ML19281B375
| ML19281B375 | |
| Person / Time | |
|---|---|
| Issue date: | 04/23/1979 |
| From: | Pappas H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Wright G ILLINOIS, STATE OF |
| References | |
| NUDOCS 7905240997 | |
| Download: ML19281B375 (2) | |
Text
T TIC p ** " 4,,*
t UNITED STATES
,., g
./g NUCLEAR REGULATORY COMMISSION j i
,,C.,)
j REGION ill f
799 ROOSEVELT ROAD sy GLE N E LLYN. ILLINOIS 60137 APR 2 31979 State of Illinois Department of Public Health ATTN:
Mr. Gary N. Wright, Chief Division of Nuclear Safety 535 West Jefferson Street Springfield, IL 62761 Gentlemen:
The enclosed IE Bulletin No.79-05B titled " Nuclear Incident at Three Mile Island - Supplement" was sent to the following licensee's on April 21, 1979, for information and action:
INFORMATION American Electric Power Service Corporation Indiana and Michigan Power Company D. C. Cook 1, 2 (50-315, 50-316)
Cincinnati Gas & Electric Company Zimmer (50-358)
Cleveland Electric Illuminating Company Perry 1, 2 (50-440, 50-441)
Commonwealth Edison Company Braidwood 1, 2 (50-456, 50-457)
Byron 1, 2 (50-454, 50-455)
Dresden 1, 2, 3 (50-10, 50-237, 50-249)
La Salle 1, 2 (50-373, 50-374)
Quad-cities 1, 2 (50-254, 50-265)
Zion 1, 2 (50-295, 50-304)
Consumers Power Company Big Rock Point (50-155)
Midland 1, 2 (50-329, 50-330)
Palisades (50-255)
Dairyland Power Cooperative LACBWR (50-409)
Detroit Edison Company l
Fermi 2 (50-341)
_THIS DOCUMENT CONTAINS POOR QUAL.lTY PAGES 79052409 n,y w
State of Illinois Illinois Power Company Clinton 1, 2 (50-461, 50-462) lowa Electric Light & Power Company Duane Arnold (50-331)
Northern Indiana Public Service Company Bailly (50-367)
Northern States Power Company Monticello (50-263)
Prairie Island 1, 2 (50-282, 50-306)
Tyrone Energy Park 1 (50-484)
Public Service of Indiana Marble Hill 1, 2 (50-546, 50-547)
Union Electric Company Callaway 1, 2 (50-483, 50-486)
Wisconsin Electric Power Company Point Beach 1, 2 (50-266, 50-301)
Wisconsin Public Service Corporation Kewaunee (50-305)
ACTION Toledo Edison Company Davis-Besse 1 (50-346)
Sincerely, hfA&
t Helen Pappas, Gief Administrative Branch
Enclosures:
- 2. ACRS Recommendations to the Commission dated 4/18/79 and 4/20/79 cc w/encls:
Central Files Reproduction Unit NRC 20b Local PDR NSIC TIC Mr. D. W. Kane, Sargent and Lundy
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III S
7 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 whien should be ected upon by ifcensees with reactors designed by B&W.
As discussed in Item 4.c. of Actions to be taken by Licensees in TER 79-05A, tha preferred modo of corc 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 (TMI) 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 emintaining a relatively high water level on the secondary side of the once through staam 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 OT5G 1evel setpoint to 50% on the operating range when all reactor coolant pumps (RCF) 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.
s Other means of reducing the possibility of void fonnation 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 PORY that was stuck open.
e
IE Bulletin 79-05B April 21, 1979 Page 2 of 4 B.
Reduce the er.ergy input to the reactor coolant system by a pronpt reactor tri increases. p durir.g transients that result in primary system pressure obj[ectives.This bulletin addresses, among other things, the means to achie Actions To Be Taken by Licensees:
For all Babcock and Wilcox pressurized water reactor facilities with an operating license: (Underlined sentences are modifications to. cod 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 subcocied by at least 50*F before natural circulation is initiated.
In the event that these instructions incorporate anticipatory fillino 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:
maintain pressurizer level sufficient to prevent loss of level e.
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; maintain pressure - ter:1perature 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
-M.
mmgWW-
IE Bulletin 79-05B April 21,1979 Page 3 of 4
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safety features will result in unsafe plant conditions. For example, if continued operativo or envirieervo parety featunes would threaten reactor vessel integrity then the HPI should Se 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 pressu re.
