ML17272A434

Forwards IE Bulletin 79-05B, Nuclear Incident at TMI-Suppl

ML17272A434
Person / Time
Site:	Columbia, Washington Public Power Supply System, Satsop
Issue date:	04/21/1979
From:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	Strand N WASHINGTON PUBLIC POWER SUPPLY SYSTEM
References
NUDOCS 7905080378
Download: ML17272A434 (15)

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+~*gW UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V SUITE 202 ~ WALNUTCREEK PLAZA 1990 N. CALIFORNIA BOULEVARD WALNUTCREEK, CALIFORNIA 94596 April 21, 1979 Docket s.

50-460 50-397 50-513 Washington Public Power Supply System P. 0.

Box 968 3000 George Washington Way Richland, Washington'9352 Attention:

Mr. N. 0. Strand Managing Director Gentlemen:

The enclosed Bulletin No.79-05B, is forwarded to yau for information.

No written response is required.

We have also enclosed copies of recom-mendations of the ACRS to the Commission for your information. If you desire additional information regarding this matter, please contact this office.

Sincerely, I

R gefken irector

Enclosure:

IE Bulletin No.79-05B with 'Enclosures ACRS Recommendations to the Commission dated April 18, 1979 and April 20, 1979 cc wgenclosur es:

M. E. Witherspoon, WPPSS A.

D. Kohler, WNP-2, WPPSS J.

P.

Thomas, WPPSS W. J. Talbot, WPPSS G.

C. Sorensen, WPPSS

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 April 21, 1979 IE Bulletin 79-05B NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Description of Circumstances:

Continued NRC evaluation of the nuclear incident at Three Mile Island Unit 2 has identified measures in addition to those discussed in IE Bulletin 79-05 and 79-05A which should. be acted upon by licensees with reacto'rs, designed by B8W.

As discussed in Item 4. c. of Actions to be taken by Licensees in IEB 79-05A.,

the preferred mode of core cooling following a transient or accident is to pro-vide forced flow using reactor coolant pumps.

It appears that natural circulation was not successfully achieved upon securing the reactor coolant pumps during the first two hours of the Three Mile Island (TMI) No.

2 incident of March 28, 1979.

Initiation of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of non-condensible

gases, in the primary coolant system.

To avoid this potential for interference with natural circulation, the operator should ensure that the primary system is subcooled, and remains subcooled, before any attempt is made to establish natural circulation.

Nautural circulation in Babcock and Wilcox reactor systems is enhanced by maintaining a relatively high water level on the secondary side of the once through steam generators (OTSG).

It is also promoted by injection of auxiliary feedwater at'the upper nozzles in the OTSGs.

The integrated Control System automatically sets the OTSG level setpoint to 50% on the operating range when all reactor coolant pumps (RCP) are secured.

However, in unusual or abnormal situations, manual actions by the operator to increase steam generator level will enchance natural circulation capability in anticipation of a possible loss of operation of the reactor coolant pumps.

As stated previously, forced flow of primary coolant through the core is preferred to natural circulation.

T Other means of reducing the possibility of void formation in the reactor coolant system are':

?

A.

Minimize the operation of the Power Operated Relief Valve (PORV) on the pressurizer and thereby reduce the possibility of pressure reduction by a blowdown through a

PORV that was stuck open.

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IE Bulletin 79-05B April 21, 1979 Page

2. of 4 B.

Reduce the energy input to the reactor coolant system by a prompt reactor trip during transients that result in primary system pressure increases.

This bulletin addresses, among other things. the means to achieve these objectives.

Actions To Be Taken by Licensees:

For all Babcock and Wilcox pressurized water reactor facilities with an operating license:

(Underlined sentences are modifications to, and supersede, IEB-79-05A).

Develop procedures and train operation personnel on 'methods of establishing and maintaining natural circulation.

The procedures and training must include means of monitoring heat removal efficiency by available plant instrumentation.

