ML20023A452

Advises That Three Listed Accident Classes May Each Be More Frequent than Commission Safety Goal in C-E Designs That Lack Capability for Core Cooling Via HPI Injection & Venting of RCS Absence of Feedwater Replenishment

ML20023A452
Person / Time
Issue date:	01/29/1982
From:	Jerome Murphy, Rowsome F NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Speis T, Tedesco R Office of Nuclear Reactor Regulation
Shared Package
ML20023A453	List: ML20023A452 ML20024A192 ML20024A193
References
FOIA-83-168 NUDOCS 8203090638
Download: ML20023A452 (23)

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~ h, ' JM 2 f 1932 c ~ NE!!ORANCult FOR: Bob Tedesco, Assistent Director for Licensing Division of Licensing, itRR Themis Speis, Assistant Director for Reactor Safety Division of Syst:ns Integration, tiRR FRCli: Frank H. Rowseme, Deputy Director Division of Risk Analysis, RES Joseph A. Murphy it Reactor Risk Branch Division of Risk Analysis, RES ~~~ " ~

SUBJECT:

FEED Atl0 3LEED ISSUE;FOR CE AFFLICAtiTS

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We have perfor.ned a quick and dirty analysis of the risk implications of CE designs that lack a capability for core cooling via HPI injec. ion and deliberate ventihg of' the reactor coolant system, in the absence of feeduatar replenish:ent. e conclude that three classes of accidents may each he core frequent than the Cc= mission's safety goal of 10-4 core melts per reactor year or less, and that the total core melt frequency for such plants could be of the order of 10~3 per year or more. The three sequences are: 1. Transient and failure of all feedwater (not associated with loss of ' AC ;cuer) (TML). Loss of offsite power, one diesel failure disabling the :ctor driven Z. AFW train, and failure of the turbine-driven AFW train. @.% Has.Been Sent to PDR re 3FI(SD). 3. 'lery s,sall LCCA and fai g . po 30.$ )

3 :...aend : o fs* ' ewing upgrades :: thesa designs- .1

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.s T T d;.r- ._2_ ~ _;- . ~ _ _.. t ~ = 's The base case plant is assumed to be incapable of feed and bleed cooling, only ~ one diesel generator is assu.ned capable of energi:ing the safety r: lated ;notor The turbine driven AP.4 train is AC-independant, but the _ driven AP.I train. Industry non-safety grade mot:r-driven AP.1 train requires offsite pcuer. The analysis that ~ average HPI. reliability and 5 -LOCA frequep. y is assumed. c ?- shows that S 0 ::iay be too frequent appitef,to other PWas as well. 2 The attached paper describes the analysis. e / &J V a.- ~ - _ Frank H. Rows:me, Ceputy Director Division of Risk Analysis Office ofiNuclear Regulatory Research <}'

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Joseph A. Murphy s / Reactor Risk Branch - - Division of Risk Analysis . Office of Nuclear Regulatory Research. Attachnent: As Stated cc: R. Barnero G. Surdick R. Mattson S. Hanauer M. Ernst t A. Thadani RRS Staff _M RAB Staff x 1 1 { t C l o

I Feed and Ciced Issue for CE Applicants Uc understand that the current crop of CE license applicants are proposing that no pressurizer PORY's be installed, that the' HPI shutoff head is to be well below the pressuri:er safety valve setpoint (around 1400 psi), that diameter re.:o te-canual vents, high point vents provide no more than two la and that th'e auxiliary feedwater systens will be ::mposed of one AC-independ \\ ~ turbine driven pump, one AC-power train, and a third non-safety grade notar driven pump. We have attenpted a back-of-the-envelope FRA in o[ der to evaivata the risk C implications if these plants are incapable of " feed and bleed" cooling. 3 meet the Commission's :ifaty 9oa15I - The resuTts suggest that they cay fail b ,N ~ f fix of a core melt frequency less than 10-4/ year and the present (< orth-o,a to enable assured feed and bleed cooling is of the order of $10 million or i Me c:nsidered five more, per plant, based upon reduced financial risk alone. loss of main feeduater, loss of offsite pcuer, groups of accident seque,nces: very s:all LOCA, transient-induced small LOCA (late start of auxiliary feed-water allows a lift of a pressurizer code safety valve which cay stick open), and station blackout with restoration of AC power just before the point-of-co-return. We did not consid'er main steam line breaks'or ATNT, although in these

• saqJences'an assured feed and bleed capability c:uld also enhance safety as well as in the sequences considered.

