ML20073J909

Forwards Gpu 830331 Response to Encl NRC , Informing NRC of Planned Change in Reactor Coolant Pump Trip Setting & Criteria for Throttling Hpi.Issues Were Subj of Testimony & ASLB Conditions.Svc List Encl

ML20073J909
Person / Time
Site:	Crane
Issue date:	04/14/1983
From:	Trowbridge G METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	Buck J, Edles G, Gotchy R NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP)
References
NUDOCS 8304190451
Download: ML20073J909 (32)

Text

-

2 SHAw, PITTMAN, PoTTs & TROWBRIDGE A PARTNER $te p OF PROFESSIONAL CORPORATIONS e

saco M STREET, N W.{ g {i' WASHINGTCN. D. C. 2od3I5 @

(zoa) e22-ioco RAMsAY D. POTTS. P C JOHN A. McCuLLCuGM. P C JErrERYLvAeLoN ANDREW D ELUS F*6'EUART L. Pf77 MAN. P C J PATRICM McCMEY. P C.

GEORGE F. TROW 8R80GE P C GEORGE P. M'CMAELY, JR., P C.

JACM MCKA RICHARD A. SAMP TM fI

. S cCORMICM THOM AS E. CROCKE R, JR.

$TEPHEN D. POTTT. P C.

J THOMAS LENHART. P C TELECO JOM

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R. JR WENDEUN A. WHITE SE1 ALD CHA PNCFF. P C.

STEVEN L MELT 2ER. P C.

PMiUp J. HARVEY STANLEY M. BARG PMELLIP D. SOSTW8CK. P C DE AN D AUUCM. P C.

(202)822-1099 & 822-1199 8

ROsERT M GORDON MRISTI L. UMBO G. TIMOTMY MANLON. P C JOMN ENGEL AC BARBARA J MORGEN LESUE M SMITM ftO ORGE M ROGERS. J R, P C.

CHARLES 5 TEMMIN. P C

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JAMES 8 MAMUN, P C.

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LSZLIE A. NsCMOLSON, JR., P C.

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RCSERT E. ZAMLER, P C.

" "^L RICHA*4D J KENDALL P.C.

ROBERT S. ROngeNS. P C. '

SETM M. MOOGASIAN JAY L SILSE RG. P C STEVEN M. LUCAS. P C.

TELEX-SMEILA MCC. HARVEY STEPHEN S. MEIMAN,N EV ALLEN PC C RD E GALEN

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p NATHANIEL P SREED. JR., P C.

VICTORsA J. PERMINS STEVEN P PtTL ER*

PMeup D. PORTER MAJM AUGENeuCM P C JOMN H. O NEfLL JR.

RICHARD J. PARRrNO MsCHAEL A. SwiGER E RNEST L. BLAME JR P.C.

JAY A. EPSTIEN ELLEN A. FREDEL' ELLEN SHER'FF 4

CA5'LETON S. JONE% PC RAND L. ALLEN JOHN F. DEALY' THOMAS A. SANTER. P C.

TIMOTHY a MesRIDE SANDRA E. FOLSOM EILEEN M. GLEIMER JAMES M SURGE R. P C.

ELISABETH M. PENOLETON CruNSEL JUDITM A SANDLER DAylO R SAMR EMELDON J. WEISEL P C.

M ARRY M. GLASSPIEGEL EDWARD D. YOUNG. til C. BOWoolN TRAIN

• seot aceestvso os e c April 14, 1983 WRITER S DIRECT DsAL NUMBf R 822-1026 Gary J. Edles, Esquire Dr. John H.

Buck Chairman Atomic Safety and Licensing Appeal Atomic Safety and Licensing Appeal Board Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Washington, D.C.

20555 Dr. Reginald L.

Gotchy Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.

20555 In the Matter of Metropolitan Edison Company (Three Mile Island Nuclear Station, Unit No. 1)

Docket No. 50-289 (Restart)

Administrative Judges Edles, Buck and Gotchy:

In Mr. Baxter's absence I am enclosing as a notification to the Appeal Board a letter dated March 31, 1983, from Mr.

H.

D.

Hukill,

. Director, TMI-1, to Mr. D.

G. Eisenhut, Director, Division of Licens-ing,-NRC, together with a copy of Mr. Eisenhut's letter to Mr. Hukill dated March 4, 1983, to which Mr. Hukill's letter responds.

Mr.

Hukill informs NRC of a planned change in the RCP trip setting and in the criteria for throttling HPI which were the subject of testimony and ASLB conditions in the.TMI-l restart hearing.

After his return 8304190451 830414 PDR ADOCK 05000289 O

PDR

.SHAw, PITTMAN, Porrs'& TROWBRIDGE A PARTNCRSwim Or PROFESSIONAL CORPORAT'ONS

' Gary J. Edles, Esquire Dr. John.H. Buck Dr. Reginald.L..Gotchy.

April 14, 1983 Page Two i

to'the office.next week, Mr. Baxter will. communicate further with

'the-Appeal. Board on the relationship of this change to the hearing

. rec rd.

Resp ctfully submitted, b4 eor F.

Trowbridge

_ Counsel for Licensee

. Enclosures

.cc: ' Service List E

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UNITED-STATES.0F AMERICA NUCLEAR REGULATORY COtD1ISSION BEFORE TEE ATOMIC SATETY AND LICENSING APPEAL BOARD In the lMacter of

)

)

METROPOLITAN EDISON COMPANY

)

Dccket No. 50-289

)

(Restart)

(Three Mile-Island Nuclear

)

Station, Unit No. 1)

)

SERVICE LIST Chainran ~

Jarres M. Cu: chin, IV, Esquire Gary J. Edles, Esquire -

Office of the Executim Legal Director Atmic Safety and Licensing Appeal-U.S. Nuclear Regulatory Ccmrissicn Board Washington, D.C.

20555 U.S. Nuclear Pagulatorf Cerrrnssion Wasrington, D.C.

20555 Docketing and Service Section Office of the Secretarf Dr. 00hn H. Buck U.S. Nuclear Pagulatorf Ccrmtissicn Atcmtc Safety and Licensing Appeal Washingtcr', D.C.

20555 l

Board U.S. Nuclear Regulatory Ccrmtissicn Jchn A. Levin, Escuire Washington, D.C.

20555 Assistant Counsel Pennsylvania Public Utility Ccmnission Dr. Paginald L. Gotchy P.O. Box 3265 Atcmic-Safety'and Licensing Acpeal Harrisburg, Pennsylvania 17120 Ecard U.S. Nuclear Regulatorf Cctm'issicn Rcher Adler, Escuire Washington, D.C.

20555 Assistant Attornev Ceneral 505 Executi m Hocse Ivan W. Smith,. Esquire P.O. Box 2357 Chainran Harrisburg, Pennsylvania 17120 Atcmic Safety and Licensing Board U.S. Nuclear Regulatory Ccanission Ms. Icuise Bradford Washingtcn, D.C.

20555

'IMI AIEET 1011 Green Street Dr. Walter H. Jordan

'Harrisburg, Pennsylvania 17102 Atcmic Safety and Licensing Scard Panel Ellyn R. Weiss, Esquire 881 West Cuter Drive Har:ron & Weiss

' Oak Ridge, Tennessee 37830 1725 Eye Street, N.W., Suite 506 Washington, D.C.

