ML20059B855
| ML20059B855 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/29/1989 |
| From: | Braun D SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20059B824 | List: |
| References | |
| FOIA-93-92 ANF-89-131, NUDOCS 9401040268 | |
| Download: ML20059B855 (37) | |
Text
J ' ,w+4--s ss. A - - -
- 6-12-32 ;12'05PM 443 7061;# 1/11 ;
StNl By:O,,r,&,
.,, A~ , , . . . . . . . . . . . . . - - -- -
COXYO.VtAU.lti tDl50N-* .
YMMSWS MF 09131 !
l w.u- o a,d fan k.n.me f rnemo 7tm l*.e ==. ll !ssus Date: 8/29/85 {
~
~ %suces u "
PCs P:C i au cec.
} -
m --
wu .
- 3ct EDS tb i I
ORC 3 Dell UNITS 2 AND 3 LOCA ECCE MALYl!5 i j llitGLE TAILURE D][5EL CEllCRAIDR [
l . :
I I
Prepared by: l'
.I .
I' . Sc'A D().1rsun.
B d(neer t l
l5 IR Safety Ana sis ,
Liesufng and lifaty egineertrg Fael Ergtr.sseing ord Technical Services : 3 i a i
l-I- i t
i i.
August 1889 i
/a I , .
i I 1
-o 06.II.92 12:04 PM PD1 !
9401040268 930524 4
N LAPfPERT93-92 PDR yJ L .
t
r
'4 StM BY:ur&A ; ti-12-32 ;12:06PM ; COMMO)M_AL'ljt tljig,,, ,,,,,, ,443 7561;# 2/11 l
MF 89131 Fage 1 l
CREEDIN l#ilT! ! AS 3 LOCA Ecc5 MALY3!I
$1MLE FAILURI D!tsE; GDEAAT0A '
I L 1.0 INTADWCT!0N i j
This report presents results of a less of-ccclant accident essrgency core ce91tng systes (LDCA.ECCS) antlysts perform 6d by Wranced Eclear fus1s Corporatten (MT) for MF Dxe 1 f'uel in the Dresden Units 2 tad 3 reacters. l i
Tbs analysts pas perforrad in satport of a proposed technical specification !
things to reduce the required Icw pressure coolar,t injection [LPCI) flow from ;
24,50g gps for tirse pumps to 9,000 ppm for two pumps.
The 11 canting aralysti of record for kNr txt 2 fust in the Dresdan Units I is reported in MF.84-192.III The most 11miting single failurs assumed in the' ar.slysts of reesti, was fallars of the LP;I valve, on the intact rectrculation l
inop, to open. Thus, with no LPCI flow tynilable for cooling the cars, tat i resJ1ts of the analysts of record (I} are act affected by the raducticn in LPCI n ow.
For other singis failures, LOCA utlyses results can be affseted by tbs reduction in LPCI flow, s*4 the pcssibility exists that a LOCA with a diesel generatcr failurs and reduced 1PCI flev say become more limiting than the !
arslysis of record. .
I I The LOCA analysis reported herein addrusses the n. i it case LOCA including crsdit for LPCI. [
51r.gle failurs of a diesel generabr is assmed and LPCI injecticn is esseled at the reduced f1w rats. Available (Cts systems are f ]i ,
]
shawn in 1stle 3.1. The cbjective of the analysts is to confirm that seat ;
claddtra temperature (PCT) for a diessi gansrator sins 1s failure analysis estains below that of the current analysts of racers.
j i I e !, l t
i
+
.)
I li d'
1 StM BY:Qr&A 0-12-32 :12:03FM ; COXMONMt.AL'IM tl)l50N- 443 7661;* 3/11 l
-- = +
essgst4I14* GCusCmWEALTri EDI50N:8 i t
Ahf II 131 3 lage 2 l
TABLE 1.1 3RESDEN R.!A; TOR AYAttAK E ECCI SY57tX$ PIA SINGLE TAILURE Sinele Failurs !
Available Efft sntees '
LPCI' Yalva 1 HPCl**.1 LPCS"*, A05+
i
.DC Power so ply!
01:301 Catarator 1 HPCI, ! LPCs, I LPCI, AD$t
- tow Pressure Coelut Injection.
- Migh Pressets Coolant infestfon. ;
'" tes Pretture Core Scray. f
+ Automatic Depressortration Systn (not credited in analyses).
i 1
With this single failure the (CC5 ' systems shon wt11 he operab1s in the Dresden Units 2 and 3 begint,ing with their Cycle 12 Operations I
I l
Cf. 12.92 12:04 W PC3
i StM BY:Qt'&A * --b-12-t!2 ~12:Ut!t'M ; CUEYOVt.AL.'ilt LIJISOY 443 7 bbl ** 4/11
- -- * ** ' sinILL114* C0tkow;ETH WBWe 5 MT-ts-351 Paes 3 i
2.0 #UPf9Utf The limiting LOCA breakIII for the Orssden Units 2 and 3 reactors was n:J:alyttd assuring single failure of a diesel generatar and the reduced I LPCI flow rate prope ed for the Technical Specifications. The LOCA calculation wa perforned las.2 fuel. at full power fuil-flow conditions and assused a full care{ of MF ,
The PCT resillt for the calculation sistulattag Cissal gansratar failors LPCI vatveiste102 open.F lower than the equivalent esis issuming single failure of the t The conclusion is that even with the redcad LPCI i ficw. '
the wcrst single fal16ra for LO:A remains failure of the LMI valve to open Thus, .
the analysts of recardN which sinced failure of the LK! valve to opsn rstains applicable and consonative. j The ;,
i planar linsar heat gensreth nte fMAPLMR) luits giveriff Mr.gg.lgg,Tjl gy.gg.tegM and XN hT 5171N !'
for MF 9xt-2 and
~BTEiE"Twel bcth do'dl coetinse to assure that W MC 10 CFR criteria are set50.46 fer l
sngle leep ooeration.
fsel fer. .. erded burnup cut to 39 GW4/MTU.The critsrts is also'ast for AM Ex0 j
P i
f i i.
II' Gg
), ! , [
n b '
f 1
i I!
l
.. {-
, t 4
StM tn:ut'&A .
...'. 6-12. #
. . . . . . . . . r u32,112:10r'.x mn : WOOMtALE Wish !
ig 31123441 9 cenewsnTH c: son;s e i
i l
5 ANT 8s-331 j
Pepe t l 3.0 Jt7 PUNP SWA Ecc5 EYMWATION MX)EL .
{
1 l 3.1 LotA DeseMet te
{
A less.cf-ecolant accident (LOCA) is deffnta is a hypothetteil rnture of g the reseter coolant syste's piping, up te and includtag the double-ended rupure of the largest pipe in the reactor coolant system or of arw line I cor,r,etted to that sylts* UD to the first closed valve.
