ML20211Q265
| ML20211Q265 | |
| Person / Time | |
|---|---|
| Issue date: | 07/17/1986 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Bernero R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20209C630 | List: |
| References | |
| FOIA-87-10, RTR-NUREG-0170, RTR-NUREG-0900, RTR-NUREG-170, RTR-NUREG-900, RTR-NUREG-CR-3301 NUDOCS 8607240316 | |
| Download: ML20211Q265 (9) | |
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[ '" - # # 3 NUCLEAR REGULATOAY COMMISSION i
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NOTE T0: Rbbert M. Bernero. Director Division of SWR Licensing Office of Nuclear Reactor Regulation FROM:
Domenic B. Vassallo, Chief l
Facility Operations Branch Division of SWR Licensing,
SUBJECT:
BWR DOMINANT CORE-MELT ACCIDENTS
( is a table of useful information, prepared by Ed Chow, on the dominant core-melt sequences given in the existing PRAs for a number oi' BWRs.
The table contains, in essence, brief descriptions and freqdncies for the plant-specific dominant core-melt sequences, together with some relevant information on plant designs for comparison. Enclosure 2 is a brief description of ongoing FRA prograns within NRC and nuclear utility industry.
The data we have ccmpiled for the table were mostly obtained torm NUREG-0170, i
"NRC Policy and Future Reactor Designs," and NUREG/CR-33D1, " Catalog of PRA l
Dominant Accident Sequence Information." The information describing various PRA Programs were mostly obtained from NUREG-0900, Revision r, " Nuclear Power Plant Severe Accident Research Plan."
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/b..enic'B. Vassallo, Chief,s Facility Operations Branch Division of BWR Licensirg
Enclosures:
As stated cc w/ enclosures:
3 W. Houston G. Lainas J. Zwolinski E. Adensam D. Muller l
W. Butler B. D. Liaw 1..EsHulesn M. Srinivasan W. Hodges CONTACT:
l Ed Chow, NRR/CSL 49-27625 g', 7 b'J-/O 4
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El4 CLOSURE 1: DOMINANI BWR CORE-MELT ACCIDENT SEQUENCES IN PRAs-Core-Melt Page'1 of 2 Frequency Estimated in
- Median, Warranted PRA (Plant Mean or AuAhorization/ Containment Product Thermal Power modifications Events Point PRA Source Configuration Line (MWT) not included)
Considered Estimate Internal /
Big Rock Point Consumer Dry BWR 1 158 1 x 10-3/RY External Mean 1
Power Company Events (fire and risk analyses included)
Millstone 1 NRC-IREP*
Mk. 1 BWR 3 1727 3 x 10-4/RY Interna)
Pt. Est.
Events Shoreham Long Island Mk. 2 UWR 4 2436 5 x 10-5/RY Internal /
Pt. Est.
Lighting Co.
External Events Internal /
Peach Bottom 2 NRC-WAsil-1400* Mk. I BWR 4 1600 3 x 10-5/RY External Median Events (fire and risk analyses included)
Browns ferry 1 NRC-IREP Mk. 1 BWR 4 3293 2 x 10-4/RY Internal Pt. Est.
Events Internal /
Limerick 1 Philadelphia Mk. 2 BWR 4 3293 2 x 10-5/RY-External Mean Electric Co.
Events (fire and risk analyses included)
Grand Gulf I HRC-RSSMAP*
Mk. 3 BWR 6 3833 4 x 10-5/RY I nterr.al Pt. Est.
Events 4
Internal /
GESSAR GE Mk. 3 BWR 6 3579 4 x 10'6/RY External Mean Events I
(fire 8and risk analyses included)
- Enclosure 2 contains a description of various PRA programs.
I
e Page 2 of 2 DOMINANT BWR CORE-NFLT ACCIDENT SEQtlENCES IN PRAs Islg Rock Point - Fire in cafile penetration area inside containment destroys all safety system cables, and fire is.not manually suppressed (the frequency is 2E-4/RY).
Transient with loss of emergency condenser, stuck-open safety valve, and failure of reactor depressurization system / core spray (the frequency is 2E-4/RY).
- Medium steam line break inside containment with failure of reactor depressurization system,e and core spray (the frequency is IE-4/RY).
Hillstone 1
- Loss of station power event, failure of safety relief valves to reclose, failure of FW, failure of automatic pressure relief (the frequency is 7E-5/RY).