The procedures must also contain a method of assuring that the primary coolant system is subcooled by at least 50 F before 0

natural circulation is initiated.

In the event that these instructions incorporate anticipatory filling of the OTSG prior to securing the reactor coolant pumps.

a detailed analysis should be done to provide guidance as to the expected system response.

The instructions should include the following precautions:

a.

maintain pressurizer level sufficient to prevent loss of level indication. in the pressurizer; b.

assure availability of adequate capacity of pressurizer

heaters, for pressure control and maintain primary system pressure to satisfy the subcooling criterion for natural circulation; and c.

maintain pressure

- temperature envelope within Appendix G limits for vessel integrity..

Procedures and training shall also be provided to maintain core cooling in the event both main feedwater and auxiliary feedwater are lost while-in the natural circulation core cooling mode.

2.

Modify the actions required in Item 4a and 4b of'E Bulletin 79-05A to take into account vessel integrity considerations.

"4.

Review the action directed by the operating procedures and training instructions to ensure that:

a.

Operators do not override automatic actions of engineered safety features, unless continued o eration of en ineered

IE Bulletin 79-05B April 21, 1979 Page 3 of 4 safet features will result in unsafe lant conditions.

For d

i i

d 2 f would threaten reactor vessel inte rit then the HPI should be secured as noted in b 2 below b.

Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actuate'd because of low pressure condition, it must remain in operation until either:

(1)

Both low pressure injection (LPI) pumps are in operation and flowing at,a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2)

The'PI system has been in operation f'r 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure.

If 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.

The de ree of subcoolin be ond 50 de rees F and the len th o time HPI is in o eration shall

.be limited b the ressure tern erature considerations for the vessel inte rit~."

3.

Following detailed analysis, describe the modifications to design and procedures which you have implemented to assure the reduction of the.

likelihood of automatic actuation of the pressurizer PORV during antici-pated transients.

This analysis shall include consideration of a modifi-cation of the high pressure scram setpoint and the POVR opening setpoint such that reactor scram will preclude opening of the PORV for the spec-trum of anticipated transients discussed by BEM in Enclosure 1.

Changes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated transients.

4.

Provide procedures and training to operating personnel for a prompt manual trip of the reactor for transients that result in a pressure increase in the reactor coolant system.

These transients include:

a.,

loss of main feedwater b.

turbine trip c.

main Steam Isolation Valve closure d.

loss of offsite power e.

low OTSG level f.

low pressurizer level.

P a W

IE Bulletin 79-05B April 21, 1979 Page 4 of 4 5.

Provide for NRC approval a design review and schedule for implementation of a safety grade automatic anticipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level.

6.

The actions required in item 12 of IE Bulletin 79-05A are modified as follows:

Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a,control.led or ex ected condition of o eration.

Further, at that time an o en continuous communication channel shall be established and maintained with NRC.

7.

Pro ose chan es, as re uired, to those technical s ecifications which must be modified as a result of our im lementin the above items.

Response

schedule for BSW designed facili.ties:

a.

For Items 1, 2, 4 and 6, all facilities with an operating license respond within 14 days of receipt of this Bulletin.

b.

For Item 3, all facilities currently operating, respond within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

All facilities with an operating license, not currently operating, respond before resuming operations.

c.

For Items 5 and 7, all facilities with an operating license respond in 30 days.

Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, llashington, D.C. 20555.

For all other power reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required.

Approved by GAO, B180225 (R0072); clearance expires 7/31/80.

Approval was given under a blanket clearance specifically for identified generic problems.