In the un1ttely event g
a LOCA eccurs in a Dresden reseter, the reitter cositat system inventory less 1 would result in a high cor,ttir.nent drywell pressure and reduced reatter vessel pressure. The high drywell pressure er low reacter vesse) level eeincident vtith befogs the withcoetant low reactor pessure provide a LPCl/LPC8 initiating 119681 j
1ajection systems 'inte operation to limit the accident certtauences. ,
I g During the early phast of the LtY.A deprgfluritation trnntitet, core ecolleg is provided by the amiting coolant inventory.
le the latter sttee of system deressurisat{on and after depressurtratica has bete achieved, the I available Icw pressure cars spray te.d low pressure coolant infection systems ,
I provice tore enoling er4 supply Ifquid to refill the 1Wtr pytter, of ths ;
I reacter vases) and reficod the core. The reflood process providet sufficient hast removal to torninate the core temperature trustent. ,
h I' 3,2 UNNR Aro14 eat te to Dnsfer l'ntta f aN! 2 i
l The ExtK/tVR ECC3 Evtluatten Model codes were used for the LOC calculatfors for the Dreseen reactcr.
R0XIZ 0) IlfLAX,(II FLEX,N and KlXY/8ULGE2.W'38) The EXDt/8WR codes consist of I '
The tritial stored seerty and ftsstse gas release calculattens for 9:C-2 feel vs:
I takee from that previously esiculated(!) with the A00 tit code. The systen LOCA depressartratine fran the time of the tirent to the time Aen the g
s.
los pressure coolant injection (LPCI) fills the ditcherpe side af the intett i Icep with tubetoled 11guld was cateulated using the RELAX code. l This time l l occure after the time of rated IDW pretture core spray (LPt!) flew witich the RttAX ccda also caleviated.
The Fit! code was used to conoute the systes I ,
l cc.12.s2 12:cs FM 7C5 4
- e ;
M I BY:Of&A 44;$ 7601:# 6/11 6- ,-12.-32.;1.
. 2.:.10!'y ; g)y.y0Vt.AL'111 till50Y i s1226442144 ccPr0NWEALTH Ec EcM8 'l !
r i
~
ANr 89 131 '
Fage i depressLrtsation from the time the intact loop was filled with subcocted Itc,uid and to calculate tes system refill and time of reflood when significtnt entrainmant securs at the core midplana during the core reflood.nreca**- n-WKf/8VLCEX code was cred to cesputs the themal transt6nt it the midplane of '
the hot or saxima power asses @ly using initial condittens tatan from previews cales1&ttens(I) using R00 Elf and system boundary conditions from RELAXj FLEX. '
HUXY/BL'LtEI slas computed clad twellleg and rupture and the extent of estal vater reaction.
The liELAI systes blowdown calculation detsmired the reactor system tishavtor dartnp the initial pcrtion of the system depresskritation transtant.
Ths R[LAI systan blewdewn nadalization ta shown in Figure 3.1.
A separata AILAX/MCT (NM[L clicalltion was used to calculate the cligding to coolant f ,
tatt transfer coefficient anc coolant thermodynnic properties for the saxtr.um pmr fuel assembly.
This calculation considers ore fuel assedly with tim-j danndent kndary conditions froc the PILAX tystan blowdown essults being apetted for the reacter vessel spper and lever planum volums.
The RELAX /MT g' l DimEL nodaltzation is given in figure 3.2. '
The RELAI systen blowdown results also provice initial ecoditten inpat at the tie.s tra disetarse side g, ,
5 :
of the intact loop is ft11st with subcDoled 11ald by the LPCI for the FLEX reft 11/rer1 cod cateulation, !I I
h The FLEX systers refit 1/reflood entlysts calculated the later portion cf l
1 i
the systes depressurizatics, tsactor vossal tent p1stsu refl11, cars nMood and the tire at which the rartooding 11cvid is tatrair.sd to the saziaus power '
plane in the core (tise of het neds reflood). ;
The time of hot no:!s inflood vns an input parameter to the hattup cllculation. !
Figure 3.3 gives the epdaltsation used for the FL(1 code calculations.
Because LPCI seteccled g; -
vater is injected into the lower plenge, the total phase separation model was esed in the lower planus in the FL(X cateutetten. g The WIf/BJ' LEEX hettup enlculation used calculated paramstera from ADDEXfW (fwel stored energy and fission gas release), RELAI (time of rate I spray, decay power, heat transfer coeffletents, and coplant conditiers), and 06.12.92 12:04 W PCB ,
\
_-_.____----A-
. l g y gy. m ; ti-12-32 .;12..:.11t'M.
- COMMM NY 312281411&* CDEOWMTH EMSCN:s I N W 89-131 Page i FLEX and (time the percent of het neds reflood) tp dattraina the peak clad tag j oxidatfo't of the cladding. erature(PCT)
Ccnsistett with prior LOCA analysts for tas Dresden Units, a syrmstrfc canter.psakad axfa1 pow vs: used.
The testo calculatica van performed at a burnup of 8 CWd I exposure has the highest pistar perar in the crtginal ar,alysts,U) a/HT14 this nd thus is appropriate fattures, for this calculatien to deterstes relatfre effect of systeet i
j The Dresden reactor single failure dissal guarater LOCA analysts performed parts assuring an entire cers of MF .txt 2 fuel astelbifat throvsheut of the smalysis.
given in Table 3.1 and Rafarenes II,Drescan raatter systam data used in this a ,
I I
I l l
1 l
l.
y 0
. a i
l I b
'I
{
e t 12. 9 2 12:04 PM P07 l
)
- o.
1 44;$ 7t361;# B/11 5tM t$Y:Qr&A I . g'.3 y..J.2.
.. .: 12. .: 4l .l PM. ~WNMEO#
3123548214* CW40WECH ED! ION:t 4 i
B ANF 89 131 98ge 1 g
g TAILE 3.1 CMsotN REACTOR lifITIAL to@lflons l
t Primary Heat Output NW 3577.5*
Total Reactor System Voluss ft3
. 12tso.33 Total Roseter Flcw Rats, Ib/hr 96.0 X 100 **
Active Cors F1pW Rgte, lb/hr 88.08 x 10 8 McMns1 Reactor Systen Pressure l (upper plenurt) psta 3,c17."
core Inlet Inthalpy, 8tu/lb g 528.3 "
Racirculation lecp Flow Rats, Ib/hr 17,11 x }c6e.
Itsar,riorRats,lb/hr 3.55 x ID8
- Tiedwater flew Atts, Ib/hr f.95 r ID Es Rate Rac1rculatten Pumo l<ssd. ft 570.
Rated Rectrctistfon Puy spass, rps l 1.870 Masnt of ]rartin,1be.ft I/rac 10,953.
Recirculation suction Pip 61.D., te.
!5.78-beirculation Dissharge Pfpe I.D., In.
