ML20204H554

Forwards Responses to Comments & Questions Provided by NRC on WCAP-14696, Core Damage Assessment Guidance, at 990224 Meeting with Wog.Informs That Pending Final Resolution of Comments & Questions,Wog Will Revise WCAP-14696 as Needed

ML20204H554
Person / Time
Site:	Wolf Creek
Issue date:	03/16/1999
From:	Liberatori L WESTINGHOUSE OPERATING PLANTS OWNERS GROUP
To:	Wen P NRC (Affiliation Not Assigned)
Shared Package
ML20204H518	List: ML20204H515 ML20204H554
References
PROJECT-694 NUDOCS 9903290098
Download: ML20204H554 (19)

Text

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i Mr. P:ter Wes Prcjeti ll;inaE:r, Gcani: Is sues :.ud Envisorunental Projects Branch Nuicir R:plaurv Conunissten Wultaatc a, DC 23555 0001 Suije:t Wettinghouae Owners Group Iypsmittal 00Rimnonses to NRf,,,Qirnments [ymi,jls,Edi,ngg.tLL4J2 jj,[ng Qtrname Asseuroent Meetine (ht,UllP.1302)

Red reoc e i l)

P. C. Wen, Memorandum to File," Discussion Topim Gr Fet ruary 24,19991.la:tes w.th Westinghouse Owners Group Regardmg WCA:' i 1695, ' Cere Darnage mmsment

~

G lidance'," Febniary 17,1999, Des *Mr. h At ::sc NB C :uid Westinghouse Owners Group (WOG) rneedng on Febinary 24,1999, the NIC provided n lisi of c.pnutnums/q,iestions on WCAP 14696," Core Damage Assessmeot G,tituice"(ref. l) Tim:

cairinnats'qccitorts and the usociated WOG responses were discusma durian :be meeting. Attachrnent A dosureitt ;11e WOG res. pons:s to these conunents/quettions. Penir.g hm! rio)htt on of these co ntin ntsigt. cit cas, the WOG will revise WCAP-14696 as necesstu~, if you itqutre fonh:r r;bmatica, fe<l fiee tii cix.tr.ct bIr, Ken Wvrek in the Westinghom e Owners Group Prcjw. Dftice at d. P. 3%I 4 302.

W r, tuly you u, 4

/ 81kgd,el h

Q,,4 g Lctiis r 1.ib:ra'eri Jr., Chairman Wr:tagh :.use Ow aers Group alla "htl1l:t 15 cc-YiO3 5 te : ring Conunitte: (IL, I A)

ViO G E ri nary Representatives (1L, I A)

ViOG An tlysis Subcommi: tee Representatives (ll, I A)

ViOG Licensing Subcomrnittee Representathes (11., I A)

A P. Dra (e, Westinghouse, ECE 5 16 (IL. l A)

J B O Bien, USNRC OWFN 9H15 (ll,l A)

P J h114 Jr., USNRC OWFN 0117 (IL,l A) 9903290098 990316 PDR ADOCK 05000482 P

PDR

99. i%i ix

'd i Attachment A Wc:::inghcus: Ow=:r: G:ce; Response to MC Co"="i"Wim an WCAP-14696. Cem Damm Assessment Gmdance M

=

Ref.

NRC Cc-- "9h.

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s.__.:_.....c._..,,,,,,*.n.-..,.

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3caacd. Sectica 1.3 some uttittics asc the care Ine source terr-mr etmte ec3 -umn mi u-no-uma - -~.

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l mmge assessment mentocciegy (and/or the post-acddren } incimimg ccic damage n caAI, in3 00r"-m rasaura m=m m-ini.-s-= = u r 1

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! 2"'1 105. Off?it? Or ""#f"..r**'"Dcal 5".fYt anatyses. Apprepnate aa!os

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!=4 n. -tcm) to sdcct th rurce

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s. uaniaew -

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g I n reire ome acccument. For those n! nts, the results of the 1

e core damage assessment guideline (percent core. damage) other sources of information that are used to estimate a source term tor ortstre cose

(-R wculd need to be translated into a source term estimate.

assessatent, per current plant conditions g

However the same fissim product (FP) behaner that w.uptk. des eme damage asscssmcr.t (c g., hc! dup m de Typica!!y, plants use either a source term hhrary &veloped from the licensing ba:n,s g

reactor coolant system, containment, and sumps) would analyses and/or the containment high range raint:en momter. Options for usmg core g

r,cc.N taccounted for in uds trant!atica The rb 2ge assessment / sary3e analysis results wrie indaded in the report for ccmpleteness T

since the WOG did not do a complete survey of all plants. In particular. Wolf Creek (the methodology should discuss how the core damage

ssessrr.cr.t results would be used to determine source terms lead plant for NRC review of this topical report) does not use the information from the

))

at these plant- :spectally if mformation presently provided core damage assessment to assign a source term for dose aesessment.

,y,,

O by the post-accident samp!c system (PASS) is no longer Further, it is beyond the scope of the core damaae assessment methodo!cgy to specify the avatbb appropnate adjastments that shouE c made for usmg core damage assessment results y

I (or any other meseus) to derme (Le >ource term tar artstte dcsc assessments.

The core cut thermocot:ples measure the temperatute of the fluid curmg a tuct assembly.

u-a P 15,

.the dacument states tnat an utdicated temper:ture of 12003 Investigations of the behav.ior cf the thermocouples re!ative to fuel cladding and,uel c.

Eara3 "F can be translated w a peck dadding temperatute of i pe!!et tempemtures were conducted during (bc devdopment of the gencnc Westinghouse i

.; u.,s g. Oncrgcrcy Re:;. :- G:i&!me-EF9). '.*fhich me d.e 13 sic far the M

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• I',in.I' CaG Uc IITJasiated 73 a peah ciOdding !Z'".p"'~'*"'e U[

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I I (and EOPs't the core exit theimucouptc3 cuc osed s thc ynincm ba cd cppmach !O

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! Cere Damage AssessmentTeam know when an ntattunwnt and1200*FIc:Fe0!irdy E

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' in WCAP-9753 shoe umuhe ccie e.ut triermocoupic temperamtes can la;; th= fu ! c.!

peak c!zddmg tempera *ures by 200*F to 400*F. The discussion of results de. scribes the 3

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cn m neathms m the modehng of these analyses and condudes that the cere exit l

l thermecouples wculd be much dnwr en the actual neak cladding temocratures than is I

w. m

Ref.