If 50 degrees subcooling cannot be maintain 3 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 /
temocrature 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 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.
Iow pressurizer level.
IE Bulletin 79-05B April 21, 1979 Page 4 of 4
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5.
Provide for NRC approval a design review and schedule for implementation
, of a safety grade automatic anticipatory reactor scram for loss of feed-r water, turbine trip, or significant reduction in steam generator level.
6.i The actions required in item 12 of IE Bulletin 79-05A are modified as follows:
Review your prompt reportino procedprn for NRC notification to assure Inat 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 unintained with NRC.
7.
Propose changes, as required, to those technical specificatjons which must be modified as a result of your implementing the above items.
Response 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 Items 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 Regionc1 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 it required.
Approved by GAO, B180225 (RD072); clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
Enclosure:
Listing of
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IE Bulletins Issued in Last Twelve Months J
g
=9 e9-
IE Bulletin No.79-05B April 21, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
s: ~
78-05 Malfunctioning of 4/14/78 All Power Reactor Circuit Breaker Facilities with an Auxiliary Contact OL or CP Mechanism-General Model CR105X 78-06 Defective Cutler-5/31/78 All Power Reactor Ham:ner, 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 with an OL, and all Priority 1
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Material Licensees 78-08 Radiation Levels from 6/12/78 All Power and Fuel Element Transfer Research Reactor Tubes Facilities with a Fuel Element transfer tube and an OL.
78-09 BVR Drywell Leakage 6/14/79 All BWR Power Paths Associated with Reactor Facilities Inadequate Dryvell with an OL or CP Closures 78-10 Bergen-Paterson 6/27/78
All BWR Power Hydraulic Shock Reactor Facilities Supprecsor Accumuistor with an OL or CP Spring coila Enclosure Page 1 of 4
IE Bulletin No.79-05B April 21, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE M0hTHS
?
Bulletin Subject Date Issued Issued To No.
78-11 Examination of Mark I 7/21/78 BWR Power Reactor Containment Torus Facilities for Welds action: Peach Bottom 2 and 3, Quad Cities I 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., Cauges specific licensees Models 7050, 7050B, 7051, with the subject 7051B, 7060, 7060B, 7061 Kay-Ray Inc.
and 7061B gauges 78-14 Deterioration of Buna-N 12/19/78 All GE BWR facilities Components In ASCO with an OL or CP Solenoids 79-01 Environmental Qualifica-2/8/79 All Power Reactor tion of Class IE Equipment
i e
IE Bulletin No.79-05B April 21, 1979 LISTING OF IE BULLETINS ISSUED 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 Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Is3and Facilities with an OL and CP 79-05A Nuclear Incident at 4/5/79 All B&W 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
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IE Bulletin No.79-05B April 21, 1979 79-06A Review of operational 4/14/79 All Pressurized Errors and System Mis-Water Power Reactor Alignments Identified Facilities of Westing-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 Ihree 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
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Enclosure Page 4 of 4
EXTRACT OF B&W C0mVNICATION - RECEIVED BY NRC INTRODUCTION Om Page 1 of 4 THE CONT!flDIllG REVIEW OF Tite SEQUENCE OF EVENTS LEADIrlG TO TIIE INCIDErli AT THI-2 DN MARDI 28, 1979 SHOWS TIIAT ACTION CAfl BE tar.EN TO PROVIDE ASSURAfiCE THAT THE P5 LOT-OPERATED RELIEF VALVE (PORV) POUNTED Dri Tile PRESSURIZER PLAtfTS WILI, NOT BE ACTUATED BY NITICIPATED TRAtiSIEffTS W!!IcIl llAVE OCCURRED OR HAYE A $!CilIFICAffT PROBABILITY OT OCCURRING IN THESE PLANTS.
TilIS ACTI0ff t1UST NOT DEG LADE THE SAFETT OF THE ArrECTED PLANTS WITil RESPECT TO TilEIR RESPONSE TO KORr4Li UPSET OR ACCIDErfT C0flotTIONS NOR LEAD TO UNREviEVED SAFETY THE NfTKIPATED TRANS!EtiTS OF CONCERN ARE:
I.
LOS' $ bF CXTERNAL ELECTRICAL LOAD 2.
TURBINE TRIP 3.
LOSS OF MAIN FEEDWATER 4.
LOSS OF C0fiDERSER VACUUM 5.
