ML20196B685
| ML20196B685 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/09/1988 |
| From: | Grimsley D NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| To: | Weiss E HARMON & WEISS |
| Shared Package | |
| ML20195G243 | List:
|
| References | |
| FOIA-87-714 NUDOCS 8802120079 | |
| Download: ML20196B685 (5) | |
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INFORMATION ACT (FOIA) REQUEST FEB - 91988 cmit
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eEQutSilm Ellyn R. Weiss. Esquire PART 1.-RECORDS RELE ASED OR NOT LOCAftD (See c% casa bonesi No agency twords subpect to the ondes have been located.
No edetonef egency recorde evtgect a ~ request have been located.
Agency recorde subsect to the roowour tat are ideetted in Append a are streedy evelable for m repecten and copyeg m the NRC P#c Document Room, 1717 H Street N W., Washegtce. T,4 Agency records subpect to the requeur tvu are adentifed. Appendia Fe beeg made es44mme lor p#c esoecten and copying in the NRC P@c Document X
Room. tri? H sersei. N w. wsonngicr DC, in a foid., ond., ih. roiA nomb.r and,ea weit., na,r..
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Accorde subject to the request that are doectibed in the enclosed Append cos e,e3.;n, e n,ioinib,,,.ni,,,,0,,,pe,,uno,roi, Esemptione end for the reesono set forth below pursuant to 5 U.S.C. 552(b) and 10 CFR 9 Sial of NRC Regulatione.
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The Wormation is coredered to be propretary informaten purnoant so 10 CFR 2 790tdHIL The informaten was submitted and recorted in confidence from a formgn source pursuant to 10 CFR 2190idH21.
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- 7. The withhold mformatson conests of t%estgatory records compted for new enforcorrent purposes and a beeg wthheed for the reeeantal hdcases itXtMPTCN D Doctoeurs would interfere wth an enforcement proceesne becesse e could revaal the scope. erecten, and focus of enforcement efforts. e,d Pius could possacey enow them to take ecten to she64 potental wrongdong or a noneten of NRC roouvements from vivestgetors. textMPTCN hA$
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ano ihat its production or deba4e e contrei to be pubhc ritarost. The poreans responsable for the denial are those cercels dent #ed beaow as conseg cPfoess and the Drector, I
Ommen of Rules and Recoros, Orfce of Admenetraten, for any Geneels that may be appealed to the inecute Oroctor for Operatene stooi DENylNG OFF6CIAL TITLE /05Fict RECORDS DENiEO apetuATE OFF6CIAL I
Secog ur, tDo l
James A. Fitzgerald Assistant General App, E X
counsel for Adjua1Cattons and Opinions P AftT f t D-APPtAL RIGHTS The derwal by each denying off6al k$entded in Part II.C may be accealed to the AppeRate Officel identrfied in that section. Am exh appeal must to in n-iting and rnust be made within 30 days of receipt of thes resporee. Appeats must be addressed as appropriate to the Enocutive De setor for Operations or to t e secretary at the Commeson, U.S. Nuclear Regulatory Consruesson, Washington, DC 20t45, and should clearty state on the omsope and in the letter taet R e en "Ap4eal tror., an innel FOIA Deceon.
esec vones ans tren a U.S. NUCLEAR REGULATORY COMMISSION coat FOIA RESPONSE CONTINUATION
~
L Re: F0!A-87-714 APPENDIX D r
RECORDS MAINTAINED IN THE PDR UNDER Tif A80VE REQUEST NUMBER 1.
Undated Purpose of Seminar.
(27 pages) 2.
Undated Enclosure C. Regulatory Analysis.
(22 pages) 3.
Undated Enclosure C, Backfit Analysis.
(14 pages) 4.
Undated Enclosure E, Draft Congressional Letter.
(2 pages) 5.
Undated Enclosure F, Draft Public Announcement.
(3pages) 6.
4/30/86 Memo for Speis from Bernero, subject: Request for Prioritization of Generic Safety Issue - MPSH For ECCS Pumps.
l (1 page) l i
7.
5/13/86 Memo for all NRR Employees from Harold Denton, subject' NRR Office letter No. 39, Revision 3 - MRR Procedures for Control and Review of Generic Requirements.
(9 pages) 8.
5/18/86 Memo for Murley and Beckjord from Jordan, subject: Loss of Decay Heat Removal Function at Fressurized Water Reactors With Partially Drained Reactor Coolant Systems.
(80 pages) l 9.
5/21/86 Memo for Speis from Bernero, subjt t: Prioritization of Generic Issue - Valve Interlocks to Prevent Vessel Drainage During Shutdown Cooling.
(24 pages)
- 10. 11/26/86 Region Office Policy Guide No. 0901, Revision 1.
(22 pages)
- 11. 6/10/87 Memo for Beckjord from Murley, subject: Resolution of Generic Safety Issue 93 "Steam Binding of Auxiliary Feedwater Pumps'.
(3 pages)
I
- 12. 7/29/87 Memo for those on attached list from Jordan, subject:
Long-Range CRGR Agends.
(15 pages) l
- 13. 8/14/87 Memo for Beckjord from Murley, subject: Resolution of Generic Safety Issue 93, "Steam Binding of Auxiliary 1
Feedwater Pumps".
