ML19319B355

Forwards Background & Current Status of NRC Review of Asymmetric LOCA Loads for PWR Facilities & Revised Request for Addl Info

ML19319B355
Person / Time
Site:	Davis Besse
Issue date:	01/20/1978
From:	Stello V Office of Nuclear Reactor Regulation
To:	Roe L TOLEDO EDISON CO.
References
TAC-10754, NUDOCS 8001150940
Download: ML19319B355 (15)

Text

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MRC PDR Local PDR LWR 1 File L. Engle J. Stolz E. Hylton H. Smith A11 m, a L i t e r,3 c n.,

r,n, ; j 9,.g n In f c tobe-of i ll1, tFa WC staf f noti fiei e li:ensee Of an C;eratin; M. F facility of r o te n ti al safetv p robi c'1 concerninn tne qesian of tne reactor cressure vessel support sys tei.

Those l ettces requesteu each I tcensee to rev eu tns des 1:n hasis for tr,e re]c'.or voss:1 i

su ;ro rt sys te--

r for each of i t s

. ' f a : i l i t i e s t') de t.sc-i ne

• nc the r certain transient l oads, whicn ar-omcribed in tv mclosure to tra letter h a a b e e r.'

anpropriataly t.it ' inta account in the desi0n.

Turther re, taese la tters indic a Y 'na ?, on the basis of the resul ts of li.:ensees' reviews, a reassess" art n' i

c. "P 3C te r vessel suDport 19s GP for eacn ope r a ti nn pap facility tc r? quired.

Licerseo rasn'v s-in tha t reques.

ndicated tna t thew restul atec asyr metric Ica is W not l'een considered i n t e 10 ;i yr t. u i s f o r the reac tv vc::'; su'v rt sys tem, re 3c to r i" t:-nil s i nc l udi a: tre fuel, sted" n+'at'-

sunports, puw suoports, e ergencj corc ccolinc system Irrc5)

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-cac to r c'olant sys ta i n ni n1 or centrol roa i

drives.

E u*; se que n tl y i-

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- l 5'6, the 'f)C s ta f f i nfr,rwe e 3c h N licens.a tNit d reassos; of tho reactor Vassel support sustem cM i ae f,, r eac h o f i t s f ic 11 i t '"..15 required.

Whil e t.ie e --h3 sit cf tre se l e t te rs

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'r ily fncused on t9c need cn re3ssass the

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sono 0rt dasian f ': r t Psient dif ferential press vc; in t: in%:3r renion oeta een *'< reEtor vestel anc the cavi;y snielo 21' anc icerss ths cor ~ S ml. oe i riicdtw tha t our g "M e i c r ov r.

ri ay extand to ot"cr

'rs in ti'c nucl e )r steam so71 syc tc -

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,*iy 7% ne required.

ene your infcrm *i

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and current s*.' v l' our rev i e'< e f fa ts ral l*

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8001150 NC

o All PWR Licennes A nuary 20, 1978 We have now deter-ined that an assessment of the potanti31 fc r damage to other 45$1 ccmponent supports (e.g., steam geneeat.or aro puno sunoorts), ths fuel assemblies, control roi drives, and ECCS piping attached to the reactor coolant system due to loadings a s socia ted v.i th postulated coolint system piping breaks is required.

Our reouest for addi tional information transTiitted to you in June 1976 has been revised both to clarify our original request and to identi fy the er tensior of our concerns to other areas in the N3SS, as identified 3 Dove.

A copy of this revised request for adai tional information is orovided as Enclosure 2.

The revised rMuest fnr 3ddi ti onal i n for na ti on identi fies a reovirement that your as;ec s'e"t of potential damage to the re 3ctor.essel ano other NSSS component sunnorts, re 3ctor vessel, f uel and internal s, att3ched ECCS lines and the cnntrol rod drives should incluae consicer3 tion of breaks both inside and o lt side of the reactor pressure vessel c av i ty.

Ini s assessment shrold be made for nostulited cre 3F s 1-the reactor coolant pi pino sys tem, ( se:ondary systems are not te ce included), including the followina loca tio,s-a) Reactor vessel bot and cola leq nozzle safe ends b) Pump discrarca nnple c) Crossover leq d) Hot leq Jt t c stain generator (S&W anr1 ^2 pl3nts on1v1 A number o f licer:c ss. Fave presented to the NC sta f f al ternate proDosal s, other than to conouct 3 detailed analyses, to resolve thi s ccccern.

Based upon our review of t"ese propos3l s, we have Concluded tha t these al ternatise proposal s de not eit3'lisn an accent 3 Die basi s for l on, term cperation wi thout a det3'f ed 3sMss tent of the risk resultioq frc~ taese postulated transient lea 1ina conditions.

We have, however, concluded th3 t the low probaDilit" for occurrence of an event which could result in these loads est3blishes in W193 t' basi s to justi fy continued operation for a short ter.m period.