20006 Dr. Linda W..Little Atcmic Safety and Licensinc Ebard Steven C. Sholly Pa".el

~

~ ' "

Union of Conceded Scientists

'5000 He ritage Drive 1346 Conneccicuc '- nue, N.W., Suite 110)

Paleigh, Nc d Carolina 27612

, Nashinecn, D.C.

20036

/

/

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JCrdan O.-Ounni". Cham, EsTJire

~ 2320 No: th Seccnd Street Earrisburg,l Pennsylvania 17110 ANGRY /SII PIPC.

1037.Maclay Street Harrisburg, Pennsylvania 17103 William S. Jordan, III, Esquire

- Ea.rrcn & Weiss 1725 Eye Street, N.W., Suite 506 Washington, D.C.

20006 Chauncey Feoford Judith H. JoPasrud Envircn:mntal Coalition en Nuclear Fower 433 Orlando Avenue State Collece, Pennsylvania 16801 i'

~

Marjcrie M. Aamodt R. D. 5 Ctatesville, Pennsylvania 19320

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TELEX 54 2355

'*mte 's Ovec: D:a* Numter:

March 31, 1983 5211-83-017 I

Office of Nuclear Reactor Regulation Aten:

D. G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Co= mission Washington, D.C.

20555

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (DfI-1)

Operating License No. DPR-50 i

Docket No. 50-289

. 4. **.

'c./:RCS Trip:: en. 25.. Subcooling Marsia.

,. : L5HP=,:9M&n..h. -

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, e v.r:o.+ :r.

G.teInrresponsab::tiin+n'e? Win'~ &p &WWt.w 4..s.

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takyour' lettertof

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,.fS f f W5'.JEn61osure$IS.discussda'it.hilb. i.sf..A3..o:5.%'indihatedis'ub'i tE.25

.J. O*'3S!'?.dritIr16:

.dr

/..aSedEctio.ndi. 2.::subcoolingc.from 500E. to

.~ -

a RC ;iu=pitrip.;

.cifid.Ee..rr.er.s. analyses.c Enclosure.2. provides.

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en.sub;c ooling ma...in4Praciosure 3: addresses--

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.ri.Jthe basis >.for.

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. f RC' pu=p?o~ eratieafdrit'arfoi forinor=alJtransient' and ' accident cenditions.

p

~

~ - These ch'anges arilin' compliance sith; tha require =ents of 10 CFR 50.46 and i= prove the plant safety =argins for certain non-LOCA events.

Further, these changes do not involve. a. change to lechnical Specifications or an unreviewed. safety question-and areg therefore, being inplenented under

- 10 CFR-50. 59. -

S,5 : m ~

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sincerely,

/

3,

, f:. '.-

o li H.' ~D. Huhill Director, Di!-1 EDH:LWH:vj f Enclosures cc:

R. C. Haynes

1. 7. Stolz J. Van Vliet 4.

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E.clcsure 1 25'F Subecoling Margin

===1.

Background===

.Since the accident at TMI-2 the NRC ano ut!!1 ties recognized the need to maintain adequate subecoling margin.

The NRC, through bulletins and NUREGs required each PWR LLicensee to assure that adequate subccoling margin was_

Through maintained and in the long term to install a saturation margin monitor.

coordlitation with tha utilities a minimum margin of 50'F was established.

Fer B&W units, the 50'F was based en an actual mergin of 5'F (which allo's for ths.

w dif ferences in loop temperature of tha -highest elevation and the point where the hot leg temperature is measured) with a 45*F instrument error.

The 50*F subcooling margin becarra the determining criterlen for allowing throttling of HPl.

Il

... Analysis...

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W f l4 liisEdinsiR[us' It { 6d$.DE TS.6 L@gM BheiidLCCAphs subeco! Ing::mai2;iii.~n:ciitter'!IslE t

"Jid%not,(reqdf}ad. forf RC pp#gTrt:monttcr. but rather-LP!'.cporation..

]

irQripsifiFurfherDPipo;isiitiss(Etee.f alls; notLdapendent?.

crr-the"saturaticer. mar specific guidance.to.ths operator,-C-PUN proposes that the 25'F subcool f.ng margin '

~.

monitcr critorie be.used and alarmad.....

go u es < ;. y ~. 7... _..... w<

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w.1IIH.3: Discuss 1on#. - M,s pr%...h.v3.0:::o d. f. ig..r J.,

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-m Since the fall.;of 1979 analyses parformsd. by G?C.M. d 3r*? have de=enstrated thsc-

~'

HPIJiniciation 'and.f throttling basedicn'subcooling =argin is adecuate to e=sure m.

.M

'subcoolina during the.threa ain events of. interast.

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Smal r Br eak LOCA E' ents :

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The reduct!on in subecol1ng margin to 25'T during 53 LOCA events allows better plant centrol during system recovery by allowing HPl thrcttiing sconer. The lower subecoling margin alicvs a breeder centrol band which permits the operator to maintain conditlens with greater-margins to

posalble overcooling conditions.

The SE LOCA events are bounded by the analysis discussed in Sectiom II (Containment Temperature

' 245'F, 4 R).

Centainment Pressure -<30 psig, RH - 100% and ecse - 5 x 10 b '.

Steam Generater Tubi Rupture Events s

e

in2_ prima y to seccndary leak rate during single and T.ultiple Tube ruptur:s Is~a functien cf primary to seccndary dif f erentiel pressure.

Tne differential-pressure is minimized wi;h re:uced subccoling margin anc ty primary' depressurl:siica operating the 90 pumps.

Figu.e 1 i!!usTraves tnat by ' changing frcm a 50*F subccoling margin ( ith pumps eff) to a 25'?

subeccling margin (with pumps On) a 50% reduction in the amcuni cf RCS leakage is attainable.

Reducac integrated eakage wi l l, as a cer.s equence,

reduca the dese to individuals en anc off site.

c.

Overecoling Transients a

During overcooling events a reduced subecciing margin provides an increased

.cperating band making it easier for the cperator To stay within the presssure temperafure limits.

for all other. transient and accident conditions, there is ne reduction in the saf ety-margin or consequences of an accident as described i n the FSAR.

Mitigation of. LOCA, tuta rupture and overecciing events is. net dependent en subeco!!ng 7 argin as m' signal to initlate automaric piant protection.

Since H?l is noc aute=acically initiated for any events ena.!yred in the FSE besides LOCA, tube rupture and,overceoling events; 'd?I thro:: ling does no: affec: the censequenceg as analyzed in tha FSM (See Table 1).

IV.

Conclusien

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. _ - k.:. Maintaining; atminimumct.25'F. in'dtcatedssubecciing margin assures that the-RCS' n'cr=al ;.transTenteand?accidenh conditlens.'Therafcra,.HPI-

$h$& b 1%ra.ryation[@irn,pt anfysafety.h@;furthagacre J.'7d?ffiWTs~ subccoles.Yd6dngs_bhEd kkE.ekpf{c((2Mffshecolitjng'. margin. without ei.

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reductioniin,: subccol ing =segin ;frem,

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i... tRC letter dated 23,s1979 9 Meeting Minutes fer Octobor; 12,. t979".

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' IN Latter-frcrn' WarrenjpHal.I:to Darre!!. G.T Usenhet dates March 31,1982, Question ie13.

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'I.

Background

Based en' initial analyses received from FWR venders, MO cencluded in NUREG 0523 that celayed trip of reactor cool. ants pumps during a small break LOCA can lead to predicted :f uel cladding temperatures in excess of current licensing li=lTs.