25.46 hel Assembly Rod Diesoter, in.*
0.434 Fuel Assesb1 l, 7 Rod Pit 'h it."* g,g73 Attfre Cors Height, sn.***
445.24 I
- 1021 of rated power.
- At 100% of rated flow.
- "AhF $d Fuel paramettra .
l-
- . . . ... l' Ot.12.92 12:04 ru FOB t
4
StM BY:UP&A : 6-12-152 :12:llPM ; CMOMt.ALR1 QW.b 443 7561;* 11/11 '
I ~ ~ ~ ** * " i 1122su2144 c;mwgg g f
ANT 89 231 l
Pope !!
4.0 MALY3!! RERlLTS A LOCA Ecc5 Itatting besak cales1stics was perf ened reactor itith a single fattere of a Ciessi er. genant for a Drstaan g break spectrum antlysit for the RWR ,
c 3 nietor (2) whi h The approved '
4pp1tes to the Dresden break af the rectreulatten pepe .gv1110ttne 1.0 (1.0 OfG/P$).
suction pipe with a '
, tal t fuel were parforzad for this Itniting reacter LOCA with bTM LOCA I t diesti ganantor with the reducsd LPL. fisw reak for a single fallsra eP I
Avsenge core blowdwn and reflood celeklaticas were sid core of ANT fr9 2 fot).
I LOCA analysis was dens at full power and full ficw (1 .
, the.
vary little limiting sieg1ovsr the range of allowed f1w eparatio . LotA result:III failure is characteristic of the ens. rea:ter since the systemfatture wcrst applie' future cyc)ts unless significant s in SMsantchanges .and in rautora !
co>ditions, or LOCA reinted reacte* setpotr.ts are madesystem hardwin, optr CalculatM event itas results reak diesel gerarator for hith thefailure lirittiv of a b I ars 91 Yen in Table a.1. fystes blowd x p*ssented ih FigJreg 4.1 thrcugh 4.11. results are l to 80 stecess; no !
tins.
The syntes blowom rthits are shewn 4.24. systes refill and nflood results are given in F ,
These syntam conditicas are used as boundary conditt rough
- I cliculation CHA$tt at a burnep condition of I CWd/MTlJ. ens for the heattp eticulation are given in poilts fu a RELWH0T terprature equit as calculated by Hun /Btttffigures 4.t3 Thtthrev) clnd 4.17.
tN of het node tsflood used is the nettup calcultti.EX 541n7.'41102.
1'ha ts shown in Figu% 4.14 en is from the nfloed Na~
=
. hI C f., 3 2. 9 2 12:04.FM PC9
, l
ti StNT BY:UP&. A . .- -.. . . ....... . i . 6. .-12..32..
-- - . ;.12..: 1. 2PM, ; COMMOM,L t,A,L,1,,H,3.,ll
, S.O,M, g , g ,443 g 7661**10/11 i
g$.L! T
,. p MF 89131
.a $
=
1 The PCT restit frca the haatup calculation for staw1 sting a single failure of a diesel geroretor wasg F lower than the KT from the equivalent h
analysis assetrig strgle failure of the LPCI n1re to open. m saatsve sotal-water resetton far any rod in the hot chenrel deterstnad from the hettup calculation f6r & diesel ger.aratar failure was .26 lower thar, the metal-water
( !
'i resetton from the equivsfont LPCI velvs fail're e analysis. The conclusion is that svan with the reduced trCI flow, the worst single fatture for LOCA Iti the Dresden plants rentins tre failurs of the LK! valve te o on. These LOCA.ECCs
[
results are in codom.tnce to the U.E. NAC 10 CFR 50.480 criteria, e i 4
1 8l .
E f ;
I
).
1 i L ,
F :
g:
O.
C6.12.92 12:04 TM P10 e
+
~
4 M T isY:QP&A ; ti-12-32 ;12:12t'N : CUXYOMtAL111 tUl509 443 70til;*11/11 <
"*'-a i
[
<*'st 3 ggggp3 t I-
' i ARF *89-131 f89013 4.1 DAIMIN skt.2 LIMIIINS SRIAK EVDff TIpets 'i
{ g . * '
Start Tifw fsael - i t
Initiate Ersak
)
Faer.%ater Flow stops Steam F10w itsps 0.53
(
S.05 LW Lcw Mfattre Love)
Jet. Pumps Do;cygr 4.5 7.6 Recirculation Pipe Dnggyges 1 10.7
- lower Pierum Flashes WK2 Flow Start:1 11.t 14.s LK1 Fin starts
.) LR s Starts Kated LPCS 60.1 g
1- 57.7 ;
D*Pressurftat1Ch Ends '
t
$ttrt of Refjesg 128.0 .
Tire of Hot Nede Aerloog 134.8 i
236, l Fett Cind Terparattre Aesthed -!-
238 .
I IIDW EIIDWd f2 RA4]fgfg '
4 .
I I
s ,' i
- i i
. i
. f kf
- 06. 12. 92 12:C4 PM F11 i
i t t
M T tsY:QP&A : ti- U-32 :10:18AM ; WN.nOP_AL'!!1 tDISO.b 443 7661:# 1/10
, Ar t a ' 32 ei LO =>@ LA1. nut' p,
%hl -
f'VS '
SIEMENS gM 8 of p*p*
- f June 2,1992 po,,.it' beand fax transmittal memo 7671 YUF;137:92 3.
gg, g rr. gg gg
" " 0 AA jVfA
- ggo '"*
72.P6 --
ru n y.g,p3ff,)
Dr. h J. Ch1n Nue: sat Fuet Servioss Commonwealth Edison Company Room 900 Edison Bldg.
? O. Box 767 ff .q ,
CNeago,IL 60690
- s 7 Q Deat Dr. CNn: ,f
/
Subject:
,/ Drescen Unh 2 Single Fa!ure Diesel Generator LOCA With No Recircufation isolation Valve O!ceure Technical Rev'tw
Reference:
Lener, Ye t).' Freak to R J Chin,'Dresden Un!! 2 Single Fallure Diesel Generator s
70CA with No Recirculation DIsenarge Va've Closure,' YUF:057:92, March 18, 1992. ,
A technical review has been completed of the Dresden Unit 2 Single Fallure Olesel Generator LOCA Witneut Recirculation Discharge Valve Clorure which was reported to Ceco in the Reference. The rey!ew covered verifica1!on.cf all appropr'ste input data and changes made to that data to affect the ant'ys?t. )J#fe71fi65fidn{ncluded the changes needed to affect the maximum average planar heat g4neration rate (10.y4 kw/h), LPCI flow (9890 gpm), and LPCS flow (4600 ppm).,D,ueJo the rei;twAILobangemes made to the Ana'ys!s_which IDoreasedfhe peak clad temperature to 21951 and max'wmlocat metal water reaction to 5.32%. These tesuts sil!! remain within the 10
- 8 crt
- eria of 2200'F and 17%.