EXTRACT OF BSH COMMUNICATION - RECEIVED BY NRC Enclosure 1

. ENNNcr-Lob Page f of 4

'IC'N Co'ITTNUKBG REViEM OF THE 5EQUErlCE'OF EVENTS LEADS?t" TO T}fx rrfCIUEnT m VRL-2 M fiNRN 28

't979 SHOWS Tffhj ACTION CAR. QE TAKES TO PlrOVEOE aSSUWHCE

'78AT THE P~t.OT-OPKRATEO BELIEF VALVE (PORV) tlOUHTEO Rt TflE. PRESSURIZER OF 858

'PWfr8 ALLMT BE AGAN'EO BY N<rICIrmTEO TnlrSrEfrXS HlfrCH ffWVE OCCVarr<O OR phd A sIGBKFIcABT pmsaaILrrv oF occvzpING IN zfEsE pmNTs.

TffIs acrrQ

> tsar CK GMRME THE SAFETY OF THE AFFECTED PLAt')TS HEST/) RESPECT TQ TfrEIB BESPO.'l"E R MRS', UPSY OR ACCIDENT corrDITIorrs rfOR k.Ehp To ui~REvr EQED sAFEt Y cot]CERNS.

VHK AWICl'PATED TSMIBlTS OF COERCE% IlAF.=

t OSS iF EXTihNi. EtzerRrcAi. Lmo 2.

TURBAN TAfP 8.

LOSS OF PAIN FFEtQATER 4.

x9sS oF aeo"r6KR VACVVH S.

EeeVEATEHT CLOSURE OF rex'YMAR ISOLATION VnLVES (nSIV).

4

'A )MBFR OF ALTKSRTKVKS MErK COHSIDERFO Zr) DEVELOf ma &E aCTrOnS PROPOSEO SENT BICLUDKBf.:

.~RlCTNC BEAUOR PLACER TO A VAL'UE RANCH RQULD assVAE NO hcTUATJON OF

. '~ vow.

mz. nEAcToR pmnEcrrm svsTEv, DEsrorr pREssURE a.lo poRv sv"-

p02rWS AErfALHEb AT THEIR MRREt]T VALUES.

-kowans THE HrGH pRFssuxE REAcmrf Tarp sETvorrlr To I vhLuz whuff HauLD ASSURE gg AQTUATIOH QF THE PORV.

THE DESI'RE5SURE QF TI]E REACTOR ArfD 7HZ SETPQIN'OR PORV IM:TUATIOB REHArrrEo AT TfrEIR cURAB)T VALUES.

46'-"EADEM THE HIGH PRESSURE REACTOR TRIP SETPOtH7 NtV I~MU5TrtfG Tl)E O~aaaTIR6 PRESSURE (Are TmERATU~Z) OF TffE RNCTOZ TO ASSu~E tfa POaV

'ACVUATEOd N$0 To PMVIDE MEgVATE fV~AGIf) TO ACCORIQDATE V/l'BritlTIONS IA omnmXNS PRESSURE.

mE SETPOr>n FOR POnV ACruATrarr nErernED AT ITS CVRmn i'.m.

Yks ALvFRrrATrvE koULO frEOUCZ rfET ELECrnrc~L OUrpur.

I A5$~LF THE-HIGH pREsslJRE Tair ue THE poav sETporrrrs To assure rro erne aCme.rai FOV THE AS OF AhzrCIPeEO EVErfTS OF CO fCEm

>HZ OESim I'BESSURK OF THK REACTOR REHAKrrED Af KTS cUAAErfT VALVE "gg AnADsIS oF THz zw~acT 0F THEsE vhnrovs Au.ERnarrvzs Ano THEIR corrrnrourro,<

TO ASSUAIHB 7MT THE PORV HIt.L NOT Ac)'UATE FOR THE CLAUS OF SWTEEIPATEO TMNstEr>TS QF COME/% HAS BEEN ~LE'fED.

'PiE REsUt Ts SNN THAT."

I.C%aihg Tr@ HECT PRESSuRE, REACTOR TRr{, SETPbIm Fmg 2~ ASM 70 2390 PSXG I

ASO Pal'5XVj~ THE 5ETPQ1%'OR "A$E PILOT:OPERATED RELIEF VALVE FRQ;=t ZZ55 mSG TO 24SO VSra PrgvlD S THE REqUIRED ASSUSWCE THIS ACT'ro'$ fNS THE FURTflER /LOVE't)Thugs Qf'.