NRC Comment!Ouestion WOG Rest,onse

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1.,.m e,eI'm vb. heet. cei-meeMAAP4 code which i

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.. mc= lek the core as a lumocd nede (does not nelet separate dadding and firei pe let l

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i ucinac:: mperantes. i'm2Hy.expe:iman pufM by NIId PONI I

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! TL -,w ~,4e T O.F. T_ PWR thn. ener.ce'" ccectttee tnat tne core cwit umimwupu my g s

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! te several tmnored oegrees lower (nan um pcan acapcaatuiu m ua, w=w =--,j

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. the evider.cc supports the cenclusion tht the thermocouple indicatiotu were inasco oy um g'

core / clainig temperatures in the upper portion of the care.

gi B=ed en th:s infb-'N, it is concluderi shat the core exit decrmoccuptcs prende an

['

adequate measure of core temperatures to estimate temperatures at which potcatial h,'

c'addin;- ima;;c (i.e., abeve abe,it 1400*F) and core overternperature (above about

'O 2400*F) may be occurring. Additionally, the difference betweert the core exit I

thermocoup'.c indication and the actual clad temperaturcs varies w.th core conditions and, o;

for the purpcses of the core damage assessment is assumed to be a constant 200'F value.

q,

' Fina!!y, since there are a large number of ccre exit thermecouples that are nearly Cl uiJformly &stributed radially thmu;;heu*. the core, the failure ofimiividaal thermnceup!=s should be easily detected based on comparisan w.th h.d:caticus from u

adjacent thermecouple lousticits. 'llus val;dstion cf instrumen:atio t indications by iu,

canparison en caher information is aheady part of EOP and SAMG trainmg and M"

therefore is not unique to core damage assessment:

L ihe encument states d at the NUREG-0737 requirements Eased en infernutien presented in the CDA tcpict! repert, the use of fixed in-plant j

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3!fc ::cm gearr rentb.e h w W on1 e4

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3 m5 Ie f!uids coes not have a valid basis usin5 the currem l amotuni eiww damas ccaipard *a aralym of 0=F :i 0 CI I2 1 Eidt Y"E h I""'I I

'3 F

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I understanr*:ing of fission product behavior and the l On the evolution of knowledge on the propression of core camaste acetcerns anu ine y;

re.atmr of ision pre incte "* 3 6 M m dd=w: m CDA 'We'es? EC: 000 s

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! p:e,;ressien ofcore damage rem v s

t in m ut=:

1 I situatics, this may bc truc. Bu: in r;;Iity niany ofthe on j rely 3 on m'ysn "! M'35 YO N '?PS * "" "i?""5' "S"'""5"'", "-

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!!nc intuments !nzy not be IMle 2nd can n0t be @e erly NWMM retiremtats for numcrous samp!= cf piant M are not W m g-l

source to obtain the stants of the core. Tne grab samples current ie,M& ige of fusic.n praduct behavicr; the analysis resu!

s from these samplet

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i I are c'irical e=ren thew!;;h the rmb==y not le av.uiable

, are not timely. may notbe accurate and ma'; not i c avaiial:Ir, evr:: ut II c bas scrin alC0r ;

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1 T%r.:Ca!!v,. n! ants (inciadmg Wolf Creek. the icad plant for the NRC renew of this

,..~..a..,e..,.,,M.., A., w.ir,m. F_, ar.ee_n.ev Ac.ti.nn I rveis (EALs) and Protective Action l

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WOG Response -

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NRC Comment / Question l

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.,E.46 4 nisme inMrstnre ta rimvide a nie-asseavas.: g 4

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e l Arte:wir i !cvel ofacticas ticercesrv to Drotect the ptiblicf This diverse set ofindicators ~ j

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, tuva cigrgicd (ne oni:et of r5rt: ri:tn:ae= artililic Emeticincy PGFD'i'.0 7

Organizadtm Will 5150c dit tc'[u e ed ESIM iw~*d IAE3 IGG 5.doIO E 00r0 D.MC ~*

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{ ts cm.p.icM escn 355s MS rensCa m:.2c-: togy pirrnmu un uus sqmts turuiu 1

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T.w. _..4.b La.. r! a. D i c. 5 cv.er.e.n.._c -...t. d..ci,en e ~ h*b-e I, Na***W i+ is b!!jeved that tt'e etmud l'_A55 Syste ris mm tne oetai.

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' of NUREG-0737 and Resulatorv Cmide 1.97 m v:ew et the ! et NUKtu.ois t anc negueatory umae i.vs. run c=upac, u. muum-= ny,,

NUREG-0737, item II.B.3 (11) recuite numerous decirsi constderations weiteri snouia oc

y -

deficier.cies identined considered (e.g. provisions for purging saatple hnes, for reducing plateout, for g+

appropriate disposal of sampics). The PASS installation at WOG utility plants, including Watf Creek, has been reviewed mimerous times by hath ut!!ity and NRC

.5

. personnel during initial licensing and/or during the subsequient years-During all of thes g

i exami::2 icus, no desigri dis -epancies in violation ofeither NORFG-0737 or Rceulatory '

.g Guide 1.97, Rev. 2 have been identified.

h The information'provided in the topical report refers to current knowledge of fission y

product behavior that impacts the interpretation of the results of radionuc!ide samples

')*

Based en the identified uncertainties in fission product b.:havior. the results of sample-analysis were concluded to be less reliable (in temts of timelmess. accuracy and y

availabihty) than fixed in-plant instrumentation which theti:tated the change = core y

damage assesserc.t methodalegv.

1. .a n

P 37, Describe what evidence exists to demonstrate that the RCS Refer to note 1.

Eg para 2 pressure / temperature relatienship provides a reliable basis n;

uRG T.10:nr0 WC CCC:ir. JuM11V C'.K i.30 l J

1 int 'tidCmq %A..%.

j j u.do ac3 not cor.:an ccascrvatisms tnat may tend to over-l j

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I predict clad failure. Sheutd addmu tha through a I

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v n:2at:ce act:rttv.

P45, l Tn: WCAP s:nt s th: for centaum:-rn radia+:en teveis j a..ietaiie.i ady of the progrenina <Ia mic dana.ec sa,ddr.ad sisuws dat Lm.cn two l

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t mro ~ l c, _ m 4.dinc in n m ni m e# 6amt ewig little or no 7P i thirds and W% ct the cierf will experience rupture before any et une soel wc:s g

rei.e.ase frem fuel prild eiustais steouki be onurrkig. Tiw l experience a hai.k temperature signific =tly in execss of 2400*F, the poirn where fitel y.

r-

- ~ m --- - e --

-r-n.-~..

a = mu-3 I

i hans ihr tha 611% value is not cicar. Wevious ciscussion i mettetttuerature releases tegn to occur tste page u et une t'.' =w sci =r-

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j rn:rne v2tue wm rnunswi nit in euw m me oiscussion un euc 4).

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.- i l aad fiyes caa.!er e 3 n " n,urt rescaec =n4 =anu en=r l

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) 1m.i. nkrr i~ei> R ! a n.W p#+a iAedd 4 wac-i. +d j ra*g 6 <ti++caJW ti. La s

• O'h;;::;nd 5.0 I =: : 'i Of th: r ~;~ " M"C"**:

h e.

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to about a 209, gap release flowever, the guideline a-d

' by non-i.OCA events. Tnerefore etwsura a fission product reicase valu-carr-,.-

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the value for CRM2 sms to be ba:ed on a t% peitet over-to 1% fuel overtemperature as the onsct ofdiagne. sis of firet metemperature, which 5

m l

temperaturc reican Clarify wheriter this is the cause of corresponds to about 35% clad dantage (middle ofpage 43 cf the topical repcrt) assu es l

l the +~-r:ty and make the text and guiMaa ce.sistent.

that the anset of fWI cw -..:~are will be conservatively duertosed (i.e., fi:ci

_1 weim.

w---

l

[ Ref.

