ML17296A359

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

ML17296A359
Person / Time
Site:	Palo Verde
Issue date:	04/21/1979
From:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
References
NUDOCS 7905080365
Download: ML17296A359 (28)

Text

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/p +~*~4 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V SUITE 202 ~ WALNUTCREEK PLA'ZA 1990 N. CALIFORNIABOULEVARD WALNUT CREEK, CALIFORNIA 94S96 April 21, 1979 Docket Nos.

50-528 50-529

~

50-530 Arizona Public Service Company P. 0.

Box 21666

Phoenix, Arizona 83036 Attention:

Mr.. E.

E.

Van Brunt, Jr.

Vice President, Construction Projects 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, Kgel en rector

Enclosure:

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

W. Hartley, APS

0

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 reactors designed by BQ(.

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

~(as 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 l<ilcox 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 feedwa'ter 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

'f 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

'ystem 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.

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 llilcox 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.

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 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 AprH 21, 1979 Page 3 of 4 safet features will result in unsafe lant conditions'or 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 actuated 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 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~

3 ~

Following detailed analysis, describe the modifications to design and procedures which you have implemented to assu're the reduction of the likelihood of automatic actuation of'he 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 B&W 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.

1 ow OTSG level f.

low pressurizer level.

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

?.

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

Response

schedule for B8M 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 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,

'Ilashington, 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 BQl COl%UNICATION - RECEIVED BY HRC Enclosure

'I

. IIWRONniph 4/20/79 Page 3 of 4 THE C'0)ITIWUEBG ~gIQg Op ~p sEqgpl)lCf of EvptiTS l QPDItlg TO 'pig IticIUptiy THE-2 ON bNRQl M. 3979 SHOWS TllYIT ACTIOti Cll PE TAYEti VO PROVIDE ASSuaW<rE.

78A~ VnE PILOT-OPaR~TEO ~ELIEF VALVE (PORV) teutnED ai THE PRESSURIZER OF eà PLAN75 NLL HOT BK ACTUATED QY AtfTICIPATEP TfW<SIE'ITS M/IClf i<AVE OCCURRED 08

'AVE 0 SIGNIFIC@%'ROBABILITY GF OGCVBPING Iti TllESE PLA>>ITS THIS ACT IO)'t l'/VS'ff.

OMMOE THE SAFETY OF PIE AFFECTEQ PLANTS HITli RESPECT TO TltKIR RESPO.']SE IMemt, UPSH'B ACCKOEm COnVITIOtlS NOR LFAO TO U~RFVIEQED SAFETY CotiCERNS

.THAN A,jTXCKPATKO iRAMIEtAS OF COtlCEQt glRK-ROSS W EXV'ihtgi itXa'eImi. l.am 2.

mmRNR 'rsiP..

8 I QSS OF PAIN FFELQATER 2.958 Qf'OBQENSKR VACUA recvee'zm'tosvRc oF reIo sTmH isoLATIoN vnLVFS (nsIv).

0 kiNSFR OF ALTERlQTKVES MERE C0%IDERKD IH DEYELOPItlG THE ACTIONS PRQ70SEQ SBW XNC4UOSHS.

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~RICTNC REACTOR PQ'AGL TO A YAL)UK MUNICH &Ut 0 ASSUAE NO ACTUATloti OF TQE PQAV THE. REACTOR PBOTKCTIO)l 5YSTEH>>

DES?GN PRESSURE AtlP PORY SET-POIW'5 AEttAIHEO AT THEIR MRREHT VAk.UES.

eowRIxs THk HIGz PRvssu~k mhcro8 TRIP SETvoItir To A vh<uc mlcll souLD ASSURE hQ ACTUATION OF THE PORV.

THE DESI'RESSURE QF Tlkj REACTOR Nlo 78K SETPOINT FOR PORV ACTUATION W&rNED AT PtEJR CURRENT VALVES.

66MRIHG TfK BKGH PRESSURE REACTOR TR3P SETPOINT NN I~MUSTIHG Tl)E OPERAiiNG PRESSURE (RNO TEhVERATVREj QF T)fE BEAl:TOR TO ASSURE tl0 PQRV

- ACTUATKOd N]0 'N PROVIOE AORgVATE tV~BGIH TO ACCORQDATE VflRIATIONS IH 0&mTInS PRESS'HE SHvOI>iT FOR Po~v AeruATIOtt REret t~ED AT ITS vomer VXI.uZ.