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EKTBPtGT OF BQ CONUNICATIOtI - RECEIVED BY NRC 4/20'f79 Enclosure 1

Page 2 of 4

+~>N N~ >~~~8<L~YIt OF PORC AND A5HE tQDE PRESSURIZER SnVH'j yrLVK

~ KTL%iE05 FOR OTHER lNGREASIFQ l'RE5SURE TRANS'IENTS.

PBKSKRVLRG Pi-SSURK RKIFF &&CITY.FORAt.t. HLCH PReS5URE 'TM~61ENTS.

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KERTRAHK8 THE P05STBkf.ITS OF INRODUCIHG Ut)AEuIB<ED SItt-En'anCFWS.

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- ABSCESS THE TbK AT WTCH VHK STAT SVSTEH >)uT Srr)K N}ULD 8K LOST r@

THK EVEST ENERGEvD'EEÃlATER 80%1 MERE DELAYED.

'fkVVIjKffTS5S GIVER XN TABLE f

".A SVemN OF @E itPACr OV THE PROPO".Eo SCTPOr>rr CW.~SE5 Oii nl.< am tCCPWED

~

SR8 PVQ65 ARK CVRRElVKY CAt'ABfE OF RUNBACK TO 15Ã OF.

FULLY POWER UPON LO55 QF LOAO Oa TRtp OP mF Tljnerm.

VnrS CI'Paa<LrrV REqurRES nCrUW.IO;t OF TtlE PILOT-GPERATE9 RKYEF vNYEs.

THE cAPRBtltTY EtfcREhsE5 YHK RLKADttlTYGF POWER SUPPLY 70 THE SYSTER 8'f RETUS)ING.THE'UHETS fQ POWER GE~1ERATIO?J ISA'UICYt.'(

Ama meSa Vfl>>SEeVS.

VHE ACTran PROPOSEO aaOV~ VILL R~qUIRe ment mZ naB ag TMPPEO FOR TIreSE KVZrZS:

J NRC NOTE:

The effect ef changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typified by th attached figure 'I. which was developed by.

BSM for a loss of feedkater transient.

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TABL'K 7 5QCNRV QF PROTECTSf ACA?F5T PORV fLCTUATIOI)

PAOVTGEQ BI PAOPOSEO SETPQiNT CHA.'~CF.S FOR ALL NA'TCLPATEQ TRAHSIEi TS EXTRACT Op BBM'COhpPNCCRTION

~ RECEIVED BY lrRC 4(zO/79..

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A'IITrorpATED TRArrsrsrrrs IIHIOH HAYE occURRED AT Usrr prArrrS IUrD IBIIcrl woULD r'oQN iv AGTIUATE PDRV AT THE cURRElrr sETPorHT. (zzss Psrc):

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VURBi% 782P tLCSS OF EXrESuu. zlzCTAL&t, Lee c.

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P>ZXCrPATEO TAAWfErnS arm HAVE OCCuRRED N'le P&SVS ee m>ICO vQ&5 MRRAU.Y AcTUATE Ponv AT Tf)F PfNPQ5ED sETPOINT {z459 PsEG)."

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-feUCKVNFO TAAHSIBa8 >n>re mVE ieV OCCVaReo AT a~@

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SOi% COHmN. tao GROS Vimann~h<S

{MODERNC TO Hrul n~CTrVIn

. BCrHH'GRQfPS far OTHERWISE PROTECTEG BY HIGH FLUX TRIP) 8.