I NRC Comn.enti?.!estie l

WOG Response I

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.j overtemperature may be diguu3ed wycu imuu is occaring, ht is high!y li:!y b be f

.dht;md

• m -=P-"Mrc%=d m

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! Thc r. arc arnue :fac==:= med:&!cr..e.r e 'r!in 1% tor.*! re:wme ncmcrea X

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I nem2 I arnices of unmics ut olmd fioids would not provide any i fm the pupac of pronding sput to th: tmcrgccy t'ianmng cecasten-m:xtng; re, j

7 l

I clarification ofthe troe and dearcc of core damage j even though :t is net required to make EAL decia:sions or PARS. In uns v. nkA l

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! =- r-c> =- d.M n-r nrmAde any darir~*- he 5

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__.-4__..,.. _rw n., n ed r~.wwt ccWe m.th.a. i itse decenbed in the topical repert.

-3 of plant fluids for the determinatica of the type and degree of core dam:ge is not required by the new WOG Cere In the longer term, core damage can be assessed by a diverse set of plant indicators.

Dam.v,e Assessment Guideline." Alicugh the results from Samples are but one type ofinformation that may be used m assessing ie plant status.

p simp!mg may not be availabic in a short time for any l Sm also the respoine to s.wmient P2 of 10 (direc:!y be!cw).

f f

n deceinns, they would be very hcipful for confirmation of ~ l the r+.gs given by the in-phnt inst:une ntaticu. Tite gi n

results of the grab sampics can be used for the P2ofl0 ' It seems that the revised guideline is relevant oidy while the Evaluation of an acciden: may be described in simp!e terms of !) diagnosis. 2) yli d

determination of the type and degree of wie damage.

escrt is watinaing to deteriorate (since CETs and RCS.

parection, and 3) exphnation / reccvery. Diagnosis and protectica are addressed via the prcmres are different then at the time ofcore damage, and Eats and P ARs. He EALs and PARS do net re:;uire information from a cere damage r !c~ger indierive of what bad cecarred earlier), ami thne

issessment However, information from a wre damage asscument may be used to 1

fc2cwmg recove:y, some cier :pe of assessment (i e.,

validate the EALs and PARS denng an escin, Addmonal evaluations cf core dxnage N

PASS er grab samples) would be needed. It is strongly using sampic data would more appreptiately be in de Icager term accidcnt explanation recemmended that the secpe of the guideFne be expanded time frame u!

5j to (1) :nclude cose Januge assessment during ba:h the ccre l I

a.

. ergpW and acdde~. rececen ph.ner, rd G} M

• 9 i ii'r N ia-e 4 a m m3F ~-

re tn=rwm ru cv me y j

' ir respective reies of CDAM and PAS 5/ grab sampics j enteing and euting the WGG bac Saca Accdct ataw6= Omccu=:Iwy c:

during, cach phase. As mich, the guideline would represent j SAMG). He SAMG are b-ia.g used daring the diagnosis and protection paases. uuring !

p>

, tt:. ::me, the 19.t. en b: at a Generai Necy irvei mai FAIu'me buh haLI I

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,.m. _ t.,..m.!. mn. r.wh. t..w. m.. e. u. m the e..are n! r! ~ c.vc w

i durin;; and fc!!cwinz corc &grada* ion

! en the be:t ava ! bie t"icr: reran hM utt!;et esmtmemman a a mme appargum j g;

df I

meshmi roi ecimaue cme damage based on rchbih:v (time! mess, accurac'/ and f*

f ava.labdity) ar.d on minimhing radiatica expos =es to plant workers. He enteria for

cutma de 5nMG are hsed en the cmM -~ cia knrg,iettn sidh A d _

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i i remem. -nce ente vinidi iulix.rmed nom tixed m-obrd uninme.utation. Ancr ra:

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j ';,:.Mr. m 6.a :.g.u u,c. n.c CAL: sa *-!y :: bc &-cs- '-d f::= a Orm j I

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i i Emergency condition and PARS a e not brim: umhetsally assessca on a tisgn pncray eg,

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basis. His is the beginning of the explanation and recovery phase of the event Durmg

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the expiaraum t recovery phase, confirmation and/or refmement of the core damage l

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-emee would be =*ie by the utiliev en assist in piarming for recovery and cicanup.

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l.....nw m Arvatan :A,-tailed core damacc assecement method for this explanation I

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-3 l

- l tanc. I[vb si sten a n_. hod are to be deveh p d, de m.-d.hy0d l,

.l l the most apprcpriate med=>i, wiuch rsy itseptms nmu u m un nn t-g gt

! ingnm e.t tire. nor*.2b'. insta.m"*n%tum. etc-i i,,.,.,

i u,o:,... _... _ _ a..m.- -. _., i....;,., - r i,. i.,,,. t.,..a n.

! See rem. cate to P37, Parn3 (Abcvel l

-i i.

e"D

,.n m -

pressurettemperature critet;a is valid in practice. Some type wi of validation ofihis concept, as well as the w uuw.ded pi values for RCP2. CET3. and CET4, appears necessarv.

ine seipmnt po. tion w_ca.c s.,r. n,.,. a ~... ~s C" imace wol be icvised to retain e

r u

-...~ ~ - -

i P13of27 Recommend retsn ng the CHI setpomt ter ice concensers all of the hydrogen hetpoints (CHI through CH5) for,ce comienser plants. vdth the acte g, i

! since this hydreeen concentration is well below leur deat containment hydagen would caly be t rehable measure of fuel overternpersrure for g

~

fiammabihty tattit and would not be tmpacted by igniter operation Also, rather than deletag the CH2 setpoint for accident sequences in which the hydrogen ig:uters were not in operat:en.

ice condensers, th TSC should ccusider whether igniters y

are actuated. and whether there is evidence of a burn.

p' P19ef27 In establishing CH2. CH3, CH4, and CHS. the WCAP Refer to Ncte 2.

u1, recommends certain assumptio.s reg:rding the amoum of 3-

' n.a.J-et.;r n i.c&,n and f=ctica cfhy&c;;ea rete: ~! to containment. The validity of the recommended ulues

[

should be itiustrated by comparing sese values with the g

r, -

rcsults from best estimate code calculatiens 81AAP, m

MRCOR, SCDAP) for representative severe accident g,

}

.w m.

i '__tsasea en genene intoutuiuut e *wa w w ~ -,.

,m.. m..,..r

,m..

2.

p....~ -ex~,-~ - e - - y 7 i..m g

7 c

.u.-

r

.~.>

P2 i Althouch there is no specific regulattert for core damage

, amo:re cf core damage) is not used in the prescnptivc tonmutae for Ine setectort s g

3.,.

' ass (,ss.Y.,~it,10 CFR 50A7w)(9) re iuires hcensecs to have

.,........,,__so,,

i a

, ctcciar.itm at Pmer.:.. enc.y Actens i.evers tg:et tvcar-" "r n n. -"

v., m

--.-.cc_ nm_,

, para 4 a

I t

.o '

~ - ~ -

a n.>., r a.: *. m o.rs.