(2 pages) i e
Re:
FOIA-07-7I4 APPENDIX E RECORDS TOTAll.Y WITHHELD NUMBER DATE DESCRIPTION & EXEMPTION 1
12/01/86 Hartin Halsch to Raymond Fraley Ex.5 re Licensee's Power to "Invoke" the Backfit Rule (8 pp.)
2 08/06/86 Hartin Halsch to Raymond Fraley Ex.5 re Application of the Backfit Rule to the Resolution of USI A-17 (Systems Interaction)
(4 pp.)
3 07/22/86 William Parler to Chairman Zech Ex.5 and Commissioners Roberts, Asselstine and Bernthal re Application of the Backfit Rule to Proposed Amendments to Part 55 (Operator Licensing)
(5 pp.)
4 06/23/86 Hartin Halsch to Commission re Ex.5 Application of the Backfit Rule to Proposed Amendments to Part 50 Requirements on Communications Procedures (3 pp.)
5 06/20/86 Hartin Halsch to Commission re Ex.5 Application of Backfit Rule's Compliance exception to Commission's Response to Guard (4 pp.)
6 06/05/86 Hartin Halsch to Chairman Ex.5 Palladino re Application of the Backfit rule to Rules on Record-Keeping and Reporting (5 pp.)
7 05/21/86 Hartin Halsch to Commission Ex.5 re Proposed Backfit Analysis for Proposed Part 20 Revision (5 pp.)
8 04/30/86 Hartin Halsch to Commission re Ex.5 I
Application of Backfit Rule to Option 2 For Response to Guard 1
(3 pp.)
9 04/14/86 Hartin Halsch to Chairman Ex.5 i
Palladino re Application of the Backfit Rule to the Proposed Insider Rules (8 pp.)
Re:
FOI A.87-714 A.PPENDIX E
R_ECORDS TOTALLY WITHHELD NUMBER DATE DESCRIPTION & EXEMPTION 10 02/04/86 Herzel Plaine to Chairman Ex,5 Palladino re Backfit Analysis for the LEU /HEU Rule (SECY-88-17 and SECY-85-284) (7 pp.)
11 01/23/86 Martin Halsch to Commissioner Ex.5 Asselstine re Application of the Backfit Rule to Relaxation of Requirements (4 pp.)
12 12/11/85 Martin Halsch to Commission Ex.5 re Application of Backfit Rule to Part 20 Revision (SECY-85-147)
(3 pp.)
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.v' HAllMON & WEISS 2001 S ST R CCT, N,W.
SulTC 430 WASHINGTON, D.C. cocoo ties O AIL MCORCCVY M ARMON TELEPHONC CLLYN R. WCISS (202)328 3500 OIANC CURRAN OCAN R. TOUSLEY ANDREA C. FCRSTER October 20, 1987 Fi E trG M UF l g glp,g g g ACI REQUdST.
Director Division of Rules and Records gg 7p 7 Of fice of Administration
/d.$,3 -
U. S. Nuclear Regulatory Commission Washington, D.C. 20555 RE:
Freedom of Information Act Request
Dear Sir / Madam,
Pursuant to the federal Freedom of Information Act, I hereby request the following on behalf of the Union of Concerned Scientists:
1.
All cost-benefit or value-impact analyses done since September, 1985 in connection with the consideration by NRC staff of generic or site-specific backfits.
2.
Any and all lists, compilations or other identifications of potential generic or site-specific backfits under consid-eration by the NRC staf f at any time since September,1985.
3.
Any and all memoranda or other documents since September 1985, from the Committee to Review Generic Requirements
( "C RGR" ) containing requests or direction to the NRC staff to perform, modify or reconsider value-impact or cost-benefit analysos regarding any potential generic or site-specific backfit.
4.
Any and all documents containing guidance, criteria or exarrples used by the NRC in deciding which generic or site-specific backfits are appropriate for cost-benefit analyses under the backfit rule and which are not so appropriate.
Y%
ggoN & WEISS Please call me if you have any questions regarding this request.
Very truly yours, Q
Ellyn R. Weiss HARMON & WEISS 2001 S Street, N.W.
Suite 430 Washington, D.C.
20009 General Counsel Union of Concerned Scientists t
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PEMORANDUM FOP: Thomas E. Murley, Director Office of Nuclear Reactor Regulation Eric 5. Beckjord, Director Office of Nuclear Regulatory Research FROM:
Edward Jordan, Director Office for Analysis and Evaluation of Operational Data
SUBJECT:
LOSS OF DECAY HEAT REMOYAL FUhCTION AT PPESSURIZED WATER REACTORS WITH PARTIALLY DRAINED REACTOR COOLANT SYSTEMS Introduction On April 10, 1987 with the reactor coolant system partially drained, the residual heat removal (RHR) pumps at Diablo Canyon 2 were tripped due to vortexing/ cavitation. As a result, the plant lost its ability to remove decay heat for 85 minutes. During that 85 minute period, the reactor coolant system (RCS) heated up and bulk boiling was present in the RCS. The loss of the decay heat removal (DHR) function which occurred at Diablo Canyon 2 is one of 37 such events that have been reported to have occurred at U.S PWRs over the last 10 years.