YHSS ALVEeimIVE kiOm.O REDUCE riET ELECrnlCAL OurpuT.

P MUSTLF,~ THE SIGH PRESSURE TRIP Nia THE PORV 5ETPOInfS To ASSURE nO

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PMV ACTrNTIOH FOR THE t.V5S OF AHTICIPATEO EVEnrs OF CO'<CERN.

THE 0Ãlmt PRESSURE OF THE REACTOR REFINED AT ITS CUAAEtli VALUE

-P4 ANLVSIS OF THE &PACT OF THESE VARIOUS ALTERMTIVES NiD THEIR CONTAIOUTIO;i 79 ASSURING THAT THE PORV MID. FlOT AC'FUATE FOR THE CLASS OF JWHCIPATEO THl,tlsrEnT5 OF %ACERB'AS BEER QPPLFTKO.

'ALE RESULTS SH(M THAT.".

I.C'%VSit8 Tm HI% PaCSSum REACT0R TRIP SETPOIN VRON 2355 PSM 70 2390 PSXG AD PA'ISING THF SETPQXN'OR AfE PILOT '.OPERATED REt,IEF VALVE BNH ZRSS PSIG m 2C5O PSIC

)'.

'gl3VIDES THE REQUIRe.O A55USWCE THIS ACTEO'l HIS THE FURTHER ADVI'tlTA<PS OV:

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FKTQCT OF BRW C08%HICATION - RECEIVED BY NRC 4/20/79 Page 2 oV 4 I

MtSCKtR THE PROBN3KETf QF PORV WD ASRE CODE PRESSURIZER 5AFETV N'L.VK ACAQVNN Faa Okra LNCREASrtQ PRESSURE TrNWSIEnrS.

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PRESERVING PRHSURE RKIEF CAPACITY -FOR At.t. HIGH PRESSURE Tel')SIFNl'5 I

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BLiHINATLKCTHE NMIBkt.I7fbF INROOUCING UNREVIB<EO SAFETY t.artCF'WS.

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8HNCNS THE Tf)%, AT kQKH THE STMR SYSTEM HEAT 5Il/K MOULD BE LOST I5 iHK EVENT ENERCENt;7 FERN<'ATES FL.(N HEBE DELVED

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TAN5$FSN ES GIVER XN TBBLF 'f 89%fARf OF THE ENACT QF'THE PROPO".ED SKTPOINf CHA,')GES OH N.L Ar<TICIPATEO v, I

. '85f PkANS ARK CUABBN.Y GtPABLE OF At@BACK TG 154 OF. FULL POWER VPOH LOTS QF'.QQ)

QH TRIP OF THE TURBINE THIS CAPABILITY REQUIRES AGTUATLOff OF 'BtE PILOY-CPERATE9 RB:REF VAlYKS THE CAPABILITY INCREASES YHK RG K~iDILITY OF PO'cfER SQPPI.V 70 THE SISTER SY RETURlltR.SHE'UHITS TQ POWER GEOFRATIQtl fQAE QUICl',L$

AFKR THESE TRN/S1BfTS.

THE ACTION PROPPED ABOVE NELL RFqulRE Tlirt THE PM'QQR BE TRXPPEQ FOR 'g)ESE EVEHT5-HRC nOTE:

The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typified by the attached figure 3. which was developed by.

SSI for a loss of feedHater transient.

=:2

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VASt.Z tt Enclosure l Page 3 af 4 S~NRV OF PROTFU ECf ACAttfST PORV ACTvaT tQI$

PAOVKIKO BY PAQPOSEO SETPQlFJT CHA.'AGES FOR ALL AN'TCKPhTF.O 'TRAl)5iFi TS EXTRACT Oj'BW COlgUNICAT10N

~ RECEIVED BY NNC 4(20PR AINICIPATEO TRAHSIEYITS MHICH HAVE OCCURRED AT BQI PLAfiTS l0lD MNICll SIOULO raeNLT AETIOATE PDRV AT THE cURRENT SETPDIHT <zzss l sir>:

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YSQNK TRAP I.css OF ExTK$NL zt.ccTArMI. Lse c

Ress 0F AH FEKMATER 5.

