ML16341C345

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

ML16341C345
Person / Time
Site:	Diablo Canyon
Issue date:	04/21/1979
From:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	Crane P PACIFIC GAS & ELECTRIC CO.
References
NUDOCS 7905080366
Download: ML16341C345 (28)

Text

9I II II<<Oi Wp0 0

++*++

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V SUITE 202, WALNUTCREEK PLAZA.

1990 N. CALIFORNIA BOULEVARD WALNUT CREEK, CALIFORNIA 94596

Docket Nos.

50-275 50-323 Pacific Gas and Electric Company 77 Beale Street San Francisco,'California 94106 Attention:

Hr. Philip A. Crane, Jr.

Assistant General Counsel Gentlemen:

The enclosed Bulletin No.79-05B, is forwarded to you 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, R.

H.

ge ken D

ector

Enclosure:

IE Bulletin No.79-05B with Enclosure ACRS Recommendations to the Commi ss ion dated Apr i 1 18, 1979 and April 20, 1979 cc w/enclosures:

W. A. Raymond, PG&E J.

D. Worthington, PG&E R.

Ramsay, Diablo Canyon 790508M@b

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 April 21, 1979 IE Bulletin 79-058 NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Des'cription 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 reactors designed by B&W.

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 p'otential 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 505 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.

Other means of reducing the possibility of void formation in the reactor coolant system are:

i 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.-

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).

1.

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 natural circulation is initiated.

2.

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 maih feedwater and auxiliary feedwater are lost while in the natural circulation core cooling mode.

Modify the actions required in Item 4a and 4b of IE 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 if i

d i

'f i

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

below

'3.

b.

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

(1)

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

The HPI system has been in operation for 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.

Th~e de ree of subcoolin be ond 50 de 'rees F and the len th of time HPI is in o eration shall be limited b the ressure/

tern erature considerations for the vessel inte rit~."

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 PORY for the spec-trum of anticipated transients discussed by Burl 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'eactor 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.

E f

l'I

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

Provide for NRC approval' 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 controlled 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 BN< designed facilities:

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 o'perating, 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.

t 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'iven under a blanket clearance specifically for identified generic problems.

'C

0

~

~

EXTRACT OF BSM CONWNICATION - RECEIVED BY NRC

'nclosure

'l

'.;:.,tMBONCTZON Page

'l of 4

'. -THK eQHTli')UIBG REVIEW 0l-" THE 5EqUEHCE 'OF EVENTS 3.EADIrlG TO 'nlE ItlCIUEtB'T

-- %BE-2 ON INCAN 28. 'f979 SHtNS TllAT ACT$0rf Cei BE TnvEN vo vnovIDE nssuatulrE

'TRAY VHK PILOT'"OPERATED BELIEF VALVE (PORV) tSUHTED (N TllK PRESSURIZER OF 85M

-"PANTS MELL BQT BE ACTUATED 8'f NlTJCIPATEQ TMtl5IEt)TS M11Cll lNVE OCCUMKD OR

-PAVE A SIG)GFICABT PROBABILITf QF OKVRPIrlG Irt ptESE PLANTS.

THIS.ACTIO'l rtvST

';.@rr. Mama mZ Shmn OV &e rFFZnEi Penis HIT>> RESPECr TO THEIR R~SPO'l.-Z

'Ko KORQ'L, VISEl OR ACC'f06% COt/DITIQrlS NOR LEAD 70 01REVIEWED SAFETY C01)t:ERrlS.

,,;VHK AWKC'LPATEO TMNSIE%S OF COHCEm A%=

4.

t.l5S N-imXhmi BXCTVlmi. t.OAO TURBINE TRIP f.QSS OF MIN FEENATER e.

LeSS OF ace"rmKS VACuux 5

ENSVERKNT CLOSURF OF HAl8 STER ISOLA'IQN VALVES (HSIV)

"A ~ASSER OF ALTENNYKvEs ME& co%IOE%0 xH DEvEMpirlt'HE ncTrorls vaoposEO

BACH LrlCLUDLNB.

~RICING REACTOR PGAER TO A VAtUE REICH >10ULO ASSUAGE NO ACTUATION OF 78K POW.