B&W examined what would happen if the reactor ecolant ;ceps were Trlpped at seme time into the accident when the system vold fracTic.. was high.

They arb1Trarily assumed that the pumps were tripped when the syster. void fraction was 90's.

At

.the time of pump trip the liquid that was previously dispersed around the

- primary system through pumping action now collapsee ccen to low points of the primary' system such as the bottom of the vesse: and steam generaters. This resulted I.n signifIcant uncovery of the reacter ccee, due to an insuf fIcient amount of IIquid being available to provide acceptable ccre cooling.

Due te design features as well as temporal behavice cf system vcid fracticn, ET,W concluded that unacceptable consequences wouic resett f rcm delayed reacter coolant pump.tripfenly fer a range of small. breaks (.025 to 0.15 f TM and a

. range of trip delay tl=es af ter, accident. initiatien.. Ease: en these findings, a y

meating eft ut!-I LtyJvondecsn andccwners was,hoId. WIT.M C in-Septerber i979..At' *

.... fj t.Mtthi sp.most } hgli.diay)de s$tlh'aktM.f.f d Q0pp,higEEA(,$ 1 intlj ;p rov [I do,cdi mal y[

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signal-resu l t's" t rt;.f ewer ^ p'um;r. trips duri ng 'everecci ing " events and If 'the'-T 1

saturatierr margiir Ts suitably lowp tns ne pump.t?ips ~ during design basistuhe~

' ruptura events cc: veryTsmall' break:LOCA?s.. Ths. rescits cf tha analys!s cantered},

aroundc a-rev Tow oftvo.id' fr. action,s,tc.r? pumps. en/cff ccnd*Tiens..;.

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. The abcyo schema-ics illustrate that The system wi: L'initicily be less vciced as well as.a ;a lower pressure with pumps running.

These resu:ts cccur because The centinued, cperaTicn cf tne RC pumps keeps tne ccc! ant circe:ating thrcughcut the lower het teg

lcops prov i di ng better steam generater heat Transf er and teeperatures.

The lower system pressure wiin The pumps running during the Initial phases of the transient results in a decreased leak flew and thus a lower voi d fraction.

However, the centinuous Operarien cf The RO pumps results in lower quality discharge through the break wnich eventually effseis The decreased pressure ef fect en the leak f lows.

Thus, the fluid in the primcry system ' ultimately evolves to a high. void fraction for cerTain break sizes as a resu lt of - the contir.ued RC pump operation.

The studies done by S&W show that the-crossover point between the pumps en and off cases cccurs at 407 system void fraction independent of the break si:e Loss.of subece!!ng margin in-the het leg occurs well before the RCS veld

- fraction can become large enough to threaten core uncovery. Theref ore, the use of -the subccoling margin is an acceptable alternat ve tc 1500 psig ESAS.

i

-I11.

Discussien Since the f all of.1979, anal yses performed by G:Utf anc ELW have demonstrated that Rsactor Ccolant Pump trip en.subecoling margin acccmp:ishes ihe erigina1 objective of tripping.ROF's without decreasing safety margin 10.- the three main events of. interest.-

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did tic'dc A ndEdgTE5Mpit fdSc5hhcES:5 dine 18c~t}shti ya; cF,itf,l p;IrdshdROPsi:M s:MiNMi$ fed th5ifk.' tre'5% L0 DAM. Fdsy3ndN.MitiItic sifia$c6{Edn gsd5htly.$fcig@tmo_.wtrips tha 25*

S:c ifg V' MWus ESASWTheicperaterc havess enti al.} t,thec same:t

- #A on s adof*satuiraticEafwit!Oha#curi enfi ESAssectu'aticrt eriterleE k raviewiF m

cf the SS LOCA. analysis: Indicates that the basis fer RCF trip was.20% veld' ~

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frectiert.

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".q r.<; Stearrr Generateri uba Rupture. Events O.'

e Curing Steac Generaten Tuba Rep,tures in.:wnich m*nirum subecdlIng-c.argin.isy L

maintainedg centinucus-RC pump'eperatierr assu es expectTieus cec * :0 ?

v. iin;s

' aiminimum: prim'ary to= secondary dif f erenrir :rsssure.

This changt "r:

2

' criterTr fer RCP trip. ~wi-l l' at low R.02's T ::e-c; era ad 'er a g e ETs. ~

~

j spectrum ef tute ru;tures. Cincluding rup ures beycnd ins design casis) and.

to reduce the offsite doses for these events.

I c.

Overecoling Transients.

For overcooling even,ts in which the pressurl:er dcas not empty, the subecoling margin does not-drep telow 25'F even' thcugh RCS pressure is

'below the ESAS initation setpoint' (1600 psic).

When ecmbined witn our propcsed RCP trip cc,iteria,.the ROF's w ill remain cperaticnal, thereby l-precluding void fermeTign in the het legs an: minimi:ing fermation and E

duratien of voies in the reacter vessei hea:.

purther.cre, ROF cperaTien l

.ailews centinued use of 14ain Feedwater in lieu cf Emergency Feec ater and e

-. ~

th9 use of pressurizer spray ic centrel RCS pressure during recevery frem

~

the overeccl i ng Transi ent.

Table i il:esTraTes Tr.z; Tne 23*F succccling cargin is net -ics? for any event in wnich f creec RC 'ico-is requi.ec fcr event mitigatien. -Therefore, the cperaTer sculc nc

s for:ed To Terminate forced flow-when it had previcusly been Taken credi; for in FSAR analysis.

IV.

Alarm in addition, a Control Rocm alarm will be-edjustec_to annunciato if either subcoollag margin monitor Indicates less than 25'F.

The plant ccmputer aise independently ec=putes pressure and temperature saturatien cargin for legging, trending and alarm.

Y.

Conclusten The change in RC pump trip f rom low pressure ESAS actuation.to loss of subccoling mergin:.

ensures RC' pumps are turned off.chen recuired which o

assures no decrease in safety.

permits.RC pump operstien f or seme everecoling events o

.which Increases plent centrol..

~

4.,,.6. O_.,.u.

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.A...

2

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'.7, %,..

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,.~

h 7

t. u...,..'.r. V.1

~

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• 47' =MDhdGC7f03SIW.dODR3$. 3"En.,cx,pediting.

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TABLE 1

-HPt Forced Flow 2**F SOM Accident-cc For Event Assumed in Reac.'.s: Caring Transient

_Mitication FSAR Event

-Rod. Withdrawal Power No Yes No Startup No Yes l'o Beren DIiution No Yes it:t Cold Water Addition.

No Yes

!./A Loss of' Flow No No N/A

+

D cppad Rc6 l

. T.. :..

... +.

...:.. mt:.+ m. n.st. ~...M,;;- NW~.::6.... :l~.Ol.,.. h..w.:..Y.e.:.s.;n,..J wuu,.2m

.:.s =.,.

... j:

R. :

of ETecQLesd5:ayv&r. =+. u.w s.........

.. 4.. ~ -.. r.r w -

r

..c.~.

-. c up$11%yW,,.Q@:ay..::gr.~5 ywa c ;. nl4/A s:'-,-":,i?- -

m:

.r:

%. %W:.s :<5{;:.. s,s A.g.smre:diz.a,q:.s%. d$mkenw,# KctM:c:

Q Losub:r.~

~

MWNdliMF#EM's M@:

- r.