1,ua_....._._.a..-.____

,I

. m s

, of..-

l

! nrncm enav ine'unie core dnuage asscssmcini afit is used as I meicdologies) er the er.;ance et t'ntecitve Acpan neca..uiuaucus. a samt.tu7, g g

EALs and PARS e deculat based primanly on pt:nt p:rameters that can be dettmuaca j g>

I hatt of the licensccs cmcigency phn to make prc".

vc I

i Tt d/ori 9

' a l rrom Itvg-tn-p atn unt.lumeritist:s%

.t.,. merg,.. op - ' .e. pm'a' "rcs n-et ec. an l

g...

umo

.y gi I action rmunmendations or to classity cvents requi.,

!, m_.

..,. -.;_ i_. i, c a i a I,

I sim <tritus of niarit svsictas.

t i

i f

i ne==ty pessible excepticu o,4 L k EAL um bu,cd ca a :::- : cc:{a-i rf" {

c radicactivity level of 300 nuerocuries per gram DEI Tne Mi oCi i gm dei imio.,~ ~ i g

i the EAL bases as an indicatcr of patentin! fue! damage of 5% Altboingh rrector g

i i.....! we,. riE!\\r 1.r.t.-A__c.a_rC nct 4v.=1 in the rwised core damaac assessment prescated m this tcd.--! report, using die ".eW CDAM would prnvide a differeitt method Of Commit to

e

~

Ref NRC Comment /Ouestic:

WOG Response i ein;iar w,nciasica (c g. Tv.c CET: :::ded 200TF).

g g

I I

I M

A t

,9 9

! '!S exe8mm et rerfor co*nt actmtv as a crect mmmmr m une tuuu.=c was tasce r

3 i

. e n, =. t. _.. _.

' cn a m vr.ew c: core cam:ge sequenren,utxumme m %ru -.- ~~r -c ~

s,..u

.~-

i e

j j ecludere niat odier EAi., crtcm -~~"". " appwM..-.

I 3

EAL decimmku in -2 so I eletas"isme frime tha. i t' e conhrft 3ctWity level 33f1 Thefe*0fe t.ne.400 rNICTOCUTte }mt l

s.

I t

! s: ram DEI =peci6 cation is redundant.

P3, Incom:ct str.tement regarding activity levels used as basis

' See response to comment P4. Para 1 (directly below).

m pa a6 fer EAL for RCS clad barrier (it is based on normal m

cooiant activity) n 5

F4 incorrm smtement regarding Gie amutmt orded damaac Ec topical rcpert vd!! he rev::ed :c read "2 to 5%"

p pc.al (5-10% should be 2 - 5%)

P5, Sheuid be updated tu reflett htest RTM (96) informanm ne topical report wi!! he revised to reference Rnt-W. No other change to the topicsi y

para 4 report is required as a result of referencing this later document m

P7 Table 1 should have another column with " Indicated Core Since most of the temperaturcs in the table are tcyond the core exit thermocouple tr-ge, 6'

Exit Temperature" it would only be confusing to add this information. Adequate utformation is provided in Ni C

the text to make the transition from the core temperature to a core exit thermocosmic LA

{

indication whcic appropria e j

P9, c5 "4 de~ v.hich FP epecies are cencid-red "nen-Dis is discussed in tFe rnp; cal repc*r in Section 2.2.

n N

l paraI velable" He last scatcoce is not truc if non-coolable O

i Eco:nc:ry fc=ts or molten poc! is retained due to extemal L

l r**r Yessei cceri1g I.

y'

• i P9, Next te last sentence seems to say tiw any senorimn (Te) l Widic the core is in-vessc!. it is expect:d (based en MAAP snalyses) that the release of g

a. r.ral ; cr m.:JJ scrucn: cf nca vei til : wc.!d indic:.te in: de

! Te and ~.T.-w!We' M "" o"4"v_mt w.w.id im ncgi;gihic fer ni! caru d.una i

i core is ex-vessel 'Ihis is no' true, as evidence:t by me car!y i sequences (except possibiy a hot legixeak ncas d.e auctor vessc: nozzic) duc c rc:ennen j l

} hi-vessel scurce term in NUREG-1465, and should be 8 in 6e reactor coolant sys+cm. Derefore, it is believal that the presence of 'l c and nc:t-

'.i' ;

1 I

i

u..:-sit., m :!w conenmment : nar a renstric in:iicator cf stei mr mperatmc. cven e A :

o.

I jci;rifh-1 g

2

case wherc &c molten poei

s retamed tr-vesse! d'x te e truui m,or vessei coniing g

g Hewever, it is recommended in the topical report that the NUREU-1463 source :erm be i kj l used to predict the comammcit! ratisatiet mwitor response. Ser.sitivity studies y

I i nr.n-r.mt.im;n, she wna en A memodotwv descioument snov,:c mat me -

e:

t r -- -

s l T:::d =cn t-

• e::e cea ai-ere =imes= n n nimr rmm=w fix core damace j

j j

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t h nr t

o o t r s o e hh aic r,

t T

n a eG m.

s E

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c r ch t

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eh e n et l

k mmf mor gi m (e a

c t

t v

n.a c - W ~,

w e mt.a s r a c e n t

f n i

o o c

,hi t

nf pt ad e y d m r

i c

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t s u. % #h o-a s

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r d g o

s s yr c :m ad ek ca onf s

t a

c n l

e l

min

e. b o ch u elo e,

d t

cr s h a et s h ot s o ie t

e !t v -

at u

nie S r

.A Gmf e Tc ae c Tdi n

a I

r e+

l

,ll!

.i ll 3

! f! I 8

• it I,ig I

, 4

+

,l

,4

,3 K a e

f 2 a 3 a 8 a I

r r

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e 1

r i

a a

F a R~

P P a P p p

n p

~

l,iljl!i }i ele, l

e ! e i,s 9.iI

I Ref.

NRC Comment /Ouestion l

WOG Rest >osise l

i r

i gi comaunticias usay astsu ac>uis att avv= inyssivsca l

SI '

meanrntinnc r:r* indes+ vr tJihr Ar-crrr-nf enre sbn:ne l u.- - s _- -

.,.r.,

! PIC i The ::cuircmem is fcr := li= and = - :9 : M 'hin 3

! Thts sist-n.ae a-tt 9a -mn in me icpect re;w: to a muy niim i.:nic s:: vm s ;

i

.._ i.

i pam3 i 1.ouis of the decmon to da sc (rd.acc:c:nt Ituttat:onJ. We i actrs et me cects:ca to sampte.

l l are not aware of a:ty plant who has a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> sample l

3,

! IcOSE'Ocnt

-r

! P72, I %e docunet indicates that a 0.1% clad damage is used i To be rensut with the resconse to t_omment ra, parai, me icpicai repon wm oc i

para 2 for EAL classification. Picase describe how dus level of revised to change the reference for the EAL basis fro:n 0.1%,1% and 5% clad damage to

's cI2d damage wou!d generaUy be dected 2 and 5% clad damace.

p P22 Should note that dose projecti m should not just take into Generically (and specifically at Wolf Creek which is the lead plant for the NRC review j

m para 3 account CDA but also PASS resuits e of diis topical report) dosc assessmcat chooses the appropriate data fer e apphcati=,

y>

regardless ofwhere the data is cbtained. The orTsite dose prcjection teeinuques generally g

do not usc core damage assessment quaritat vc dra er PASS results as the primary g<

method ofdefining the soarce term for the effsite dose projection _ PASS does not y

provide timely, useful information; it is point in time data and is, by its nature, always g

historical. See also response to comment P4 above.

