n......

....a......-

J u s t i f y t h e "u s e 'o f an amplification f a c t o.r.c f less t.han 1.1-to account-for recound.

m 4

-~

.v

.=

z.

Pages 3.6A-33 Through 3.6A-36 Section 3.6.2.2.5A A thrust coefficien: of 0.7 was usec wnich is.Less than 1.26 as specified in Section 3.6.2.2.3A anc recuired by Standard Review Plan 3'.6.2.

Please provide justification

1. r -his discrecancy.

Also, the Loac are ceferna: 4 :.a of :ne noneycome panel are found without referer.ce

ne

..c n e y c :n o stiffness.

Please clarify these catcula:icns.

Page 3.6A-38 Section 3.6.2.3.LA.4 Please provide clarification and justification for the use'of shape factors of less than unity.

Page 3.6A-39 Sec: 1:n 3.d.2.3.2.2A PCea'se clarify tne statement regardin'g' the oct:cming out;of s o m e, c o m p r e s s,i,v e; a b s c r b e r.s..

In c a r t i c u l a r.,.. h o w ' a r e J.

es: air ':a:s de:e- ' ed un:er - ese c: ci-': s-

~

c

...Page.3.6A-39 Section 3.6.2.3.2.2A Is :ne " retaining recess" in the bumper pipe used to restrain the moving process pioe in the lateral direction?

If so, provide analysis or data to suppor: this.

q

~

_ _ -. :. ~

. =,.

Page 3.6A-40 Section 3.6.2.3.2.2A Provide explanation of the methods used te design L'

... ' ; ~ n

• omni.irecti'onal restraints and l i m i. -

d t stops.

Page 3.6A-40 Secticn 3.6.2.4A The reference should be te MES 3-1 not MES-1.

Tables 3.6A-1 Through 3.6A-11 Plea'se provice'a schedule for completien of the-stress analyses and updating of tnese tables.

Will c,umulative usage. factor.s be limited to less than 1.0?

Table 3.6A-5 P; ease ce'iee C '. a s s '. '. i ; " - e a e r ; y ci:49; as ertiened-in t h e 'f o o.: n o t e.

~ ~ -

~s,

~

Ta:Les 3. 6 A - 12 T *: r c u g n,3.6A.-20, 3.6A-27 inecugh 3.6A-42 anc 3.' o A - 4 5 Tnrougn 3.6A-5L

.c u.

~~

'ProvideJa schedule f o'r com:letion of these: tables.

Figures 3.6A-12 Through 3.6A-19 and 3.6A-21 Thrcugh 3.6A-33 Provide a_ schedule for. completion lcf pipe stress analyses as.they.effect pipe break location. selections.

t l

S

,A L.

.~_

._ J

~

./

I Page 3.68-1 Section 3.6.2.1.19 Is tne reference to Section 3.6.2.1.1A correct?

... 3

,'"v

e

-C Page 3.68-3 Section 3.6.2.1.2.45' Are any considerations given te relative cice cia eters anen exempting impactec pipes of equal or heavier wall thicknesses from rupture?

If not, provide justification for not costulating the rupture cf a ::i c e when i :ca c t ed by a cipe of' larger ciameter, as required by Standard Review Plan 3.6.2.

Pages 3.68-4 and 3.62-6 Secticn 3.6.2.1.2.55

.. # r C,V.i c e,,j y s,t ; f i c,a t i c r i c,e nct, ;; c s t ; l a t ir g. c. e a ( s

r. e n.

... ~

j.:,.

j t h'e cumulative usage f a'e t'c r i s ' g r e a t e r 'than C'.1'and'the

~ ~ '

, s t r,e s s r a n g e, c a lc u l a t e d.:f ro m. E q ua t i o ne. (10.h. i s. L e s s than 3-5.

Page 3'. 6 5 -9'

.Section 3.6.2.2.1.2B

.Please crovide-.astist.of all instances where crack Eull cre sk s%

propagation times or p et. ; opening._ times so in excess'of

~

one millisecond were used.

e e

+

*

• h_ _ __ _ __ _ __ __ _

t m.

..__.._.;- ~.

Page 3.6S-10 l

Section 3.6.2.2.1.29 Provide.i.ustification for using a thrust coefficient.of, l

.. v;..

.. 1ess than 1.26 for saturated steam and 2.0 for subcooled water.