INADVERTENT CLOSURE OF HAIN STEAM ISOLATION VALVES (MSIV).
A REGER OF ALTERNATIVES WERE CONSIDERED IN DEVELOPING THE ACTIONS BELO'd INCLUDING:
1.
EESTRICTING REACTOR PoriER TP A VALUE Wi!CH WOULD ASSURE NO ACTUATION THE PORV.
THE REACTOR PROTECTION SYSTEH, DESIGri PRESSURE AND PORY SET-POINTS RE?tAINE0 AT THEIR CURRENT VALUES.
LOWERING THE HIGH PRESSURE REACTOR TRIP SETPOINT TO A VALUE MilCll WDU ASSURE NO ACTUATION OF THE PORY.
THE DESIGli PRESSURE OF Tile REACTOR NIO THE SETPOINT FOR PORY ACTUATION REMINED AT THEIR CURRENT VALUES.
LOUERING THE HIGI PRESSURE REACTOR TRIP SETPOINT NID ADJUST!?fG T OPERATING PRESSURE (AND TEMPERATURE) 0F THE REACTOR TO ASSURE fl0 PORY ACTUATIO.Y AND TO PROVIDE ADEQUATE fnRGIN TO ACC0h?pDATE VARIATIONS Iri CPERATING PRESSURE.
THE SETPOINT FOR PORY ACTUATION REMINED AT ITS CURRENT VALUE.
THIS ALTERNATIVE WOULD REDUCE liET ELECTRICAL OUTPUT.
ADJUSTING THE HIQ1 PRE 5SURE TRIP AND THE PORY SETPOINTS TO ASSURE NO P03V ACTUATION FOR THE CLASS OF ANTICIPATE 0 EVErfTS OF CONCERN.
THE DESIGN PRESSURE OF THE REACTOR REMINED AT ITS CURRLhT VALUE.
Jf ANALYSIS OF TWE IMPACT OF THESE VARIOUS ALTERMTIVES Afl0 THEIR C0tlT 1
ASSURING THAT THE PORY WILL NOT ACTUATE FOR THE CLASS OF NiTICIPATE F CONCERN HAS BEEN COMPLETED.
THE RESULTS SHOW THAT:
LO'4ERIftG THE HIGH PRESSURE REACTOR TRIP SETPOI?fT FROM 2355 PSIG TO 2300 PSIG MD R5! SING THE SETPOINT FOR THE PILOT OPERATED RELIEF YALVE FRQ;i 2255 PSIG TO _2450 PSIG 00VIDES 1ME REQUIRED ASSURANCE.
THIS ACTION HAS THE FURTilER ADVAflTAGES OF:
EXTRAET OF B&W C0mVNICATION - RECEIVED BY NRC 4/20/19 Page 2 of 4 1.
AE00CJftG THE PRDBABILITY OF PORif NIO ASHE CODE PRES ACTd6 TION FOR OTHER li1CREA5tfG PRESSURE TRNISIErfr5 PREhMVING PRESSURE RELIEF CdPECITY.FOR E.
8, tt!HIMkTING THE POS$!B1LITY 0F IfiTR000CING U!iREVIEWED S ftEDUCING THE T!!E AT e1ICH THE STEAM SYSTEN llEAT SITIK 4.
THE EVENT EMERGENCY FEEDWATER FLOW WERE DELAYED.
A StP0fARY OF THE THPAcT OF THE PROPOSED SCTPOIlfT C TRAILSIEITIS IS GIVEN IN TABLE 1.
SEM PLNfTS ARE CURREff7LV CAPABLE OF RtilBACK TU 15% OF FULL LOAD OR TRIP OF THE TURBillE.
THIS CAPABILITY REQ'JIRES ACTUATIO ( OF Tile PILOT-CPERATED RELIEF VALYES.
THE CAPABILITY IllCREASES THE RELIADILITY OF POWER SUPPLY TD THE SYSTEM BY RETURNIf4G.THE UNITS TO P09ER GE7lERAT FTER THESE TRNISIEffTS.
REACTOR BE TRIPPED FOR TilESE EVEt(TS.THE ACTIOff PROPOSED ABOVE WILL REQU NRC NOTE:
The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typiffeif by the attached figure 1. which was developed by B&W for a loss of feedwater transient.
