ML14097A394
| ML14097A394 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Crane |
| Issue date: | 04/21/1979 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML14097A395 | List: |
| References | |
| IEB-79-05B, IEB-79-5B, NUDOCS 7905110186 | |
| Download: ML14097A394 (13) | |
Text
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 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 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.
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.
7905// o /
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 50OF 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 operation of engineered
IE Bulletin 79-05B April 21, 1979 Page 3 of 4 safety features will result in unsafe plant conditions. For example, if continued operation of engineered safety features would threaten reactor vessel integrity 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.
The degree of subcooling beyond 50 degrees F and the length of time HPI is in operation shall be limited by the pressure/
temperature considerations for the vessel integrity."
- 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 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.
- c.
main Steam Isolation Valve closure
- d.
- e.
low OTSG level
- f.
low pressurizer level.
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 controlled or expected condition of operation. Further, at that time an open continuous communication channel shall be established and maintained with NRC.
- 7. Propose changes, as required, to those technical specifications-which must be modified as a result of your implementing the above items.
Response schedule for B&W 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, Washington, 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 B&W COMMUNICATION -
RECEIVED BY NRC
__DUCT-10 4/20/79 Page I of 4 A SIGNI AEVIE OF THE SEqUENCE OF EVENTS LEADVG TO TlE ICIUE AT 1-2 ON PL4RcJ 2!8 1979 $IqOWS THAT ACTION~ CAT 13E TAK.EN TO PROVIDt ASSLJR~trE YHAT THE PILOT-OPERATED RELIEF VALVE (PORV)
PlOWITEtD OW THlE PRESSURIZER OF Bpi iLA?4r VILL NOT-BE ACTUATED SY ANTCIcPATED TRiAUSIEtiTS I'iIICl lHAVE OCCURED OR
'RAYEA swiGNIiCAF PROBABlILIT-Y or GCCURRIING IN TIIESE PLANPTS.
THIS ACrzO,8 MUr~l OT GRADE THE 5APETYCOF THE AFFECTE PLANtS WITS RESPECT TO TIR E
ATIPL9 ASTOR ACCIENT CONITIORS NOR LEAD TO UNREVIEWED SAFETY CONCERNS.
- AwrTclPATE0U TRM~S IEWkfS Ot COUCER71 ARE:
LoSs 0r AETE AL ELECTRICAL LOAD
- 2.
'S.
LOSS OF RAIN FEEDLATER
-4.
LOSS OF CONDEfSER VACUUM IMADVERTENT CLOSURE OF MAIN STUA!
ISOLATION VALVES ($TEA)M WER OF AtTEfUATIVEs WERE CCnSIDER0 IN DEVELOPING THE ACTIONS PROPOSED t~f4INCLUDING:
6 VSTRlCTXP REACTOR PU TO A VALUE WN1CH WOULD ASSURE NO ACTUATION OF llir PORV.
THE. REACTOR PROTfECTIONl SYSTEM,. DESI.G PRESSURE AWfQ PORY SET VZO1iTS REMINED AT THEIR CURRENT VALUES.D SSURE E HIGH PRFs RE REACTOR TRIP SErPOINT TO A VALUE 14I[ClI WOULD ASURENV.-ACTUATIOrN Or THE PORY.
T14E thSIGti PRESSURE OF TilE REACTOR AN~D
.. NE SETPOINT FUR PORY ACTUATION REMAINED AT THEIR CURRET VALUES.
.b. AKMttERT!G TME HIGH PRESURE REACT09 TRIP SETPOIriT ArmU nAUsUTfING TlE
-O0PERATING PRESSURE (APD TE)'?ERATURE)
OF THlE REACTOR TO ASSURE NO PORV EACATINGA AN TO PRSVSE E AEQUATIE MARGIN TO ACCOM7MtODATE VARIATIONS In
,OPERZATIN PRESSURE-.
THI SETFOINT FOR PURY ACTUATION REPM1INED AT ITS
'URR~EfT VALUE.
THIS ALTERNATIVE WOULPO REUCE FET ELECRICAL OUTPUTr.
Ans THt Dl EE PRESSURE TRIP ANU THE PORY SETPOINITS TO ASSURE N
~POV ACTUATIo3, FOR THE CLASS OF A?TICIPATED EVEHrS OF COUCERN.
THE DESIGN' liRESSURE OF THE REACTOR REMAINED AT ITS CUENT VALUE.
T9,,;AMLYSlS OF~ THE IN'AC OF THESE VARIOUS ALTERATIVES AtID THEIR CONTRIUUTIO13 TO ASSWMrG THAT THE PGRY W~ILL Nr ACTUATE FOR THE CLASS OF ;AVTICIPATED TRANSlEIETS OWF-tCER9 HlAS BEEN CW~LErE,.