Table 1 presents a chronology of these 37 events. These events have the potential for leading to more serious events.
Table 2 presents a chronology of 'iRC and industry actions in the area of DHR system losses.
This memorandum p(resents a corposite set of NRC and industry recomended remedial actions Enclosure 1) which are based upon the 1985 AE00 case study report C503 dealing with decay heat umoval problems for PWR operation, analysis of events subsequent to the case study, and related industry recomendations including INPO SOER 85-4, INPO SER 79-84, and NSAC-52.
In its transalttal letter of the case study to the Director, ONRR, AE00 reconenended that the recomendations contained within the report be considered in the resolution of Unresolved Safety Issue A 45.
In a response the Director, ONPP believed that the AE00 reconinendations were not directly applicable to the resolution of A-45, but instead planned to include them in the resolution of Generic Issue No. 99 'RCS/RHR Suction Line Interlocks." T fically concerned with loss of the RHR system during cold 'his issue was speci-i shutdown or refueling.
GI-99 was subsequently modified to evaluate these issues.
Loss of DHR during shutdown is clearly not a new issue. However, the continued i
eccurrence of loss of DHP events, the apparent lack of effectiveness of licensee corrective action in response to past NRC and industry actions, re assessment of the estimted risk of such events, and the dependence of the risk estimates on haman performance, all indicate that prompt regulatory action is now needed to minimize the loss of DHR during periods with a partially drained-primary 1
avstem and to help assure its rapid recovery should it be lost.
Gwau w-ns
i Thomas E. Murley -
Discussion U.S. PWR experience has shown that loss of DHR events have been occurring at a rate of approximately one every 3 to 4 reactor years, and in particular there have been 7 loss of DHR events in the last 2 years when the RCS was in a drained-down condition.
Human errors were the root causes of most of those events.
Plants may be subjected to relatively high risks when they undergo partially drained (mid-loop) operations.
It is standard procedure for PWRs to drain down the RCS during shutdown to a low for steam generator maintenance, inspection, and tube plugging, and/or reactor coolant pump seal inspection / maintenance.
i 3
Factors which contribute to the accident risk during such operations are:
i The containment is likely not to be isolated (the equipment hatch is often 1.
open).
3 2.
Plant design may dictate a ver drained down operations (e.g.,y narrow bar$d of allowable RCS levels du at Diabic Canyon 2 the range of acceptable RCS levels was only a few inches - the constraints being the elevation of the steam generator nottle and the suction head required by the RHR pump to prevent air binding.)
3.
RCS level seasurement during drained down operations frequently depends upon jury-rigged equipment which is unanalyzed and prone to errors which may exceed the required control band (e.g., at Diablo Canyon the level seasurement. error was on the same order as the range of acceptable i
operation possibility 3 to 12 inches).
4.
Generally, procedures for operation during modes 3. 4, and 5 are of an ad hoc nature, scant or even nonexistent.
from a loss of DHR are not necessarily well thought out.Similarly, procedu l
operators say not be trained in recovery from a loss of DHR.In addition, During shut-down opustions, operators may not be fully aware of what equipment is out i
of service vs. what alternative equipment is available for racovery fros a loss of DHR.
Operators are not necessarily aware of time available for j
recovery froe loss of DHR events.
For example, at Diablo Canyon 2 operators
)
thought that if the OHR function was lost the RCS heatup rate would be l'F/ minute.
However the RCS heat up rate was 2.7'f/ minute. Therefore, the optrators were not expecting bulk boiling to begin as soon as it did.
t 5.
Plants may not have adequate instrumentation available to determine RCS temperature in the reactor during a loss of DHR event.
For example, i
Diablo Canyon 2 had disconnected the core thermocouples prior to the loss r
of DHR event in anticipation of head removal.
In January 1983, the Electric Power Research Institute's (ERPI) Nuclear Safety 1
I i
Analysis Center (NSAC) published a report on RHR experience at U.S. PWRs I
(NSAC-52). NSAC 52 provided data on loss of DHR events, as well as recommenda-tions to industry to improve the situation.
Similarly, numerous industry reports 2
(e.g., INP0 Sets 17 86, 79 84, INPO SOER 85-4 mation on loss of DHR events, including recomm)endations for improv i
i situation.
ficant industry w!As improvement in OHR loss experiences.Nonethele I
I Thomas E. Murley
-3o i
i i;'
9 In July 1984 EPRl's Nuclear Safety Analysis Center published NSAC 84, a PRA i
which addressed operation at Zion 1 and 2 during shutdown. That PRA utilized l
eaintenance and operation records and control room logbook inforsation to i
estimate equipment availabilities and recovery times.
To our knowledge, it I
was the first comprehensive PRA to address operations at U.S. PWRs during modes 4, 5, and 6.
That study shows that the likelihood of a core damage event in non power modes is comparable to that during power operation.
NSAC-84 notes that,10 days after shutdown, if the plant is in a drained down 2
(eid-loop) condition, fuel damage
- can occur 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after losing the DHR func-i tion.
Assuming the same decay heat curve, we conclude that if a loss of DHR i
were to occur during drained-down operations at Zion 4 days after shutdown, i
fuel damage could occur within about 80 minutes.