The IPC st3f f will consider an analysis tha t i s 3rnli;abi. ta more than one specific cl3rt if i t can be aderlua tely der'ons tra ted tha t such an analysi s i s e i t"0< roaresent3 tive or boundinn for each ni Pt concerned.

Adni tional qui o rce renarding loading conbinations ( sa fe soutdes.n erthou F e inads, loss of c nlart accident Icads), will be provioed by about "arcn 1, 19 76, f ol l owi ~1 tc conclusion of st3 f f investi,ations in this area.

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qnuary 20, 1G78 please respond wi t"in 30 days of recei t of th15 letter. indi;dtinc f cur D

intent to proceed t.i th an ev31uation of the overall asy%3etric Ioss Of cnol3nt accidont ( l.PC a l loads 35 desc ribed herein.

In 31d i ti o n. pleise suD'it to us, within 01 days, your detailed sch-:quie for rewid1nc the requi red eval ua tion.

/wr schedule should be consistent wi th our cesire to resolve this crenle' wi tFin two years and shobid c l e /i rl ;. state your intent to desenst 3te the safety of lona te m continued oceration.

We are transmi ttino in f ormation copies of thi s letter to tr'e,lestinghouse.

Combustion Ennineor:nc and Babcock & :lilcox Cn73nies, if you have any cuestions or v.a,t a v cl ari fication on thi s n 3tter. olease call your NT,C Projec t 'lanage -

Sincerely,

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~' ic tor Stell o, Jr[?i rec tnr Division of Oceratica ce ac to rs Of fice of Suclea-re ac tor Degulation

Enclosures:

1.

Background and curre it Sta t'is 2.

Revi sed Relaes ' + 'r Ad ii tion 31 Informati,n CC W/ enclosure See 3 tt 3C hP1 1 1 3 '.

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Docket No. 50-346 Toledo Edison Company ATTN: fir. Lowell E. Roe Vice President, Facilities Development Edison Plaza 300 Madison Avenue Toledo, Ohio 43652 cc: Mr. Donald H. Hauser, Esq.

The Cleveland Electric Illuminating Company P. O. Box 5000 Cleveland, Ohio 44101 Gerald Charnoff, Esq.

Shaw, Pittman, Potts and Trowbridge 1800 M Street, N. W.

Washington, D. C.

20036 Leslie Henry, Esq.

Fuller, Seney, Henry and Hodge 300 Madison Avenue

~ ledo, Ohio 43604 a

1 l

January 20, 1978 r 05Uof 1 13ACKGK 4*O A'D C'P IE':T 5'A'US OC THE N?: SIVF RE'!Ir4 Oc a n " r 'r 71 C L.O_C_ A L 0 a D.5_ _F OR P WP__F AC I_'_ '_ T I_E_S_

On Maj 7,19 75, t"e 4 " aas in f oe'ed by 'li rai ni a Ciac tric 7 ?o er :onp3ny that an asymmetric loidiac on the re3ctor ves sM i cports resul tinq from a costulited re3ct" cool 39t pine rupture at i specific 1 0 4ti11 (e.c.,

the vessel no::le) haq not been Considered bv de - t l o fiqu s' or Clone 1 d ?M ter in the cricin3' dasian of the reactor vessel succort s/ stem for Uneth Anna, Units 1 3nd 2.

It had been identifiad that in tne ave 9t of 3 postulated instant 3nrous, double-ended of fset LOCA at the vessel nozzle, asjmmetric loadial ;ould resul t f rom for:es induced en t"o reactor inter-qais 5y tr3nsient di'f-ential pressure across the core 5ar el and by forces on the /essel due t) transient different131 cressures in the re3ctor

. t v i ty. Wi th the adve ' af more sophisticated Co*? uter Codes 31d t"e acco%anjina acre d?' 3iled analytical models, i t bec ame a0:arent that such differenti11 presss m, althou".h of snort dur3ti r ewl d nl 1:a 3 si qi-ficant load on the eictar vessel sucoorts and ca other comr vents, tnere-by possibly 3ffec'in3 thair i ntegri tf.

Al thouq this rotenti il sa#etw cancern nas firs ide tified during the review O' tne '; orth anna f cilities, i

i* has geqeric implications for all PnDs.

Uron closer exr,1oati' af

• ni s si tua tion, it w33 determine 1 t93t oostu-lated breiks in e r:r:t,r coolint rice at vessel nozzlas acre not the only 3re3 of Concern but r3 tile' that otner pi De bre31 3 in the reactor C00I 3nt sy s te'l Coul d Ciuse in ter"al and esternal transient lo3d; to i:t u?on tne re3cter vessol lod ct"e c04nonents.