Mn

P24, The discussion seems to over-cmphutze the amount of The infomution in this discussion is meant to provide background en the core heat-up g

para 2,3 information that can 1 e inferred from cose exit signature and is used to suppc:t the assu=ptien th:t core meki.ag generally dor s not therrwm,Mee (C ETs) DN tevel of dHcussion seems

, occur untd 60 to 90 % of the fuct red claddmg is failed as discussed in the response tc u

N unnecessary since the cpercors will not be able to discem Comment P45 Para 2.

the ddrerences between sequences based ca the CET data yi (given that it is cr ly avadable up to about 2 IOOF), and since ti.c guidelines do not include such

'j

]

2mh&_m_v term i Pu-as i Figures 2 and 3 present " core exit thermocouple

i ne uue or tcese ngures wm oc uwingul to bad.carca usat tncy arc pt

a:cted ccm j

e' 8 i 1 temiteratures, inh cf core exit thermoccuple indicari.ww l

p2 -

i ndicattons" -ttues which wa.dd not be possible for the I

i i

a, i

c j

l e hti ;;iestn-dc. Sh-mid inicate that these are l

. gI l

I " predicted pcci: sci tcmpcrdums" raict $= " core exit j

j

f; i

iek c_y...qe i.,A.wiec '

, P29 The ciscussion seem : rn impiy th:n operators woula ass:g::

- Ice tast scenence wiii be deicted frem Clie puaffaph since it provider. informatica Sat ]

h; I

no r, i I Airerene i,ve: nr rrm*n;ia, en crT,,miin.o.rkm,1,=

1 mw he mkter.dma and docs not directly supoort the case for de rev.scd ccre d: mage i

~

n a

~

s 5

t r

.. b y.

OkO Ic*.~:1 =f

• 21i*""'

w i

j :mee tne gmceime: ce net -:1c s :ch j

i i

i guidance! instruction:

Should include some ciscussion oz nydrcRen product:o a As safrd in the response to Comment P2 cf 10, this core dm=ge wsment 3

~

- i:n= r=&niyr,is and corro ncn, cnd he.. it cc= pares to: (1)

==thodelegy is n-for use during the transicr.t portions ofa cme damage accider t P29 piuduction Lum s.iad over-teutperature, and (2) the 1%

until the plant is broudt to a contm".ed stable conit:en. For purposes ofdeveloping a w s.

o 4

t A

Ref.

NRC Ccmmen!!Ouest:ca WOG Response _

t hydrexen concernrmion value d1wimed in Swnm 6.1.

numerka' nwthodo'cgy, this tim; period is :2==cd te be != i:n 6: fire. 24 hcurs cf

! ! rum er an ecce m, esmr. c-,rwive <ic6en bris assumoucas, is reremiiv a f.*cian y

)

A--riin.: hinwen omduction frem radiclysis and corrosion durin the first 24 9

e

., i 9

9 w

l cr 3 to w tess man tne tycregen precucaen man z.n-waxi tem.i:= e,mves cerc i

t l untuvery.. Ad&tinuily, wnm ac WUU CDA mdhcMogy, cre.tamnten I

i 3'

l l

i

, ~~%1 Wde ef tha 2mc2n! Ofcere ove6m+mr due in large nnecriainis in the

[

e l<

s

{

I' f

I amount et hytogen produced t>y zire water reacuens for any iridividual cae dm%c l

l g

! accident sequence. He hydrogen production nom radiolysis is within the uncertainty of E!

the amount ofhydrogen pruduced from zirc nter reaction ucrcfore, hydrogen g[

omd-+ica from radicivsis c:n be neglected in this methodclogy.

3 s

c 1

~~-* C - 6 he centainment has y'

P32, ne armun orcesium hyds odde in the wnta;nment

. Ccsium hydroxide is a hygrosecpic species.

t para 1 atmosphere is said to be much higher in sequences with a large impact on the depletica rate ofcesium hydrcxide frum the containment g

I sprays or fans. Confinn that this is correct.

atmosph:re. R: grantat'= ' sett!ing cf cesium hydreside is enhanced when water x

(steam) is absorbed. He depletion rate is higher for accident sequences without comain nent spray or fans due to the higher steam contcat in the ccataimnent atmosphere.

g'

P39, The discussion scems to overstate the impact ofinstrument histrument errors are typically provided as a percentage of the full scale indication.

i; i para 3 accuracy on low end readatgs. Wt accuracy given as a Thus, m the case of the hydrogen monitor (full-scale is 10 percent hydrogen, masmom

].(

function of measured value? %c mesage could be ctrer is 10% cf full-scale), en mdication of 1% means that (bc cer.tainment hydrogen is t

inen ectly interpreted to n can that hydrocen cencentration between 0% and 2%

. i values !:ss dtan 1% are not high enough to be taken ti !

seriously and used in the assessment P2cfl0 ht the event of an SGTR or ISLOCA, containment

. I For SGTR and ISLOCA, the core exit thermocouples provide the only reliable method 7*

radiatica monitors vcli not provide us:Fa! inforrmdica. %:

for estimating the amour.of core damage These :wo accidents result in the direct ro

.... s..

s

m-W of p crm. we

a % q pmandy m

} tcenc or ramcacm: ma: nm n :nc =vuur.nr= unu un u ; :wimm um.e azu :

c n

5i j

i t c a s. Expiain why use ofudier radimion momiots (e.g, j wotdd i.c madc taasc4 oa pl ant parmne: rs, cmerg:ncy prcccourcs in effect, and spt:ms j l

, b the s:ctm line or M)is not sugg~W to provide I availabihty brdicating tou of(nr potentet :oss ot) barncts. Reicase pcint monitors are pf i

a t i.

erimure. nt enre e.

nusge

sc circeny usee m re_i_ znc rms. asen- >>. m m u.c w w...._,s, -.a t

am.

i

FAF.3 would be !econwended prior m enre tiamrage; iiie use of die sc!c.sw uumt iad i

EI j

j i

.s I

i momtors and castic / cttsite radz.arton surveys scr.es crectof as a cacep m the t:.AL ano j g;

PAR assenments. CDA is not needed for the ca:!y r::: pense to these events. Release l

j ooint meni:crs canic b: usc5:1 m vaW6g dte zelmiive a uvuni 90wie dareagc (in bro i

i i

i i can-mried e dimaini num tire s. ore ent utermocoupics. but aue wt iMdd in thc j

r s

4;;gy 4; r.nhcd ini :::i:21 ::0 -(ncr ini exiriing CDA

i l

j j ev..~d COA ud:

i i

i methodology approved by the NRC in l'im h '>

I P4cfl0 it is not c! car whether agreement within 50% can See rcspcase to Coma P19 of 27 gl

- "" be achieved. His should be ev=1=ad through

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Ref.'

