ML19291A427
| ML19291A427 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/21/1979 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Hines E DETROIT EDISON CO. |
| References | |
| NUDOCS 7905080492 | |
| Download: ML19291A427 (1) | |
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4, UNIT ED STATES
- t NUC LE AR FIE GU LATOIfV COQMtSSION j g.;C.3 i HEGION lit g 'A j
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799 ROOSE VEL T ROAD gr g
G L E N E LLY N, ILLINOts 60137 o,,,,,-
April 21, 1979 Docket No. 50-341 The Detroit Edison Company ATTN:
Mr. Edward Hines, Assistant Vice President and Manager Quality Assurance 2000 Second Avenue Detroit, MI 48226 Gentlemen:
The enclosed IE Bulletin No.79-05B is forwarded to you for information. No written response is required. We have also enclosed copies of recommendations of the ACRS to the Commission for your.in-formation.
If you desire additional information regarding this matter, please contact this office.
Sincerely, M
James G. Keppler Director
Enclosures:
- 1. lE Bulletin No.79-05B with Enclosures
- 2. ACRS Recommendations to the Commission dated 4/18/79 and 4/20/79 cc w/encls:
N Central Files
'J[fjg Director, NRR/DPM AIEi!T CONT,y;[3]
2 Qll/vilTY pg Director, NRR/ DOR L
PDR Local PDR NSIC TIC Ronald Callen, !ilchigan Public Service Commission Eugene B. Thomas, Jr.,
Attorney 70050804 93
3 U.S. NL' CLEAR REGUIATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III April 21, 1979
-]
IE Bulletin 79-058 NUCLEAR INCIDENT AT THREE HILE 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 TER 79-05A. the preferred mode of core cooling following a transient or accident is to provide 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 (THI) No. 2 incident of March 28, 1979 Initiation of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of noncondensible 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.
Natural 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 DTSGs.
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 enhance 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 hduction by a blowdown through a PORY 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 tri increases. p during transients that result in primary system pressure 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 raintaining natural circulation.
The procedures and training must include neans 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 inccrporate 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 res ponse.
The instructions should include the following precautions:
reintain pressurizer level sufficient to prevent loss of level a.
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 criterios for natural circulation; maintain pressure - temperature envelope within Appendix G limits c.
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 procedurts and training instructions to ensure that:
Operators do not override automatic actions of engineered s.
safety features, unless continued operation of engineered
IE Bulletin 79-058 April 21,1979 Page 3 of 4
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safety features will result in unsafe plant conditions.
For example, if continued operoLivo uf cog erseered safety feStures
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would threaten reactor vessel integrity then the HPI should he 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 HP1 1s in_ operat_i_on shall be limited by the pressure /
temperature considerations for the vessel intsgrity."
3.
Following detailed analysis, describe the modifications to design and procedures which you have implenented to assure the reduction of the likelihood of automatic actuation of the pressurizer PORY during anticipated transients.
This analysis shall include consideration of a modification of the high pressure scram setpoint and the PORV opening setpoint such that reactor scram will preclude opening of the PORY for the spectrum 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 Yalve closure d.
Low OT5G 1evel 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.i The actions required in item 12 of IE Culletin 79-05A are modified as follows:
Review your prompt reportinc procedures for NRC notification to assure that MRC 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 meintained with NRC.
7.
Propose changes, as required, to those technical specifications which rust be modified as a result of your impiementing the above items.
Rasponse schedule for B&W designed facilities:
For Items 1, 2, 4 and 6, all facilities with an operating license a.
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 operation.
For Iters 5 and 7, all facilities with an operating license respond c.
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 infornation 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.
Enclosure:
Listing of IE Bulletins Issued in Last Twelve Months J
e eD
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IE Bulletin No.79-05B April 21, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MON 111S Bulletin Subject Date Issued Issued To No.
78-05 Malfunctioning of 4/14/78 All Power Reactor Circuit Breaker Facilities sith an Auxiliary Contact OL or CP Mechanism-General Model CR105X 78-06 Defective Cutler-5/31/78 All Power Reactor Ha=mer, Type M Relays Facilities with an With DC Coils OL or CP 78-07 Protection afforded 6/12/78 All Power Reactor by Air-Line Respirators Facilities with an and Supplied-Air Hoods OL, all class E and F Research Reactors with an OL, all Fuel Cycle Facilities sdth an OL, and all Priority 1 Material Licensees 78-08 Radiation Levels from 6/12/78 All Power and Fuel Element Transfer Research Reactor Tubes Facilities sith a Fuel Element transfer tube and an OL.