THE RESULTS 'SHOW
-MAT; NLOtiEkr THE lli(-
ESSuRE REACTOR TRIP SETPOIIT FROM 2355 PSIG TO 20 PSIG A17D P5i THE SETP00T FOR THE PILOT OPERATED RELIEF VALVE
THIS ACTION HAS THE FURTHER ADVANTAGES Or
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EXTRACT OF B&W COMMUNICATION - RECEIVED BY NRC 4/20/79 Page 2 of 4 m
RG T FE PROBABIRI0Y EOF PRY ND M E CODE PRESSURIZER SAFETY VALVE 4gACjMIUM POR OTHiER INCREASING PRE5SURE TRAflSrEWis.
PRS VING PRESSURE RELIEF CAPACITY -FOR ALL HIGH PRESSURE TRAWSIENTS.
Min riks ws$sitD TY O) I~NTRODUCING UNREVIEWED SAFETY CONCERNS.
(WU ING THE TIM AT tP4CH THE STEA SYSTEN HEAT 5 tlK WOULD HE LOST I TH EVENT EMERGEN4CY FEED IATER FLOW WERE DELAYEDL RA1Y OF ME IWAC. OPL THE PROPOSED STPOINT CHANGES ON ALL ANTICIPATED NYSETS is GIVES It? TABLE 7.
L A
E CURRET LY CAF ATt F OF Rt8ACK TO 15% OF FULL POWER UPON LOSS OF LWW OR TRIP OF VAE TURBHIE.
THIS CAPABLITY REQUIRES ACTUATIOrf OF THE PILOT OEltRATEo REL.IEF VALVES.
THE CAPABILITV IrICREA$SS THE P-ZLABILITY OF PURJER SUPPLY TO THE SYSTEM BY RETURNING THE UNITS TO POER GEER.ATOI OI E QUICKLY AMER THESE T SIElTS.
THE ACTIOn PROPOSED A80VE UILL REQUIRE T EAT IKE MA~CTR BE TRIPPEO FOR THESE EVENTfS,
':RCOTE:
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 8&W for a loss of feedwater transient.
TA-tE lI Enclosure I StiVtfRY OF POTECTIOn AGAINST PORY ACTUATrtri Page 3 of 4 PROVIDED BY PROPOSED SETPOInT CHANGES FOR ALL.
A"TICPATEu TRASIEjT5 EXTRACT OF B&OCQUNICATION RECEIVED BY NRC 4-20/79 ICIPATED TRAMSIENTS WiICH HAVE OCCURREti AT B&W PLANTS A'D MIf1 WOULD Dit.u ACTIVATE PORV AT THE CURRMNT SETPOrT (2255 PSIG)*
A. TW1N TRIP LOSS OF EXTEWAAL ELECTRICAL LAD LoSS OF MIN FEaEaTER LOSS OF CO'DENSER VACUtm fiADVERTEffT CLOSURE OF MIV AWTICIPATED TRANSIENTS WHICH HAVE OCCURRED AT B&W PLANTS AND MtCH
,uLD KORRALLY ACTUATE PORV AT THt PROPOSEo SETPOrr (2450 PSIG):
ffIMflCIPATED TRANSIErff HICH HAVE NOT OCCURRED AT B&W PLAN'IrS (LOv RD0AILITY EVENTS)
AND HNICH kWOULD NORMALLY ACTUATE PORY AT THE CUR SETPOINT (2255 PSIG)
CSOM coNTROL VZD GROUP WITHDRAWALS (MODERATE TO HIG11 REACTtVITY
. OThGROrPS fIT OTHERWIsE PROTECTED BY HIGH FLUX TRIP).
t0IEWRATOR DILUTION.
rATED T.-.51ENTS MHICH HAVE POT OCCURRED AT B&W P LAITS~
LUTTA EVERTS) AND WHICHM OU1LD ACTUATE THE PORV AT TJJE PLOPOSEDj SETPOIT 4(2450 PSIG)
L. SOME CMuROL ROD GROUP IITHDPAALS (HITG REACTIVITY 'VORTH1 NOT OTHERMISE PROTECTED BY HIG FLUX TRIP).
0 g
Page 4 of 4 EXTRACT OF B&W COMM1NICATION RECEIVED BY NRC P
4/20/79 r~pc se nct oe r
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UNITED STATES NUCLEAR REGULATORY COMMISSION AOVISORY COM.MITTEE ON REACTOR SAFEGUARDS WASHINGTON,;D. C. 20555
,pril 18, 1979 MEMORANDUM FR Chairman Hendrie Commissioner GilA sky Commissioner Kennedy Commissioner Bradford Commissioner. Ahearne FROM R. F. Fraley, Executive Director Advisory Cowmittee on Reactor Safeguards Attached for your information and use is a copy of the recommenda tions of the Advisory Committee on Reactor Safeguards which were orally presented to and discussed with you on April 17, 1979 re garding the recent accident at the Three Mile Island Nuclear Sta tiol Unit 2.
.R.
Fraley Executive Directov
Attachment:
Recommendations of the NRC Advisory Committee on Reactor Safeguards Re. the 3/28/79 Accident at The Three Mile Island Nuclear Station Unit 2
~
~
UNITED STATC-4 MUCLEAR PREGV~LATQFRY COMMISSION Aivisony CO.' MIrfEE DN REACTOR SAI-EGUADA Apt~il 20, 1979
- A~t1~victor' ciL1nsky
- aoMngon, DC 20.55S5....