NRC Cnmment!Quctin i i

WOG Response -

i' t.

x m not seg anc in conc seg. y transiem, p) ms a n.

t

. } PI20:27, Oscit: nhc "CS,..a. i='::- ':: :in -- l Si=ce RV'.IS i: sed 2s a ca"8'=2%'i~iM'a, "" Pdim d'he leve! with ran:ct l

~

I icaetc-wessd icxet irsinanc. ara:ica Tc r c; nfur.atin;,ince ! rc cc:: ?- mre i- ~ * ~9p "re. ~&2 c~e"M-p'*". '"mrmacen m oc g

.~...

i it would affect swoiica icvel. Confinn what picssure was I conststent with tr.c trlu 5evere Accident Managemen:_tecanreat nas s ameumem ti n-

=

t he ma j.

3:

1 assunni in determimng the reccatmended value, and

. l 101h69) descriptiert of f.he OX piant damage state, Tabic 2-1 of that wint i Etes d.m"er c! wen at a p Mye in&2 tor that the care s nee n'enver :t t

j v.Aa.mhis v!"c-be d= the P.CS pressure ef! bet r

g

=

! Poof?4 NtW 3 ed 4 ; w!ic.ve ht the enre should be nartially i nc topical retinct will be revised to state that the core exit themtecouple, toop K i ts, 3

uncovered if clad damage has occurredL His is true ifcore reactor vessel level and SRM histories accd to be considerect in addition to current C;

~

~

damage is occurring at that moment, but the core may have values 3;

m ;

uncevered previously, and then been recovered. The

'S.

discus < ion simuid indicate that the reacter vessei level and'

-e SRM histones need to be considered rather than the 3

mrtantreous values. (Same wemnent en P18 of 24) g-Editorial Comments

.o,

P2, Change " Criteria II.B3" to " Item II.BJ' He topical report will be revised to read " Item I[.B3" is.

M 4

Para 2 n

P37, Change "if the RCS temperature is less than 1050 psig" to

%c topical report will be revised to n:ad "if the RCS pressure is fess than 1600 psig",

aj para 3 "If th-RCS pressure is less ebn 1050 psig "

per this emnment and our response to comment P5or27.

' P44 Eip:rca 5 :nd 6 shcu!d b; =p!: ::d en -

' d-" wculd Since p!=t :pecific verriens of the=c figures are developed u part ofdeveloping a plant-f; penmt interpolation specific CDA, Figures 5 and 6 only serve as examples. Ecrefore, there is no need to ne-4 plot these figures to permit interpektion s

P46 The wonis "fhcl rod over-temperature" and "fuct over-Re topical report will be revised to state that the descriptive phrases are meant to be

.m Table 7 terr.perzture" seem to be used interchangeably, as are the compicte:y mterchangeable y,

l t n&ds %:c c'.mcittpe.:r. urc" c d "Mi cva-

.f C

. temperature". Please clanfy in the dccument whether these j'

3

'~

, i wania an: inn:nded te be synonymous

!_F7ei24. I 'ilee md3 C151'4 oud CET) as switdvd ia sc sccA.4 I E::cpics! mf=: will bc =vi::d ic==:::tly id r.j cgu 3r:.3 CET4 ce ;N3 page j

g, I

E a

u 6

s narah I centence i

j m

f4 6 w

b)

L

.E u

.T, 9.s*

Ao s

n O

V Fi 16 i M 9 til: 1 o

(w)-r1A,org ppOJECTS 412 374 4902 10 8 0 01415:3 2 P. a/ ts T

9.-

Not: 1:

NRC Oxn neat P3't, paragraph 3:

Des: rice v.imt nidence e>.ists to demonstrate that the RCS pressuic/tcinpt rafurs relatoriihip provides a :6a le b;u is forjudging whether clad rupture will occur. Justi:9 tha.t the cod:s do nr.contain s

am;er satisrrs that may tend to over. predict clad failure. Should abess th;s throug,h a 6..didit:oo at ti vit i'.

Wt)G F.es;onse:

i Tts rrixichn3 of fuel rod cladding burst using MAAP3D and MAAP4 coniputer ccxle tralytus, for a w de ;pnge cf core damage accidents, was used to (evelop the conelatton fcr clad dur.:ye in the WOG Cme Damye Assessment Guidance OVDPA-14695) htAA P (soth t be 3b and 4 code versions) models fuel red claddag capture luul on the stie!!cs i

dn oloped in a fuel rod ard the burst strenSth of the fuel rod cladilaq. The tratet contaizu 5,retal d:taili that; i,rt entiN to the nudeling of fuel rod cladding rupture:

T be be rst streogth of the fuel rod claddmg is a function of cla jjing terr,r erature; a realistic a

aise:ismett of the strength of the fuel rod zircatoy, from NEREG/CR.C497,is us<d.

'I hc st,ss developed in the fuel tod is compared to the burst strength to pred.ct Ibel rc d faihue, th:

a adimg tress is a function of the delta-pressore across the tu ding, t:n claddian thichms and c adjiqt nameter.

i

'lhe red internal pressure during a core temperature transient n; 4 fun: tion of the atlial fill

+

r ressura, :he xenon generation and the fuel rod temperature; tne xenoa generation i: a funct t'it fuel ri4 burnup i

j 1;lastio atd plastic strain ts evaluated for the f ni rod condit.ces. Thn irl red dheru:tu and e'

t:ic!cieis and internal pressure are adjusted to account for eluric am. plas:ic straen bince pla stic stram is a rat: phenomena, the ft el icd failure is r. fmedon cf the fiiel rid a'

t imperature and tinM at temperature.

b esults ol'a number of MAAP analyses for a typical 4 loop Westinghcuse PWR, at d ffnmt n:aeter coutaat system pressures, were used to deselop a generic profile of < bij burs.t vs TCS pt:snne. Tif s is reproduced in Figure L i

i l

i 9kg11M c

TR 16 1FM bl: 12 FF (WM1A. TOR PMJJECTS 412 374 4902 10 31201:sts2002 i>,1aets l

a e

Typicci Fuel Rod Clad Failure Temper.r ut es I

2 0 0..-.,,,

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7

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P.C5 hesme (pa.ig)

T his figure shov s rod burst for a large LOCA at a clad temperatur: of aton 14757. Th s car re:poads te a core exir. thennocouple indication of just over 1:TT (Pagn 37 of the WCAP-14696).