THE. REACTOR PROTECTBPit'SYSTEH, DESIGr( PRESSUREAND PORV SFT-

.Porms AErmIrlEb Av THEIR mRRFrn. vhluEs.

~,2.. aowaIrls mk Hx~o pavssuns REhcTQN Talp sETPQINT To h vhLvc ~IC>> HovI.O ASSURE HQ ACTUATIOH OF THE PORY.

THE DESIGr) PRESSURE QF THE REnCTOR Nl9 THE 5ETPOKHT FOB PORV ACrUATIOB &HArnEO AT TllEJR CURRY)T VALUES.

-: 8.,:46':ERIN TVE VIGH PRmStNE RBtcTOR TRIP. SETPOIm ue rMVSTInG TilE O~ENTrRB VRESSuaK (Ne VEmERATVm~j Ov THE REACTOa TO ASSURE tiO VORV RCVUATKO(9 Arm To-PBOVLDK MEgUATE fViRGIH TO ACCOh igghTE VARIATlGHS IH

'mNrxrls pREssvRE mE sETPOIrlT FOR poav AcrUATIorl nc~mIrlzo AT Ivs

cvR&m vs.VE.

THxs mxzRrwvrve koULo REoucz rlET E<Ecrnlc~L OUrj uT

-4.

AMUsTxr," nE HIaH pBEssuRE TRIp ee THE povv sEv'POIrlvs To nssune no PMV AC'nNTIOH FOR THE CV5S QF AHTICIPATEO EVErlTS OF CO')CERN.

THE OESIQt PRESSURE OF THE REACTOR REHAINED AT ITS CURRANT VALUE "P4 @fANSKS OF THE APACE OF'HE5E VAREOVS ALTERNATIVES ArlD THEIR CONTRIUUTI'OB To ASSUBJM THAT THE PORlJ MILL NN ADUATE FOR THE CLASS OF JWI ICIVnTEO TRhr}5IEr)T5 OF ARCED HAS BEN CG%'LF'fEO.

THE RESULT5 SHM THAT."

ILQ'%RNQ THE HLER PRESSUlK ABACTOR TRIP SETPQI%

F805 2MS PSM TO 2390 PSXG N.O PZi1'SXVj~~ THY $ETPQXN FOR TlfE PllOT '.OPERATEO REl.-REF VALVE FMH 225S PSXG TO 2450 PS'EG t

pt@vTDEs THE AEQUI'Reo A55vfUwcE THIS AcTIoil HIS THE FURTlfER hDvnHTAGES QF.

I

t I

EKTBPCT OF BIIM COtOUHICATIQH - RECEIVED BY NRC 4/20/79 Enclosure l Page 2 of 4

'kEggDR THE PROBABD.KTY QF PORV AND A5HE CUD'RESSUREEFB 5AFETV VP~l.VK fCANYKOS FOa Oman NmmSrm PRESSuRF TeelSrexrS.

PKS@Vrie Vi~SSVRE REuEF'iWiETY.FOZ mL Hrm PRCSSUaE TrullSIEnTS.

8.

EiIH)NATik8THE POSSIBILITY OF IPfrROOVCInn U'lRFVIFl>ED SnVEn ConCF'mS.

4 4.

QENKTNS THE TIP% AT >.,lTCH THE STUN 5'tSTEH llEAT $ ErlK MOULD BF LOST IN iHK EVENT Ef%RCENCY FFFS"ATER FLPJ: NERF DELAYED.

'A 89%%BY OF %E'tPACf OF THE PROPOSED SETPOINT CHN>GES OH ALE. NiTICIPATEO

'7$VSEKSTS ES GIVEN XPk TBBLF. 'f

'-'648 PKAHTS ARE QJABENl.Y CAPABk.E OF RNfBACK.TO 154 OF, FULL POHEA UPON LOADS OF'k.QN)

OH TRIP QF, THE TlJABIBE THIS CAPABILITY REQUIRES ACTUMIOtl OF TllF. PILOT-

QPNrTEO REt

IEF V%YES.