Page 3.65-7 Sectier. 3.f.2.1.2.62, Ite.- 5 Please provide a list of all instances where mechanistic apprcaches were used to reduce break areas.

9 e

C..-

.i e,a

...:.t

_e-T' C.

s,.

Tables 3.oE-1 T h r o u g h - 3. 0,5 - 4, rig'ure 3.65-4 Provide a sched'ui,e for completion of needed'information.

i..

i.

.., : :1 : -...

.- c ;- :..:-.

Please. provide a.sc.necule for. completion cf t h e. failure moce analysis it cice c r e a r. s a r. c cracss ter tnese s.fste I.

F i g u r e ; 3 '. 8-4 '

Please provide informa* ion on the locations of.alllshco and field ~ welds to' precess cice.in the guard pipe region.

f Al'so provide more information en the mic guard restraint.

e.

.E p9* '

L

+

4 I.

- _... ~.

~

Section 3.9.1 Appendix 3A and Section 3.9.1.2.28 NUREG-0800 reauires :na: computer programs used in analyses of seismic Category I Coce and ncn-Code items have t h e.,f o l l o w i n g -i n f o r m a t i o n pr.ovided.to demonstrate their applicability and validity:

a.

The author, source, catec version anc facility, b.

A cescription anc the ex:ent and limita:icrs c' its acctication.

c.

Scs :' ens :: a ser"es of tes: : ::,e s.- :-

s.a.,

ce ce Ors:ratec :: :s s.: star: a.sy s;-.ar :: ::..:::rs 00:31nec fren anf cne of scur:es 1 :n cug-

, ar:

source 5:

1.

Hand c a l c u l a t i c r. s 2.

Analytical results cuolishec in :ne litera:are 3.

Accectacle excerimental tests 4

Ey a MEE acce::able similar Oregram 5.

The cencnnsrk proote-s crescrice: in Re er: NURE3/

CR-1677, "Picing Eencnmark Prc Lems".

Please demonstrate compliance with these recuirements and provide summary cenparisons for the computer programs used in s' i s :i c Category I analyses.

Section 3.9.1.2.3B kni:- 00 uter coces.ere usec :: analy:e :ne ec'r:.laticn Oa :s?

Prosice verifica:icn.cf tr.ese ccces.

Section 3A.18 Pl. ease provi.ce serification,f.er the GHOSH-WILSON coce.

See:ic.-

.25 kJ - O h co 00 iae":5 Were se:.; s t a : Ins ' :. S Y S cc e?

Table 3.CA-1 Please. rovide $ustificat ien by :urcine s:ce valve closures are ne: in: s :ec in t r. e'

.;:: cf transients.

Aise, has the alternate shutcown cooling moce been inclucec in the design transients?

Section 3.9.1.3.12, page 3.95-23 Descricticns of :ne supper: a.d whip restrain: tests could not be found.

Please provice clarifica:icn as to

{

l their location.

• ""*%aa 4

/

~-,

Section 3.9.2

~

The discussion of*the preoperational testing program does not-discuss the acceptance limits for steady state and transient vibrations.

What criteria will be used in developing these limits.

If a stress limit will be used, what basis will be used to determine the actual stress from the measured values?

Pleasr provide a list of flow transients and a lis:

'~ of selected L6 cati'ons f o r s v i's u'a l inspections snd measuring

~

devices.

It is the staff's position that all essential safety-related instrumentaiten lines should be inclucec in the vibration monitoring program during pre-operaticnal or

_ start up testing.

We require ina either a visual or instrurentec inscecticn (as acccccriate) be conductec to icent ify any excessive vicration that will result in fati;.e failwre.

Provide a list of all safety-related small bore piping and instrumentation lines inat will be incluced in the initial test vibration monitoring program.

The' essential instrumentation lines to ce inspec:ec should i ncluce (put 'a're not linited tc) the following:

a.

Reactor pressure vessel level indicator instrumentation lines (used for monitoring both steam and water levels).

b.

Main steam.instrumentaticn lines for meni:cring main steam flow (used to actua e main steam isclation valves curing high steam flow).

c.

Reac:ce ccre isolatice eccling (RCIC: ins: wren a:icn tires or tne PCIC steam i ne cursice :: :a -

e-- (use:

to cn;itor-hign; steam" flow anc-actuate i s c l a t i o n-).