(NSAC 84 data indicates that, I
for some maintenance outages, drain down of the RCS to mid-loop operation was i
reached within 4 days from time of shutdown).
1 i
Recent experience at other U.S. PWR5 has shown thet there have been many loss of DHR events during drained down conditions which were caused by level measure-ment errors. Many of these events lasted more than 80 minutes. There have also been many siellar shorter duration events which resulted in the initiation of i
bulk boiling (see Table 1).
Review of plant operations during modes 4, 5, and 6 have shown that the key to prevention, sitigation and recovery from loss of the OHR function depend strongly on operators and their ability to perfore certain tasks. Because of the strong dependency upon haan performance, and the large error bands inherent in quanti-fying human reliability, the results of risk assessments for operations (estimated to be in the range of 1; to 5 x 10 5/RY) in modes 4, 5, and 6 are subject to large uncertainties. This is noted in both C503 and NSAC 84, i
i While there may have been over a hundred loss of DMR function events that have l
been successfully sitigated in the past 10 years at U.S. PWRs, the potential for a serious event is apparent particularly during drained down conditions, The frequency of such events continues to be several per year even after exten-t sive NRC and industry communications; the estimated probability is in the i
range of 10 5 co.e damage /RY and there is no assurar.Je that containment would be available; and often the operator, being the key element in less of DHR func-l tion events, is not provided with adequate information (instrumentation), or l
well thought out procedures, and training.
The cost-benefit analysis for the implementation of remedial actions shows that I
laprovements can be made at modest cost and that the cost / benefit ratio i
justifies action (Enclosure 2).
The total cost range free $13 million to a savings of $321 million.
The benefits from averted doses range from 59,000 l
person-rem to 177,000 pe rson-ree, i
l "NSAC-84 assumes that fuel damage occurs when the RCS boils off to the ald-plane of the core.
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4 Conclusion Adequate justification exists for an appropriate generic communication requiring
{
prompt corrective action to minimize the loss of RHR during periods when the RCS is partially drained.
We trust that the composite list of recomended remedial
)
action and the sost-benefit analysis will assist you in preparation of the generic communication.
AE00 is ready to assist your offices in the preparation and implementation of the generic comunication.
fww kJ 4 3 is t0Sor.Va8 Edward Jordan. Director Office for Analysis and Evaluation of Operational Data
Enclosure:
As stated Distribution:
V5tello JMTaylor JH$niezek FMiraglia R5tarostocki Dross TSpeis Wussell JMGrace JGKeppler ROMartin J8 Martin EJordan CHeltenes TMovak V8enaroya JRosenthal Plan Hornstein AE00 R/F ROA8 R/F DCS OSEE PREVIOUS CONCURREEES ROA8:DSP: AE00 ROAB:DSP:AE00 *ROAB:OSP:AE00 *D:DSP: AE00 %: AE00
- D: AE00 Hornstein:md Plan JRosenthal TNovak CHeltenes EJordan 5/14/87 5/14/87 5/14/87 5/14/87 5/14/87 5/15/87
Table 1 Chronology of 37 loss of DHR Events Attributed to Inadequate RCS Level Cocket Plant Date Duration Heatup 344 Trojan 5/21/77-55 min.
Unknown 3/25/78-10 min.
Unknown 3/25/78-10 min.
Unknown 4/17/78-Unknown Unknown 334
. Beaver Valley 1 9/4/78-60 min.
145 - 175'F
-366 Millstone 2 3/4/79 Unknown 150 - 208'F 272 Salen 1 6/30/79 34 min.
Unknown 334
-Beave:' yalley 1 1/17/80' Unknown Unknown 4/8/80-35 min.
0 4/11/80-70 min.
101 - 108'F 3/5/81-54 min.
102 - 168'F 344 Trojan 6/26/81' 75 afn.
140 - 150*F 369 McGuire 1 3/2/82 50 min.
105 130'F 339 North Anna 2 5/20/82-8 min.
Unknown 5/20/82-26 min.
Unknown 5/20/82-60 min.
Unknown 7/30/82-46 min.
Unknown 338 North Anna 1 10/19/82-36 min.
Unknown 10/20/82' 33 min.
Unknown 369
- McGuire 1 4/5/83 Unknown Unknown 339 North Anna 2 5/3/83-Unknown Unknown 280 Surry 1 5/17/83 Unknown Unknown
- 328 Sequoyah 2 8/6/83 77 min.
103 - 195'F 370 McGuire 2 12/31/83 43 min.
Unknown 1/9/84 62 min.
Unknown 344 Trojan 5/4/84-40 min.
105 - 201'F 316 DC Cook 2 5/21/84 25 afn.
Unknown 368-ANO-2 8/29/84 35 min.
140 - 205'F i
295 Zion 1 9/14/84 45 min.
110 - 147'F 339 North Anna 2 10/16/84-120 min.
Unknown 413 Catawba 1 4/22/85 81 min.
140 - 175'F
- 327 Sequoyah 1 10/9/85 43 sin.
<1'F 296 Zion 2 12/14/85 75 min.
- 15' 361 San Onofre 2 3/26/86 49 min.
114 - 210*F 382 Waterferd 3 7/14/86 221 min.
138 - 175'F 327 Sequoyah 1 1/28/87 90 min.