DarDr+ Gilirisky:
T.6s-letter is in rCes1,r.9 to yourfi ofL Ap1
- 97 -Mc,,uea-e t
th
- CA r~otiry the Cm s 1 i mdiaftey
- i. L yo ouJr ort rcc, -r,3t in o f Ap'ril 27-' should be ac.t ' Upon hafore Our
-nec rjularl y sh~ce-i eetzirr-at; vHAch t oldpr r r~t)
-lefter.
The Co>mittee discus_-d this - pic 1 y cori ;eren>
e te2phora al oTVA!zpri1 19 and offers the flIowiny ;
.4 o I tha racori;.-r 6tfons mnade by the J~~In It meting Wi th thle
- ~ kicson kipril 17, 1979t are qirneric in nasture =t to ly to all
~ks ?~na
-ee -intende to rezuire ngT&it hm~ nor~i-pro 1;trmw, or plemnt mdiflcations of oce)ia'tirg PR. s1uch Cb~xes shouJd h6 =~de on3y after Etta' o their effects on overa~ll szlfety.
spch,.tud ie-iLi~ x~m.ide by the licenear and Lh; ir supplisrs or cuar bWl1y the~ MC
- Staff, 7b* Coite b'aieas that th.2s-_ :tud5ie,- zoi).
bs b-v in thL-re -r future on a 1tzrw_ Sz:ale thazt will riot; d lvtrt the
')K Staff or the inydustz-y rtprli -ent-ative_8 from their tixks ralatir~q tto:7:
th ~
~
J Uo~or ~~e~NIehad nit 2, Buwevertbe tt be 13 Survey 04 OP6 ratirjg P~rocedurez for aqhivrG nauval Circulatinr cluiirtg thL-cA-,e when 0'f:Fs'1*Ee pow-r is 3.ost, ar thce roCle of th pre7=
'Lvrize2r heaters in sucaf.~6cdrs 7;'. its cn A0-1 16 and 27.,'1979, thq CoMUtt-tee 41scu~SS;d.%.ithb th z St'aff thbe raatr o'f natux'al Cimile.~don for the 1Thr e ai~. Is*
Unit 2 pla.nC.~ The C.mitttt b,_&i(,-VeS that thi, m rtter is receiv inorC-fUj. StZErtjon by tho %'jG Staff an-d tbeL 2iC;(-njee......
T L ED3 *or Appr-opriate Action. Distribution:
DPi'D 01A.
E. Cse.
7-17..
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April 17, 1979 RECOMENDATIONS OF TH8 NUCLEAR REGULATORY CO'4ISSION ADVISORY CQMMIrTTEE CN REACTOR SAFEGUARDS PGARDING THS MARCH 28, 1979 ACOIDENT AT THE THREE MILE ISLIAND NUCLEAR STATION UNIT 2 Presented orally to, and discussed with, the NRC commissioners during the ACRS-Commissioners Meeting on April.17, 1979 -Washington, D. C.
Natural circulatiqn is an Important mode of reactor cooling, both as a planned process and as a process that may be used under abnormal circumstances.
The Comittee believes that greater understanding of this mode of cooling Is required and that detailed analyses should.
tbe 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 assurance 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 4pressure 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 to provide redundant capability.
Presently operating PA plants should be surveyed expeditfously to determine whether such Zarrangements can be provided to assute this aspect of natural circula tion. capability.
The plant operator should be adequately informed at all times con cerning the conditions of reactor coolant system operation which might affect the capability to place the system in the natural cicu lation a
nE 6perationn or to sustain skxch A mae.
Or piArr-inr
-.-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 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.
In 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.
t
~The exit temperature of coolant from the core is currently measured "by thermocouples in many PWRs to idetermnine core performance.
'The
~Committee recommends that these temperature measurements, as currently available, be used to guide the operator concerning core status.
The range, of the information displayed and recorded should include the full capability -of the thermocouples.
It is also reconvnended that other eXIsting instrumentation be examined for its possible use in
~assisting operating action during a transient.
'The ACRS recommends that operating power reactors be given priority
.with regard to the definition and imiplementation of instrumentationf
- which provides additional information to help diagnose ard follow the
~course of a serious accident.
This should include improved samnpling
~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 adopted shortly after they are
~judged to be appropriate.
The ACRS recommends that a high priority be placed on the development sand implementation of safety research on the behavior of light water
.reactors during anomalous transients. The NRC may fird it appropriate
~to develop a capability to simulate a wide range of postulated tran-
~slent and accident conditions in order to gain increased insight into measures which can be taken to improve reactor safety.
The ACRS
~wishes to reiterate its previous recommendations that a high priority
~be given to research 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 ACRS is continuing its review of the implications of this accident
-and hope to provide further advice as it is developed.
.~..
.