THE CAlaBILKTY IrlCRWSES NF. ALAI'OiLITYOF POWER "SULU" 'N THE SYSTER 8V RETVi%')LING.THE Ui)ITS TO POWER GE~IERATIOW PNAE qUICKLJ E]RLR THESE TRA)SLm'S.

THE ACTION PROPOSED ABOVE MILL REqu1RE n)AT THE "tKAQMBE TfQPPFO FOR 'H)ESE EVENTS; f

~

NRC "NOTE:

'The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typified by the attached figure 1. which was developed by.

.BSW for a loss of feedwater transient.

~ ~

~

I'.".-":.

j. ~.~ ~ ~

7"9P l'hat;e 7

Enc'tosure

'l Page 3 of 4

$~%Kf OF PBOTKCTIOtf ACAll'6T PORV ACTUATtoil PAOVXOEO BY PAQPOSEO SETPOIHT CHA.'AGES FOR ALL MKICIPMEQ'TRAl)5IEi T5 EXTRACT QP 85M: CQQUNQQTION

- RECEIYE9 BY NgC 4(20/79.,

"."QVKEPATm TRANSEEfAS &rat HAVE OCCuRRED

@T eau PuulTS mlo elIVl VOuLO

."mdiv Acnvlim iahv m ~F. cuiUimr sun aiiir <zz~s r srG>:

'1 A

VQRBENR TRAP

~4.

I.OSS Oe Enzeuu..a.eCTAEmi.

Lme

.~c<

4055 0F RMH FEHNATER IL&S OF COit'~rEA YACC KECDVERTEfA CLOSURE QF HSIV

,"PZALCDATED TfQNSLENTS WHICH HAVF OCCURRED AT S8Q PLANTS elD REICH

'."L'Q@tl &iARAl.t.VACTUATE RORY AT Tf)E PAQPOSE6 SETPOINT {2450 PStG) =

3,,;RhTSCKPATEO YfNHSIErPS Sn>rCH HAVE NOT OCCURRED AT B~Q VLA.~rS.(LOW

';PMBPBit BY FVEHTS) rND HHECH NOI.O temALLY ACruATE PORY AT THE

~

~

l

.;.CtfBxnY. SETPOXRT (ZZ55 PSIQ):

iA

'."88% CNTfKH. AGD GROUP HITHDHAMALS (NOOFRATF TO llIG!l REACTIVITY

. BCRTH'GROfPS F07 OTHEf5tISE PROTRACTED BY HIGH FLVX TRIP)

<<8 MQ~c TOR DILUTED'.

S l

SORE CONTROL ROD GROUP l/ITHDPA'PALS (llIGll AEACTIYlTY '..'OBTll l,'OT I

'OTHEfQISE PPOTECTEO BY HECT FLtlX TRIP).

l Y

I4

. AN~CE!AZO Ta><S?EN'S t.alCH HAVE riOr OCCiSarxg AT DSM PLAr>TS-KUH-PKH/93IQ7r--

jlEVKHTS) ANp REICH 'hOUt D 3"GTV/LTE TIVE PORV AT TllE PAOP05FD Sl. TPOIslT

~;l'ZCSO P576) ~

I~ ~

)

EXTRACT OF 85Ã COHNUNl'CAT)ON -" RFCELYED BY NRC 4rZOl'7g Enclosure 1

Page 4 of 4 ical

)

I r

~

)

~

~

~

4 e

~

j

~

~

~

1

~wt

~

I 1 1

~

~

~

~ ~

~

~

'eer

}

I '

~

~

~

)

~

)

~

)

~

I.

0 1 1

1

~

I

~ 1

~

~

,e.