95 - 115'F 323 Diablo Canyon 2 4/10/87 85 min.
100 - 220'F j
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Table 2 Chronology of NRC and Industry Actions A
Chronoloqy of NRC Actions USI A 45 (circa - 1980) originally focused on all phases of shutdown for PWRs and BWRs redirected in 1986, no longer concerned with modes 4, 5, and 6.
IEB 80-12/IE IN 80-20 requested licensees to review Davis-lesse 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loss of DHR (4/19/80), and to analyze their own plant's procedures, focusing on redundancy, administrative controls, and technical specifications.
Generic Letter 6/11/20 - Requested licensees to review St. Lucie's upper head voiding event, amend technical specifications regarding DHR capability.
IE IN 81-09 discussed Beaver Valley's loss of RHR (drain-down - Tygon).
NUREG/CR 4005 (Parameter, Inc., 6/85) closeout of If Bulletin 8012 - Stated that the issue of DHR operability was closed out at 75% of affected facili-ties (did not address operation during drained down conditions, Tygon etc.).
AE00 Case Study C503 (12/85) - Addressed loss of DM, included 32 events during drained-down conditions (1976-1984).
Indicated that the situation is not improving.
Five major recomendations were made, including:
reliable level measurement, operator aids, taproved procedures for DHR operations, improved procedures / training for recovery froe loss of DHR events, improved technical specifications.
In response to C503, NRR noted it would include the recommendations of C503 in GI-99 (interlocks).
To resolve this issue, Brockhaven National Lab is to extrapolate the Zion OHR PRA (NSAC-84) to other PWRs and assess the ef fect of laplementing C503's recomendations. A preliminary report is due in June,1987.
Preliminary results indicate that core melt frequency due to shutdown may be as high as 5.4 x 10 5/Ry (which is three times higher than NSAC 84's result).
Brookhaven's preliminary results indicate that implementing C503's recommendations may reduce the core pelt frequency to about half that value.
I IE IN 86-10112/86 "Loss of DHR due to Loss of Fluid Levels in RC$"
discussed events at SONGS 2 (3/86), Zion 2 (12/85), Sequoyah I (10/85),
i and Catawba 1 (4/85).
Referenced AE00 Case Study C503, IE IN 81-09, i
NSAC 52.
AE00 is presently contacting g foreign country for information on improved r
level measurement equipment.
IRS report #659 (8/86) indicates that a l
foreign country is testing improved level gauges based on "different physical principles."
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Table 2 (Contined) 8 Chronology of Industry Actions NSAC 52 "Residual Heat Removal Experience and Safety Analysis, Pressurized hater Reactors," January 1,1983.
NSAC 52 reported on 96 loss of RHR events tnat occurred at US PWRs from 1977-1981.
It concluded that procedures are the key to RHR system perfonaance.
The report provided many suggestions for improv-ing RHR operations.
The suggestions addressed procedures and administrative controis relating to:
maintenance and evolution planning; nonitoring of reactor vessel level during partially drained operations control over plant status, saintenance decisions, and outage coordination. ;In addition NSAC-52 suggested improvements in human engineering and hardware, including:
control room indica-tion; audible alares for low RHR flow; redundant independent RCS level indica-ting systems; improved instrumentation; and improved data collection for e
shutdown operations, i
INPO SER 79 84 "Loss of Shutdown Cooling Due To Inaccurate Level Indication" -
November 1984. The SER discussed numerous events in which the DHR function was lost due to inaccurate RCS level indication and air-binding of the RHR pumps.
The SER noted the need for accurate RCS level indication and discussed methods for improving RCS level control.
The SER provided comments on the problees associated with using tygon tubing.
It also dis:ussed air entrainment and R
vortexing, and it noted that methods for recovery froe loss of DHR cooling should be included in operator training and procedures.
NSAC-84 "Zion Nuclear Plant Residual Heat Removal PRA," July 1985.
The report presented a PRA for Zion during modes 4, 5, and 6.
It indicated that there were large uncertainties in the estimates of risk for shutdown operations.
It concluded that modes 4, 5, 6 say present significant risk relative to operating modes 1, 2, and 3.
Core seit frequency for shutdown operations was estimated at 1.8 x 10 5/Ry.
INPO SOER 85-4 "Loss of Degradation of Residual Heat Removal Capability in PWRs,"
August, 1985. The 50ER noted that probabilistic risk studies had identified loss of RHR as a significant contributor to the potential for core damage.
Other areas addressed in the 50ER were automatic suction valve closures and f
i' loss of RHR pumps. The report stated that analyses had shown that under adverse conditions with a partially drained reactor it is possible to uncover the core within 15 to 30 minutes after loss of DHR due to boiling off the RCS.
The 50ER noted that controlling RCS level in the "required narrow range is a dif ficult evolution." It referred to INP0 SERs 60-83, and 79 84 which point out the need for reliable RCS level information.
The SOER stated that the use of certain procedures, operational controls, training and hardware could have prevented many of the referenced loss of RHR events.
Specific recommendations addressed training, operating procedures and emergency procedures relating to drained down operations.
INPO SER 17-86 ' Loss of Shutdown Cooling Flow," May 1986.