K.OSS OF Cc;:trmEA vhcNS XNQSERTEfA'LOSURE QF %lV I

~I AmcrPATEO TRANSFmTS HHrcH HAYE dcGURRED PT 88M PLANTs e<D NlicH vQQEQ tRRAU.Y AcTUATE PQRv AT Tf)C PAQPQsEQ sETPQINT (z459 PsKG)

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.BHTSDPATEO'7RNSIEMS 49/XCH HAVE NOT OCCVRRFD AT BrM PLANES (LCM PROBA9kiQTf EVNTS) AND HrfICH HOUL9 RQRHALt.Y ACTUATE PORV AT THE I

CORREllT. SETPOINT (Z255 PSIG):

A; SOi~ amTeR. fan GROS VirADrui~eLS (hb~RWC m t~ratt n~CTrVrTY

. PORTAL GROtPS I'r OTHERMrsE PRIECTEO BY Br'LuX TALP).

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PM'lCKK~<D T~A~SrEhfS FILCH ~AVE <ST OCCDR~O W OSM Vg~r~rS-~<+ZmrmKm--

EVENTSj ANO krlICH MOULD ACTUATE THE PORT AT TllE PllOPOSEO SLTPOINT

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go;~ go)~ggt, goo GpotJp tlr~otNMALS (}lIQlf REACTIVITY "QBT))

tlOT'THFWrSe Pr.OTE'C'reO BY HIu FL4X TRIP) ~

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BSM COHNUHICAT)OH - RECEIVED BY NRC 4/20/79 O

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Peak. pressunzer pressure as a functi'on of RCS pressure trip setpoint for a loss of feedeater transient; for. expected condftions and. various

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UNITED STATES MUCLCAR BEG ULATORY CO4VilSSfON AOVjSORY CO/i'its'ilTTEE OMIAEACTOR SAFEGUARDS VYASHINGTON,io.C. 2b555 April 18, 2979 1

I

'EMORANDUM MR.

Chairman Hendrie

'ommissioner Oil&sky Commissioner Kennedy Commissioner Bradford Commissioner Aheax'ne was:

. R. P. Fraley, Exeputive Director Advisory Committe'e on Reactor Safeguards'Attached for your information and use is a copy of the recommenda-tions of the Advisory Committee on Reactor Safeguards munich vere orally presented to and discussed with you on April l7, l979 re-garding the recent accident at the Three Mile Island l4uclear Sta-tion Unit 2.

Executive Director

. Attachment".

Recomm 'ndations of the. NRC Mvisory Committee on Reactor'afeguards.

Re. the 3/28/79 Accident at The Three Mile Island Nuclear 'Station Unit 2 A

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MUCH.EAR FfEGVLAYORY CGA'!'AlSSIQN AD'v)SORY CONMtV7EE ON ABACTOR SAf=EGU'55 Vfb~ggi1tGTQH,.Do 0 Ma&5

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. AprLl 20, 3.979

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H Msgr-hie Victor Gilf.nsky Q'ti~ CQaixman U. S. Qpcleay Regvla".exy Co~ssien Kv 5inpton, DC 20555.

D sr Dy, GiiinSky:

T.'-.is letter is in Xeapvn-"e to your~ of April 1S, 1979

~~mich re;,ue-.eo.

"we he FCN notk@ Ne Cowi88fot~c'.C" ~a9iet@v if vc believe any of Q'Jr orgy x Bcchiiv nQBcioj& of E.p=iX l7:shou' b

actini u~n before our n m re~u3.=-rly scheduled m e-i~ a" Mien <Ie could pre," re r fozuel pe@Dr.

we coritpe dfs "u".SM this topic bg con"eren=e re3. epP~ne

~ c&lon'April 19 ard oZ ers the ollowing cow.ants.