T.wx o.my1..w m. 4;m w p..:

2pp4;'aw.2:4 cp u--

ur..

%ngungq?if&sgym,qi?.w :.:y. - :..d;fis:; 5.5.l5 t..

i iQyp:

.y,; ::

. " ~

G$.OhW6MiU%Y@$@an'4E%fas3%?%,fi#j}$#@@d%EMd45M M T: 5tackout52f? B M NM$$53MENyj%

.N.E: e' k M5teamt.Ine. Fai!~ur,0.$2/6Yosf.ii.f.5N6p;d!$.7i3k,$4@@$5er Dis =uss '.

iGric.dD, I:t ;eT _' ~,-

". f',

~. "

E d

' Tc:.,ks$..!G. t.r...:z,:%i$..:F..e#.v$.f..shi$.wo!8

.?

-2.

- Z.c.6 2 @>

.~

... ~.,. m. e..>. =.s.:. ~. v., : e

.;,.w.

1.

~

....n.

.c Tube-Rupture Yes

.-Yes

'~ ~

t r;/h u...

.. ~. -

.u...a,.... ;... N/A -.

- F0s.!: Handling. '. 4.

No.

s..

.~'

'Yes Yet Yes-.

- Rod Ejo.ction

~

~

...,,.. ~.

2 ',".

. 94 ~

. Feed ><ater Line c-

- No-

, e Break

' <res; irasts Ges: Cacay-Tank Ru-ture No N/A e

.SS LOCA

.Yes See Discuss.

Sea Cisc ss.-

1.arge Break LCCA

.No No "Es

• Analy:ed with and without RCP U.kration.

6 d

/

/

s..-

.-a

-.c..

,ew yy y-.

--.,--..,.---,n-,-%

a e

.r f.

a.

E A w U h e i

e L -== =.= c p.,. C.,

m.t. u

.L w;..=.,=.....

n

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p e

= ; *.

-* ". - (".;. *:*

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. _.,s -,. -

e-t q v.

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s

a.,u. d.....,c. _

~u

. m3 t

.c.

a

'180 I,

{,

l.

j j

j j

j i

'.16 0 - -

1: 25 ! SC, T JMPS ON 11.: 5 0*F SC, PUMPS ON

~~

v*

/

c.

!!!: 50"F SC, PUMPS OFF

.~- -

}4Q y

A (lll:

8

+

- = =

I jjj r==

a b

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glll=, L:,

ll a

1UU

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=

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' *[

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s

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e-.-

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h.Y. :

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• h' :,tY.5,f fb$.I-

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2

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- n, - :..s-- -,~+-... :. x.'. r M.4

..'. =,...

e :....m,,-c

... a.s,

..

. Qf. BTIME'.(MINUTES) -

-. M.. M $,,._

96, F.y.;me;p'.;..

.c

......,..w.....g..

.,m ;.,

. w.,s

,. ~,- e,

+:a.;;.;,.._......::.....

.n.

~..

n...-.

..+ ;.

u.

..w

...-..w c:

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..m....

yl.

SC - Subecoli::g.

4 l

S p C

o

~

Encicsure.3 RC Pump Operating Criteria i.

RCF Operattens - Trip.

A.

. Set ' poi nts f or RCP Tri p-

~1.

Forced Flow Table l' of enclosure 2 provides-an accident / transient se: mary matrix fc which iferced flow is -assumed.. 'Nete, specifIce!Iy,..that RCS fIcw is assumed fer design basis tuba ruptures and that fc a 23*F. 'subecoling cargin trip setpcInt RCP trip does 'not : cccur.

Further, tripping RC pumps en 23*F subeccilng margin reduces trips fer ncn LOCA events for which continued RO ;.: p c;srar:en !s desirstle.

. 2.'

Volded System

l. Jsyste.rppinOJepsA25?r.suec=sMMrkepest.ic cf. c purpsKin a vetdae s..

.de syste+o,5,k.p.,mcahm2.v$.Ifcoe uds e e w @- 2 p > % Q

. s wnenmpessvns

~

yghmp rc'?y s'YTJWWTOV.**Q*wr M.F.

.u.~;c -

9. n

.w*

d m%.. j D.'R'Ws i.;84 [s.$.=:

M aW4M M.

-gW sn S

.Q;. u gf W M M % w % n i M'A'3; M '2's u;c-r. < ? ratisoo b '@^ M%~

't [

b"~ x

& m n FG W. ch d

GW:.w:sh 1Q

- 4.weM

• y A h*I M.m g N z.*;""p~.Thd.~..

wh chz._E;[-[,cT.3'$Ckb4YU 'trIpr;j+ty'*ifp

1. M

4. %

$Q MT fof *7IE46 0704M%.

'" 3 r

avchs0 nacvo-3 ant'(

eaGll4TMLWwy.fDusar;S32'%y.._ fergRCP'tr(p;y$- gd3,s1E tem =1ud cpa he.i..t$.rs advertant ' c.TcEJ.r.5. *F.lh.<.i.di b!

is'd.aEt' dent.t...i.?sys.t.

S..I.@.S M

m j

~,, %9.has' showrn,thsthi,n....,1 pstatern. rtilt.FatTon. asra yscy'. royprebab tj.,lty,.ofb.,9yg.,.y n

4 ye u....

E s

'~Yoedusrenco'sidirlp)Tr@RCFr d6Fr225?lrfsdtic661$igha;rgIrifaduces.!spurreurs RWQi i

s trips,...thereby reducing'chall'ailgesste the PORUf ern nen-!.00A event.s L-f rice '

• tc.centrol RCS pressure.'

p~ressur!zer.sprmys.is ava1Iabiec,d o.

. < 6 u

s

. -. 9....w : rc. r.

w :&m.:

L Het-Stagnentd FL ut dshA.p;g:f.x,s%..

. W.,,.., -Gh,4.,n. ::.... ?. :.h..

. ;. 1. ~.'..w.C

....f.

t

. m., m...,..w:eyg

7..-

3.

a.

Tha. addttlen. ofr 5?E. margin, to the. instrument, e r:,r calcuj atica sur ar;Ist l'r

.enciosurai:ot this. letter [pt;ovidassdtticisn't =ar;li-c p. ec* uds satratico ir

~

~

?

conditions. TEltha? RCS. dutin tF ;;hysicai?iccatich.c.i Tns' sen.sces.

Oc rce";!EM '

L amcrgency prccedures-discusPcieratiene. irtz a ycT:sc ' cec? ?'enin E:c.~ a l ' eti

~

vcids that result frcm tiashing..J.Operater-training; spic!f: cal.ty adcresses Thess ;

procedures.-

5.

RCP Services

-As desc-Ibed in secticn 2.1.1~.5 cf the TMI-1 ResTar Re?cri injection water te the.RC pumps 1s capable of being restcred follo ing ec 4tainment isciatien. _

st tha. event that tha cool ing water to. the pu=ps !s -lesT, tne injection water provides adequate cooling f or' the seat s.an: The p :p can be operated indefinitel y if The seat in'jo.,ction is funcTicning e. crc 2:'iy.

s.

t

/

e I

I

a LTha saturation margin menitor which is part of t. e 00 instrurentaticn system M

wi!! be used fc-indicarica of 25'? subeccling merg:n f:- E'F Trip.

B.

JustifIcaTienL fer Manual RCP trip 1.