- Loss of station power event failure of safety relief valves to reclose, failure of FW, failure of LPCI/CS, and failure of containment cooling (the frequency is 4E-5/RY).
- Loss of station power event, failure of isolation condenser, failure of FW, and failure of automatic pressure relief (the frequency is 3E-5/RY).
- Loss of station power event, failure of isolation condenser, failure of FW, failure of LPC1/CS, failure of containment cooling (the frequency is 3E-5/RY).
- Loss of station power event, failure of isolation condenser makeup, failure of FW, and failure of automatic pressure relief (tha frequency is 3E-5/RY).
- Transient with failure to scram and failure to remove reactor core heat (the frequency is 2E-5/RY).
Shoreham
- Station blackout event with failure of ECCS to remove reactor core heat (the frequency is IE-5/RY).
- Transient with failure of ECCS to remove reactor core heat (the frequency is 1E-5/RY).
- Flooding in the reactor building with failure of ECCS to remove reactor core heat (the frequency is 4E-6/RY).
Peach Hottom 2 - Transient with failure of ECCS to remove reactor core heat (the frequency is 2E-5/RY).
- Transient with failure to scram and failure of ECCS to remove reactor core heat (the frequency is IE-5/RY).
llrowns Ferry 1 - Transient with failure of power conversion system, failure of torus cooling, and failure of shutdown cooling (the frequency is IE-4/RY).
- Transient with failure of power conversion system, and failure to scram (the frequency is SE-5/RY).
- Loss of station power event, failure of torus cooling, and failure of shutdown cooling (frequency is 3E-5/RY).
limerick 1
- Loss of station power event, failure of flPCI and RCIC, and failure of LPCI and LPCS (the frequency is 6E-6/RY).
- Transient with MSIV closure, loss of FW, failure to actuate ADS (the frequency is 4E-6/RY).
Grand Gulf 1
- Loss of FW event, with failure of power conversion system, and failure of RilR (the frequence is IE-5/RY).
- Loss of offsite power event, with failure of power conversion system and failure of RHR (the frequency is 6E-6/RY).
- Loss of FW event with failure to scram (the frequency is SE-6/RY).
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- Small-break LOCA event with RilR/SW failure (the frequency is SE-6/RY).
GESSAR
- Loss of station power with failure of RCIC (the frequency is 4E-6/RY).
4 DESCRIPTION OF PRA PROGRAMS To visualize the interrelationship between various PRA prograns in NRC and nuclear utility industry, two figures are provided. Figure 1 indicates how infor=ation sources on all PRAs are pooled together. Figure 2 indicates the ongoing PRA programs in NRC for doing risk evaluation and sequences analysis research.
A brief description of the existing PRA programs in NRC and nuclear utility industry follows. The description is focused especially on the distinct features of each PRA program.
(1) WASH-1400 The Reactor Safety Study (WASH-1400), published in 1975, was sponsored by NRC/RES to estimate the public risks that may be involved in potential accidents in commercial nuclear power plants.
The main objective of the WASH-1400 study was to reach some meaningful conclusions about the risks of nuclear accidents using current technology.
The specific objectives were to perform a more realistic, quantitative assessment of probabilities and consequences of reactor accidents, to develop the PRA methodology needed and understand its limitations, and to identify areas for future safety research.
The WASH-1400 study only addressed two plants, Peach Bottom Unit 2 and Surry Unit 1.
(2) RSSMAP The Reactor Safety Study Methodology Application Program (RSSMAP),
sconsored by NRC/RES, was conducted to apply the methodology developed in WASH-1400 to an additional group of plants for identifying risk-dominant accident sequences for a broader group of reactor designs.
Among BWRs, RSSMAP selected Grand Gulf Unit 1.
i Sandia National Laboratories was contracted to perform systen analyses and I
Battelle Columbus Laboratories was contracted to perform analysis of the accident process.
l (3) IREP i
The objectives of Interim Reliability Evaluation Program (IREP), sponsored by NRC/RES and conducted mostly by Sciences Application, Inc., were to identify, in a preliminary way, the dominant core-melt accident sequences, to quantify their frecuencies, and to develop insights and procedures for conducting similar PRA analyses in the future.
The BWRs addressed in the IREP are Browns Ferry Unit 1 and Millstone Unit 1.