The SER discussed errors inherent in the tygon tube manometer system that was used for RCS level 4
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Table 2 (Continued)
I - Chrono _ logy of Industry Action measurement:
the routing of the tygon tubing, and the lack of operator awarene potential for vortexing.
The SER also presented potential corrective actions.
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Table 3 Recent loss of DHR Events Which Occurred Durino Drained down Operations Attributed to Inadequate RC5 Level Measurement Date Duration Boiloff initiated 7/14/86 221 minutes Yes 10/16/64 120 minutes No 1/28/87 90 minutes No 2
4/10/87 85 minutes Yes 4/22/85 81 minutes No 3/26/66 49 minutes Yes 8/29/84 35 minutes Yes I
(NCLOSURE 1 Recommended Remedial Action for Reducing Risk from DHR Operations (Based Upon NRC and Industry Sources)
(1)
Licensees should maintain containment integrity to the maximum extent practicable during periods of highest DHR risk (i.e., early stages of shutdown and drain-down operations).
It is recognized that the containment equipment hatch must be open to allow major inspections or repairs during maintenance and refueling outages.
Never-the-less licensees should take actions to minimize the risk to the public by:
i delaying the time of opening the equipment hatch following shutdown, and improv-ing the procedures and training to minimize the time required to re establish containment integrity during a loss of DHR event.
For example, task analyses to integrate equipment hatch opening with the maintenance and refueling operations should be performed.
Measures to permit reclosing of the equipeent hatch during outages should be developed based on the task analysis.
It should be recognized that operability of the containment purge valves is relied upon during shutdown operations. We also note that during an accident inoperable containment purge valves could compromise containment integrity.
Therefore the task analyses should address the containment purge valves and any other valve whose operation is needed to re establish containment integrity during periods of highest DHR risk.
This iten reflects the staff risk analyses based on NSAC-84 and $NL's on going work in support of GI 99.
The risk analyses contained in Enclosure 2 focused staf f attention on the laportance and benefit of containment integrity during shutdown operations.
(2)
Licensees should improve planning, coordination, procedures, and personnel training during shutdown to ensure the availability of DMR.
NRC C503, INP0 $0ER 85-4, NSAC-52, INPO SER 79 84 all recognized the importance of this issue and contained reconnendations, suggestions and observations to this offact.
We t>elieve that significant improvements in OHR systes availt ollity and reliability can be achieved by focusing on human factors aspects of plant shut-down.
Emphasis should be placed on detailed planning of test, surveillance and maintenance activities, and the equipment or systea interactions which have f requently caused loss-of-DHR events.
In addition, plant practices regarding the procedures and training of personnel for performance of normal (non-energency) operations during shutdown should be evaluated.
For example:
all operations and maintenance staff (licensed and non-licensed) should receive training to assure that they become sensitized to the risks associated with plant shutdown.
Emphasis should be placed upon understanding the risks and high vulnerability associated with times of high decay heat rate, drain and fill operations, disabling redundant safety equipment, etc.
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(3) i,1censees should have a_ reliable method of measuring and monitoring re tctor vessel l_evel during shutdcen modes of operation and corresponding technical specification requirements for operability.
~
NRC C503, NSAC 52, INP0 SER 79 84, INPO SOfR 85 4 and 1NPO SER 1186 all addressed the importance of reliable level instrumentation.
Common industry practice using unanalyzed makeshift devf:es such as tygon tube sight gages to monitor RCS level during plant shutdown should be modifled or i
di scontinued.
Reliable, redundant level indication should be required during modes 4, 5, and 6 to ensure availability of trending data, and to warn operators in advance of unacceptably low RCS level, in addition, plant procedures should be modified to assure that the frequency of RCS level monitoring is coamensurate with plant status (e.g., as noted in section 4.1 of C503, one plant could have sonitored vessel level as infrequently as once every 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, whereas fuel uncovery could occur only a few hours af ter a loss of DHR).
As a sintaua, each plant's safety review committee should review the instrumentation and procedures used for RCS level measurement during modes 4, 5, and 6 to ensure that a high i
level of reliability is achieved.
(4)
Licensees should perform a task analysis of DHR operation.
MSAC-52 recognized the need for laprovements in human engineering.
Performance of a task analysis per se is a specific AE00 recossendation.
we recognize that all DHR losses cannot be totally eliminated by good planning, good procedures, well trained personnel, etc.
We believe that if all licensees would perfore human factors analyses of their plant's OHA operations, (including normal and abnormal conditions) and modify their plant practices and san / machine interfaces accordingly, the risks from DHR losses would be significantly reduced.
A model to use for such human factors analyses is one used by NRR (Ref.1).
Reference 1 requires licensees to perfore specific task analyses, and to inte-grate instrumentation, alares and annunciators into normal and energency proce-dures for transients and accidents occurring during power operation.
Licensees should be required to perfore similar reviews for shutdown operations, with i
emphasis on detection and altigation of loss-of-OHR events.
The operators should be provided with information outlining the time margins available for recovery from postulated loss-of-OHR events as a function of tice from reactor trip for a representative set of DHR loss transients (such as Figure 4 of C502, parametric curves of uncovery time vs. shutdown time).
Examples of such transients are: primary system filled at maxista DHR systes temperature primary systes drained to minimum level and open to the atmosphere; RCS at refueling temperature, etc.