At a ; 200 psi;; reactor ccolant system pressure, the correspondtr4 clad te nperatue at which future wo 2k be pr:dicted to occur is about 1800*F. this correspendsf.o a 1500*F core exit thirrocouple in her tion Takir g iato ai count variations in fuel rod strer.gth, clad temperatute vs. core exit iodi:s licn, burnup, time 2.t terqie:ature for plastic stram to progress, etc., the values of 1200"F and 1600T were chosen to repreaent l'.i gb and low RCS pressure conditions at the time of clai ovenmtirg to the durate point.

j A itai:w of care damage sequences from PRA stuslies shows that ino :t stqueaces citt.cz nt dn above

a. vdue of aXut'1600 psig or fall to at least the steam generator s::ondiuy side pies.surc r:0ief setpoint (e.g.,1050 ps g) before fuel rod overliesting occurs.. For thme r.4nts abo ve iis vdue (1500 psigl, oxur t1 civulation in the reactcr coolant system aod reflux coolin.; scen Ib3 stearn genemtos s is pner ally oc cntring which slows the core beatup rate and results teh: a "typac:ll heatup r ue, s qar :lless cf RCS pressure. In addnion, the core n:<it thermo:ouphs would be exp:cied imc.re close y trt.6cihe cladding temp <!ratures (i.e., less than 2000F). ~llas, a "alue of 1603 ps:g br the ILCS pnsit re to distinguish bet,veen high and low pressure con daTutge ewou is appr:'pite.

-Note?

l i

i N 3.C Coirarent P16e 19 of 27:

In est able.hinj CH2, CH3, CII4, and Cil5, the WCAP recommeuls certs.in asxnptbui septrdmg the in ci nt ofinttal-water reaction and fraction of hyitrogen released 1o coattin:nent. De i.alidity of the ir.w imended values shculd be illustrated by comparing these values with de iesults &on I: cst i

- cutiirc.te cak calculations (MAAP, MELCOR, SCDAP) for iepu:senta.tisc p en: ac:iJent n+ences.

i 99n015 cm

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De esl' mles of fission pnxluet behavior are based on a large nun inr of annlytes, prhauity with the l

u MAM3D ar.d MAAP4 computer codes. The MAAP3B a. rye: acre, in irgt part, used in the ihrnkpirent of the WOG Severe Accident Management Guid:mce T1 e MAAP4 analys us v.:re used to valida:e the MAAP3B analysis results. These ansdyses l roside key evidence th.t the contiinmer.t i

hydro;pn ami radiation !cvels can be impacted by the integrity of th: reactar cooltat systi:rt d. iring core

.hrg;e ud can be correlsted to the reactor cothi system pressum !!urher, these analynes s,ere

.t:3 t1 develop the estimates provided in WCAP-14696 for the ty drogea, noble gm, cetiam and iod no ruent on in the reactor coolant system for core dan. age accidents et high an:! low reactor cwlant sys:er t prcat utes. In addition, the various in-vessel hydrogen gen: ration estirr.ates picv ded in WCAP-l'109( wem al so derived liom thesc WP analyses as discus sed b(low.

A ru;;e of milyses are readily available and are summarind m Tatne 1 beimv. The w OG Core D.ny:e A me:sment (CD A) Guidance, WCAP-14696, is based en an ass:ssm(nt of these ar.alyses.

IEmiancnial to the WOG CDA is the philosophy that the core d unap,;c esitate should be as, realistic a s po :sible, bt.t on the " conservative", or high, side. Frem the nual:s bel.nv there is :iigdfic.mt holdup of cesiun and iodino in the reactor coolant system regardless of the reactcr coo'. ant systan preisure.

Additionaby, there is signitica.ntly more retention in the reactor coolmt syriem when the BCi is at hig.hit pites ire, ne #n the increased residence time for th:se fission prom:ta in f e scactor coBant : yr.em is also noted that depressurization of the reactor c(olant system cut but <lces tot necessiui]y. result hi the tr mspoit i>f significant amounts of cesium and iodine acredo 21o t.1c (omahment. 'Nih the conte st of des cloping a core damage assessment methodology thm i:s cas> to ute (i e., does tot require expeit kiwledge of ses ere accidents), the amount of cesium eat cline f siion produ:ts ia (Le contaiamcar vras directly related to the RCS presstire. Thus, if th: RCS pressure was lun 98% of the cetiu n and icline was assumed to be retained in the RCS;if the eactor (oolaret systern pietsure is lN,10% of11e cesium and iodme is assumed to be retained in th RCS. These best :stma o " bounds" s ro s'aprcatsd by the analyscs.

71:e noble g at es behave in manner similar to hydrogen in that the cnly ictenuon is due to the pues that teu pressur:zias the reactor cotht system The hydrogen / no'cle su icteation in -he RCS La illustrate try the two an:dyses presented on page 29 of WCAP-lG. %e re analyses ne ansistent with the iusults of numeious other analyses reportal in individual plant severe accide.at ti.:alpes

'n ess ;ualyas show that for thetransient case"where the ree.(to coolant syttem pre:are is high (e.g. 2250 prig), as much as 50% (e B, at 25 minutes after caset of:rire wates reactions in the table) of the h drogtn (and consequently noble gases) can be retained m 1.h: reacic r coo:. ant syst:m. The i

analyscs illar. rated in the report also illustrate a significant dificoace in te in ves:sel b pdrepn pr od;.ction that has beci observed in other annlysts, which can N oorrelat:d to RCS perss.te at the line of hydrogen production. Dere is also applicable informatioa on i.aoet sel bydtcgen pralu: tion piesunted ia F.xpert D's elicitation on PWR In vessel bydrogen its the NUlWG-1150 m.ew ;ianel QaFWG/CR-4551. Vol. 2 Rev. I, Part 1).

La RCS yressure cutoff for high and low pressure is based or eni;ineenig,itdarneut ficm a large

un :ict c:i se vere accideat analyses.

or fissue product retentien in the RCS, the specific value of i6}0 p.iin vias chesero represent i

~

' hose con e damage sequences in which there was residence i ac in du RCS for 6ssi.au icoduct.s.

l This pres sure would only include events that can be classilini as tnuuitsits (non-LOCM and very

.imah LOCAs whkh depressunze the RCS suf5ciently to pemte an Si s [pul.

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Fc t hydnj en the spectfic value of 1050 psig was chosen to i mesent thme a;ciden; secp:nces th.c renais at or above the steam generator secondary side relief setpcint. Je these u::lu: noes, the l

hWnon traer.. tion is increased (ccmpared to sequences at iner pitusures) due to alluung ficm th : steam pnerators or natural circulation processes, both of v1ica tesult i, coatinued ulcam supply to t2e werheated core. Core damage sequences that v: cur at f CS pressues bilcw 1050 l

peig wc to have bizber *rnm losses from tlie t eactor cooluu estem ugl tic hydioger. genemtica is limite411y steam availability j

Table !

Comparisc.n of RCS and Containment Volatile Flusi.)n Pro hats fcr Core Damage AccidentScertarios I

~.. -- ;

t Timo Event RCS Pressure Ceiirn

!alide (brs' (psig)

RCF C,cptiirtrient St5 tim B ackcur; No EFW (iter 1) 1.2 3G Dry-out 2250 0

0 1>

Core Uncovery 2250 0

0 2 Si-:

fust Befort Vessel Failure 2250 96 4

i 4

After Vessel Failure 15 92 8

Statioc 8 ac:cout; TD EFW Available for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; RCS CooldowtdD< prets.trizat:en w/ SGs T.Ref 1) 7.3 SG Dry-out 200 0

0 9.I Core Uncovery 2250 0

0 11.0(-)'

Just Hefore Vessel Failure 68 98 2

11.0 After Vessel Fa.ilure 15 92 8

LI)FW;liFW 8mtilable; RCS Cooldown/Depressurization w/ SGs tit:f t) 72 Core Uncovery 175 0

0 9.3M Just Befon: Vessel Failure 200 90 to

!!. )

After Vessel Failure 15

____SE 12_._

RCP SeaI LOC 4. ECC Recirc Failure; EI'W Lost % 3 lirs (Ref 1) 9.5

. Core Uncovery 175 0

C 1... E.)