MKMMRnELUTlOH.

p Ap <<kv

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$~ff'fl78~h v'oP.L~~ IQfQ KVKHTS) ANO 4~aICH >'OULD (z~sa ps>6):

tPrlICH HAVE 7$i GCCURREO AT DSM PLIrPJTS fCUH P7KlEN3i'QYY ACTUATE THE PORV AT TriE PRUPD5EO Sl:TPOIllT

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80,'CE CONTROL ROD G OTHERWISE PROTECTE Il RatP tfITHDRAMALS ()lIGll RE'ACTIVtTY r,-'ORTl) r:GT 0 BY HECT FLOX 7RlP).

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UihlTED STATES NUCLEAR BEG VLATORYCOl"ifdlSSlON AOVJSORY COli'its'ilTTEE OM!REACTOR SAFEGUARD'IASglNGTONiD.

C. 2D555 April 18, 1979

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If' FRQR:

HEMORANOUH HER:

Chairman Hendrie '

Commissioner Oil&:ky Corranissioner Kennedy Commissioner Bradford Commissioner Aheagne I

R. P. Fraley, Exeputive Director Advisory Committee on Reactor Safeguards Attached for your information and use. is a copy of the recommenda-.

tions of the Advisory Committee on Reactor Safeguards shiich vere orally presented to and discussed vith you on April 17, 1979 re-garding the recent accMent at the Three hile Xsland Nuclear Sta-Uon Unit 2.

R.,'P.

Fraley'xecutive Director

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Attachment:

Recomm ndations-of the NRC Mvisorj Committee on Reactor Safeguards He. the 3/28/79 Accident at The Three Nile Xsland Nuclear'Station Unit 2 k

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'P',f i wp y ~ *a UNlTBD STATES NUCLEAR BEGVLATORY CGVifvllSSIGN AbvlsORY co'iQAl'PfEE os REAcTdh sAf-EGU'bg Yfh.~l\\1fiGTOH tie e ~M5 I

I

. April 20, 1979

'. 8~mr-N,e Victor Gil5nsky

~i'tiling

~irma%

U. B. Hpclear Regula".cry Cma6ssion

)i~=55naeon, DC 20555.

D ar Dr+ Giiinsl;y:

T..is letter "s in x'Capon e to youf's of April 18< 1979 <e~cQ rerve~~e4

." "wC 4c.

PCRS notify the Cceaia8fonc'.e imaiiatelv 3f >e hei'eve any 6f ou" org3. x'eco~noacion-"

aE g'.pril l7:should k act~ urn before our n m regal,=-rip schMuled.

m a~i~

aC iMcn <re cou2.8 pre," re n &ra@l

.lq9er.

RB Cc~t.ttQQ QiSCU".SH ttll.s toPxc 5g Con'eran=e re3ePhOne col on'Pp<il 19 ard of ers the o3,3owir~g comments.

"Pal of the reco->~

nuations neQe by the h~w in its <<"~-"ei~ v$ "4 the Cc.siggioncrs an April. l7, 3979, et'e g~weric in na'~ore end. ro-ly to Q3, P~Rs.

H>no ve"e 'intesded to recruits

'>w:&Bi te ch~e",

in o crete-q p",o-

<QQL'as or ply uaQificaQKons of o. ex'at1rig 9'~i, Rich chmqes

~~gulp h~ a=de pray a e'er tv8g'K their eCZec~

on over'Ql safety.

5u-h tud-ie ohaL>LC bz mae by ~&a licensees rid tbeir supp".ebs or can=-ulr~z

- 'Mls bg the?RC Staffs

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lieve" that it imuld be ~ssible asm desirabla W in1 e L~~i~q$y ia survey o~ op-rating p".ocedures Zor >chin'.$~ natu-43. ciJ.culaeiOh, jn-

. c1u9irg th". qg "e shen ofE~'i'Ee poua1 5.s 3.oct, aced the role of the pe~

serf.~~M h~aters in sufi procedures.