Ccnfermanca with 10CFR 50.46 A generic ~ analysis has been performed by ETW wnich envereps TMI-1.

This.

analysis is su=arized in Section 11 of enclosure 1.

2.

l/ cst Probable Best Estimate Since there ls no substantial change in time to trip RO?'s f er the worst case SB LOCA ' under the 1600 psig c-Iteria er the 25'F subeccling margin and in the time.

for the operatcr to_ respond to the initiating signal, this analysis is nct cons 1 dared. necassary.

C.. Other Considerations 1.

Instrumentstion The saturation margin meniter and alarm are addressed Ir. secticn 2.1.16 cf the TMl-1 Rastart Report..' This instrumentation - wi ! i be upgraced to satety, grado

_ during tha Cyct e 6: refus!!ngd- - sh.

.. ;.,&Q4..n. g.w.....;;Q.s.n.,-., u.mc:.nw;z' t v,.... a-c. t.

Ly q-f:

M 4,.7.14%

?.dy. i3@Q:t&22.**Y yg,57y.,&m.m w.,3,',.~.

c4.

s.

. ~. ~..

m:.w..

.nz 5

:. :.-;;.X:. 655?

S f.9;.

-y

~

'i

~

JW4 ig SaaI L2 hcoslRi!0.CMandi.cthcrs c=2rgoneyt precedurosOit i b:.,ba supdatsdr.toJreestab th5iRCWr' start.icittcii If

?'htR&?Ms$$

d%W[NdIN[$$NEN$((@&. iik?50*WW25?P'h.WS&$%%[}k}.YlM%n'-l;hM o

NUp M$ %

Y[IN N.MN6h3(@$

3M~ =IMd*:[gy";;SWE!!4Y}[glEkh-&,gg,d$d.:;$$;q$y$.'

{$g:n @b

. :my.. d MN 5

~

-ynq... r 7:.c. : p-y mq :n.yg gty. p.

' Training. on RCP operation during transients'anc ac:tdents; is. ;integra!"tc ther.

training progras.cn emergency preced'ures which inclades simulate-training.

~

~

.y

, 2 3.,

c.q ' r..n-;.

1 1.-

JSurz:ery of TCCFR.50:59 E at ustien-v c-

wnyp.Q:

~.,. :

_~

.U The probahfilty of'. cccurrence er ths ccisecuences cf an accicent cr

' malfunction. ofC eqdipment;i::ipertant te saf e.y. - e, icus t y eva: aatsde ini

~. '

7that,Safsty Ans!ys.islRoportih'aslnct ice eass d Enciesbres !-and 2 3

" provide-a.summarr.-analysisJ ct'these ne Ts cf ccncern a c Tatie *

~

address al1. accidents, eval usted In.Tne FEAR CnapTer 14.

in fact, by tripping'RC Pumps en 25'F subecoling margin', greater conTeci anc reduced consequences occur.

No additicnal ec.ripment is addeo hy this change ner 'af fectec' which would lead tc a mal fu,cticn.

2.

The possibil.ity f er an accident c : a:f enction cf a dif fererer type.

-then any ' evaluated previously in the Saf ety Analysis Report is ncT created.

All credi ble events analyzed are er.ve!cped by existing FSAR events and no new unanalyzed events a e creavec.

3.

The margin of ~'saf ety as definec in Tne basis cf ry Technical Specificatien is ncf reduced.

RCP Trip anc pressure Tempe.+ature limits

.i-

/

.m-

$ ^N

f.:

Toch_-hs: bases have been carafui!y evti-v e4 at i W 0a76C jn

.Enclo rss 1 and 2 with ne -resuMir4 ren: *. 'n Jaf ETY. marg m.

r c.,

certain. eveds tne Jsef ety mergin has bear inc EBSEC-s 1

s

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d UNITED STATES f,n

.-q/ - c NUCLEAR REGULATORY COMMISSION 3.. - C,. j -

WASHINGTON. D. C. 20555 g, ^ ' ';; jr 9.,,,.*'

March 4,1983 Docket "o. 50-289 Mr. Henry D. Hukill Vice President GPU Nuclear Corporation P. O. Box 480 Middletown, Pennsylvania 17057

Dear-Mr. Hukill:

The purpose of this letter is to inform you of (1) the staff's conclusions regarding your analysis of LOFT Test L3-6, (2) the continued acceptability of your ECCS evaluation model for predicting small break LOCAs with Reactor Coolant Pump (RCP) operation and (3) criteria for resolution of TMI Action Item II.K.3.5, " Automatic Trip of Reactor Coolant Pumps."

We have completed our evaluation of your analyses of LOFT Test L3-6 and.

conclude that the evaluations acceptably predict the test results.

The refo re,

we find the currently approved. B&W evaluation model for small break LOCAs in continued conformance with Appendix K to 10 CFR 50 for the case of limited RCP operation after reactor trip and for the range of licensed B&W reactor designs.

. We have reviewed industry analyses and performed our own analyses to determine

. -whether RCP trip is necessary duri.ng LOCAs, and evaluated the desirability

'of centinued RCP operation during non-LOCA transients and accidents, including

s. team generator tube ruptures.

We have concluded that there is a wide range of transients and LOCAs where it is beneficial for the operators to maintain forced circulation cooling and mixing through operation of the RCPs.

However, some of the calculations show that for certain small break LOCAs, primarily those with only one of the two High Pressure Safety Injection (HPSI) Pumps assumed available, continued operation of the RCPs or continued coeration of the RCPs followed by delayed RCP trip could lead to core damage.

Some uncertainty in these conclusions remains.

Specifically, there is a complex interrelationship among break size, break location, RCP trip delay time, available safety systems, and peak cladding temperature (PCT) for each type of NSS$ design.

Moreover, although the staff's and each vendor's calculational models adequately predicted LOFT test L3-6, there appear to be subtle differences enbedded in the computer models which, when applied to large, commercial, PWR designs, yield differing results regarding the necessity for RCP trip during small LOCAs.

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Mr. Henry D. Hukill (Because.cf this,' we place substantial weight on the views of the reactor

' designers and the utilities which aref almost unrnimous in asserting that for some small LOCAs with:less than the maximum available HPSI flow, delayed RCP. trip could lead to core damage.

Some utilities indicated their preference (to keep the RCPs running for all events; however, this view appeared to be based solely'on the desire to; maintain forced circulation and did not consider

-the-consequences of. delayed RCP trip.

WhileLacknowledging the industry's general conclusion that the-RCPs 'should

- be tripped for small-LOCAs, both the staff and the industry recognized that-

there are other accident sequences of much higher probability than the small LOCA where the absence of forced circulation makes' the operator's job more difficult and.can' increase the likelihood of operator errors.

For this reason, we believe;that a balance should be struck between. the competing risks associated with. tripping the RCPs early ar.d leaving. them running following transient 'and accident events.

Based on our' discussions with both licensees and the reactor manufacturers, E

and= our. internal evaluations,- we_ believe that appropriate pump trip setpoints

can be. developed by the industry tnat would not. require RCP trip for those transients and accidents where forced circulation and pressurizer pressure control is a major _ aid to the operators, yet would alert. the operators to trip the RCPs for. those small LOCAs where continued operation or delayed trip might result in core damage.