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-2 (4) ASEP The Accident Sequence Evaluation Program (ASEP) is an ongoing PRA Program that estimates the frecuencies for dominant core-melt accident sequences and identifies their major contributors for six reference plants (Peach Bottom Unit 2, Grand Gulf Unit 1, LaSalle, Surry, Sequoyah and Zion).
The results from ASEP, together with the results from other NRC PRA programs, will be used to prepare NUREG-1150 as an updated version of WASH-1400.
The purpose of NUREG-1150 is to provide the bases for updating our perception of risk from selected plants, developing methods for extrapolation to other plants, comparing NRC research to industry results, and resolving numerous severe accident issues.
The ASEP information base may be valuable in supporting regulatory and licensing applications for plants that have no PRAs.
(5) EMI-2104 The BMI-2104 program, "Radionuclide Release Under Specific BWR Accident i
Ccnditions," an ongoing PRA program sponsored by 1RC/RES and performed at Battelle Columbus Laboratories, is to update the WASH-1400 source terms with the latest knowledge.
The source terms for the fission product releases from accidents in WASH-1400 were based on research report TID-14844, " Calculation of Cistance Factors for Power and Test Reac* ors Sites" published in 19o2.
Since then the behavior of fission produ;ts has become better understood.
The purpose of the EMI-2104 program is to develop i.pdated release-from-plant source terms from accident sequences for fou.- types of nuclear power plants, using improved computational tools for predicting radionuclide release, transport and deposition, and to provide in-plant time / location-depencent distributions of radionuclide for use in equipment qualification.
The BWRs incluced in the BMI-2104 program are Peach Bottom Unit 2 and 4
Grand Gulf Unit 1.
(c) SASA The Severe Accident Sequence Analysis (SASA) program, an ongoing PRA program sponsored by NRC/RES and performed at various national laboratories, selects and examines risk-dominant sequences ' rom ASEP, IREP, and RSSMAP.
f The purpose of the SASA program is to determine the minimum safety features needed to prevent core meltdown, the influence of the plant control system, the effects of alternative operator actions, and the improvement of emergency procedures.
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3-One of the important feature of the SASA program is accident management which examines the impact of operator actions vs no action, as well as the effectiveness of early venting of the wetwell airspace.
The BWRs selected for the SASA program are Browns Ferry Unit 1, Limerick and Grand Gulf Unit 1.
(7) SARRP The Severe Accident Rebaselining and Risk Reduction Program (SARPP), an ongoing PRA program sponsored by NRC/RES, is used to evaluate the risk reduction potential of proposed new safety features designed to reduce the frequencies and/or consequences of severe accidents and to perfom cost and feasibility assessments for safety options that have promising risk reduction potential.
The effects of abcut 20 safety options were evaluated. Risk reduction calculations were performed for the individual safety options and for combinations of safety cptions. A procedure was developed for evaluating the financial risk from reactor accidents.
The BWRs selected for the SARRP are the same as those for the ASEP.
To interface with other PRA Programs in NRC, the SARRP assembles the front-end analysis results fron ASEP together with back-end (containment event-tree) analysis results from the SASA program, the BMI-210a program, and other PRAs in the nuclear industry to provide the complete risk profiles and evaluate the risk reduction alternatives.
(8) IDCOR The Industry Degraded Core Rulemaking (IDCOR) Program, which was set up by the nuclear utility industry to develop the technical basis for determining whether changes in regulatory requirements are needed to reflect severe accident considerations.
The interaction between the IDCOR program and the staff programs is to focus on the definition of the most important technical issues of relevance to severe accidents and to compare the results from the IDCOR program on independent models and assessment of severe accident behavior with that sponsored by the NRC. The BWRs initially selected as the reference plants in the IDCOR program are Peach Bottom Unit 2 and Grand Gulf Unit 1.
i Based on a review of the selected BWR reference plants, the conclusions drawn from the IDCOR program are that the frequencies of BWR core-melt accidents are very low. In order to determine if this general conclusion is applicable to specific BWR plants, the sponsors of the IDCOR program have developed the Individual Plant Evaluation (IPE) Methodology.
1 The metb.cdology provides a mechanism by which important insights resulting from the IDCOR program, NRC evaluation, INP0 review, other BWR PRAs, and plant operating experience may be assessed for each BWR.
At present, studies have been performed based on applying the IPE Methodology to Peach Bottom Unit 2, Shoreham and Susquehanna.
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