78-09 BWR Drywell Leakage 6/14/79 All BWR Power Paths Associated with Reactor Facilities Inadequate Drywell with an OL or CP Closures 78-10 Bergen-Paterson 6/27/78
'"All BWR Power Hydraulic Shock Reactor Facilities Suppressor Accumulator with an OL or CP Spring coils Enclosure Page 1 of 4
IE Bulletin No.79-05B April 21, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS
?
Bulletin Subject Date Issued Issued To No.
78-11 Exa=ination of Mark I 7/21/78 BWR Power Reactor Containment Torus Facilities for Welds action: Peach Botto= 2 and 3, Quad Cities 1 and 2, Hatch 1, Monti-cello and Vermont Yankee 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-13 Failures In Source Heads 10/27/78 All general and of Kay-Ray, Inc., Gauges specific licensees Models 7050, 7050B, 7051, with the subject 7051B, 7060, 70603, 7061 Kay-Ray, Inc.
and 7061B gauges 78-14 Deterioration of Buna-N 12/19/78 All GE BWR facilities Co=ponents In ASCO with an OL or CP Solenoids 79-01 Environmental Qualifica-2/8/79 All Power Reactor tion of Class IE Equip =ent
s Facilities with an OL or CP Enclosure Page 2 of 4
IE Bulletin No.79-05B April 21, 1979 LISTING OF IE BULLETINS TSSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To
?
No.
79-02 Pipe Support Base Plate 3/2/79 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 79-03 Longitudinal Weld Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manufactured By Youngstown Welding and Engineering Co.
79-04 Incorrect Weights for 3/30/79 All Power Reactor Sving Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-05 Nucinar Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CP 79-05A Nuclear Incident at 4/5/79 All B&V Power Three Mile Island Reactor Facilities with an OL 79-05B Nuclear Incident at 4/21/79 All B&W Power Three Mile Island -
Reactor Facilities Supplement with an OL and CP 79-06 Review of Operational 4/11/79 All Pressurized Errors and System Mis-Water Power Reactors alignments Identified with an OL License During the Three Mile except B&W facilities Island Incident Enclosure Page 3 of 4
IE Bulletin I:o.79-05B April 21, 1979 79-06A Review of Operational 4/14/79 All Pressurized Errors and Syste= Mis-Water Power Reactor Alignments Identified Facilities of Vesting-During the Three Mile house Design with an Island Incident Operating License 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System Mis-ing Designed Pressurized Alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating License 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactors Identified During Facilities with an OL Three Mile Island Incident Enclosure Page 4 of 4
EXTRACT OF B&W C0mVNICATION - RECEIVED BY NRC JNTRODUCTION 4/20/79 THE CONT!!!OlilG REVIEW OF THE SEQUENCE OF EVENTS LEADING TO Tile IflCIDEtli AT THI-2 DN MARCH 28, 1979 SHOWS TilAT ACTION Call BE tar.EN TO PROVIDE A55URAfif;E THAT THE P5 LOT-OPERATED RELIEF VALVE (PORV) POUtlTED ON Tile PRE 55URIZER O PLAtfTS WILL NOT BE ACTUATED BY NiTICIPATED TRAtiSIEllT5 W!!ICil llAVE OCCURri MYE A S!ONIFICAffT PROBABILITY UT OCCURRING IN TIIESE PLANTS.TilIS ACT1071 fluST NOT DEG LADE THE SAFETY OF THE AFFECTED PLANTS WITil RESPECT TO TilEIR RESPONSE TO KONL; UPSET OR ACCIDEffT C0tIDITIOfl5 NOR LEAD TO UNREVIEWED SAFETY C0flCERN5.
THE A4TKIPATED TRkiS!Eff75 0F COHCERN ARE:
I.
LOSS hp tXTERML ELECTRICAL LOAD 2.
T11RBINE TRIP 3.
LOSS OF MAIN FEEDWATER 4.
LOSS OF CONDENSER VACUUN 5.
INADVERTErfT CLOSURE OF MIN STEN! ISOLATION VALVES (.*f5IV).
mPGER OF ALTERNATIVES WERE CONSIDERED IN DEVELOPIrlG THE ACTIONS PRO BELO'd INCLUDING:
EESTRICTING REACTOR POWER TO A VALUE WHICH WOULD ASSURE NO ACTUATIO THE PORY.