In ' summary,- we have concluded-that the need for RCP trip following a transient

.or accident should be determined by 'each licensee on; a case-by-case basis, considering the Owners Group input._ However, the staff must ensure that Lwhatever, decision is made regarding pump operation, it will result in safe, reliable operation of reactors and will not adversely affect' the ability ef licensees to comply with the Commission's rules cnd regulations.

-The enclosure to this letter provides guidance for -the development of either (1) satisfactory setpoints for RCP trip or'(2) the technical bases for allowing continued RCP operation in.the event of a small LOCA at a licensee's - facility.

As stated in the enclosure, manual tripping of the RCPs for a LOCA can be allowed'under certain conditions.

We recognize that -possible differences exist between the requirements of 10 CFR 50.46, which assure ample core cooling capacity, and the approaches

! described in the enclosure which are based upon assuring proper operator /

system-response under' conditions that may be faced during accidents and transients.

Accordingly, in such cases, we'will consider a request -for exemption from specific r'equirements of 10 CFR 550.46 pursuant to.10 CFR 550.12.

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Mr. Henry D. Hukill 3-For the purpose of providing uniformity of setpoints and methods and for minimizing potential confusion that could arise because of diverse actions by individual licensees, we strongly urge that licensees work collectively with owners of similar plants (i.e., owners group) and propose setpoints and methods consistent with other licensees.

If a licensee elects to trip RCPs, when RCP trip setpoints are developed which are believed to substantially meet the guidance provided in the enclosure, we encourage licensees to begin implementation of these new setpoints at operating plant (s)*.

We caution that careful judgment should be used when developing proposed methods and setpoints in accordance with the guidance in the enclosure.

If RCPs are to be tripped, we recommend that the licensees utilize evenc trees to systematically evaluate RCP trip setpoints to minimize the potential for undesirable consequences due to a misdiagnosed event.

Specifically, we recommend the setpoints be evaluated for events where the RCPs could be tripped when it is preferable they remain operational.

We further recommend the setpoints also be evaluated for the case when the RCPs are not-tripped early in the event and for which a delayed trip may lead to undesirable consequences.

We are not requiring a formal submittal of the analyses which support either RCP trip setpoints or the decision to leave the RCPs operational for all events.

However, once the technical bases for the decision are established,

. we intend to conduct inspections of individual licensees led by Regional pe rsonnel.

During these inspections, we will examine the translation of the 10 CFR 50, Appendix K, and RCP operation mode evaluations into plant procedures.

,'We would expect the evaluations to include consideration of the guidance con-tained in the enclosure to this letter.

Copies of these evaluations should be made,available to the staff at these inspections.

Alternatively, a licensee may choose to make either an individual submittal or reference a generic (i.e., owners group) submittal which provides the technical justification for treatment of RCPs during transients and accidents.

In that case, an inspection would not be necessary.

The requirements set forth in this letter supersede the actions required in IE Bulletins.79-05C and 79-06C.

"Unless implementation entails a change to technical specifications or an unreviewed safety question, which require NRC approval prior to implementation.

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Mr. Henry D. Hukill Accordingly, within 60 days folicwing receipt of this letter, please provide your plans and schedules for resolution of this issue for your facility.

You should also indicate whether you desire to make a submittal concerning this issue.

If you cannot respond within 60 days, you should indicate within 30 days when your schedule will be submitted.

The information requested should be sent to Mr. D. G. Eisenhut, Director, Division of Licensing, Wash'ngton, D.C.

20555 pursuaat to 10 CFR 50.54(f).

This request for information was approved by the Of fice of Management and Budget under clearance number 3150-0065 which expires May 31, 1983.

Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Room 3208, New Executive Office Building, Washington, D.C.

20503.

If you believe fu:ther clarification regarding this issue is necessary or desirable, please contact Dr. B. Sheron (301-492-7460).

Sincerely, h')'f

. [l.c.E' Darrell G. Eisenhut, Director Division of Licensing

Enclosure:

Resolution of TMI Actio'n

-Item II.K.3.5 cc w/ enclosure:

See next page

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l ll l

GPU NucleariCorporat' ion 50-289, TMI-1 Mr. ;R. J. Toole Jordan D. Cunningham, Esq.

Manager, - TMI-l.

Fox, Farr and Cunningham GPU Nuclear Corporation 2320 North 2nd Street P. 0.. Box 480

~ Harrisburg, Pennsylvania 17110 Middletown, Pennsylvania 17057 Ms. Louise Bradford TMIA-

- ' Board of Directors 1011 Green-Street

P. A.~N.'E.

Harrisburg, Pennsylvania 17102 P.'O.

Box =26B Middletown, Pennsylvania.17057 Ms. Arjorie M. Acmoct -

R. D. #5

-Coa'tesville, Pennsylvania 19320

-*Decketing-and Service Section Earl B. Hoffman U. S.' Nuclear Regulatory Commission Dauphin _ County Commissioner Washington, D.

C.-

20555 Dauphin County Courthouse Front and Market Streets Chauncey Kepford Harrisburg, Pennsylvania 17101-

- Judith H..-Johns rud -

Environmental _ Coalition'on Nuclear Power Union of Concerned Scientists 433'Orlando-Avenue c/o ~ Harmon & Weiss

! State College, Pennsylvania 16801 1725 I Street, N. W.

Suite 506

• Judge ^Reginald L. Gotchy Washington, D. C.

20006

-Atomic Safety & Licensing-Appeal Board

- U.S. Nuclear Regulatory -Commission Mr. Steven C. Sholly Washi.ngton, DC '. 20555 Union of Concerned Scientists 1346 Connecticut Avenue, N. W.

7J[B.;Lieberman,Esq.

Dupont Circle Building, Suite 1101 Berlock, Israel -& Lieberman Washington, D. C.

20036:

26 Broadsay -

' New; York,_ New Yor.k-10004 Mr. 'Ron'ald C. Haynes, Regional Administrator

,U.~S.'N.~R.~C.,; Region-I

631 ~ Park: Avenue.

King "of. Prussia, Pennsylvania 19406 i-

• Gary J.- Edl es, Chairman Atomic Safety & Licensing' Appeal Board U.S. Nuclear Regulatory Commission

ANGRY /TMI -PIRC

,1037 Maclay~ Street

-Washington, DC 20555-Harrisburg, Pennsylvania 17103

',,Dr. John H. Buck

. Atomic-Safety & -Licensing Appeal Board John Levin, Esq.-

.l.

.U.S. l'ucl ear Regula tor.7 Commission

Pennsylvania Public Utilities Washington, DC 20555 Commission l Box 3265

/

- Harrisburg,LPennsylvania.17120-I

=.,..__. ~

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- Mr. Thomas Gerusky.

ATTN:

C. xet Clerk

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Bureau of Radiation Protection 1725 I. Street, NW Department of Environmental Resources

- Washington, DC 20472 P. O. Box 2063 Harrisburg, Pennsylvania 17120 Karin W. Carter, Esq.

SG5 Executive House

'P. O. Box 2357 Harrisburg, Fennsylvania 17120 G. F. Trowbridge, Esq.

Dauphin County Office Emergency Shaw, Pittman, Potts & Trowbridge Preparedness 1800 M-Street, N.W.

Court House, Room 7 Washington, D. C.

'20036 Front & Market Streets

~

Harrisburg', Pennsylvania 17101 Mr. E. G.-Wallace Licensing Manager GPU Nuclear Corporation 100 Interpace Parkway Parsippany, New Jersey 07054 William S. Jordan, III, Esq.