The original analysts was done WM &n aH SNP 9x9 2 core; the re!evant Diesden 2 Cycle 13 core loc!udes 104 SNP 6xS fuel bundres. The orldinal anays!: was determ'ned to be unaffectsd when taking into account the 184 bundles of SNP h6 fuel.
Thte work was performed at the request of C Co. An involes will be issued in July for the add:tional work. <
Very truly yours, Y
9 (Mrs.) Ude!! Freak f/f,kf Contract Administrator pe$lm D. R Zahakayfo (CECO)
RAPlFAX _41.1- t c:
Slemens Nuclear Power Corporation EA0 yp s Co w re o*ces
5t.NI' 57:Uf&A ; 6- 6-32 ;10'16AM ; LM.MOMtAl.!!M tDl509 44E 7661:# 2/15 y - -a
~
- m. " M SIEMENS s:
I
(,M March 18,1972 YUF;057:92 p#
Dr. R J. Chin Nucleat Fuel Services 0
Y Commonwealtn Edison Comps.ny Room 900 Edison Bldg. l P. O. Box 787 Chicago,IL 60690 Deat Dr. Chin:
Sub>ct: Orosden Unit 2 Single Fa!!ure Diesel Generator LOCA w'th No Reoirculation Discharge Va!ve Closura 7
The attachment reports the resutts of a e!ngle fa!!ure diesel generator LOCA analysis without closing the rectreulation discharge valve for the Dresden Unit 2 reactor. This anatysis was requested by CECO for a past operating cond!!'or%. .was disccvered wh!!e attempting to close a roo:rculation discharge valve during Cycje 13 operaTCn. The discharge va!ve would not close '
at the reactor operat[na prejsare.
M performed and provided on an accelerated bas!s at the request cf Mafk Wagner. The anafys!: has not undergone normal SNP qua!!ty assurance review; thus, the resu '
are provided for inf0rmation and should not be used in any fortr,1! Ilcensing applicatto -
Thus werfwse per'ormed under CECO Purchase Order Number o00130. An involos will be Submittod in mid April.
Very truly yours,
+
(Mrs.) Udell Fresk Contract Administrator tkn Attachment c: D. R Zahakaylo (Ceco)
RAPlfAL tVV~
PAGE._/Of Siemens Nuclear Power Corporation come o me
- - . . . . . . . em . . e. e.a... wg ce*N A m *rd D % A', u m Fu @ ($'l-4448
~ 1
- ATTACHMENT IMPACT OF NONCLOSURE OF THE RECIRCULATION ISOLATION VALVE DRESDEN UNTT 2 LOCA ANALYSES
SUMMARY
Commonwea'th Edison Company (Ceco.) has fequested G!emens Nuolear Power Corporation (SNP) to evaluate the licensing impact on the Dresden loss of. coolant accident (LOCA) a of a past operating condition of the Dresden Unit 2 reactor. While attempting to restart a teoircuistion pump in August 1991 during Cyc's 13 operation the reactor's d:acharge isolation va!ve in a racirculation line (RDV) was unable to be closed at the operating pressure. The non.
clesute of the RDVs diroctly affects the ECCS low preasure oootantinjection (LPCI) system during I a LOCA sinceMCI flow is Inleeted diroctly Into the intact toclrculation loop. Without olosjag c
.the RDV mcs1 of the LPcLftew woutdao back throug hmump, out the suction intet of t_h, jniactjectreulst!on loop, and then out the break in the brekiirrTe6icujation,loopjnjt_Ilmitingj .
I break LOCAM.
~The nonclesure condition has no effect on the lim' ting LOCA analysis of record" because LOCA has a fa!!ure of the LPCIInjection valvo and an added fa'fure of the RDV therefore has no effect on h. The added failure does impact the LOCA with an assumed single failure of diesel generator (SFDG)" since the LPCI system operates in this LOCA. An 8na!ysis of the SFDG LOCA with an added tal ure of the RDVin the intact loop has been performed to verify that the SNP fuel remabs in conformance to the U.S. NRC 10 CFR 50AB cr toria during the LOCA.
The Ilmiting LOCA break for Dresden Units 2 and 3 hts been reana'yred assuming fa' lure of .
d:ese! generator and ROV with Cycts 13 speci'ic data. The peak c'ad temperature (PCT) calculated is 2163'F and a maximum local metal w6ter reaction of 4.81% which mew int 1 50.46 criterla. Therefcte,it can be concluded that even with an added failure of in+ RDV no analyzed LOCA's for the Dresden Unit 2 reactor exceed U.S. NRC limbs during the Janusty to )
August 1991 time period ci Cycle 13 operation.
i ANALYSIS DESCRIPTION l The EXEM.BWR oodes RELAX, FLEX, and HUXY/BULGEX were used in the ansfysis performed.
The RELAX code was used to calculate the system response from the time of the brosk to the '
time when the Lntact loop was f!!!ed with water by the LFCI system. The FLEX code was used to calcu' ate the system depressurization from the time the intact loop was f;!!ed with subcooled water unt:1 significant entralnment occurs at the core midplane during the core re$ood process.
A RELAX hot channel analys's was performed to determine flutd condittons in the limking assembly. Heatup cateutstions were performed to determine PCT using the HUXY computer-code. ,
l nog _Jkorf.'
l
5tM BUur&A ; b- 5-32 ;10:17AM LWYOMtAllH tl)lbOM 443 7661:# 4/15 e *:c. ..;u n u u. }Qy y Zgf Gp , .
/j a., Jaf 'fi n. ub 2/ Y -c "' ~-
c ATTACHMENr 2 The LOCA calcu'ation was performed at full-power and full Sow condit ns or a full core of SNP ex9 2 fuel The LOCA ana!ysts wu analyzed with the some plant da kd in References 1 and 2 except for piant measured low pressure coolant injection flow M - p,m" and low pressure of 5 n the water rods in the core spray of 4600 gpm", a U.S.
h9 2 fuel during the spray oooiing phase, '
NRC approve 10,64 kW spraf5bstf:o's or dximum ,%average p!anar lineat heat generation rate obtained in the resc( e uary to August 1991 time period.
ANALYSIS RESULTS
\
Calculated event timos from the BFDG LOCA are provided in Table 1. RELAX system and hot che.nnel bicwdown results ate prosented in Figuree 1 through 14. Figure 7 shows the intact loop -
4 Jet pump drive flow out to ntnoty-two eeconds w!th a statie flow into the jet pumps of 360 !b/seo. 7 Wah a LPCI flow of 1340 lb/see at that time into the intact loop,
- f j.t t the jet nozzels and 73% out the suction in!st.