THE REACTOR PROTECTION SYSTEH, DESIGri PRESSURE AND PORY SET-POINTS RE?tAINED AT THEIR CURRENT VALUES.
LOWERING THE HIGH PRES 5URE REACTOR TRIP SETPOINT TO A VALUE WHICil ASSURE NO ACTUATION OF THE PORY.
THE DESIGfl PRESSURE OF Tile REACTOR NID THE SETPOINT FOR PORY ACTUATION REHAlf!ED AT TifElR CURREfiT VALUES.
LOWERING THE HIGl PRESSURE REACTOR TRIP SETPOINT NiD ADJUSTIrl OPERATING PRESSURE (NiD TEXPERATURE) 0F THE REACTOR TO ASSURE 110 P ACTUATION ARD TO PROVIDE ADEQUATE PMRGIri TO ACC0h?t0DATE VARIATIO OPERATIttG PRESSURE.
THE SETPOINT FOR PORY ACTUATION REMiriED AT ITS CURRENT VALUE.
THIS ALTERNATIVE WOULO REDUCE liET ELECTRICAL OUTPUT.
ADJUSTIPG THE HICH PRE 55URE TRIP AND THE PORY SETPOIHis TO A55UnE tio PG'!V ACTUATION FOR THE CLASS OF ANTICIPATED EVErfTS OF CONCERN.
THE DESIGN PRESSURE OF THE REACTOR REMINED AT ITS CURnErlT VALUE.
fJfALYSIS OF THE IMPACT OF THESE VARIOUS ALTERMTIVES NiD THEIR C ASSURING THAT THE PORY WILL NOT ACTUATE FOR THE CLASS OF NIT F CO*;CERN HAS BEEN COMPLETED.
THE RESULTS SHOW THAT:
LOWERIffG THE HIGH PRESSURE REACTOR TRIP SETPOI?fT FROM 1355 PSIG TO 2300 PSIG MD MISING THE SETPOIffT FOR THE PILOT OPERATED RELIEF YALVE FROM 2255 PSIG TO 2450 PSIG
.3VIDES THE REQUIRED ASSURAfiCE.
THIS ACTION HAS THE FURTilER ADYNITAGES r,F:
EXTRACT OF B&W COWiUNICATION - RECEIVED BY NRC 4/20/79 Page 2 of 4
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1.
ftE00CJttG THE PROBABILITT OF PORY AND ASHE CODE PRESSU ACTil5 TION FOR OTHER INCREASIfG PRES 5URE TRNISIErfTS.
~
E.
PRESERVING PRESSURE RELIEF CdPECITT.FOR ALL HIGH PRESSURE tLtH!fikTING THE POSS!BlLIYY 0F INTRODUCING Ui1 REVIEWED SA 3.
ftEDUCING THE TIME AT HHICH THE STEM SYSTEN llEAT SIllK W 4.
THE EVEffF EMERGENCY FEEDWATER FLOW WERE DELAYEO.
A StM4ARY OF nlE IMPACT OF THE PROPOSED SCTPOIffT CH TRANSIEfiTS IS GIVEN IN TABLE T.
SEM PLANTS ARE CURRENTLY CAPABLE OF RttfBACK TV.15% OF FUL LOAD OR TRIP OF THE TURBIffE.
TNIS CAPABILITY REQUIRES ACTUATIO!! 0F TIIE P!t.0T-CPERATED RELIEF VALVES.
THE CAPABILITY IffCREASES THE RELIADILITY OF POWER SUPPLY TO THE SYSTEH BY RETURiilf4G.THE UNITS TO P0uER GEllERAT FTER THESE TRNISIEffTS.
REACTOR BE TRIPPED FOR TilESE EVEf(TS.THE ACTI0fl PROPOSE 0 AB NRC NOTE:
The effect of changing the reactor coolant sy: gem pressure trip letpoint upon peak pressurizer pressure is typifieff by the attached figure 1. which was developed by 6&W for a loss of feedwater transient.
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TABLE 1 Enclosure i e
5thtuRY OF PROTECTION AGAltf5T PORY ACTUAT!D*l PROVIDED BY PROPOSED SETPOINT CHNIGES FOR ALL NTTICIPATED TRNISIEti_TS EXTRACT Of B&W C09fJNICATION.. RECEIVED BY NRC 4/20/79 I.