~ 4>>

~

0 I

~

4 '

err

~

~

1

~ e

~

ZB,'OP

~

~

~ er

~ <<<<e v) 1

~

~

~

~ ee I

~re))L ~ <<e e>>

0<<7

~ 1 AM v rw 1

<<0 1

<<,Q e <<re))<<11 <<e 1 ~

gp rr<<e I

~ e

~ r

~

~

~

1~

)

I..f Pe- -emu>>I ivyH.

l

'ASMmo0 AS'f N/c7~

4tl~~-"--

~

~

I I

~

~

~

1 I

Vr

~

' 1 r 1

~

~

e

~

~

~

~

.I

~

~

~

e ei

~

1 1

~

e

~

~

~

~

~

~

~

~

~

\\

0 er f e

~

~ >>

~

~

~

~

~

r e

1 1

e 1

~ ~

~

~

~

e'

~

~

~ ~ 1

~

~ e ~ 1 j-

~

)

e

~

)'

0 ),-

~ ~

e

, ~

1

) 1

)

~

~

• ~

~

~

~

~ )

~

<<10

~

~

~ ~

~

~

~ V e 11

~

e i

1 e) reer

~

~

~

~

~

~

~

~

~ )

~

~

~ ~

~

I

~

)

~

I

~

~

s e ~

t-1 ~

1 1 v

~ ~ ~

~

~

1

~

~

~ ~iM I

~

~

~ e

~

z

~ ~

1

~

~

~

~

~

~I

~

~

)

~

~

)

~

~

~

I

~

~

g 1

~

1

~

l l

~ 0 1 e

~

~

~

~

4

~

~ >>

~

~

v ~

~

~

~

~

~

~

~

~f 1

~ 0 ee

~

rr e'

~

I

~

~

r

~..

~ }

~

~ I I

e ~

~

~

~

eee ee

~

~

~ 1

~

~......e

. ~.l

~

~

~

e

~

~

g..

}~,) ~.

~ e <<~

1<<<<e e<>e e1 e

1

~

~

~ I

~

~

~, r I

I

-neN<

~

} ~ ~ 0<<

f

~

po~w.

~

1 fe<<~

1 ~ VI.

~ ~ >>e

~

~

~ <<

~ ~

e !~

~ 0

~

~'

4

<<e

~ ~

~ re<<e

~ e

)

~

~ I'4

~ e<<

~

T

<<1 er 1

~ 0

~

1

~

e

~

~<<<<

~

e

~

1

~

)

~

ee 1 t

I

~

4

~

e<<e>>

e

~

~~1 e<<

~

1

~

f

~ q ee 0

}

~ 1

'} ~

r

~

~

e ~

}

)

~

~ )

~

~

1

~ e

~

1

~

ee ~

~

~

~

e)

~ ee

~

I

~

~ ~ ~

)

~

~

~

~

ee tee

~

~

~ 0 1

~

~

~

~

~

~

~ 0

<<0>><<e r ) 1.1

~

~

~

~

~

~ ~ ee

~ ~

~ ~

1

~

~

~

~

~ j ~ r

~

ra l

~ re

~

~

~

~

1 e ~ ef 1vre

~ ~

1 0<<<<

e}.

~

~ r<<4. ~

~

~ ee

~

~ 1 1

~

~ ~

~

~ e

~ e

~.1 1 )>><<1 1

~

~e ~

r 1

~

~

~

~

~

~

~

~

I e

'I

~

~

~

1 <<

~

~

~ 1 I 1

~

~

)IS

-L ~

~ e

~

~

1

~

~

~

e

~

~

~

1

~ L e

p<a8 e~

T4( 0 $ El I os'.z~~

RKPQ

~

~

~

e ~

1 I

1 ~

~

~

~

1'

'e

~ e ~

~

~ r 1

~

1

~

~

~

e I e 4)

~

~

~

~

~

~ '

~

V

~ r ~

~

~

go@

t

~

~

~<<4 eS P

~u PRG4'Q 6'WWMWx.,

~

I

2400

~

~

)

)

~

~

~

~

~

~ )

~ }

e*'

'eak. pressurizer pressure as a functi'on of RCS.pressure trip setpoint for a -loss of feedwater transient for. expected conditions and various initial pressures.

Figure 1

r g4'~ ".

gyve-vy O~

"<.e,a*+

UihlTED STATES NUCLEAR REGULATORY CQl"rVilSSlON

, AOVlSORY COVI,',lTTEE OM'REACTOR SAFEGUARDS WASHINGTONio.C. 20555, April 18~ 1979

~

FAGER HEHORANDUH FOR.