P the postulita1 cire bee 3i in the cold leq. asymnetric cressure chinges could t3ke nilce in 'he annulus betseen the C0re 51r E'

191 the vessel. lecomped iiol "0411 Occur on toe side of the vesse' rnuiu n e a - ? ; '- tie pipe break 'lefore tce rossare on the opposite side o' the sessel coinges.

'hi; 1 ) "e n ' ej li'ferential pressure across tne.o-e Nierel could induce 13te91 I n i t h a

11 the core barrel 3nd on

'.,e rea;ter vess?l.

'!e r t i : 31 loads cou!1 11 50 De appl i ?d to to cor f internals and ta the vessel due !? the vertical fl ow resist 3nce throuqh

'h' care and asyretric avi 31 1ac 1"10 e ; : i v o f the vessel.

Sinul tim w;1 v, for vessel q)::le 'creik s, the annulus netw?en

'he reactor ind bi Co" ?l shi el d alli coul d eco"9 1;j v 'fi:1'lj p mssurized resolt'n1 11 a di f ferential ertssure acros s tr' /cssel ca ninc additiona' vi: ental and vertical exter,a! Icads on tnc v?ssel.

In 3ddition, the vnssel cou11 bn l o 3de l 'if the e' f ec ts of i ^ i '. i il !?n-sien release ind $1'. N o tiruir it the nipe nee 3i.

Inesa 1 5:- coul d occur simul tanecu s'y.

ce 3 reactor ve s 01 o itl e ' ve ti

'"e t re tjoe r

of Icalings could ccv, but tne internai loads aquid be needo'in31tly vertical due to 've ici 1 dec )'cres ;io, of t'ie upper pl e o.

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-2 Al though the PC t ', fs original ea phasis and concern.ece focused primarily on the intearity of tne reactor vessel succor'. sy-tec. i tn respect to post elated breaks inside the reactor cavity (i.e., at a nozzle), it han since become apparent that significan* as,wetric forces can also be generated by postulated ipe break s cutside the cavity and that the scoce

^f the problem is not limited t? tre vessel support s is teal itself, rer suco cutside-cavity postulated brens, the aforeuentioned concerns such as the integrity nf fuel assemblies and other structure. ne d to be examined.

n June 1976, the VC reques*ed all ocerating o',? !icensees t: evaluate the adecuacy of t"c reactor system corronents and their sucocrts at their facilities :i*h esrect to these newiv identified loads.

In resconse to our renues t, mes t licensees wi

• n '.~es tin *cr a piants oreposed an au7 m'9ef irservice inspection pro 7e v (:5l; e< tre reactor vessel sa'e-eni-to-end pice welds in lieu o' ccm. N n; an i

en uation of ons t'. lated piping failures.

Licenseos sitr Co % s* ion ngineering plante jb,itted a probability stacy (pre:3 rec b.- 5:ienca

-polications, fr., in saccort of their ccnclusion :na t a b ri,4

-+ a

,3ssel nozzle) has sucn a low crobe ili*$ o' particullr 10c3ti De occurrence that n? further analysis is necessarj A si-ii3r study nas been recently su 'i

• ted by Scier.ce Applica ticns, !nc. (SA!

f o r BM.'

plants.

<; hen the Westi" 90 'a e 1 CE owrers group recoc h aere receised in Sep* ember 1976, t'-

pc formed a special revie<, tas; grN to evaluate tnese alternative oc:resals.

In aodition, EG E !canc, Inc.

was contricted ts pv 'm an independent review of the SAi Orctatility study subnittei 06

'he TE ov.ners group.

This review ef' ort recul ted in a substantial ru ber o' 7.esti:ns

..nich previousi / h a '.

been provided to representa*.i'ees s' each group.

335fd on the na' re of these questions and o*her '1: tor; t: te sis'ussed later in thic report, we cannot accect these rennets in their preser.t fo~ a a resoluticn for the asymmetric W: lead generic ist Je.

Based or e r review, ae have Concluded that 3 suffici3nt d3!3 tase does nnt C<es9e'.If exis t wi tnin the nuclear indnS try t: Drovide 53tisf3Ctory anS'.!Crs tc *hese inforF3 tion needs.

Severil

'*';-term o.nerimental rewa :s aculd be required to provide Toch V this i n f orma tion,

a' theijan the probabili ty study recon tly cW 4'ted by SAI for certair OW Tc.mes does respond to some of tne,

• 4 ' "a l questions raised during our review of the SA: recort m m=~

by CE plants, the more fu.'damental cues tions remain.

Therefore, tnis conclusion also applias to the SAI tooical report for M':1 plants (SA!-050-77-PA).

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- i recnnd - and caua'l* iroortant - reason for - ? accartian ornnan*

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'c' aDoroaChes as a %Olution at this Doint concernt our need and in19 ster's n ed e

to cain a better unde stiadino of the problem.