~ ' " -

-d.-

'C o n t r o l' r o'd drive' lines inside containment '(not normally cressurized'but required for scram).

See: ice I.9.1.15 St'ancard Review Plan 3.C.2 of NUREG-0500 equires that five OBE's be assumeo.

The numoer of cycles Der e a r t hcua ke-shcutd be'chtained from a c:yn:Te-Lic iime.his::ry Awd:n a minimum duration of 10.seconos)'usec for a system analysis, or a minimum of 10 maximum strese eles per earthcuake may be assumed.

Please provide just,

.ation,for using only 10 stress _ cycles or commit to using 50 maximum stress cycles.

In addition, the. number of cycles for main steam anc recirculation ciping are missing awai:ing ccccletion of the -

~

new loads program..

Please provide a schedule for completion of this information.

K &

a

. M~ A* l ' +

~"

t s

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TO ALL APPLICAf4TS:

Due to a long history of problems dealing with i n.c o e r a b l e a n d-.i n c o r r e c t l y instal.Lec snuceers, and due to the potential safety significance of failed snubbers in s a f e~t y retated systems and components, it is requested tha maintenance reccres for snuccers ce cccumen ec as felicws:

Cre-ser ice Exaninstion cre-serisce examination sr.:vtc ce mace on atL snubbers listed in Tacles 3.7-La and 3.7-4b cf Standarc Technical Specifications 3/L.7.9.

This examina:icn should ce mace after snuccer i r, s t a l L a t i o n cut nc: more than six months prior to initial system pre-operational testing, anc should as a minimum v e 'r i f y the following:

1.

There are no visicle signs of camage er impaired c'peracili y as a resut: cf storage, handling, or installation.

2.

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a r. c configuratien (a::achments, extensicns; etc.) are according,to oesign drawings a.nd s p e c i f i c a.t. i c n s.

3.

Srucbers are nc sel:ed, fec:en er s-ec.

4 Acecuate swing clearance is provicec c allow snuccer

. movement.

5.

If acclicable, fluid is to be recommended level anc is not Leaking from the snubber system.

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Structural connections such'as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter eins are instalLec correctly.

If the period between the initial pre-service examination anc initial system pre-operational test exceecs six mcnths due to unexpected situations, re-examination of itens 1,

4, anc 5.snait be performec.

Snutzers wnick are instat e:

incorrec:Ly or etnerwise fail to meet :ne acove recuirements must be recaired or replaced and re-examined in accordance with the acove criteria.

Fre-Operational Testing During pre-operational testing, snubber tnermal movements fer systems whcse c;eratine tempera:ure exceecs, 250 F should be verified as follows:

a.

During initia.L system teatu; anc cc:l:Osm, at scecifiec

e ce.ra ure inter'.aLs f:r anf system w r. i c.m a : : a ' n s.

operating' temperature, verify the snutber ex;ec:ec

..t h e r m a l. m o v e m e n,t.,.

c.

• : :se syttens wn :n' ce r:-

st ain cre a:in; temperatur,e, verify via observation and/or calcualtion that'ihe snuocer will a'ccedmebate :Ne

[ec:ec :9ermal movement.

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Verify the snubcer swing clearance at specified heatup and cooldown intervals.

Any discrepancies or inconsistencies shalL be evalusted for cause and

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corrected prior to proceeding to the next specified intervat.

The acove described cceraciti y cregram for snu bers snoute ce inctucec anc c:cumentec => :ne pre-service inscectice anc pre-c:eratierai tes: :r:gra.s.

The cre-service' inspection must be a prerecuisite for

ne pre-coerational testing of snu ber thermal motion.

This test program should ce scecifiec in Cnapter l' o'

ne FSAR.

Sec. tion 3.C.25 PLease s

. pecvice a sta:emen; as. :: the c c m : t t a r. c e. w t n.

NUREG-Oc".C,

'BWR Feec. ate-Nc::le a-d CO. ret ::c Crive Re: urn c i r. e !;c :Le Cracs ng.'

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Section 3.9.3 Table 3.98-2 Much of the data included in this table will'not be avaliable until the New Loads program is completed.

Our review can not be completed until this information is provided.

Please provide a schedule for submitting this information.

Not all of the criteria for limits are incluced.. Please provice assurance tr.at all limits are in ecmpliance wit h.*,'UREG-0800.

Also provice more detailed information en :he analysis of the recirculation cumo case summari:ec in Table 3.93-Zi.