D Results from the RELAX / HOT CHANNEL ca'culation are given in Figures 17 through 19. The resuits from the RELAX ans.!yses are Used as boundary cond!tlons to perform a hestup ana'ys!s at the lim! ting exposure of 5 GWd/MTU. System refi!! and reficod results are shown in Figures 15 and 15. The t!me of hot node reflood used In the heatup analysts is from the FLEX reflood calcu!ation when significant entrainment occurs at the core midp!ane. Cladding temperatura calcu!ated in the hestup ensJysis (HUXY) Is shown in Figure 20. ,
The PCT calcu'ated for the SFDG LOCA with no ROV closu trotahwatst res: tion for the SFDG anahs s is 4.81%. The results for the SFDG LOCA with no 7 , 3 RDV cicsgro demonstrate that no 10 CFR 50.45 criteria are viotated for the Dresden Unh 2 reactor F during the January to August 1991 Cycle 13 operation.
[
i p
REFERENCES -
< m
-1. "Dresden Units 2 and 3 LOCA-ECCS Ani'ysis MAPLHGR Results for ANF h9fut!? ANF4519L Advanced Nu:!sar Fuels Corporation, De0 amber 1955. i e mm- :
- 2. 'Dresden Unita 2 and 3 LOCA-ECCS Analysis 6!ngle Failure Diesel Generator,' ANF49131. ,
Advanced Nuetear Fus!s Corporation,gustg.
- 3. Letter Udell Fresk to R. J. Chin,' Analysts to Support increased Recirculation Discharge VsJve Closure Time at Dresden Station,' YUF:250,91, September 27,1991.
- 4. Dresden Operating Survel!!anca,1500 5, Rev.10, Dated 11/25/91. l I
- 5. Diesden Operating Surveillance, i 4001, Rev.18, Dated 124R2.
- 6. 'Susquehanna LOCA ECCS Ana?ysis MAPLHOR Resuits for exe Fuel,' XN-NF4643. Econ Nue!sar Corr.pny, Inc., May 19M. j
- 7. Telephone Pete Weggeman (CE0o.) to D. J. Braun (SNP), Febnsary 7,1992.
RAPirAX b l
m . . . . -
. ATTACHMENT 3 TABLE 1 DRESDEN LOCA UMmNG BREAK EVENTTIMES ht.d 3!11.IHd Start 0.00 Wtlate Break 0.05 Feedwater Flow 8 tops 0.55 -
8 team Flow Stops 5.05 Low Low Mixture Level 4A Jet Pumps Uncover 7.5 Recirculation Pips Uncovers 10.7 Lower Plenum Fisshes 11.7
~
HFCI Flow Staris 1 14.4-LPCI Flow Starts 39.0 LPCS Stans 30.9 Rated LPCS 56.3
. Start of Reflood 198.0 ,
Time of Hot Node Reflood 226.0 Peak Ctad Temperature Reached 226.0 s
1 No HPCI fiow allcwed in analysis i
l M
. -i RAfifAX - R W !
PAGEIOf -
4 i-n -e--
- , -muu - uuw eu,v w,o.
ATTACHMENT 4 FIGUME 1 SYSTEM PRESSURE Detsm 2 1.0 cro/Ps rou. sra eetc _FD0 NO RDV CtC$'AC ___
g _
i 1
t E
t g
I
~
l, .
t .
f ,
f ..
I f f f a0 E @ H N SN CN te lN TIE ISCC) ,
FIGURE 2 TOTAL BREAK FLOW encSetw. 1.e eram _ ris.t sig eenc _ arco wa nov easuet g
Mg .
l- "
\
lig -
, , .- =W , , ,
m m e = = in m se II T I S )
- RAP 1 FAX nu l.W.E. ;
~ 5tNI' tlY:Ur&.5 ; 6- U-32 :10:16AM ; COMMOMtAL1M tDISO.N-443 7561;* 7/15 ~
- ATTACHMENT 5 t
FIGURE 3 CORE INLET FLOW en*ocN-2 s.o.orcers ruu. ers ccee sros wo ev etesunc g _ ,
- t p
li -
Y \
ji t v -
t 4
h . f f n f f b = = = = in is is in-TINC (SCCI
?
F;GURE 4 CORE OUTLET FLOW g cacS0cN-2 1.0 eta /Ps riu sxt c Rr, sros No Rev ctosm i o
l .
si .
g g
i -
=' - 1
+
I , _. . . . . i Pa a e = a = := e. in T!!C ($CCI RAPlFAX PAGEIOFM l
n up,wg ,,,,,,,,,..
,,it m u m ;usmte u p ,
@m em l
^
ATTACHMENT 6
\
FIGURE 5 HOT CHANNEL INLET FLOW l
oncsco a m t.oes/rs are .ect to.s4 wurt sm mRW-_ l j
. . e i a
, i
-s
- ' l
.A
^
hn I
k .
e
, , ,_ , . - , i h a a a e a d M N TIPC ISCC)
FJGURE 6 HOT CHANNEL OUTLET FLOW ORCSOCW-) HC !.CC2/Ps gr0 CCRC 10.84 K .WTT SM W RW
, . . > =
g , ,
9
+R g ,
r
=
G
= =~ .
_n n
- 9 *
- n
% . . . . = = = =
TIPC t$tm RAPlFAX not 7 erd _b
ATTACHMENT 7 FIGURE 7 INTACT LOOP JET PUMP DalVE FLOW g
eatseces q.e ecses rutt ers eene, vos y ev ewstm 4
E l -
lg hg .
1 -
ist -
b
'i .
t . . . .. . .. .
TINC ,5CCI FIGURE 6 INTACT LOOP JET PUMP SUCTION FLOW cats 3cN-2 1.0 ets/Ps nn vr9 coRc v;s No Rev car.at _
g ,
E[
li -
e -
eg .
E E -
si a
t . ,s,. -
RAPifAXkNN noe forg.
S
- ATTACHMENT 4 NGUME 9 LOW PRESSURE COOLANT INJECTION FLOW g
catseces !.c etws nn pro coat, erce wo e ccem
~
fE i
.g .
8 Eg -
d N 4 N N lE lE $O M T!?1C ISCCI RGURE 10 LOW PRES $URE CORE SPRAY FLOW w setw-g 7.o ecurs ruu. srs cent stoo_ uo nov etes s (I
g _
Eg -
, i , _
se N # M N in IN le IN RAPifAX IkY PAGE 1 0f M .
M.
._ o, ..
I
- 'r ATTACHMENT FIGURE 11 LOWER 00WNCOMER LIQUID MASS cacsxw-2 1.0 Ts/P5 ' ru.L tr100tC W ,
y D YN --
3l ~
l '
~
"l .
- g. :
Ig .
l
, , e
- y y a a se is 5# 8#
TIPC iSCCI l
Figure 11 L
FIGURE 12 LOWER DOWNCOMER MIXTURE LEVEL ;
OtCSCO-2 1.0 N G/PS FV.1 . trl CCRC . SPDG WO ROY CLC9URC -_
g
$A '
E - ;
.s .
c.
4 8 . !
I= .,
]
. i
_v
- _ y , , e us as le t# .