ISTIChATED TRAftSIDiT5 MtICH HAVE OCCURRED AT B&W PLNIT 00RNd.LY ACTIVATE PORY AT T11E CURRENT SETPOINT (2255 PSIG):
A.
TURBINE TRIP B.
LDSS OF EXTERiLAL ELECTRICAL LOAD C.
LOSS OF HAIM FEEDLIATER 6.
LOSS DF CC:lDDtSER VACUW 5.
IMADVERTErfT CLOSURE OF M51V N!TICIPATED TRANSIEf(T5 MIICH HAVE OCCURRED AT B&W PLANTS NiD WIIICH L'0Ui.D RORFALLT ACTUATE PORY AT Tile PROPOSED SETP0f ffT (2450 PSIG):
umE AhTICIPATED TRANSIEfi75 tulICH HAVE NOT OCCURRED AT B&W PLNiTS (LOW PR03ABILITf EVENTS) AND WHICH h*0ULD fiORfulLY ACTUATE PORY AT THE CtrRRENT SETPOINT (2255 PSIG):
A.
$0hE C0ffTROL ROD GROUP WITliDRAWALS (MODERATE TO 111Q1 REACTIVITY
.VORTH GROUPS NOT OTitERWISE PROTECTED BY HIG1 FLUX TRIP).
9.
EDDERATOR DILUTION.
kiTICIPATED TRNiSIENTS WHICH HAVE NOT OCCURRED AT B&V PL 1T5 (LOW EVEliTS) CNID WHICH WOULD ACTUATE Tile FORY AT TIIE PROPOSED SETPOINT (1450 PSIC):
A.
SOME C0flTROL ROD GROUP UITHDMVALS (HIGil REACTIVITY UORTil NOT OTHERWISE PROTECTED BY HIGH FLUX TRIP).
4
r Page 4 of 4 EXTRACT OF B&W COMJ' NICATION - RECEIVED BY NRC 4/20/79
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Figure 1
r-April 17, 1979 RIXXNMENEATICNS OF WE NtX' LEAR REULNIDRY COMMISSION ADVIScrtY COMMITTEE Cri RTK70R SAfTGLRRDS RMARDING THE MARCH 28, 1979 ACCIDDrT AT TriE WREE MILE ISLAND NL) CLEAR STATION UNIT 2 I-Presented orally to, and discussed with, the NRC
~
Cbnmissioners during the ACRS-Commissioners Meetirxj on April 17, 1979 - Washington, D. C.
Natural circulatiQn is an important trode of reactor cooling, both as a planned process ard as a process that may be used under abnormal circumstances.
De Committee believes that greater understanding of this mode of cooling is required and that detailed analyses should be developed by licensees or their suppliers.
D e analyses should be supported, as necessary, by expc-iment.
Procedures should be de-veloped for initiating natural cis-ulation 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.
ne 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.
%Is overpressure may be assured by placing the pressurizer heaters on a qualified onsite power source with a suitable arrarrgement of heaters and power distri-bution to provide redundant capability.
Presently operating PWR plants should be surveyed expeditiously to determine dother such arrarx3ements can be provided to assure this aspect of natural circula-tion capability.
Se plant operator should be adequately informed at all times con-cerning the conditions of reactor coolant system operation Wich might affect the capability to place the system in the natural circu-lation mode of operation or to sustain such a modo.
Of particular frnpod:ance is that information @!ch--might irdicate that the reactor coolant system is approaching the saturation pressure corresponding-to the core exit temperature.
Wis imperding loss of system over-pressure will s!gnal to the operator a possible loss of natural circulation capability.
Such a warning may be derived from pressur-frer pressure instruments and hot leg temperatures in conjunction with conventional steam tables.
A suitable display of this information abould be provided to the plant operator at all times.
In addition, consideration should be given to the use of the flow exit tempera-
- mres from the fuel subassemblies, dere available, as an additional indication of natural circulation.
i-
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Thd exit temperature of coolant from the core is currently measured by.-thenx> couples in many pWRs to determine core performance.
De Ccdtnittee recomands that these teinperature eseasurements, as currently available, be used to guide the operator concerning core status. De range of the infonnation displayed and rer orded should include the full capability of the thermocouples.
It is also reecmmended that other existing instrumentation be examined for its possible use in assisting operating action during a transient.
'me AGS reccamends that operating power reactors be given priority with regard to the definition and implementation of instrumentation dich provides additional information to help diagnose and follow the course of a serious accident.
his should include improved sampling proccoures under accident conditions and techniques to help provide improved guidance to offsite authorities, should this be needed.