Tacle 3.98-4 NUREG-0800, Section 3.9.3, supersedes Regulatory Guices 1.67 and 1.48.

Please update this table to snow compliance'witn NUREG-0800.

-Tables 3.99-7, 8, and 9 Verify that the limits list'ed in these tables are in comcliance wi:n NUREG-0600, Section 3.9.3.

Previce a list of,a'i insta.ces where

. :he asteriskec. et;uatiens.were usec. an:: ;:revic.:r.e neeced justifica:icn.

for 'thsir use.

S.e c t i o.n.3. 9. 3. A A.

Ptesse crevice v e.; r.

in:ercre:a:ien of juri;dic: d:nas

- cC'uncarle! a$ they cer:a'n to ivF succor:s.

.s:-ty-,. : v r D o s i,; i c'n.

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Page 3.9A-2i'i Item 6' Please provide justification for combining vibratory loads for anchors from two component systems by the SRSS method.

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Section 3.9.1.4A and Section 3.9.1.48 This section does not accress the criterien usec to assure the functional capability of essential systems when they are subjected to loads i n excess of those for which Ser.vice Limit B limits are s:ecified.

By essential systems are meant those ASME Class 1, 2 and 3 and any ciner picing systems whicn Tre necessary to shut cown the clant following, or to mitigate the consequences of an accident.

Please provide such criteria.

In particular, have the criteria in l

NEDO-21985 been net?

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Section 3.9.6.2A

't A t There are several safety systems connected to the sg.

reactor coolant pressure coundary that have, design cressure i

cetow the ratec reactor coolant system (RC2) cressure.

..nere aae aise so e systems eicn are ra:ec'at futi reacter pressure en the cischarge sice of cemes but have cu o

"s u c t i o n below RCS pressure.

In orcer to protect these s stems fron RCS pressure, two or more isolation valves are s

s pla:ed in series tc. form the interface between the hign pressure RCS and the low pressure systems.

The leak tight integrity of these valves must be ensured by periodic leak

, testing to preven; exceecing tne design pressure of the lcw s

cressure systems thus causing an inter system LOCA.

• • esso re iscia

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v a h v e4 are requir.ec Oc de a:escri or AC,per IWV-2000 and to mee': the accropriate requirements of'IWV.-3420,of Section XI of the ASME Ccde except as w,

ciscussec cele..

Limiting Conditions for Operation (LCO) a.re requirec :o

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s wnien will rec. ire correction action; i.e., shu:cown or system isolation when the final accroved leakage limits are not met.

Alse surveillance requirements, which will state the accectacle

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Leak rate testing frequency, shall be provicec in the technical specifications.

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Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and'for systems rated at less than 50% of RCS design

' pressure each time the valve has moved from its fully closed position unless justification is given.

The. testing ir,terval shoulo average to ce approxinately one year.

Leak testing shoulo also ce performeo after all oisturoances to, the valves are complete, prior to reacning power operation following a refueling outage, maintence, etc.

The staff's present position on leak rate limi ting conditions for operation must ce eoual to or less tnan 1 gallon per minute for each valve (GPM) to' ensure the integritysof the valve, demonstrate the, adequacy of the recundant pressure isolation function and give_an indicaticn of valve degradation over a finite pericoHof time.

s Significant in, creases cver this li ' ting.alve *outc ce an' i nd i,c a t j o n. o f valve degradation from o n e..t e s t t o -ano t n e r.

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' L e' a k rates higher than_ 1 GPM will be considered if t h. e.

leak. rate changes are below 1 GPM acove t r. e previous test Leak ratefor system' design precludes measuring 1 GPM.with s u f f.i c i e nt "a c cu r a c y. Jh.e s e fi tems -wiil b e r F e v i e'w'e d' o n a '.c a s e'

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by case basis.

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a The Class 1 t,o Class 2 boundsry will be considered the isolation point which must be protected by redundant isolation. valves.

In c a s e s-We r e p r e s s u r e isolation is provided by two

-valves, both will be independently leak tested.

When three or'more valves provide isolation, only two of the. valves need to be' leak tested.

Provide a list of all pressure isolation valves incluced in your testing program along witn four sets of Piping and. Instrument Diagrams which describe your reactor coolan't syst.em pressure isolation valves.

Also discuss i n f

detail how-your leak testing program will conform to the i

above staff position.

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