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The Cgam~t>ea b liGves that th s ~C'ter is receiv-inp ireful atvention hy the ~WC Staff ~ t-he "ice~ee.-

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0 April 17, 1979 "RKCGMHENDATIMS QF THE NUCLEAR RHGUlATORX CQ"dlISSIOM ADVISOR'Z CGC4XTl'EE

~ REAClQR GAFEGUARDS REGARDING 'THE MARCH 28'9/9 ACC'IDENT A~

TiiE THREE lhIEZ ISLAIK) NUCLEAR STATIOM QUIT 2 l

1 Presented.orally to, and:discussed with, the HRC Commissioners during the ACBS-Comoissioners Meeting on Apzil. 17, 1979 - washington, D

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, Natural circulatiqn is an important mode of reactor cooling, both as

'a planned process ar4 as a process that'may be used under abnormal circumstances.

The Committee belidves that greater understanding of this node-of cooling Is'*required and that detailed analyses should

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be developed by licensees or their suppliers.

The analyses should be supported, as necessary, by experiment.

Procedures.

should be de-veloped for initiating natural circulation in a, safe manner and for providing the operator with assuran'ce that circulation has, in fact, been established.

This may require installation of instrumentation to measure or indicate flow at low water velocity.

'The use of natural circulation for decay heat removal following a loss of offsite power sources requires the maintenance of a suitable over-

.pressure on the reactor coolant sgsten.

This overpressure may be

-assured by. placing the pressurizez.

heaters on a qualified onsite power source with a suitable arrangement of heaters and power distri-

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bution to provide redundant capability.

Prese'ntly operating H&

plants should be surveyed expeditiously,to determine whether such arrangements can be provided to assuie this aspect o

natural circula-tion capability.

The plant operator should be adequately informed at all times con-'

cerning the conditions

- of reactor'oolant system operation

~~Rich might affect the capability to place, the system in the natural circu-lation ado af oZerahxon ar to sustain evch

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hF. ~et->c >sl az.

iniportance is that information which-might indicate that the reactor coolant system is approaching the saturation pressure corresponding'o the core exit temperature.

This impending 1oss of'ystem. over-

, pressure will signal to the operator a possIble loss of natural

'irculation capability.

Such a warning may be derived from pressur-izer pressure instruments and hot leg temperatures in conjunction with conventional steam tables.

A suitable display of this Information should be provided to the plant operator at all times.

ln addition consideration should be given to the use of the flow exit tempera-tures from the fuel subassemblies, where available, as an additional indication of natural circulation.

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"-.The exit temperature of coolant frtm tl1e core is currently measured thermocouples in many Hgs to:determine core performance Ihe

-Committee recommends that these temperature measurements, as currently available, be used to guide the o abator concerning core status.

Me range of the information. displayed and recorded should include the

":full capability.'of the.thermocouples.

Xt is also recommended that other existing instrumentation be examined for its possible use in

assisting operating action during a,.transient.

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The ACRS recommends'hat operatirg power reactors be given priority

'ith regard to the definition and implementation of instrumentation

=.which provides'additional information to help diagnose and follow the course of a serious accident.

This should include improved sampling procedures under accident conditions and techniques to help provide improved guidance to offsite authorities, should this be needed.

The

, Committee recommends that a

phased implementation approach be em-

'ployed so that techniques can be adapted shortly after they are judged to be appropriate.

'Ihe ACRS recommends that a high priority be placed on the development and implementation of. safety research on the behavior of light water reactors during anomalous transients.,

Tne HRC may find, it appropriate

. 'to develop, a capability ta simulate a wide range of postulated tran-sient and accident conditions in order to gain increased insight inta

',.- measures which can be taken to. improve reactor safety The ACRS

-wishes to reiterate its previous reco~ndations that a,. high priority r =% giveh to research to improve reactor safety.

II Consideration should be given to the desirability of additional equipment status monitoring on various engineered safeguards features and their supporting services'o help assure their availability at 811 times i'Ihe ACRS is continuing its review of the implications of this accident

'and hope to provide further advice as it is develo~.

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