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Cata frcm the precursor Auxi.11ary feedwatar reliability is also uncertain. program suggests that the pWR average experience has been a fai,1ure probability This av'erage includes early-in-life experience as well as of 10.-3/ demand. mature plant experience and two train as well as three train experience. ~ System reliability analyses hav.e suggestad that the best of the three train systems can apprcach - at maturity 5 per demand. Mcwever, these analyses failed to c:nsider seme ccamen mode failure mechanisms so they,can be regarded It is.not unc :: en early in plant life to find as having,an optimistic bias. instances of repeated, consistent, auxiliary feedwater pump failures while

s The record s0ggests that the failure the system is being debugged in service.

probability of'the AFWS is~substantiELliy higher d'ui-ing' the first core than in. ,_..e, v -; - ~.3 , e,., : maturity. A Ndta'mYith~twodiversisafat grade'AFJ trains,and a third full -e ~7 -. m - t.- -r ~r-- m capacity non-safety grade train will probably~ achieve failure probabilities of: N 3 x 10-30, first core F(L) = 1 x,10-'91,. at maturity These estimates result in loss-of-all-feedwater frequencies of: O.9 x 10T3O* /yr, first ccre 1'= 1 x 10'* 3*4/yr, at maturity + C The uncertainty range is thus: 2.3 x 10-2 & 1 :n % 3.5 x 10~, first core c 2.5 x 10~* 5 1 & 3.9 x 10~7 at maturity cm e 4 4 b

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_g. 9 Assume for c:nyenience that diesel generator A is configured to ener'g safety grade ATJ motor driven train. As we shall see, the core melt frdquency predictions are sensit{ye to whether or not diesel generat:r 3 The event tree for can energize the non-safety grade APd train or not. less 'of offsite power can be drawn: DG's AFW 1-) akay no failures' Jg4 ), melt at 4 x 10~0/yr 95 ,-4 okay B fail s-

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• The higher failure rate applies if.ene of the diesel generators (we have ca it 3),cannot pcwer a motor dr,iven AFW train; the icwer failure. rate app both diesel, generators can pcwer a ct:r driven AFW train.

Note that the Cc=aission safety goal of 10~.t./yr for all c:re melt sequences may be violated by Test of offsite pcwer and a single diesel generator failurt if enersi:e a :t:r-driven there is one diesel generat:r that cannot be aligned t: .:arginally AFW ' train. This high c:re melt frequency c:uld be reduced t:

acceptable value in either of die ways

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_7 In the CE plants, both feedwater and ECCS (HPI) are required for successful ccre c:oling. Main feedwater may remain operable or be restartable in sc=e of these. The probability of HPI failure cn demand was found to be 3.5 x 10-31*I in Surry (WASH-1400). Most FWR PRAs are finding a failure probability for the whole multi-train HPI between 10-2 and 10-3/ demand. We.shall assume that the probability of HPI failure en demand is E x 10-3 I/ demand for the CE plants. A reugh cut at frequency estication suggests: HPI AFW MFW u 7 suC sss )sucpess ~ 39~ ~ ,3 x jpS S LOCA. 2 ,...el, a,g/yr ( _ 4 10-'il - 3 x -107p'3 ~ S x 10~d ' mel t at 1. 5 x 10' II

• I/yr The value of an assured f.eed and bleed capability here is t: eliminate the need This wculd eliminate the smaller (10-6/yr) path to cere melt for feedwater.

withcut affecting the more preminent path via H?I failure. Note that small LOCA w th total HPI failure is predigted to result in a core melt frequency above the C::missien goal for all c:re melts.- T~ne previsten of feed and bleed capabi-lity or of an improved AFW systam will not help this. It is a pr:blem generic to PWRs and not unique to the CE designs. It appears : hat the high frequency of very small LCCA revealed by his rical experience and the marginal HPI system raTiabili-f es revealed by.any FWR FRAs are c:=bining to yield unacceptable c re melt frequencies through S 0-type sequences. We suggest that NRR tackle this c n s 4

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..a .g. r sidered. - The c:re melt, cute:me frcm loss of all feedwater has already been s negHgWe at 10hyr. It can lihood of 52 - The increment in the like in the vast majority of scill be mitigated by HPI, if HPI works, as it will do Cases. f ten, (10-5/yr) With a PORY we will get transient-induced,LOCA ten times as o t of these but the block valve can be expected'to termin' ate'all but 1 percen

1. yr.