)
T!!C 13CCI RAPIFAX INE- !
not/.d.or/f l
, a- w--- m, l
10
- ATTACHMENT i 1
4 FIGURE 13 UPPER PLENUM LIQUID MASS oncterw-a i.e ecus ruu. ers conc, stoo wo ev eteiper g
I I} l 3 * -
(I -h
=
EcE el, .
f f f
f a f t
$ \& hN $
N N N Y N f!it 15GI FIGURE 14 LOWER PLENUM LIQUID MASS c*r*ctN-2 1.o otsg riu exo ecRt, sres N,o Roy cLefurz __
l E -
il n ea li .
Ig .
f 9 I 9 D
t f MC 680 N in SN 8 e N IIPIC ISCCI
~
PAGE 2 0F d [ .
3 11 ATTACHMENT l
FIGURE 15 REFILL /REFLOOD LOWER PLENUM MIXTURE LEVEL rs --
gu. 1 i
E 1- l g 9-
)
l l
j we soo 2e sao
' es ics iso we oc M NIN l.
FIGURE 16 REFILL /REFLOOD RELATIVE CORE MIDPLANE ENTRAINMEN I i
Li$-
i.
\
..n.
i J
i ._
,1 / !,
4,. ...
.m rw nu/ dor /E
~ 12
~
ATTACHMENT ;
FIGUME 17 HOT CHANNEL HEAT TRANSFER COEFFICIENT
_C*fS D -2 SC 1.DCU/N
. . SY9.CORC. 10 H KWr7 SPC8 WD ROV g5-E! .
I-s W
W h :
l :
Y .
M i
= w a a = a
'k is a w z tsccs FIGURE 18 HOT CHANNEL CENTER VOLUME QUALITY 4 *-
ontsxw-2 H: 1.CC9./n ersene spH my SN ,
2_ . .
f h0 l s .
t ,
l lo
\ . l d, . \
k !
h '
y
)
wa
\
r l w
- w. a a
e u
a a
a TItt 19201 RAPI Y PAGE Of/1 ,
1 I
-, ms ,eu: arm tRFelmMMP.
i l
.. . 13 ;
' ATTACHM.ENT l 0
F10URE 19 HOT CHANNEL CENTER VOLUME COOLANT TEMPERATUMI 1
ORES"CM-2 NC t.DCS/P1 sit CORE 10.84 KwrT $P00 W ROV ,
cg -
$ ' ~
l .
g l
g -
g e
Wg -
to to as It to a a e se
- T!?1C IRC)
FIGURE 20 HOT ASSEMBLY HEATUP RESULTS crc 50CN 2 WXY Met *LNGR.t c.s e $ ScnfN EX9 CORC SPD9 NO RDV
[7a3 f
4 eerae 8 it il 18 la 141818 tr j3 il t$ st as as 38 30 $4 4 la is 38 M 8F se 3030 5 15 30 m 18 M $3 N N 6 4ealF M Maf9438 F ll M N 38 ff 40 48 sa 0 SB E3 35 h 90 4 0 44 e it 64 te M Je 43 se et.
1 .
g.
4M -
w c; /
hN M s T14:838C8 g
ens evtute Y RAPlFAX PAGE80/[
O 30STON EDISCN COMPANY - NUCLEAJL 25 BRAINTREE IILL OFFICE PARK 3RAINTRKE, MA 021G4 TELECOPIER COVER LETTER 4th Floor Fax: (617) 449-8764 Verify: (617) 849-8700 3rd Floor Fax: (617) 849-8965 Verify: (617) 849 8964 Please deliver the following pages to:
NAME: b M N TELECOPY # h O/) G Y .78 9 @
FIPR: N C--
CITY: / Ma M C- -
FRCX:
NAME: [/l/ M b )
FIRM- 3rCO i
C:TY: Entneree (6/7) 8t'9-f'?J9 l To:a1 number of pages including cover letter l
DATE: / / 97 \
/
l T1ME: 3 $ i IF YOU DO NOT MOEIVE ALL TIE PAGES, i i
PLTASE f*1Y.L BACK AS BOOM AS PORSTRLE.
1 i
l
._________________l
l NRC WATER LEVEL DUESTIONS Does Pilgrim have a timer.in the LPCI system which must be timed out
- 1) l before flow can be diverted to containment spray? What is the time? '
Why is it in the system, and what components does the timer control?
No timer is installed in the Pilgrim LPCI logic which prevents diversion of flow to containment spray. If the containment spray valves are opened, water level below 2/3 core coverage or drywell pressure below I l
psig will signal automatic closure.
However, the Pilgrim LPCI logic does have a 5 minute seal-in timer which prevents the open injection valve from being throttled to an intermediate position. This ensures that full flow is available for >
injection into the selected loop. Cancellation of the timer signals '
after 5 minutes allows the operator to divert the water for other post +
accident purposes.
- 2) The current Pilgrim design flow of containment spray is 25% of the flow of the RHR pump. Is this the original design flow, and if not, what was the original design flow, and why was it changed?
The original design containment spray flow condition was 5000 gpm (4750 gpm to drywell sprays and 250 gpm to torus sprays). 5000 gpm was maintained to ensure full flow through the RHR heat exchanger and therefore full containment cooling capability.
As a part of the PNPS Safety Enhancement Program (SEP), a fire water to RHR system crosstie was provided in the event of loss of all RHR pumps.
Using the fire water system to support core and containment cooling required the plugging of 6 of every 7 spray nozzles to create optimum spray patterns. In doing so, the containment spray flows from the RHR pump would be dramatically reduced. Analysis by GE demonstrated that spray flows as low as 300 gpm would still perform necessary containment cooling and depressurization functions. This flow rate can be attained from the fire water or RHR pumps. To ensure that 5000 gpm continues to flow through the RHR heat exchangers, operators must direct flow to the reactor and/or torus while containment sprays are being used.
- 3) On page 5 of the operability determination, appendix K analyses are discussed that resulted in 182' deg F peak clad temperature (PCT). Is it correct that this case credits 2 core spray pump and no LPCI flow?
What is the apaendix K PCT calculated for the battery failure case? For the cases in w11ch LPCI flow is credited.which of these results in the highest PCT. For that case, what systems are injecting, and what is the '
resultant calculated PCT 7 The casa discussed on page 5 of the operability determination is the-Appendix K calculation with single failure of the LPCI injection valve.
With this failure, all LPCI flow is diverted through' selected injection paths that are dictated by the LPCI loop selection logic system. This system would divert all LPCI flow towards the unbroken recirculation loop through a single injection valve. Since this valve is the assumed ,
a i
single failure, no LPCI flow reaches the vessel. The analysis that ,
calculated 1821 of PCT assumed only HPCI, 2 LPCS, and ADS are available for ECCS. For large recirculation line breaks, HPCI is negligible and ADS is redundant. Hence, only 2 LPCS pumps were credited for this calculation. :
For the Battery Failure Case, the calculated Appendix K PCT is 16940 F.