Just Befbre Vessel Failure 175 90 10 1S 3 After Vessel Failure 15 89 11 Siaticui E tacLat.t; No EIW; PORVs Open @ 1200*F CET (Ref 1) 1.2 SG Dry.out 2250 0

0 1J Core Uncovery 2250 0

0 19M Just Beicre PORV Openmg 2250 0

0 2EH Just Before Vestel Failure 175 59

?!

4 After Vessel Failure 15 Sti 62 PDF.V LOCA; N) EFW (Ref 1)

Core Uncovery 1070 0

0 l

23(4 Just Befbre Vessel Failure 900 89 i1 i

4 After Vessel Failure 15 87

.3 l

SETR;Uo ECC; PORVs Open @ 1200*F CET (Ref 2) 3?

Core Uncoverv 1800 0

0 3 l'(-l Just Before PORV Opening 1800 0

0

,4 (4 Just Before Vessel Failure 120 50 42 9%gn t!.a c l

I I

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o 6

Af:et Vessel Failure 15 49 42 No e: Reteases do r.ot all to 100 lue to iclease to atmosphem vi:t !GTR Medium site ISLO CA; No ECC; PORVs Open @ 1200*F CET (M?

0.9 Ccre Uncovery 1100 0

0 l.l(-)

Just Before PORV Opening i100 0

0 1

2.0(-)

Just Befon: Vessel Failum (20 37 it 4

After Vessel Failure 15 34 16 No e R:kases d3 pot add to 100% due to release to atmosphere via ISLOCA LaIc Eot Leg F:reak;No ECC (Ref 3)

DA Core Uncovery 20 0

0 1.2 5(-)

Just Before Vesscl Failure 20 18 62 4

After Vessel Failure 20 16 61 Laq;e Col:1 Lc;;; Bicak;No ECC (Ref 3 0.4 Core Uncovery 20 0

0 1.25(4

.fust Before Vessel Fa rc 20

$8 42 4

Uter Vessel Failure 20 S6 41 Stukn DJacLar;No EFW (Ref 4)

N/A Core Uncovery 2250 0

0 l'

N/A fust Before Vessel Failure 2250 93 7

N/A Mter Vessel Fatlure 15 93 7

, N:,t e Not us:d in origual CDA developrnent, but incit.ded here for cong:t:ncm l

Sta: ira H act:cx; No EFW; Loop Seals Clear (Ref 4)

N!A Com Uncovery 2250 0

0 N'A lust Before Vessel Failure 2250 92 8

NiA After Vessel Failum 15 92 8

i Stuion B ackout; No EITV; PORV Open @ 1200*F CET(Ref 4)

N,A Core Uncovery 2250 0

0 i

N, A Just Beforu Vessel Failure 400 77 23 Ne A After Vessel Failure 15 M

23 Strinn B ne<out; No EITV; Surge Line Creep Failure (Ref 5) i L7 Core Uncesery 2250 0

0 31 Just Defore Surge Line Failure 400 95 5

47 After Vessel Fadlure 15 82 l8 ~ ~ ~

~ ~ ~

Std:o Blackett; No EFW (Ref 6)

N%

Core Uncovery 2250 0

C N%

Just Hefor: Veuel Failure 2250 91 9

NS After Venel Failure 15

.. _8__9. -. __ _1. 1.. -..

Sins]l T.0CA in Cold Leg; No EFW (Ref 6)

N%

Core Uncovery 2250 0

0 N%

Just Before Vessel Failure 2250 91 s

NM After Veuel Failure 15 89 11 Lmsc. LOCA.o ( old Leg; No ECC (Ref 6) 0 0

NM Core Uncovery 20 N/A Jast Befere Vessel Fadute 20 39 61 N/A Aftei Vesucl Failure 20 3tt 62 Lup LOCA uiIlot Leg;No ECC (Ref 6)

N/A Core Uricovery 20 0

0 N/A Just Befere Vessel Failure 20 29

?!

l N/A Mter Vesnel Failure 20 23 72 i

n e p n.n

% 1;. L$r/s ; 4: L5 FR (W)-NiJfR PROJEC15 412 374 4M2 10 3:.2011t'32032 P.10'tG e

)

o l

l Lair.c L.OCA b Cold Leg; No ECC (Ref 7)

N/A Core Uncovery 20 0

0 N/A Just Before Vessel Failare 20 25 7S N/A

' After Vessel Failure 20 25 75 Note ArmIgg wee; for Zion-like pir:nt ustrig MAAP4 Lo a of AI Feedwater; No ECC (Ref 7)

N/A Core Uncovery 2250 0

0 N/A Just Before Vessel Failure 2250 93 7

N/A.

After Vessel Failate 15 92 8

Ncta AnaIgis,,,.3u for Zion-like plant usu)g MAAP4,

Ref 1 IttrgJtalc Unit 3 Severe Accidant Analyses to Support Develci: atut of Serre Accident Priachircr; WC A?-11607; Westisghouse Electnc Corp.; Septernbei 198

Ref 2 Itir ejitis Unit 3 Severe Accident Analyses to Support Developta:nt cf Secte Aco:kat Pnudures;WCA? l1607, Athleudiun 1. Westingliouse Electric cot.; Febnnry 1988.

Ref 3 Un subin:le! MAAP4 Analyses for D. C. Cook SAMG Drill; Westngloun Electric Cc.

Jaraan 1;i99.

Ref 4: Co itmlling Phenornena in PWR Diackout Sequ(nces; M. G. l'in, ct. at. Ir.tamatictial Sya posiu n on !!evere Accidents in Nuclear Power Plaats; Sorrento, I n:y; Marc 11988.

Ref 5 Cseep Rup:ure Failure of Primary Coolant Piping Prior to Roctor Venel Failure fa: Severe Accic::ats. WC AP 11910; Westinghouse Electric Co ; July 1983.

Ref 6 Vo;;tle IPE Source Tenn Notebook; Fauske and Assoc.; Septerner,1592.

Ref 7. Wuti.whouse Owner s Group Core Darnage Assessment Guidauc. Wt: AP-146%.

ByggEly:tric Corp.; July 1996.

r.,, ewn 4t TCi10. FVCE.18 **

Westinghouse Owners Group- ~

cc:

Mr. Nicholas Liparulo, Manager Regulatory and Engineering Networks Westinghouse Electric Corporation Mail Stop ECE 4-15 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew Drake, Project Manager Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 5-16 P.O. Box 355 Pittsburgh, PA 15230-0355 Jack Bastin, Director Westinghouse Electric Corporation 11921 Rockville Pike, Suite 107 Rockville, MD 20852 Mr. Hank Sepp, Manager Regulatory and Licensing Engineering

, Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA 15230-0355 i

L L.

-