If / for a frequency of transient-induced and unisolated LOCA of 10 h is alnegligible i anything, the.PORV' helps 'rathier than aggravates w at- _.. ~ d LOCA.; c:ntributor to the overall Sg freqtie~ncy via' transient-induce. .u d .,.-3.- 3.. ,,_,,-1,. urieur "cpen" We should also censider the command fault LOCA's due to sp The frequency of cecurrence is a sensitive func'.icn o c:= ands to a PORY. It'cculd he' made as small as we wish by' the valve centrol icgic, design.- If we censider the C'rystal River experience suitable relia tity engineering. ge of 3x10-3/yr for as one failure. in h00 FWR-years, we get an indu x \\ N FORY c:= mand. fault LOCA. his frequency can easily experience of ~the three PWR venders suggests that t 7AN I cenclude frequency of 3x10'O'/yr. i be*made much less than the overall S ible.effect on the likelihcod 7 that1having a-?ORY cr not having a pCRV has a neglig lt, provided LOCA or of the likelihood that SiLOCA may lead to ccre me ./ It 2 is the-only censideration. of S that system or compo'nent functicnal reif ability d upcn a design with antici-gees without saying that this analysis is predicate rizer relief valves, Jatory trips saj that[rcutine transients do not lift pressuY bicek and that the,:perat:Es are trained to cicse the FOR ~ e

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s \\ _M - ~ a shorter time window to sa liary feedwater leads to uency of roughly ~ Slackout without auxi This can be expected at a freqhe point-o i. the core by AC recovery. scenario, as the time to t m pressure will be high, the reactor ceciant syste d the level will" he In either 5x10'I/yr. core cooling approaches, safety valve set point), an steam generators Refilling'the n the effectiveness (around t,he pressurizer tive core. fa.111ng tcward the tcp of the ac ufficient, depending upocoolant s A necessary but may not be sd the extent of reacto ecolant system e will be reactor ~ 1 the of reflux.condensaticri an e'n'able.HPI. to refil.witheut core damage ' feed and bleed capability to windcw for AC rec very evaluation of the h A quantitative fairly quickly might extend t e

ld require melt by tens of minutes, per saved by feed and bleed wou more.

haps he itkelihood of AC ld be of melt sequences,that coulic analysis and a or i fractico the improvemen extensive thermal hydrauHowever, it is clear thatTh f res:cration i t'of no return. ble to feed and bleed i vs time. times are befort any po nuence frequency attributa in the blackout melt seq igns with icw.HPI lyr,or less. ~ order of 10 cerrit regarding the CE des

he To summarize, the principal c:nappear te be:

ih shutoff head and. no p0RY's feedwater may be unaccep ~ melt via icss of all is questionable. i Risk of core very s' mall 1.CCA mitigat on 1 adequacy of the design for t r behavior issues. 2. The may be ceupied with cpera o s This W

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1 With Feed Without Feed and Sleed and Sleed. A cm 9 x 10-4 9 x 10~6 TML (first cere) 1 x 10 7 v.. 1 x 10 5 TML (mature) 1.4x10'i 1.8 x 10~'* 1.2 x 10' N, LOSF Casa 1* 1.8 x 10~* LOSF Case Z* 1.5 x 10-4 1.509 x 10'4 S0 2 _ 'Cas5.1 - one of the diesel generators cannot energi:e a.:otor driven AFW train il Case 2 - both diesel generators can energi:e a mater driven AFW c The ecenemic incentives can be calculated by taking the e.xpcsure t .i The econcmic the first ccre as one year and for mature operation as ten years. incentive is essentially the reduction in the present worth (at star They are shewn en the follcwing prehetedmonitarylessasduetoaccidents. diagram:. Case T $13.tM Case Z s -no!F&B ' no F&B. $23.3M 510.7M d v. V $15M ~ Im: rove HPI Case 1 1660.000' Case Z ^ ~~~~Y Reliability s / F&B ~~ ~~ F15' e e e e

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P 35 y. pressurizer valve opens,the pressurizer quench tank rupture disk b1cws, and If the valve sticks open (and cannot be isolated), a small spill cccurs. the.cperaters must restart HPI. Spuricus HPI actuations are quite cc==cn. "e assume here that the frequency of spuricus HPI actuatiens which remain en 3 icng ancugh to challenge a pressurizer valve is one per year. Sorrewing frca the prior analyses we can draw the folicwing event trees for the ' high head HPI design: 3! Without PCRV (or PORY left blocked) Safety Valve C1cses HFI Restart Upon HPI Shutoff ? small spill at 1./yr Spurious HPI 0 Actuation ilarge spill at 10 /yr 1./yr 10,y 10~3 > ccre melt at 10-6/yr With ?ORY installed and unbiocked PORY Closes Upon Stock Valve HPI Shutoff CTeses HPI Restart , sma11 spill at 1/yc Spuricus H?! . Actuation ' s= ail spill at 10~ /yr 1./yr 10,g ' large spill at 10 '/yr-10, =elt at 10~I/yr 0 er 4 J m