Information on time sargins available would assist operators in recognizing the potential seriousness of the event, and i
i assist them in choosing appropriate methods for restoring the DHR function.
U.S. Nuclear Regulatory Commission, "Clarification of TMI Action Plan j
Requirements," II.F.2 Instrumentation for Detection of Inadequate Core Cooling, (NUREG 0731), November 1980.
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3 (5)
Autoclosure interlocks should minimize loss-of-OHR events.
NSAC-52 and NRC C503 both address interlocks.
in order to prevent inadvertent DHR suction / isolation valve closures (during DHR systee operation) it is recomended that NRR consider either requiring the removal of the autoclosure interlocks t, ;he OHR suction / isolation valves, or requiring removal of power to the; OHR suction / isolation valves when valve motion is not required.
Prior to implementing this recommendation, it is necessary to ensure that there is adequate relief capacity to prevent over-pressurization of the DHR system.
(6)
Plant technical specifications should be modified _to ensure that the OHR system is available during mode 4 and the early staQefof mode 5.
While INPO SER 17 86 acknowledged shortcomings in plant technical specifications, modification of the technical specifications was ret omended in AE00 C503.
Even though NRR's generic letter of 1980 on DHR addressec DHR systes redundancy, plant technical specifications do not require DHR redundancy throughout periods when it is most needed (mode 4 and the early stages of mode 5). Since test, mainte-nance, and other shutdown activities can be initiated durSg these periods it is apparent that as a result, a DHR loss could occur at a time when the risk is highest.
We recommend that NRR address the DHR system operating requirements and that plant technical specifications be modified to:
Ensure all plants have proper shutdown mode definitions (as discussed in sections 4.3 and 5.3 of C503); and i
Ensure that both trains of the DHR system are operable during periods of high decay heat load, i.e. ; mode 4 and the early stages of mode 5.
(The 1980 generic letter permits one train to be inoperable during this time.)
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)
Since the loss of 0HR experience has not greatly taproved following the issuance of NSAC-52 and NRR's generic letter, we believe that technical specification modifications are necessary to ensure adequate redundancy.
(7)
Licensees should analyze the hydraulics associated with drained-down operations.
Level measurement errors observed at Diablo Canyon 2 (April 1947), p: liminary information from the Ali assigned to Diablo Canyon, INPO SER 79-84, and SER 17-86 which reported on problems resulting from gas entrainment in tygoq i
j measurement equipment etc, all indicate that this issue should be addressed.
Large errors in RCS level measurements have been observed during drained-down j
operations because of air or 2as entrainment which resulted from draining or venting operations, RHR pump vortexing, etc.
At man the steam generator nozzles, pressurizer surge line,y plants the elevations of reactor hot legs, and reactor coolant pump discharge are such that the there is little margin for j
measurement error prior to gas entrainment/vortexing.
The Diablo Canyon
4 licensee ran tests which indicated gas eatrainment caused erratic level sea s urements. We recommend that licensees perform a detailed hyeraulic analyses of their plants' drain-down configuration to assure that the RCS level seasuring equipment remains accurate, and operators are aware of the range allowable RCS levels which will assure reliable operation of the RHR pumps.
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ENCLOSURE 2 Cost-Benefit Analysis for Proposed NRC Generic Communication toss of Decay Heat Pemoval Function in PWRs 1.
Introduction This analysis provides an estimate of the costs and benefits associated with implementing plant and procedural modifications intended to reduce the likelihood of loss of the OHR function in modes 4, 5, and 6 at U.S. PWRs.
The analysis was performed based on the NRC's value impact methodology and it employed data which was extrapolated from the most corprehensive probabilistic risk assessment i
presently available for pressurized water reactors during shutdown (NSAC-84 July 1985).
NSAC-84 presented the results of work that was performed by Pickard, Lowe and Garrick to quantify core melt frequexy for the Zion nuclear plants during modes 4, 5, and 6.
It reviewed operating experience at Zion 1 and 2 during shutdown. -It utilized detailed plant and maintenance logbcok records to estimate availability and performance of systees and subsystems during modes 4, 5, and 6.
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Preliminary results from an NRC contractor working on this issue Brookhaven National laboratory), and AE00's review of recent operating exper(ience indicate that the core melt frequencies appearing in NSAC-84 may be overly optimistic and the value of DH9 system improvements recommended by AE00 may be signifi-i cantly greater than the values listed in this cost-benefit analysis.
{
II.
Analysis Benefit - averted dose:
Bared upon NSAC-84:
Core melt frequency due to operations during shutdown:
i 1.8 x 10 5/RY Installing a "perfect alars systee" to guarantee the operators are aware of loss of cooling would halve the core damage frequency to.9 x 10 5 The benefit of such a system is quantified as follows:
The equipment hatch is assumed open 1/2 of the time while the plant is shut down. The release is either a category 2 or 3 release.
4.8 x 10' person ren/ accident 1 avg. _ 5.1 x 10e 5.4 x 10' person ren/ accident f l
Averted Dose = (.9 x 10.s) x (.5) x 5.1 x 108 = 23 person-res RY l
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ENCLOSURE 2
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Alternatively, per Generic Issue 99's prioriti[ation in NUREG-0933, the core seit from a loss of RHR system would proceed as T 3 MlV of the 0conee RSSMAP analysis.