A~1 of the redo->~ nuohions maQe by C?)e h~~

Kn fts*p~~eirg vJ,"4 the L>~aisg$ ohcrs on Azzj3, 27, 397g>

re gDaeric in nature: M apjly m Ql Pwks Non@ 48 8 LntepQed to recruit'e

-'>aeedi te chases in oper(:tire pro-egura" or ply m~Qifiealions oZ oeex'atSrg P~,

Suck cN~pes gjould MQ2 prQ y a ter tv8$~ of the Lr efrect on over z3.1, sa eely.

ie hoU<a bz ade by De licen ceo vzd their supp:ez ox coa"ulcer hrd by the M~ Greffi Ne Ca>~>., ttbe b lieves ~Net &. s

~WuDieD ~D..+g b b~ in ehe ne:;r utu e an a i:ire s"ale that will.not div rt the --

.>i~ StafD oy ~e if'~>~"ry yepz'c"-e~tatives from th>:ig t~sks zelpzi~ to

. 'h=- coolcove 'o-. Tare

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"- -.'.icons;" th'at it ieuld be ~@sible aru desir'~le to inftf Le i Mi~~>y

.a survey a~ op-rater@ procedures Zo achier~

natuva3. eiycula~ioh, in-c1upirg th'-. ~e shen o-"2-iS po~r is last, hard the role of the pe~

'wci~~r heaters in suM pra'cedvres; 1~ its rn"tJ~g cn %ril 16'ad 1'7J 1979, the Comu.ties di~-ussed.a>+

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April 17, 1979

'REX:CM~TXOHS QF THE NJCfZAR REGULAKIRY COiHllSSIOM ADVISORY CG%4ITTEE C8 REACTOR GAFEGUARDS RERRDING

'THE lVJKH 28@ 1979 ACCIDENT AT THE THREE MILE ISIAlK) HUCLZAR STATION &iXT2 I

Presented.orally toand discussed with, the NRC Commissioners during th'e ACRS-Com.nissioners Meeting on April.17, 1979:Yashington, D. C.

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Natural circulatiqn is an important mode of reactor cooling, both as, a planned pro ess ar4 as a process that may be used under abnormal circumstances The Committee'elieves that greater understandin'g of

,,this mode-of cooling is "required and that detailed analyses should

'be developed by licensees or their suppliers.

The analyses shou1d be

'upported',

as necessary, by experzment.

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'emoval 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 onsite power source with a suitable arrangement of heaters and power distri-bution tp provide redundant capabil ity.

Prese'ntly operating P%

- plants should be surveyed exped itiously to determine whether such arrangaaents can be provided to assuie this aspect of natural circula-tion capability.

" '?he plant'perator should be adequately informed at all times con-

. cerning the conditions ef reactor:

coolant system operation which

~-'".'ight affect the. capability to place the system in the natural circu-,

latxon mado csf eporatxon ar tn sustain eLv:h * ~R~.

AF. ~ri.ic t>lsz importance is that information which-might indicate that the reactor

. coolant system is approaching 'the saturation pressure corresponding'.to the core exit temperature.

This impending loss of system over-

',pressure.wil3.

signal to the operator a possible loss of natura1

'circulation capability.

Such a warning may be derived from pressur>>

I 'zer pressure instruments and hot leg temperatures in cnnjunction with

'.;': conventional steam tables.

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

En addition

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.consideration should be given to the use of the flow exit tempera-tures from. the fuel subassemblies,

&ere available,'s an additional

'ndication of natural circulation.

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The exit temperature of coolant fr

'.by thermocouples in many HRs to

'Committee rec'ommends that,these tern avaS.Iablei be used to guide the ope range of the information

. displayecf the core is curcently meaenred determine core performance.

9he perature measurements, as currently gator concerning core status.

The and recorded should include the

foll capability 'of the thermocouples.

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

~;assisting operating action duzing a.,transient.

I lhe ACRS recommends tahat operatirg power reactors be given priority

.with regard to the definition and implementation of instrumentation

'~" -:rwhich provides'additional information to help diagnose and follow the course of a serious accident.

This should include improved sampling procedures undex 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 adopted shortly aftez they are

, judged to be appropriate, The ACRS reccmmends that. a high priority he placed on the development:

and implementation of safety research on the. behavior of light water reactors during anomalous transients.

Tne NRC may find it appropriate to develop a capability to simulate a wide range of postulated tran>>

sient and accident conditions in order to gain increased insight: into measures which can be taken to improve reactor safety.

The ACRS wishes to-reitezat:e its previous recommendations that a high priority

-be given to rem'arch to improve reactoz'afety.

'Consideration should be given to the desirability of additional "equipment status monitoring on various engineered safeguards featuresi and their supporting services to help assure their availability at

'all times e

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,',i Th'e KRS is continuing its review of the implications of this accident

; '-. and hope to pzovMe'uz'ther advice as it is developer).

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