'le consisor it essential that an understardiaq of tne inoortant breaks and issociato1 ceasequences be known before arrivian any reme dy - be it nine restrainen. nrobabilitv.

ISI. or some cor%ina tion of these measuros.

r.n l y in this way will we have a basis on which to judne the imoortance of the remedy with respect to what it is desianed in prevent.

Althouqh we have mine quee,tions on each of these tori:31 cencets, this t e s no c me 3 n th 3 t se / i m..

the prohahj]istiC /ISI 3Dpr3aCh at'3mnlat.ely

, I thout meri t.

In far+

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o,.

a prooabi. ' i s tic ey 31Jation serves

.2 as the basis for cn"tinued operation and licensim of nuclea-plants du-ina this interin recicd,; bile additional evaluations can be perforced by vendors and lice" unc.

He believe th3t t'o s ti'ic 3 tion f oi continued plant nnorati,a has as its basic foundation the fact tna* the event in question. i.e., a hvrothetical double-ended insta"tarews ructure of the main ccolart ninc at a carticular loca tion, has a vo-1^w rcha 9 i l i ty of occurrence.

The disruntive f ailu e Vrngbility of,a reactor vessel itsei' nas been estimated to lie be t.~ 10 ' and IT' ner reacter voa

- so l ~, that i t is not considered is 1 desian basis event.

The rupture promaaility 0#,

cines is estim3ted te be hi @er.

U SH-1200 used a re1ian valun o f l)"

for LOC A i ni ti a ti".7 ruttures per plint-year for all c@es sires 6" and q r.' a t e r (with a Ic,er and upper bound of 10" and IT ', res,e.+ivelvi.

We believe that co"sids-i"0 the 13rce si:e of the rires i n.2estion (un tn 50" 0.D. and 4

'N tnicvi, the lower bouad is -r e anne cri ate since these pipes are ra m like vessels in s i.' a.

!" addi ti an. t" e ~.11 i t'. con-trol of this pioin7 i-the best available and so EJ at 5ettor than tnat of the oipino used ir the 1554-1400 study.

These f actors. CMri ed.i ?H the f acts th2t (1) the b re a'. r' re"ary con-cern rust he ver. law. (2) i* must occur at a sreci'ic lec iti,n.

f31 the breik must oc u ncte-ti 311y inst;nt3aenusly.

3r+ i t ', t e: wei cs 3re currentlj suhio:t to inservice insrection bv val.

-t i c W. sur' ace tecnniques in accordanco <ith AS"~ Code tection VI. 1:3d us t" conclu'e that the probahilit. n' a pipe break resultino in subst3ntial transient laids on the vessel wnort system or other structures is 3ccent351y s'"311 such th3t conti Ued reactor oper3ti 7n iad continued licensina of facilities for op m 'ic, can continue while tnis -a tter is beirq resolved.

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In suorort of tre lhesn. the staff has deveinned a short

• er-interin cri-tarion to deter,.i e if 3n acceptable level of sa'ety amist: 'nr crerstlec Pads under. corditir s of a rostulated rice break.

Thi s i nte ri, cc :er;so is based cn 3 simplified rrobabilistic model that inco-enrite s el astic f rac-ture mechanics tec niquec to estimate the croh351 1itv c' a rire break.

r, r i ti c a l flaw size Tr# tubcritical flaw arowth rites aer 1eter~inec issu9 inn the Presence of a sarf ace flas loc 3ted in a circumferentill weld of a rnick -walled nine. Cetermination of the critical fl aw size aas based or an estima ted f rict'n e tcuchress value of KIC at a mini,un 'e nera ura of 200 C 3nd a unifor-tensile stress equal to the corsideratier of various creritino conditi7rs nreducinq elastically calculited strescas r3naina in silue from 1 to 1

• "vs tha material min 1 mum yield (trencth.

Tnen, usion tne c i':u' ited cri tic al fl aw size, tha subcriticii 1 rc.> t n rite. 3nd an es * ; -ed probabili ty di st-i bution of an u 'e tec tec fl iw in thici -walled pine # c ' ~s, the unner nound probabilitc o' oire 9ee3k was e s tima ted to De 1. '"

Ibis value is also sunearted nv a race

• Outlica-tion by

'r.

5.

1;'b' ahich states that actuli

'3 ilurn s t a ti s t.i c s 4

confir9 rates of I '.

to 19-D per re ac to r-yo a r in larne nines, witn h17her rates.15 'h? -ice siza decreases.

Considerinn thase analyses, wa conclude t'iat nu r ce,servitive estimate on.a nina n eak i-tne ori,a y rance of IC-' to 10'D Tris estimated n; e caelant syst3m is ia

• "o b re at. probabil i t.

i: considered 3cceptably low to.iu s ti fy sho-t-tern creration of nuc'm' n~ or clants.