Ms. Lennie Prough Harmon & Weiss U. S. N. R. C. - T!!I Site 1725 I Street, H!d, Suite 505 P. O. Box 311 Washington, DC 20006 Middletown, Pennsylvania 17057 Ms Virginia Southard, Chairman Citizens for a Safe Environment 264 Walton Street Leihoyne, Pennsylvania 17043 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodnunt Avenue Bethesda, Maryland 20814 Mr. David D. Maxwell, Chairman Board of Supervisors Londonderry Township RFD#1

Geyers Church Road Middletown, Pennsylvania 17057 Mr. C. W..Smyth Supervisor of Licensing TMI-l GPU Nuclear Corporation Regional Radiation Representative P. O. Box 4S0

-EPA Reafon~III Middletown, Pennsylvania 17'057 Curtis' Building (Sixth Floor) 6th and Walnut Streets Philadelphia,' Pennsylvania 19106 Mr. Richard-Conte

f Governor's Office of State Planning

. Senior ~ Resident Inspector (TMi~1) and Development ATTN:

Coordinator, Pennsylvania

.U.S.N.R.C.

P.l0. Box 311-State Clearinghouse Middletown, Pennsylvania 17057 P. O. Box 1323 Harrisburg, Pennsylvania 17120

6 r,

RESOLUTION OF TM? ACTION ITEM TI.K.3.5 The NRC, its licensees, and the PWR vendors have been evaluating the reactor coolant pump (RCP) trip issue since the accident at TMI.

The technical understanding of the industry and the requirements of NRC on this issue have changed twice in that period.

As a result, there have

~

been extensive studies to better understand the dynamic response of all classes o f PWRs to small break LOCAs.

Although some confirmatory infor-mation is still to be received concerning some models, we conclude that the analytical models are sufficiently reliable to be used by licensees to choose their own best method for RCP operation upon indication that a LOCA has. occurred.

In developing methods for RCP operation (i.e., trip or leave running) during all transients and accidents, we recommend addressing the following items that have been identified by the staff as part of our review of this i s,s ua.

We have separated these items into two groups:

Those associated with RCP operation criteria which could result in RCP trip during transients and accidents, and those associated with pump operation criteria which

~

allow the.RCPs to remain running during transients and accidents, including a

small break LOCAs.

s I.

Pumo Goeration Criteria Which Can Result in RCP Trio Durinc Transients and Accidents The staff has concluded, that, if sufficient time exists, manual action is an acceptable means f$r?l tripping the RCPs following a LOCA.

We have based this conclusion iq part upon our;cwn probabilistic assessmant.

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.showed.that the~ failure of a' designated-operator to trip the RCPs within five minutes ;follcwing receipt of a RC? trip ' signal is-approxi-mately six times more likely than is the failure of an automatic trip.

Our probabilistic assessment was limited by a 1ack of comprehensive information about the: complex interrelationships among break size,

' break location, RCP trip delay time, ' avail,able ECC systems, and peak cladding temperature (PCT) for each type of NSSS.

A complete map of' T

this. interrelationship for each design would be prohibitively expensive uto-generate -(tens of computer runs for each design at thousands of dollars per run and hundreds of hours of analyst time).

Without such a map, we cannot accurately define the bounds of the regi6n where unacceptable consequences lmight result.from delay in RCP trip.

However, based on.

our understanding.of the phenomena in question, analyses performed by

,. the NSSS vendors, limited'. independent analyses performed.by the staff, te'sts performed in 'b' th Semiscale and LOFT, and c'ur' probahility assessment, o

- we conclude that allowi.ng manual RCP trip is acceptable provided certain conditions are satisfied.

Our guidelines for RCP_ trip setpoints and methods are' set forth below.

In developing RCP trip setpoints and methods,

'thereiare.two. potential problems with RCP trip.that. continue to show up

'in.-reactor operations.

The first problem is caused by the fact"that"the

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. loss of pressurizer sprays 'upon RCP tri2 for transients and for small /'.

7, break LOCAs results-in a need in.some $lants to use power opiraged. relief' valves (PCRVs) for primary pystem pressure control.

Despise extensivei testing of prototypes and im, proved reliability engin~eering, these valves '

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Al though ' the '

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' question of. PORV. functionality has heen bett'ar charadterized by.the EPRI

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. valve testing program since the accident at THI,-there does not appear-to be significant progress in improving the overall operational reliability of PORY. systems. - A second problem associated with RCP trip is that it tends to produce a stagnant region;of coolant in the upper elevations of sthe reactor vessel.

In a number of recent operational events, this hot, stagnant fluid -has flashed and partially voided the upper vessel region during depressurization or cooldown situations.

Despite wide dissemina-l-

' tion'of information about these operating events and the learning opportunites that the'y present, we still perceive that operators (1) are not completely familiar with the significance of a steam bubble in the. upper. head, (2) have difficulty controlling coolant conditions so as to avoid or control flashing where.possible, and (3)'may have a tendency to take precipitous. actions when a steam bubble exists.

.,, In developing your RCP' trip setpoints and methods, the follcwing guide-.

li,nes should be considered:

'l.

S'etooints for RCP Trio a). The setpoints should be designed to assure that the RCPs will.

g.

,.be tripped for all losses of primary coolant in which RCP

.t[',

tri'p-is considered necessary.

The setpoints should also ensure J

continued for'ced RCS flew during steam gen &rator tube ruptures

• ' up to 'and including the design basis tube rupture.

Safety analyses should be performed to demonstrate t,he achievement of x.,

these goals.

The symptoms and signals used to alert an operator

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-.of the.nced to mandally trip RCps should be, to the extent y'

possible, uniquely attributable to LOCAs and not other depressurizing transients and actions for which centinued pump 3

operation. is des trable.. In 'this regard, consideraticn should

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_4 be given to partial.or. staggered-RCP trip schemes (e.g.,

in two loop, four pump plants, trip.one pump per loop Limmediately and trip remaining pumps once the" existence

.of a LOCA is confirmed). -If selected pumps are tripped during.the initial phase.of the transients, 1icensees should. assure that training and procedures provide direction for use of individual steam generators with 'and without RCPs in operation..Your evaluation should be capable of demonstrating and justifying that the proposed RCP trip setpoi.nts are adequate for small LOCAs but will not result in RCP trip for other non-LOCA transients and accidents (e.g.,

steam generator tube ruptures).

b)

The RCP trip setpoints should be selected so as to exclude extended RCP operation in a voided s.pstem (e.g., pump head degradation >10%) unl'ess engineering. analyses or tests are

~

available to justify that RCP and~ RCP seal integHty will be maintained under those conditions.

c)

If, for some transtents and accidents. within the current design basis, and with offsite power available, the setpoints selectsd will lead to RCP trip even though it is neither required-nor' desirable, it should be assured.that these events r

will not result'in challenges:, either automatic or from the operators, taw the PORVs to accomplish depressurizing actions normally a complished by pressurizer sprays. Heated

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e 5-auxiliary spray capability not derived frca RCP discharge pressure could be considered as one possible means of eliminating this reliance on the PORVs. On the other hand, if FORY operation is continued to be recommended for use in depressurization, then a program for upgrading the operational reliability of the PORVs should be developed.

d)

For any conditions which require or result in RCP trip and the establishment of a hot, stagnant, fluid region at high points in the primary system, emergency procedure guidelines and emergency procedures should specifically describe symptoms of primary system voiding due to flashing of stagnant regions of hot coolant.