The releases would be as follows:
Category Probability Dose (person-rea) i 3
.5 5.4 x 108 l
l 5
.0073
- 1. 0 x 108 7
.5
- 2. 3 x 108 Averted Dose =.9 x 10 5 x.5 x 5.4 x 10' = 24.3 person rea RY PVR population (present plus future plants)
W 55 reactors 1785 RY B&W 10 reactors 298 RY CE 15 reactors 485 RY 80 reactors 2568 RY Total Averted Dose = 23 person ree x 2568 RY = 59,000 person-res RY averted Cost:
NRC Labor:
from NUREG-0933 Sesolution of Generic Issue-99 (Interlocks)
For W only: 8 san-wks = $38,000 Assume CE & B&W require similar efforts 2 x $38,000 = $76,000 Total cost for interlocks
= $114,000 Assume a similar effort is needed for level measurement $114,000 but that issue is more complex, and plant specific inspections will be Each plant will need to be inspected, procedures reviewed, etc.
neces sary.
Assume 300 hrs / plant x 80 plants x $50/hr = $1.2 M Total NRC labor cost = $1.4 M
ENCLOSURE 2
. Cost:
Industry labor and hardware:
from NUREG-0933 Resolution of Generic Issue-99 (Interlocks)
NUREG 0933 estimated resolution of interlocks At W plants the cost would be $47,200/ plant (including hardware, licinsing, review, technical specifications, etc).
Assume this cost would exist at all PWRs 80 plants x 47,200/ plant = $3.8 M Assume other hardware would also be used"perfect alarm," level instrumentation, improved planning, procedures etc. -
these items cost 2 x as auch as the interlocks assume (add $7.6 M)
Total industry cost and hardware labor = $11.4M Benefit:
Onsite property damage cost avoidance
- $2 x 10'/ core-melt x.9 x 10 5 core melt RY x 2,568 RY
= - $46 M ; however the present worth assuming 15 yrs avg and 5% discount rate is
- $23 P Benefit:
avoidance of non core-melt loss of DHR eventsCost reductio Shorten outages due to better planning estimate 3 hous/RY Avoidance of non core-melt loss of DHR events - frequency of non core-melt losses of DHR is one every 4 RY assume such losses cause ou average a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay (extension of outage for a more severe event which includes investigation of the causes of inoper ability and the i
effort required to assure that adequate corrective action is taken) -
the net delay = 1/4 x 4 = 1 hr/RY.
4 hrs x - $500,000 replacement Total
=
F x2568 RY power cost per 24-hour day i
= -$213 M present worth (15 yr. avg & 5% discount rate) l
= -$107 M l
l Onsite dose and Onsite dose avoidance are neglected.
L.M UbuGL L 4
Uncertainties 1
BNL has reviewed NSAC 84 and has added one or acre accident i
scenarios and has reexamined the models used for NSAC 84, BNL has found that the core seit frequency presented in NSAC 84 is low by a factor a 3.
If 8NL is correct then the benefit from averted dose should be 3 times that listed in this analysis.
69 serson-rem 177,000 person-res total LY 2
Time available for successful operator actions to recover from loss of DHR.
NSAC-84 data indicates drain-down during maintenance outages were completed in 4 days or less from time of rod insertion.
The decay heat after 4 days is such that the drained-down systes could heatup and bolloff to the fuel mid plane (criteria used for core damage in NSAC-84) in under 80 minutes!
However, the loss of cooling event trees assume operator recovery in 1-8 hours with mean error rates of 1x10 5 to 2x10 8 These rates appear to be overly optimistic for actions which allow as little as 80 minutes for recovery free a high l
stress situation especially if the operators have no procedures, no training and inadequate information regarding the status of equipment availability...
Recent experience has shown that there have been sany severe loss of DHR events during drained-down operation which lasted l
sore than 80 minutes and there have been many shorter duration events l
which resulted in the initiation of boiloff. For example:
Plant Date Duration
- Waterford 3 7/14/86 221 sin North Anna 2 10/16/84 120 min Sequoyah 1 1/28/87 90.if a
- Diablo Canyon 2 4/10/87 85 min Catawba 1 4/22/85 81 min
- San Onofre 2 3/26/86 49 min
- ANO-2 8/29/84 35 min 3
NSAC-84 assumes that operator recovery imprc ves with shift change, i.e., if there is a shift change, discovery / recovery from the casualty is assured.
This assum DHR loss event experience; e.g.,ption does not agree with recent on 3/26/86 SONGS 2 had a loss of DHR event which was exacerbated by the shif t change.
- Denotes initiation of boilof f.
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ENCLOSURE 2 III. Summary Cost Benefit NRC labor:
$1. 4 M Industry labor + equipment:
$11.4 M Offsite doses:
59,000*
Sum:
$12.8 M person-ree averted Propert
- $23 M Damage:y (could be as high as (could be as high
- $69 M) as 177,000 person-rem averted)
Replacement
- $107 M Cost:
(could be as high as
- $321 M)
Total Cost Total Benefit
- $321 to $13 militon
$59 - 177 million 7____
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