O e
O a
see
TABLE 1 e
St>tuRY OF PROTECT!0'l ACMIPf5T PORY ACTUAi! Oil PROVIDED BY PROPOSED SETPDINT CHNIGES FDR ALL ANTICIPATED TRNISIEifT5 EXTRACT Of 8&4 COPts]NICATION RECEIVED BY MRC 4/20#9 I,
INT!
ATED TRANSIDIT5 Mf!Of HAVE OCCURRED AT 88M PLRITS NID MIIC11 WOULD
$DRNdLY ACTIVATE PORY AT THE CURREMT SETPOINT (2255 PSIG):
A.
TURSIME TRIP t.
LDSS OF EXTERNAL ELECTRICAL LOAD c.
loss 0F MAIN FEEDUATER 6.
LOSS DF CDlDEMSER VACU1.rt 5.
IMADVERTEffT CLOSURE OF PGIV A!!TICIPATED TRA'tSIErn3 WICH HAVE OCCURRED AT B&W PLANTS NiD W!!!CH L'0ULD f;0RRALLY ACTUATE FORY AT Tile PROPOSED SETPOIrlT (2450 PSIG):
KDfE ARTICIPATED TAN (SIEriTS mlICH HAVE NOT OCCURRED AT B&W PLNHS (LOW
~
PR03A91LITY EVENTS) AND WHIOf WOULD NORt%LLY ACTUATE PORY AT THE CtJRRf,NT SETPOINT (2255 PSIG):
A.
Sche CONTROL ROD GROUP WITHDRAWALS (MODERATE TO If!Gl REACTIVITY
, WORTH GRotT75 NOT OTHERWISE PROTECTED BY HIGI FLUX TRIP).
B.
EDDERATOR DILUTION.
h4TICIPATED TRNi3IENTS Mi!CH HAVE NOT OCCURRED AT B&W PL 175 (LOW PRO EVEins) AND WHICH VOULD ACTUATE TifE PORY AT Tite PROPD5ED SETPOINT (reso PsIs):
A.
s0E C0ffiROL ROD GROUP UITHDftAVALS (HIGil REACTIVITY t.VRill r:0T OTHERWISE PROTECTED BY HIGH FLUX TRIP).
o Page 4 of 4 EXTRACT OF BW COSUNICATION - RECEIVED BY NRC 4/20/79
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e April 17, 1979 R!%%2t'INEATIONS OF THE NUCLEAR RILULATORY cob 2iIS$1CN ADVISCTtY COMMITTEE Q1 REACIOR SAf1LIARDS REGARDING THE MARCH 28, 1979 AccIDctrr AT THE THREE HILE ISLAND NUCLEAR STATION LNIT 2 C
Presented orally to, and discussed with, the NRC Cbmmissioners during the ACRS-Conenissioners Meeting on April 17, 1979 - Washington, D. C.
Natural circulation is an important inode of reactor cooling, both as a planned process and 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. 1he analyses should be supierted, as necessary, by expariment.
Procedures should be de-veloped for initiating natural circulation in a safe nunner and for providing the operator with assurance that circulation has, in fact, been estab11shed, This may require installation of instrumentation to measure or indicate flow at low water velocity.
%e use of natural circulation for decay heat receval following a loss of offsite power sourcos requires the maintenance of a suitable over-pressure on the reactor coolant system.
nfs overpressure may be assured by placing the pressurizer heaters on a qualified onsite power source with a suitable arrange:nent of heaters and power distri-bution to provide redundant capability.
Presently operating PWR plants should be surveyed expeditiously to determine *other such arrangements can be provided to assere this aspect of natural circula-tion capability.
Se plant operator should be adequately infomed at all times con-cern!ng the conditions of reactor coolant system operation which might affect the capability to place the systern in the natural circu-lation mode of operation or to sustain such a mode.
Of partleular imicrtarx:e is that information which-might indicate that the reactor coolant system is approaching the saturation pressure corresponding-to the core exit temperature.
His 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-frer pressure instruments and hot leg temperatures in conjunction with conventional steam tables.
A suitable dicplay of this information should be prcvided 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, dere available, as an additional indication of natural circulation.
0
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.- 2 --
Th4 exit ternperature of coolant from the core is currently measured by,-thernocouples in many phas to determine co a performance.
1he Ccdmittee recorrnands that these te nperature measurements, as currently ava.11able, be used to guide the operator concerning core status. ine range of the information displayed and recorded should include the full capability of the thermocouples.