They should also contain specific guidance on detecting, managing and re$oving the coolant voids that result from flashing.

Operator training programs should specifically address the significance of primary system voids under non-LOCA k

and LOCA conditions.

e)

Transients and accidents which produce the same initial symptoms as a LOCA (i.e., depressurization of the reactor n

and actuation of engineer'ed safety features) and result in containment isolation may result in the termination of systems essendia'1 for continued cperatien of the reactor coolant pumps (i.e., ccmponent cooling water and/or se?l injection water).

It uns the intent of TMI Acticn Plan

6-

. Item II;E.i.2 to have licensees reevaluate essential and ~non-essential systems with respect to containment isolation.

In particular, if a facility desiiin terminates.

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.wadr services essential for RCP-operation', then it should

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be' assured that these water services can belrestored in a

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timely manner once a.non-LOCA situation is confirmed, and

.1 prevent seal damse or failure.

It should be confirmed that containment isolation with-continued RCP operation'will'not lead t.o seal or pump -

= damage or L fail ur'e.

f)

Parameters used to determinc when RCPs should be tripped

• should provide. unambiguous -indicators of a LOCA.

The inade-

~ quate core cooling instrumentation required by the Commission

~-

and described in NUREG-0737 should be factored int': the emergency procedure guidelines where useful in indicating 1

the need for RCP trip.

x

2.. Guidance for Justification of Manual RCp Trio Our review of this subject leads us to conclude.,.that, Wl1en tripping the pumps is recorrr. ended by the licensees, it is preferable to manually (rather than automtically) trip the reactor coolant pumps where it is at all possible to justify it.

However, our'

~

review indicates that there may be a few plants for which it is not possible t6 justi'ff mnual trip because of reliability considerations.

Tne guidance stated below is intended to assist

7 those plants that can and should rely on manual trip to d evelop complete justification for and to clearly identify those few plants that may not be able to rely on manual trip..

a)

Based on the RCP trip setpoints developed according to the guidance in item one above, analyses

• should demonstrate that the limits set forth in 10 CFR 50.46 are not exc e

for the limiting small break size and location. For the purposes of showing compliance with 10 CFR 50.46, operator action to trip the RCPs should be assumed no ear' lier than

~

two (2) minutes following the onset of reactor conditions corresponding to the RCP trip setpoint.

Allowances should be made for instrument error.

~

b)

If manual RCP trip is proposed, then for the limitino small 4

break size (s) and location (s) identified frem (a) abov a most probable ** hest estimate analysis of the amount of time available to the operator to trip the RCPs follcwing the existence of the RCP trip signal should be performed If this tima is less than that recer.T. ended,in Dra ft ANSI Standard NS60, the acceptability of this t.ime should be justified.

An evaluation of operating experience data should i-

• Generic analyses of general reac' tor types is acceptable in lieu of plant specific analyses.

bound plant specific evaluatiensThe ' generic analyses should be shown to "Each licensee should identify and justify the mast probable' pl conditions.

of justifiable most probable plant conditions. Conservative estima ant

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L be included when addressing, this justification.

The consequences if RCP trip is delayed beycnd this time should also' be addressed.

Contingency procedures should be ~

developed and be available to the' operator in the event the RCPs are not tripped. in the preferred ' time frame.

If

~

the time available is in excess of the standard, no further -

justification is necessary.

3.

Other Considerations

. Although acceptance criteria in the following areas are no,t speci-fied, assurance that they have been.. considered and good engineering practice has been followed will be required.

af For the parameter (s) employed in the RC? trip setpoint, the level of quality for the instrumeritation that will signal the need for RCP trip should b'e' established.

In pdrticular, the ~ basis for the following should be identified:

o The design features chosen for the' r.ansing instruments (e.g., seismic and environmental qualifications, reliability, etc.)..

o The degree of redundancy in the sensing'. instruments.

g.

In general, credit may be taken for all equiprent available to the operators and for whi'ch sufficient confidence in its

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operability, during conditions under which it is expected to operate, has b' en established.

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b): It should,be ensured.that emergency eperating procedures

- exist for the timely. restart of'the reactor coolant pumps when conditions which wil1 ~ support' safe'pu5p operation

-are established.

e).The. training program should instruct operators in their responsi.bility for performing RCP trip in the event of a SBLOCA..,In particular, the' operators should be trained in prioritization of actions following engineered safety

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features actuation.

II.

Pumo 0ceration Criteria Which will not Result in RCP Trio During Transient and Accidents Itiis recognized that an evaluation could lead to the conclusion that, based on competing risk's, both the preferred.and safest method of pump oparation following a transient or accident phent would be to have the

-pumps rmnni.ng.

In order to substantiate this conclusion, the following

. evaluation guidelines should be considered:

1.

Evaluation of Inventory loss The industry analysis mode 1~ comparisons against LOFT test L3-6, while providing gd'od agreement with the experimental data, require

~

additional verification to support continuous ' pump operation for all transients and accidents, including small break LOCAs.

These '

include:,

Completing evaluaticds of LOFT L3-6 through the ECCS recovery o

phase, if not alreadh completed; Q

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o Evaluating all modeling cifferences which are expected to exist between the LOFT prediction and the large plant

. analysis (e.g., for B&W plants, hon docs' priihaiy system geometry affect conclusions?

Can smaller scale, two-phase,

,side entry pump performance data be confidently extrapolated to large,. bottom entry pumps, in particular in the high void fraction regions?).

2.

PumoIntecrig a)

During periods of extended two-phase perforrance', pump ' integrity-

~

is a-chief concern ('during the TMI-2 accident, one of the operating RCPs was finally tripped due to excessive vibration).

The evaluation should conclude why RCP seal and RCP structural integrity will be assured during extended'~two-phase flow per-If RCP and/or RCP seal int $grity cannot be assured, for=ance.

thin the consequences of their failure should be considered-i~n the analyses.

b)

If continuous RCP operation is expected in the presence of a containment isolatiyn si nal, the ability for continuous RCP 5

operation without essenti:al water servi'ces, should be addressed, or the capability to rapidly restore essential watar services,.

should be provided.

c).The ability of the.'RCPs to operate in the accident environment (e.g., containmabt temperature and humidity) should be addressed.

If continuous operation in the; accident environment cannot

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!be -assured, then the consequences'of failure at any time during the ' course of the acciden't should be addressed.

3.

Acceotability of Results :

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Analyses should.be performed which demo'nstrate that the ECCS acceptance criteria of.10 CFR,50.46 'are met using an analysis model which ecmplies with the requirenents of Appendix K to 10 'CFR'50. "These analyses should assume (a) continuous RCP operation, and (b) assumed RCP-trip at various' times,during.the-accident if continuous pump operation cannot be assured.

If the analyses indicate compliance with the criteria of 10 CFR 50.46 cannet be achieved..the-staff will consider a request for an _

exemption to the 10 CFR 50.46 requirements if (a) it is concluded that compliance with 10 CFR 50.46 would require operating the

• plantiin a less safe condition, (this sho,uld be supported with risk /

~-

" benefit analyses), and.(5) it is concluded that design' modifications (e.g., additional HPI capacity)- would not Be cost-effective to implement from a safety standpoint.

5The risk-benefit analyses can.take credit for all equipment for which there is. confidence that this equiprent will. remain operational during the accident.

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