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9 _17 ( There appears to be no econcmic penalty (other than first cost) in providing HPI pumps whose shutoff head is at normal RCS pressure, i.e., around 2250 psi. 1 In su=.ary, then, this limited risk analysis cannot distinguish a dif.ference instali in safety among the several ways to achieve feed and bleed capability: one or more large PORY's, raise the HPI head above the pressurizer safety valve setpoint, or install a smaller PORV and raise the HPI head t's near These choices must be made on the basis of design normal operating pressures. si adequacy or thermal hydraulic considerations, preferably c'onsidering AT'AS as 'well as the design to assure that very. small LOCA's can be mitigatad even thodsh HPI or AFW may be late in starting or might be throttled temporarily by We have, hcwever, found a plant availability incentive to ~ the operators. fio. avoid an HPI head so high that it can lift a pressurizer relief valve. such penalty accrues to 'dPi designs with a shutoff head at the normal RCS e pressure. 9 e 5 W t i 1 n

AIIAC*iMENI 4 UNITED STATES sase, f'e "e, NUCLEAR REGULATCRY COMMisslCN wAmu==u. o. c. ::sss ?%) a e s%{c% f l a. I .....f January 22, 1982 l v MEMCRANCUM FCR: ?. Eisenhut, DL, NRR

5. Hanauer, OST, NRR R. Mattson, CSI, NRR C Michelson, AE00 T. Murley, ROGR H. Thcmpson, OHFS, NRR R. Vollmer, DE, NRR FRCM:

Rchert M. Bernero, Director Ofvisicn of Risk Analysis Office of Nuclear Regulatory Research

SUBJECT:

ACCICENT SEQUENCE PRECURSCR PROGRNi CRAFT l(E?CRT' r. w ;; y The. attached Accident Sequence F.recursor-report is: currently being edited) '. hy by ORNL With' expected publication in~ late March 1982. We are providing a ~ l - limited distribution of the draft report for infcrmation purposes. The techniques and methodology used in the report are somewhat concreversial. For examcle, a question has been raised of whether the correct prehabilities (absolute vs. conditional) were calculated and used to determine severe core melt pechability. We are reviewing this and other methecciegy questiens within ORA. The Precursor Report tends.to indicate a core melt probability higher than calculated in typiegl PRAs. probability in the range of 10- /re The report i /reacter/ year for typical PRAs. The precurser program tentative findings were presanted by CRNL (Jee Minarek) to NRC in meetings en 9/18/81 and 12/9/81. Two earlier draft versiens of this report werit given limited distributien within NRC, the first in early 1981 and the second draft report was distributed fellcwing the 12/9/81 meeting. We have indicated to CRNL that ye will previde them timely cc=ar..s befcre report publication. pTease provide n with any ce=ents ycu may have en this report by Fabruary 20, 1982. J. Yf L. = Robert M. Bernere, Director Divisica of Risk Analysis h Office of Nuclear Regulatcry Research t

Attachment:

As Stated cc: R. Cennig, AE00. D. Okrant. ACKS' ~

0. Ross, RE3 L. Teng, RES A.. Dadani, REA3, NRR

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1.. ENCLOSURE SS a no j og% utnTED sTATas Q crf,g NUCLEAR REGULATORY COMMISSION WGINGTC N. D. C. *0555 - p ~f i E 5 ?Mv y! g ms Cocket No. STN S0 470 FEERUARY ! MS 1 Mr. A. E. Scherer, Directer Nuclear Licensing Cembustien Engineering Inc. 1000 Prespect Hill Read Windsor, Cenn. 06095

Dear Mr. Scherer:

In its letter to the Chaird.an dated Cecer.ber lE,1981, the ACKS expressed Tne NRC cencern over certain aspects of the System 50 design (CESSAR). staff is currently addressing these ACRS cc=ents. Specifically, we are addressing the cencern that the System 50 design dces not include ^ cacabi.lity for rapid direct depressurization of the primary system := allcw feed and bleed coeratiens and the reliance placed ucon the seccndary The Divisica of Systams Integration system for heat remval cacability. (OSI) in the Office of Nuclear Reactor Regulatien (NRR) will 5e generating a Supplemental Safety Evaluaticn Report for Palo Verde and(Enclosure 1-). CESSAR on this issue. Tneir current draft report is attached. Since CE'has been reviewing the benefits of PORV's for the Systa'm 50 and since this issue is not ccmoletely resolved, we recuest that you review the attached infomation and provide us your analysis of the need for PORV's in the System SO' design. Tnis infcrmaticn is being requested in support of the CESSAR review and will be the subject of future discussiens with the NRC staff., 4 Sincerely, \\ n !Ldb M<.I L.C U. s.r ..f u sa...x, ...ector c Division of Licensing Office of Nuclear Reactcr Regulacien ~ Enclesure: 1.- NRR Supol. Safety Eval. Report. f* kl 10 fd.... 7 +- aoV 'o ^ g}}