This case, for which 2 LPCI pumps are available, is lower than the LPCI injection valve failure case. The battery failure case is the highest' calculated PCT for Appendix K calculations in which LPCI flow is credited or available. For this case, 1 LPCS, 2 LPCI and ADS are available.
All the above information is avai_lable in NEDC-31425, a document ,
provided earlier to the NRC.
- 4) Also on page 5, it is stated that " Operator actions to divert LPCI flow would not occur until the core had been reflooded ...," Provide more ,
specific information regarding when, by emergency operating procedures, operators are instructed to initiate containment spray? This information should include the containment temperature /sressure vs time plots for the limiting containment response event, and tie time that the operator would see a signal directing initiation of containment spray?
Operator actions to divert LPCI flow to containment spray are described -
in E0P-03 " Primary Containment Control". The Primary Containment Control section will direct operators to start torus sprays when torus ,
aressure is greater than 2.5 psig and prior to exceeding the torus :
>ottom pressure of 16 psig. The icon to the right of this step directs operators to assure adequate core cooling prior to spray initiation.
When torus bottom pressure exceeds 16 psig operators are directed to saray the drywell using those pumps not required for core cooling. If tie containment integrity is threatened by pressure approaching the
" Primary Containment Pressure Limit", operators are directed to vent the torus. If drywell/ torus pressure exceeds the pressure limits then regardless of core cooling requirements the drywell is sprayed to preserve the containment pressure boundary.
The actions described above are for events beyond the PNPS design ',
criteria, events requiring multiple failures for both the containment '
design limits and core cooling capability to be compromised. The event-used in the FSAR for containment analysis is the LOCA inside the
- drywell. The analysis shows o neak drywell pressure of 45 psig which is
'well belcw the maximum allowable of 62 psig. The FSAR arbitrarily
- assumes drywell spray (containment cooling) starts 10 minutes into the
- event although it may be delayed as long as 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before the containment design limits are reached. After containment cooling is
- established pressure and temperature drop steadily towards the lower
' valve.
i t
'EOPs also djrect operators to spray containment when air temperature jexceeds 280 F. This temperature is quickly exceeded for liquid and
- steam line breaks. However, operators are directed not to initiate isprays at the expense of adequate core cooling. ,
i.
=
The E0P's will direct the operator to maintain adequate core cooling until the containment design parameters are about to be exceeded. At-that time preserving containment as a barrier to fission product release is a priority even at the expense of core coverage.
h b
i i
I I
e 9 s l
I I
w y v w-n p- w
SENT BY:0ECo ;12-23-92 ; 4:09FM : BRAINTREE 4;a ft.00R- :
1 1
BOSNm EDISON COKPANT - NUCLEAR >
25 ERAINTRER EILL OFFICE PARK BEAINTREE, MA 02184 TEt2 COPIER COVER LETTER (617) 849-8764 4th Floor Fax:
Verify: (617) 849-8700 -
3rd Floor Fax: (617) 649-8965 Verify: (617) 849-8964 Please deliver the following pages to:
NAME: 74 m {24/) 3 / ~2 2[0O TELECOPY :
FIRM:
MNb CITY: [.(JM M "~
FROM:
! NAME: /d l FIRM: [
CITY: /4 fh' ('(,( 7_-) B f f $ $ 3 f Total number of pages including cover letter:
DATE: /'//h3!I2_-
/
TIME: 41 t .
/
q
' ~
IF YOt" DO NOT RECEIVE ALL TEE PAGES, I i
,PLEASE CALL BACK AS SOON AB POSSIBLE.
w
, j
S B T 9 0,0 , _ , , , , , , , _ . , _ .
- 12-23-92 :'4:10PM t BRAINTREE dth FLOOR- ;a 2 l
3.3.1 OtjecQive of Desien Chance l' Tte new ontainment spray header nozzles are designed to f optimize the use of containment drywell spray under
! severe accident conditions and to provide the operator !
l wjthsigpificantlygreateroperationalcapabilitythanis permitted by previous Emergency Operating Procedures.
a i
ThedesignchangeforreplacementofthecontEinment spray header caps / nozzles is related to the following SEP plant changes:
1 3.4 Additional Sources of Hater for RPV Injection and Containment Spray
- 2. 3.5 Diesel Fire Pump for RPV Injection and -
- 3. 3.6 Diesel Fire Pump fuel Oil Transfer System 3.3.2 Desian Change Descrintion j l l This design change (Figure 3.3-1) replaces the 104 upper i and 104 ilower containment spray header nozzles inside the d'rywell.lThetorussprayheaderwillremainasis.
l I I Tpe replacement nozzles are identical to the existing except that the replacement 1-1/2*-7G-25 Fogjet riozzles have one open spray cap and six blanked off spray nozzles c'aps, whereas the existing 1-1/2"-7G-25 Fogjet nozzles haveallsevencapsopen. The new nozzles have the same classification ,as the existing nozzles; no system i boundaries are revised.
l Replacing the nozzles will reduce the capacity of the
- drywell!spraysystemforallcontainmentspraymodes
! during a design basis loss of coolant accident (LOCA) and i i design basis small steam line break. The reduced drywell! spray capacity will provide increased operational
- flexibility and permit containment spray operation over a wider range of containment temperature pressure conditions.
l }
Qalculations were performed to determine flows through the replacement nozzles for various severe accident and LOCA opprating modes. For a LOCA, the calculations-tonsidered one RHR pump supplying water to the upper
- Containment spray header nozzles. The results of the dalcula tions were used as input to an analysis of the RHR :
i :ontainnent spray flow which concluded that the calculated spray flow of 543 gpm would maintain the drywell. structural and atmospheric temperature's below the '
- design limits of 2810F and 3400F, respectivelj. '
.l i .
m 3 f r u/ /- CJ /d d C.-
I 7/e/e;2 : 1-fSO/ w
SENT BY:BECo ;12-73-92
- 4:10pM
- BRAINTREE 4th FLOOR * :: 3 .
dprayhfadernozzlaswillnotprev'entthecontainment 3
l spray systea from t.aintaining the design temperatures in the drytell durinT a LOCA and, therefore, maintaining the design basis of this mode of the RHR system.
i This change also has an effect on the small steam line break accident. The operator can use the containment l
. spray tb mitigate this accident as described in the
[SAR. The reduced flow will again minimize the possibility of damaging the containment structure by sudden decompression following the initiation of containhent spray. For this design basis accident the
- RHR het t exchangers must maintain design flow to assure adequate heat removal from the suppression pool. To
' ^- accomp115h this, the torus spray bypass line must be throttl ad to maintain design flow through the RHR Heat Exchangars.
l Thered)cedsprayflowalsoreducestheriskof containment structural damage from an inadvertent spray initiation into a hot dry atmosphere.
! I I i This improvement poses no new loads on the piping system i and, therefore, need only comply with the existing seismici design, including support design.