It is also recommended that other existing Instrumentation be examined for its possible use in assisting operating action during a transient.
1he ACRF recarimends 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.
This should include improved sunpling procedures under accident conditions and techniques to help provide improved guidance to offsite authorities, should this be needed.
We Committee recervnends that a phased imple:nentation 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 developnent and implenentation of safety research on the behavior of Ifght water reactorc ducirg anomalous transients.
The NRC may fird 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 Wich can be taken to improve reactor safety.
De ACRS wishes to reiterate its previous recom.endations that a high priority be given to research to improve reactor safety.
Consideration should be given to the desirability of additional equipnent status monitoring on various ergineered safeguards. features and their supporting services to help assure their availability at all times.
1he ACRS is continuing its review of the implications of this accident and hope to provide further advice as it is developed.
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'I UNITED 5TATES
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g NUCLEAR REGULATORY CO.*/. MISSION B
ADVISORY COI.'.f.'.lTTEC Off REACTOR SAFCCUARDS wassincTon, o. c. rosss
-*e..e April 18, 1979 MEMORANDLM FDRs Q2 airman Hendrie Comissioner Gilinsky Commissioner Kennedy Cbmmissioner Bradford commissioner Ahearne FROM:
R. F. Traley, Executive Director Advisory Committee on Reactor Safeguards Attached for your information and use is a copy of the recemwnda-tions of the Advisory Comittee on Reactor Safeguards Wich were orally presented to and discussed with you on April 17, 1979 re-gardfrg the recent accident at the Three Mile Island Nuclear Sta-t!on Unit 2.
A R. F. Fraley Executive Director
Attachment:
Recorrnandations of the NRC Advisory Comittee on Reactor Safeguards Re. the 3/28/19 Accide.nt at *Ihe 'three Mile Island Nuclear Station Unit 2 9
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UNITED STATES
.tD[v'-('*'g NUCLEAR hEGULATORY COMMIS$10N A.T
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ADVl30MY COMMITTEE ON REACTOR $AFEGUAnu$
nAmul0 TON. D. C. tor.;ss
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Aprii20,1979 lbrarable Victor Gilinsky Act!rg Che.frman II. S. Noelaar Regulatory Cocmission bashin.gt.on, DC 20555
Dear br,
Gilinsky:
T:is letter is in recpanse to yours of Ap:11 18,1979 uhich requested that the ACKS notify the Commissioners incaadiotely if We believe any of our oral rococundations of April 17 should be seted upon before our next regulbrly =chedided meetirrg at which we could prepare a formi
. letter. The cce.wittee discussed this topic by conferenee telephone cc11 on April 19 aM offers the followire coments.
All of the recori.Tsndstrons made by the ACRS in its meeting with the ccc..lacioners on April 17, 1979, are generic in nature and cpply to all PGs. J4ne were interx*ed to require imTediate, changes in operoting pro-ccdares or pir.nt modifications of operating PW,ts.
Such charges should be cade on.1v aftar study of their effect.s on overall safety.
Sp h stud-ies chotic b2 cede by the licensees and their suppliers or consultant:
Und by the NRC Staff.
The Conmittee believes that thase studies should be begun in the near future on a tirne scale that will not divert the -
NRC Staff or the industry representativas from their tasks relating to the cooldown of Three Mile Island Unit 2.
Ibwever, the Cemittaa be-11evt.S that it s.tald bs pas.mible arr! desirabia to initiate iumliately a survey of operet!re procedures for achievirs natural circulation, in-caudire tha tese when offsiEe pwer is lost, arx! the role of the pren-autla.et heaters in such pYocedures.
At its saeth2 cn April 16 and 17,1979, the Cosmittee. discussed.trith
- the ERC Staff the Iaetter of natural circulation for the ihree Mile IA-lam Unit 2 plant.
'the cocaittee bellovos that this' matter is receiv-ing careful attention by the NRC Staff and the licensee.
'To ED3 for Appropriate Action.
Distribution:
Chm, Durs. PE. OSC, OCA, SEGY, SDR, DIA. Rapifaxe.d to EDO PA, E. Case.
79-1117..
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Mirablu Victor 0111nsky 2-Mril 20,1919
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'ths rosittee's own recmaardations to the Com131. ion un A;hil 17 htre sct intaded to apply to ' Duce !.111e Is10nd Unit 2.
W plan to write a furtherr reprt on these mattura et our May 10, 1979 acetirm.
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