!n view of all prt icos discussions corcerninn this issue, the N?-

t3ff ris concluded t at e'sa'uation rust be undertaker tc issess tne desico

= ".

3dano 3cy of the reic*nr vassel sunrorts anc ciner 3f f ec ted s

• ructures and 5 es *ams to wi ths ti

>I asvm etric LOCA loads, includinn an 1ssess ent o' the ef fects of is ~~etric Icads oroduced by va-iour cine break s hotn inside and entsid"

• -o relctor cavity. On ner'ormina tnase es11oatinas t' e staf f will re -i

• the Orcupinq Of ni ants, w9ere adeouate lusti fica-tion for suc9 arnurirc exists. in order to limi t the nu-tar of 013nt; to be analyzed.

Alt =r"3ti velv, tha st3ff will nerait the i 3'/ zinn of a

o rctotyni c a l ' clint, <ricn is sufficientis ronresanta*ive of a

-enar of clints. rrovile the necessarv information.

?nt" n' thesa concents have baen 11 cus sed.vi tn the '.la s ti n nhau te an d ~~ V :rs ~ircuns, a-d we believe t i' sa:S acoroaches could save a sic-ificar*

4,' ant of ti,e 3nd ef f o-t i n -ntain'n' resul ts en whic-te as:s 3 e

-'a c"

carrec tive reisures.

e 'Af staf' is neerirea *:

ee*

't-

- ; li e -

secs to discuss sa:" !cer,3ches and has aleca t-d:ne 50.

a a-s, a ret with the..es*

19. l o r ic-inc cure,ce of di scussi",'"# ruse owners arcun on Oc t:ner a ceneric solution fer broak s outs, na

• m reactor t ivitv.

It is evrnctad tha t 3 similar meetinc will be held in the near

• " Critical ractors in Elewdc.vn loads in tre Fca:

Guillot ne N-':le i

Break (Volune 2 -

t"r-a y'vetric Lead Problem'.'y@

0 1977 s

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, future to address brei s locatei inside the :ivi ty.

Inis 'enised" k

arproach is accerthle to us, provided that it she ts liint n

a n d serves to expadita c^nsi dera tion of the 1 ore li,itina ine.i.1a.

Civity bre'ak5.

~

For your information, the 'R' h3s a technical issi, tin:a ;ont-act wit *i EG'.G Idiho, Inc., to independently model represoqtitive,ectioihouse.

ail major

'AW, and CE pl3"ts for the purpose of assessino the 1031% oa structures and co ponents resulti11 fro 1 a sy 199 '.ri c LOCl-load;.

'4e believe that the resultt of this progran which # 11 include

.a q s i ti vi ty studies, will provide signi ficant confirqatory information re' a te 1 to thi s generic safeN concaro.

Al though, as state'1 e arlier. we believe tnat contia ied oneratian and licensing of f 3rili ties fo-tne shor t-t.-1 is justi F eq. we l'sa Selieve that ef forts to resci se this issue shoold procee i i t, mt del is, wi th the Objective of bot *1 Co"0leting the necessa'f 3s'e t: Ma te 41 1' i q s t i' lino any necessary pl ant edi fications wi thin two ye3r.i.

!" ma'<. i n th i s s ta te-Tent, de wi i to 'ne i t : lear tna t pl 3,t modi'ica ti ~s, i f indic3ted bv licensee assas;.,ents, is the prefererl,ippen;n.

at t,e sa,e *ime, we '

recoqnize that tnere 13j he ci;=::.pierein apnropriate toiiri:itions,,3v be Judgad to be uno int ?1 hisad on the cons:1? itian of a e-gli r i s:<,'.

In such C3ses, and Or'lj in such Cases, we Will be pr97ar ? ) I' 1 de firther consideration to al tirn3ta aonroaches, such as prob.4bilitj!;",1.

We feel,

1 wever, that ISI techniq;9; 1s thgy exi st todgy ;991 d 3e ce,-gijgr33;j 1mprave1, an1. to the e tent that suct i 1:]e;ve n?

wid hav? a direct bearing on thi s pr ocle, a; well M an imoact of nuclei-saffte in 03neral, we woul d welcone the - de.eloo nent.

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Ja nua ry 20, 1977

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ENCLPSURE I RD I 5ED penE 5i~ FOR 30DIT!07.'st !*!F0D"UIO'l Recant analyses have ". Fawn that certain reacter sys tem cc-conen+ s and their supports ma;, be subjected to previously underestimated as crir.etric loids under the C0rditiors that result frCm the pos tulation of ruptures of the reactor c00lart cirirg at various loeitions.

It i' the re fo re ne<mssary to reassoss the capabili t/ of tLe,e reactor sys te'v com;;cnen ts to 3ssure that the calculated dynamic asymmetric loads resultinc from these oostulated pipe ruptures <ill be within tre bounds necessary to provide high assurance that the reactor can be brou;"! safelv tn a cold shutfown condition.