Chairman Hendrie

~ Commissioner Gili'osky Corravissioner Kennedy Commissioner Bradford Commissioner Aheagne I

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

tions of the Advisory Committee on Reactor Safeguards,u,"sich vere orally presented to and discussed vith you on April I7, 1979 re-garding the recent accMent at the Three Mile Xsland Nuc'ear Sta-tion Unit 2.

Executive Director

. Attachment".

Recon@ ndations of the HRC Mvisorj Committee on Reactor Safeguards Re. the 3/28/79 Accident at 'Ihe Three MiIe Xsland Nuclear 'Station Unit 2

~

~I

~

~

~

~

~

P 4~

Nk

'~+ A

'l=~~.j..

E'A y4 UNJTEO STATES f'lUCLEAit BEGVLATORY CGViMlSS)QN AovlsonY cohtMCTEa os REAcToR sr.rvsU-'rie $

nw~rlnlGTQH. 0 4 ~m5 II

. Ap~~l 2Qr 3.979

~ ~

C I

P ~

4

,Mraz'-hie Victor GilLnsky h'bsp Chnirzen AJ. S. HpeleeY ~egQXB"a~ Co~ssion Kh-NfngM~, DC 20555.

.B Pr DY+ Giiinskg:

P.-.xs letter s in X'0"pan"e to yours of April 18, l979 ~a'ch re'ue-eg

"~e, he ~CrB not5,(y the Co~~iaeior~c'.e-im~iaiely if ve believe any of oJ." ogpu z'eco~naation" oE Ao il 3.7-'s-auld M ectm u~n befog@ o<~

nm regula,=-"ly "chMu1ed m e"i' Q.icn

<re cou2.d pre,"

Ye r. @tung

,leBex.

Re Cc~"-'ttoe dfs"ussM.this topic bg con er'er"e te3ephoge col on'~p<il 19 aud offers the follovirrg comp~nts.

A~1 'of ge'eer"p r.deCTaM neQe by tive hM~ fn Its P.aatirg b'i~J1 the Cont.is8$ oh" ~te g~weric in nature aced GDQly to'Q3, I PINR5)

H)no v&g0 il)QBJYAQ to revuirl 'Nlrldia"t6 chic ries in op<:w0Qzq t;MUlos oe pl~" mvQificaQians of operating Pm~.

Such eh~yes Wou>8, bs aude orQy a ter 5V5p~ of their effect on over~le sa ety.

Sp-h

~Qe by ~g licen~eeg

@jQ th'QLz'upp~; QY5 of con-ultenjc.

Grd bJ the 1R~ Staff>

N~~ CoMttpe 5"=3.ieves R4at &. se

~&@Dies ~~. iQQ

~z-n in the: ne,:g couture an a l:hr-.e <~"Ble that vill now aiv4rt the

.)i~ Htaf~ or the indw~"."ry repro"-e,.tatives from theb t~sks relatiryp go th= cooloomi o=, Tax'cw ~tile Xslanl Unit 2; Mvever, th Ccwi~iigie<.be

.llew(c thai h skuld te ~@sible aru desir'able N xnan L'i~ate ia~MiaQp$y e survey af op-ratirg p."ocedures So." achievit@ natuek1 ciyculee.4g, j,n-cXuPirg th'-. gg ~e ~%en of%a'ice ~~ r is 3.a"t, arr3 the role of the pea~

s>ri.~~8 heeters in suM procedures.

i~8 re "trig cn M~il 1G and 17, 1979> t'e Co~~'ttee di~uss~d.c>

the KRC S~~<f ~bc ram ter cX natu.al c rculatfon for'ne i'hYP.

Yig.e Xs-lc.c) Unky 2 p3.ant.

We Cenm~ttee b litotes that Wi~ mtter is receiv-inp c-reful stvention by the:&C Staff ~ aha liccr ee.

L

~

~

C'::.'::::::

To EÃ far.r'.pp>wpriate Action.