We Committee recommends that a phased imple.mentaticsn approach be em-ployed so that techniques can be adopted shortly after they are jwiged to be appropriate.
We ACRS recommands that a high priority be place $ on the developrnent and implanentation of safety research on the behavior of light water reactorc during anomalous transients.
%e NRC may firr$ it appropriate to develop a capability to simulate a wide range of postulated tran-sient and accident conditions in order to gain 1.3 creased insight into measures which can be taken to improve reactor safety.
De 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 quipment status monitoring on various engineered sat'eguards features and their supporting services to help assure their availability st all times.
We ACRS is continuing its review of the implications of this accident and hope to provide further advice as it is develnped.
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6-UNITED STATfit E
g NUCLEAR REGULATORY COMt.ilSSION 5('
- . e Aoviso.2.y cou.r. TTEE Oft REACTOR SAFCCUARDS WASHfNCToN, D. C.10555
- *ees*
April 18, 1979 b
MDORANDUM FDR Chairman Hendrie Comissioner cilinsxy Commissioner Kennedy Cbnnissioner Bradford Chre.issioner Ahearne TROM:
R. F. Traley, Executive Director Advisory Ccmittee on Reactor Safeguards Attached for your information and use is a copy of the recor.enda-tions of the Advisory Co.mittee on Reactor Safeguards dich were orally presented to and discussed with you on April 17, 1979 re-gardfrrj the recent accident at the Three Mlle Island Nuclear Sta-tion Unit 2.
A R. F. Fraley Executive Director
Attachment:
Recomandations of the NRC Advisory Committee on Reactor Safeguards Ra. the 3/28/79 Accident at % e t ree Mile Island Nuclear Station Unit 2 e
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UNITED STATES k.. r (y', f h'
1 NUCLEAR FsEGULATOAY COMMISSION t'
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ADVl3ORY CO.'.tMITTEE ON RE ACTOR $AFEGUA60$
manoros. o. c.==
April 20,1979 Ebrarable Victor Gilinsk,f Actirs Chairran U. S. Noelaar Regulatory Cocenission hashingt.cn, DC 20555
Dear X. Gilinsky:
Eis letter is in respnae to yours of April 18, 1979 uhich requested that the AGS notify the Co.e.iscionerc incadiotely if W believe any of our orci rocosanda: ions of April 17 should be acted u@n before our next regulbrly :cheduled meeting at which we could prepara r. fortul
. letter.
De Co.mittee discusse.'. this topic by conferenee telephone ec11 on* April 19 aM offers the followire commnts.
All cf the recoecodstfons made by the ACRS in its meeting with the coc..iccicners on April 17, 1979, are generic in nature and cpply to all M s.
June were intended to raquire irrediato, chan2es in operotirq pro-f:cdures or plant modifications of operating Pem.s.
Such charges should be rzde onh. aftar study of their effects on overall safety.
Sp:h stud-ies cheule b2 code by the licensees and their suppliers or consultant::
und by the NRO Staff.
The Comittee believes that th?se studies should be begun in the near future on a tire scale that will rut divert the -
hM Staff or the industry representatives frem their tasks relatirg to the cooldown of Tnree Mile Island Unit 2.
Bowver, the Cemittaa be-11evt.c that it muld bs possible 3rd desirable to init34te isriediately a sunrey of opqretire procedures for achievirs natural circulation, in-cludirs the case when off'siEe pwer is lost, ard the role of the pres-surlier heaters in such procedures.
At its syset'ini en April 16 and 17,1979, the cemittee. discussed.srith '
the KRO Staff the laatter of natural circulation for the ihree Milo Ia-Ja,=4 Unit 2 plant.
De Comittee belloves that thid matter is receiv-ing careful attention by the NRC Staff and the licensee.
To EJ.) for Appropriate Action.
Distribution:
Chm, D::rs. PE. Osc, CCA. SEth SDR, DIA. Rapifext.d to EDO. PA, E. Case.
79-1117.
L
/
+
Inimrablo Victor Gillnzky Mril 20,1979
- 1h3 CcDalttee's own rec 4vicardatlons to t he Ca:rnaissi n un April 17 htre o
seit ints.sded to apply to 'Ituce fille Island Unit 2.
W2 plan to write a furt.%r report on these mattura at our Mer/ 10, 1979 meetirq.
Sirv araly, W.
rbon Chairman S
S e
e O
e
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