I 3.3.3 Desion ChanoLEvaluation l
3.3.3.1 Systems. Subsystems. Conconents Affected i The system that is directly affected by the l j change is the Residual Heat Removal System 1
- (RHRS).
i :
l ! The subsystems that are directly affected by l
the change are the drywell spray, the torus
- spray and the RHR suppression pool return valve. The torus spray flow will be affected
' ! only slightly, and the suppression pool return valve needs to be open during containment spray so that rated flow through the RHR heat
!. exchanger will be eaintained.
i I I The components that are directly affected by j the change are the RHR Drywell Spray Header caps. Six of the seven spray nozzles are blanked off.
The systems that are indirectly affected by the change are the torus-to-containment vacuum breaker system and the reactor i
building-to-torus vacuum breaker system. The
- response time of the vacuum breakers is
- affected by the reduced drywell spray flow.
SENT BY:BECo ;12-23-92 : 4:11PM : BRAINTREE 4tn FLOOR * *: A Residual Heat Removal System [
l .
, The RHR cools the suppression pool water and I ;
provides for containment spray cooling. It is !
used for a wide range of postulated LOCAs as l
- j well as HSIV closure, struck open relief valve, ;
and alternate shutdown events. Impact on i i previous safety analyses is limited to those i events that utilized the containment' spray mode !
, of the RHR operation (FSAR Figures 5.2-2 to i I
5.2-7). .
Drywell Soray and Torus Serav Subsystem l The Drywell Spray Cooling subsystem provides l
water to spray header systems located in the I
drywell and suppression chamber. Under
! post-accident conditions water pumped from the i suppression pool through the_ heat exchanger may i , be sprayed into the drywell and the suppression I :
l chamber to remove the energy associated with j the steam in these regions. The containment i ! spray is used for a wide range of LOCAs.
l Impact on previous safety analyses is limited
' to those events that utilized the containment 4
spray mode of the RHR operation (FSAR Figures 5.2-2 to 5.2-7).
, Containment Vacuum Breaker Systems ,
\
j jThesafetyfunctionofthevacuumbreakersis i ! to equalize the pressure a1nong the drywell, !
i i ! suppression chamber and reactor building so l l lthatthestructuralintegrityofthe ,
! j containment is maintained (FSAR Section
- 5.2-3.6). For accidents such as those i
- presented in FSAR Figures 5.2-2 through 5.2-7, I
' a reduced drywell spray will mean a lower rate of drywell depressurization, resulting in
. ! delayed opening of the vacuum breakers. This
] ldelayisincludedintheanalysesdescribed below and has no deleterious consequences. '
i i Eolential Effects on Safety Functions 3.3.3.3 Drywell Resconse The effectiveness of the reduced dr)vell spray
, in controlling drywell temperature for small :
break LOCAs needs to be evaluated.
it .
1 J. .
O
SENT BY'8Ecc ;12-23-92 ; A*11PM
- BRAINTREE 4% FLOOR * :: 5 l RHR Heat Errhancer Efficiency Because-ef the reduced drywell spray flow, the j heat removal capacity of the RRR heat exchanger
! when operating the RHR system in the l l containment spray mode needs to be evaluated.
i j .
3.3.3.4 Analysis of Effects on Safety Functions l'
, fffect on Drywell Resoonse ,
} i TheFSARcontainmentresponsewasre-analyzeg
.. ; for break sizes ranging from 0.02 to 0.5 ft.
!~ l assuming the reduced containment spray was
. - initiated 30 minutes after containment pressure i
reaches 10 psig. It was determined that a containment spray flow rate of 300 gpm is
, ! sufficient to reduce the airspace-temperature
! to below 281*F for all break sizes analyzed.
Holding airspace temperature below 281*F
- eliminates the driving force for the wall l temperature to exceed 281*F, the design l temperature of the containment liner.
l l l The containment spray flow with the proposed
' design, with one header operating, has been l calculated to be 543 gpm when the suppression
! I pool bypass valve (1001-36A, B) is open with
- , total RHR flow limited to 5000 gpm and 1150 gpm l
,lwhenthevalveisclosed. The containment
, , spray fror: one header will deliver sufficient flow to maintain the design temperatures in the I drywell during a LOCA.
. I l Effect on RHR Heat _Exchancer Efficiency i
When operating the containment spray, the operator will be instructed to open the RHR suppression pool bypass valve (1001-36A, B) so that rated flow through the RHR Heat exchanger i ! will be r,aintained. This ensures that the heat l lremovalcapabilitythroughtheheatexchanger will not be reduced, i
1 e e n
4
SENT BV:BECo ;12-23-92 ; 4:12PM : ORAINTREE 4%h FLOOR * :: 6 3.3.3.$ Des 1cn Chan_ce Evaluation Conclus90ns There art two potential safety issues arising l* -
from replacement of drywell spray header caps:
i I (1) effect on drywell responses and (2) effect t.
on RHR heat exchanger efficiency. These safety j ! issues were addressed by (1) performing 5
analyses to show that the reduced drywell spray flow will still maintain drywell atmospheric and structural temperature below thstr
- respective design limits and (2) req'uiring an operator action which will be incorporated into I'. the revised operating procedure,to keep the suppression pool bypass valve open while
., ___ operating in containment spray mode in order to maintain rated RHP. flow through the heat exchanger.
l l
l b
f a
s 4
t .
m
~
1
seur er:eeco l CONTAINME NT SPE AY
- 32-23-s2 : 4: 12n : esAiutsee 4tn rtous--
E
- 7 j
~
HEADER NOZZLES .-
Q
. - IG!UPE 3. .5 -1 Q~ - m -
T HE SIX OUT E F C AP c. ON THE SPP AY (' v/t c i
riJOZZL E S ARE PEPL AC E D WITH SIX L id -
B, L A N V. C A P S ! ,
- - k i
j d
, l .
=. - 2 eL ##
l l if u e rfd ! l t n 3
l 3 l l h -
x l
I fu y 41, F
'%C =
t **cw j __ go!
l l
l h i l - 1 , ,. ir l ,,
,. Q f\\ ;L. W $
' l\ '- A '
- 37J ,
Q ' \, m A r &,2 A,,/ / Y
/
i
'j bla@
(' )
- U CT**
cLe s.J petstr.u q IU
, ( i) 4'OfA. -.
I i l
- 1 I -
j f ~7G C5 f~tC,/C, ' MEZ'L[
1 l ! .
FD A' O.;~ TA t ou f _7G ZZ Ft>a./[OtzLA ,$CE NOU5MIAL i
ccucm -
i
- hCDClimO84 dPEAYING SYSTEM'S 00.
i % ,o ,, r-fa a w e F 6 /f Terzz.r ,,,,,,,,y,,,,,,,,,,,,,,,,, ,
i DA 8Y 4. A*, DWG. peo.
DAT1t .P_ - 2 4 '/7
-24A.
i l.