Distv~butign;

-'j}P~,. GIAN R'p" faxed CO fDQ PA e

.C~se

~C 4

P> ~

'I '

J~l, J'-

~

V a)

I

'T~r>>>ereblo Vietay Gills +

gait 20,.19,"9 4~pg JLtgM5 Q pW?1 WBC('Lf4l8[MBCXOp9 r~r.~ mten8ed to apply t,o Three HVe X&and Co~oiQs.'.oh. un by<iX 37 %are Vnit, 2.

~"'""",

-z'& pith hO QX'i'd a fu"t')eZ Va~IT. vTL &e"8 ZGGC~BX's G" eu hay 10~ 3979

<<gC~4 j.Ap~

BLflCCZ'Kl.pg

!ham li.

"z'ban D'.Ql.MBA "e"

I V 4 tl I>

~

Apr 17> 1979

'RECOHHEMMTXMS QF THE NUCLEAR BHGUkAKIRY CCiÃXSSXOM ADVISORY CQi%4XTTEE N REACIQR SAFEGUARDS. RECABDXNG THE PARCH 28'979 ACCiDENT AT TrE THAEB HXLE ISQQK) N0CLEAR STATXOM UNIT 2 l

Presented orally to, and:discussed with, the NRC Commissioners during the ACRS-Com.nissioners Heeting on April,17; 1979 '- t ashingtonp D

C.

Natural circulatiqn is an important mode of reactor cooling, both as a planned pro ess and as a p'roress that may be used under abnormal circumstances.

The. Committee believes that greater understanding of this nede of cooling is required and that detailed analyses should

.be deyeloped by licensees'r their. suppliers.

The analyses should be

".supported, as necessary, by experxment.

Procedures should be de-

"veloped for initiating natural circulation in a safe manner and for

.. providing the operator with assurance that circulation has, in fact/

~been established; This may require installation of instrumentation to

,'measure or indicate. flow at low watex'elocity.

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

" system.

This overpressure may be

<<assured by placing the pressurizer heaters on a qualified'nsite "power source with a "uitable arrangement of heaters and power distri-

"bution tp provide redundant capabili ty.

Prese'ntly op rating H&

plants should be surveyed expedi tiously to determine whether such arrangaaents can be provided to assure this aspect of natural circula-

tion capability.

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

'erning the conditions of reactor: coolant'ystem operation (v'sich, might affect the capability to place the system in the natural circu-

.lation

@ada af oporatxoa ar ho unhain ccv..h

~ neRs AF. ~rl. ic

>s'tatimportance is that information which-might indicate that the reactor coolant systera is approaching the saturation pressure corresponding' to the core exit temperature.

This impending loss of system over-pressure will signal to the operator a possible loss of natural circulation 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.

Xn 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.

'4

,'We exit temperatuxe of coolant frhn the coxe is currently measured thermocouples in many HRs to,determine cox'e performance.

The-

'Committee recommends that these temperature measurements, as currently

~vailable, be used to guide the o gator concerning core status.

'The range of the information displayed and x'ecorded should include the full capability of the thermocouplhs.

Xt is also recon"vended that other existing instrumentation be examined for its possible use in assisting operating action during a,.transient, I

The ACRS recommends. that operatirg power reactors be 'given priority with 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 gprocedures undex accident conditions and techniques to help pxovide improved guidance to offsite authorities, should this be needed.

The

.'Committee x'ecommends 'hat a 'phased implementation

'approach be em-,

ployed so that techniques can be adopted shortly after they are

'judged to be appropriate.

The ACRB recomm nds that' high priority be placed on the developnent and implementation of safety research on the behaviox of light water

x'eactoxs during anomalous transients.

Tne NRC may find it appr'opriate

'.to develop a capability to simulate a wide ra@pe of postulated tr'an-

,sient and accident conditions in order to gain increased insight into meas'ures which can be taken to improve reactor safety.

The ASS wishes to reit~rate its previous recommendations that a high priority

be given to rematch to improve reactor safety.

Consideration should be given to the desirability of additional equipment status monitoring on various engineered safeguards features

.and their supporting services to help assure their availability 'at

,all times.

"'The ACRB is continuing it" review of the implications of this accident

.;and hope to provide further advice as it is developed.

s>

I