Or the purpose of tnis request for 3dditional i n fo r-Na tion the reactor sys '.'" C00ponents tha t requi re rea es sre"t sndil include:

a.

Peactor Dressure -les sei b.

Fuel Assemblies. I'~1 91i rg Grid S tructures c.

fcntrol Pod Drives d.

ECCS Piping trm: is i ttached to the Primary Ccolant Pipinn e.

Drimary Coolant Ficir.a f.

Reactor Vessel, S 'c v 'lenerator and Pumo Supports 9

Reactor Interrals h.

Biological Shie'd.3;i and 'ieutron Shield Tank (w ere sealicable) h i.

Steam Generator r~

v+ ent wall The followina in ft r:' a'ico should be included in your reassess.'ent of the ef fects of oostulated asymetric LOCA loads on the above-nentioned reactor system corpenents ard

'M roaCtOr Cavi ty 5 truCture.

1.

?rovide arrange on' dr3.cincs of tre reactor vessel, the stea generator and pump sucCCrt systems to show the geometry of all princieal elements and materials n'

nn:truction.

2.

If a plant-spaci'i analysis will not ce submitted for your plant, orovide suppor'ing infor-ation to de'"onstrate that the generic olant 3nalysis under consideraticn adequately bcurds tFe restulatec accidents at ycur facili'"

Irclude a compariscq of the gecw tric, structural.

echanic31 and t'e ral hycraulic simil ari ti es be t"..nen vcur 'ecility and the case an'l. M

'iscuss the ef fects of am d i f 9 r9"Ce c.

3.

Censider costulaud Sr=aa at the reactor vessel Es' a~'

':': leg rozzle safe ends, cono discharge nozzle and crossever leg tw re-su' t in the mcs t se /ere Icading conditions for the above e"'10ned D

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o m q,N I-j h [,

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-2 5/ stems.* Provi de an assessment o f the e f fec ts o' asy re* ric ores-sure di f feren'i a !- er *nese sys tems/compenants in combir a.icn with all external lea iinas including asy':retric cavit;. pressurimi cn f;r both the reactor ressel and steam generator which micht r m it from the required rectu' ate.

This assessment should consider:

a.

limited cisc;>ce ert break areas where arolicable b.

cons ide rw er of fluid-structure interaction c.

Use of actual t; e-dependent forcing function d.

reactor supD0rt sti f fress e.

break coerinc 'ime-d.

If the results of t N-assessment re uired by 3 above indi: ate 10 ads leading to irel v ti: M*. ion in these syste s or d'solacer'ert exceeding previous desian 1-i's provide an evaluatt or o f "'e follew'ng:

a.

Ineiasti c bana.'er (including strain nardening) of the "?terial used in tro s" t e" design and the ef fect or. toe lead transmi-*ed to the bact r

-'r 2r tures to which these s /! te:"s are a tt acned.

5.

For all analysis rerfe ed. include the Tet"od c' analysis, tne struc-tural and hydraulir c~puter : odes er'aloyed, d r i ngs c' the odels employed ard co~ par'sens of tne calculated

• r allowabl= tresses and strains or da rlecticns i th a basis for tne allowable values.

6.

Provide an ect' 3te c' total arcunt c' per anent defer ~3ticr a

sustained by t": fv' scer grid-Include a riescription of :ne i"Cact testin g tha!

.m perferred n succort o' veur es t in a te.

i.ddress the e# rect; n' nee ra ti ng tei.1ra ture s, seconca r.- i Dac*.s.

and irradiateo aterial croperties (strengtn 3nd Juctilitv' cn the amount of prec :ted defor ation.

De onstrate that the fuel.vi l l remain coolat'e '~ ali credicted geo etries.

7 De:ronstrate t"at =-tice ccTt':ments will perfor-Meir 3afet/ function e."en subjected to

'"o rostulated loads result an fecn a rine break i

ir the reactor :er!'at syste-8.

"enons t ate furcti ability of any essential ir4 a..u ro ervic=

la/el B lini', v: m e=ded.

ir :rder to revien the etn-d: e cic/ed to cc~.ut= ' - ai -

H: 31 pris s ure di fferer.;es 3crcs s t"e core suc cet b arrel duri"; - 2 ::cle portier of the blea h." 3r.alysis. the fcilowin; i n ' 0 1 *. i r r 's e;ucs'.ed:

~

• E&W and CE plant ' licensees should also consider breaks in the hot leg a t the stean gener3'or inlet, D

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A complete des ription of the hydraulic code (si u!cd B:lud: q the develoD'ent of the equations being solved, the assuretions and s implifica tinns used to solve the equations, the li:ni t picrt re-sulting fr om those assumptiens and simolifications and the c.e..e ri c a l methods used to solve the final set of ecuatiens.

Provide comparisons with experi rectal data, covering a wide range of scales, to demonstrate the apolicabili*y of the code and of the mndeling precedures of the subcocied bicw1 owr 'crtion of the transient.

In addi tion, discuss application o f the cot

.o the multi-dimensional aspects of the reac tor ceome-trv If an approved serdor code is used to obtain t'e asyr~etric cressure difference ac.cs? the core support barrel, s tate the name and version of the code utod and the date of the NRC acceptance of tne coce.

2.

If the asses; Tent of t' e a syr:me tri c pres s ure d i f fe renco across the core sucrert barrel is made without the use of a hydraulic tiowdown code, oresert 'he methodolooy used to evaluate the asy7"etric loads and crov de justification that this assesImert provides a conservative i

estiriate n' ?"9 ef fects of the postulated LOCA.

A comoart"'ent Tul ti-nade, space-time pressure respcnse ar3 1vsis is nocessa rv tC 9'e r"'i re the ex te rnal forces and moments on comDonents.

Analyses should te nerformed to determine tho p re s sur tensient resultino fro"1 postulatec ho* leg and cold leg reactor coclart syste Di;e ruptures v;i thin the ront nr cavit;. and any 01De penetrati m s.

I' a:Plicable, similar analytes should be rerformad for stea generatcr

7 art ent!

that may be u yi?-' to pressurization where siani ficant r^rconent suDoort loads may result.

This information tan be provide" te enc. rMs a group cf similarly desicred clants (generic approacn) er a curei, plan scecific (custom plant) evalv tion can be develcoed In ei ther cate, the proposed rethod of evalsation and crinciD3I assumDtiOM to be used in tDe analysis should be oret' 9d fcr eview in advance of the fiaal lJac assessment.

For generic o.al ?3tions, perform a survey of the O! ant to be inclu:ed and ider'ify y reirci ie parameters which may sacs from Olant to Diant.

D For i ns tarce,

• nis s"co id include blev.down r3 te and gecrei" cal. aria-tiens in princic? 'irens icos, volurre s, ven' a rea s. anc vea

• loc at ons.

i A typical ci lead

'2"' srould be selected tc co-for-s er e ' ti /i tj and envelope cal

ui3t:'nc The;e analyr.c: should 4ociuds:

(1) nodal m; del dc<e!co. eat for the configuratier c'

ser *:nc tre mes t re s ' r i c 'i;e ceome try ; i.e.. recui ri ng

• .9 cr: 2'est nedalization; (2) the mcst es 'ric tive con figura tion regarding vent a reas and obstractiers to flow should be analyzed; and.

(31 53nsitivit. t ^ ccde da t.a incut

'ould 5:e 'a: Wet <

loss coe f ficien t s, inertia terms, vent areas, ocdal volume. and any other inrut d3 ta v.here there may be vari a ticos f rc~ ^: ant to pl an t or uncnr '.a i n ty for the given plan' p

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4 These studiet should be directed at evaluating the mav.imun la teral and vertical force av ~orert time functions, recognizing that r?dels may be di f ferent 'ci letar11 as opposed to vertical load de fini t ier.s.

fhe followinc it tbc t<pe of information needed for both generic and custem plant evaluatienc.

Although this request was primarily developed for reac tor ca vi t, 3 cal /ses it may be applied to other corconent sub-cor'cartments tu qerecal acplication.

(1)

Provide and iu:'if/

i the D pe break tyee. arei, and loca tion for each analy;is.

Srecify whether the Dipe bre3L wa Doc tulatea for the evaluation

' the compartment structural Gesign. CCrDonent supports deci'm. cr both.

(2)

For each t

.v *"e-t, provide a table of blowdown rass flow rate and ancrc.

e ' : 3 5 r a t a as a function of ti e for tbo break wnich reSuitsi"'b ? maximum structural load. and for the brc3L which was used 6'

• Lo ~~rpnrent suDDorts evalua t ion.

(3)

Provian a scN at t draning shcwing the cerrartment ner:lii:ation for the deter.. iNtien of maximum structural loads, an d for the componon' ~ r ts evaluation Provide sufficiently de' ailed plan and ;

inn drawings for several viewc, including principal direns ions. shcr. ing the arrangement o f the compartmen t s tructure, major comteror+. piping, and other rajor ob tructiers and vent areas to pe mit. ori fication of the subco"'cartnant nedali zation and vent i m 3t'en (J)

Orovide a ratu'a*ien of the nodal net-free vclu~es an] 'n'erconnectin; flow catt Trea-For each flow pa th, prov ide e." LfA lf*-') ca*io, wPve L is the r.erace distance t"e fluid 'le..s in 'ha* #10w cath a-J A is the e fective cross sectional area.

?rovi de and jus ti fy r

values of vent icss coef#icients and/or frictico factc-; used to calcul ate fles tetween nodal volumes.

When a loss 'oe ficient con-f sists of ore than one component. iderti fy ea9 ccTronent. its value and Ge r' em area at nich the ' ass coef ricient applies.

(5)

Cescribe tne udaliz3*iu sen.

i vi ty s tut. :,erf orc ed to ac te rmine the minire nu ter of -lume nt e; r'.uired to :enterva'ively predict the maxi cressure load act

, nn 3 corTa.' er*

ucture. The i

nod 31 i z 3 ti c a,ensitivity study sh.uld include cen

.)e,e. ion of SDdtial pressu e v3riation; e.g.. Dressurc.a ri a ti er circu M erontiall.

axially and r3413113 within the compartment The rodai -odel develeprent s'udies should shCw that a spatiall cnn.c Jent differeil-y tial pressure distribution has been obtained for the selected evalua-tion model.

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I ut h a. " c

_5 D?scriba ti o i n 'i fy ' bc nodalization ser.si'ivits stud < performed for the ra 'c" correnent supports evaluateJ. i f di f fe r:-t 'r:c the structural anal fs1; model, where transient 'c e:es 3-0 T3"'er:: 3:: 4-'

on the cccccreats are of c a nce rn.

Where co conent load; cre t #

orimary interes t, show the effect of noding variations en tae transient force ; ar.d moTents.

Use this informaticn to j ustify the nod 31 mr.ds: telected for use in t.ne cnrpenent suoDerts evaluatie,.

If the press uri zation of subvolumes locatec i" ragicos away from tne break location is of concern for plant sa fety, snow that the selec-tion of pa r r e ter-which af fect the calcu ations have been conserva-l tively evaluated.

This is particularly true fcr pressuri:ation of the volur? beneat" the reactor vessel.

In tni case, a recel wnicn predicts tne hianest oressurization bele.. the vessel trould be selected 'cr

'"a e'. a l u a t i o n.

Il0TE:

It has Seer our excerience that for tre reactor cavitv. tnree recicr. should be censidered (i.e., nec31i ec' < nen develocinc a ro'a'.

"mic i.

These are:

(i)

'"a fol u~e around or in the vicinit, c' t*e brea loca-

• im' cut to a radius approximated by the ad'3 cent oc: les, acd including portions of the ceretration volume fn-s me plants; (2)

"= volume or region coverinc the u;per redctor c3vity, cri yliy the RPV no :les other than th.

t' r e 3 k no Zie; w

(3)

"+ ragion enco passing the lower react:.c cavity and

^t er certions of the reactor ca,i*;. not in:iudec in Itc s (1) and (2).

(6) Discuss the Pa"ner in which Tov 3ble obstructicr' t: ' e" ' fl o.s (sucn as insulation. duc*irj, clugs, iid seals).ere trea*ed.

Provice analytical and e=ceritertal jus ification ina:

en' 3reas will not be partiall, ur ccwriete:

ol',ged by discla:ed ct"ect?

Discuss hcw insulation fr :iainc ano components <;as :rnsiderec :n cetermining Volumes and

.5"*

3"eas

1. 7)

OraO"iC3 Ilv s h 0.. 'h9 Or295 arc (Osia) and diffe"?nt'3'

"'ure

[;s4' res DCnie 3 s f un".ti cr 3 O f ti7 e for 3 reOreSe='3tivo r,~bi

• nCces to indicate t'*P Sca tial pressure r?sConse.

D'scus? !"c c3 sis fCr establishing * - di f ferential pressure on structure?

a-'

enc one n ts.

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(R)

For the co ry + ent structural design pressure evaluation, provice the peak calcu'ited dif ferential pressure and ti'*e of PP n pressure for each neda.

Discuss W' ether the design di f ferential reemre is unifor',1, acolied to the cortpartrent structure er whether tt is spatially varied.

If the design differential pressure varies depending Jr t"e proximity of the pipe break location, discuss how the vent are?s and flow coefficients were determined tc assure that regiors reccicod from the break location are conservatively designed, cuticularly for the reactor c3vity as discussed abcVe.

(9)

Provide the reak and transient loading cn the m3jor ccmponen's usec to establish tne 3decuacy of the suppcet desi :n.

This should include the load f ccing functions (e.g., f (t), f (+1 f:(t)) and transient x

~y mo"'ents (e.G., "s(t), M (t), M (t)) as re5olved about 3 specific, v

g identi fied cro -dinate system.

The centeriine c' the bre3k no::le i s recorrre-d:d as the X coordinate and the center line of tre vessel as t"e : a.i t, i'rovide tne projected area used to calculate these loads and identify the location Of the area c."o'ections On plan ard sectic-d.a ings in theselectedcocedinatek.s*em.

This infor'9ati:r svuld be presented in such a ranner tha t confi rnatory evaluations of

• hn loads and moments can te ~1do.

O Sunu

• Bl Wl T D F I

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