ML20217A814
| ML20217A814 | |
| Person / Time | |
|---|---|
| Issue date: | 10/20/1997 |
| From: | Apostolakis G, J. J. Barton, Powers D Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-3072, NUDOCS 9803250258 | |
| Download: ML20217A814 (14) | |
Text
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ADVISOR OT4kITTEE ON REACTOR SAFEGUARDS JOINT MEETING OF THE SUBCOMMITTEES ON i PROBABILISTIC RISK ASSESSMENT, PLANT OPERATIONS, AND FIRE PROTECTION MEETING MINUTES - AUGUST 28-29, 1997 ROCKVILLE, MARYLAND INTRODUCTION The ACRS Subcommittees on Probabilistic Risk Assessment (PRA),
Plant Operations, and Fire Protection met on August 28-29, 1997, at 11545 Rockville Pike, Rockville, MD, in Room T-2B3. The purpose of this meeting was to continue the Subcommittees review of matters included in the Staff Requirements Memorandum dated May 27, 1997:
- 1) acceptance criteria for plant-specific safety goals and deriving lower-tier acceptance criteria; and 2) the use of uncertainty versus point values in the PRA-related decisionmaking process. The Subcommittees discussed the proposed action plan to improve the Senior Management Meeting (SMM) process and the voluntary approach proposed by the industry for reporting reliability and availability information for risk significant systems and equipment. The Sutcommittees also reviewed NRC programs for risk-based analysis of reactor operating experience.
The entire meeting was open to public attendance. Mr. Michael T.
Markley was the cognizant ACRS staff engineer for this meeting.
The meeting was convened at 8:30 a.m. each day and recessed at 6:30 p.m. on August 28 and adjourned at 3:00 p.m. on August 29, 1997.
ATTENDEES ACRS G. Apostolakis, Chairman D. Miller, Member J. Barton, Co-Chairman R. Seale, Member D. Powers, Co-Chairman J. Garrick, ACNW Member M. Fontana, Member R. Sherry, ACRS Fellow T. Kress, Member M. Markley, ACRS Staff NRC Staff D. Allison, AEOD* A. Madison, AEOD P. Baranowsky, AEOD T. Martin, AEOD R. Barrett, AEOD H. Martz, LANL*
(f W.
G.
Gaylean, INEEL*
Holahan, NRR*
S.
P.
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M. Johnson, NRR G. Parry, NRR g)
D. Kelly, INEEL E. Rossi, AEOD [L 5 ) /
T. King, RES* M. Sattison, INEEL R. Lloyd, AEOD 9003250258 971020 PDR ACRS 3072 PDR
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Industry Reoresentatives T. McHenry, INPO T. Pietrangelo, NEI AEOD Office for Analysis and Evaluation of Operational Data NRR Office of Nuclear Reactor Regulation RES Office of Nuclear Regulatory Research INEEL Idaho National Engineering and Environmental Laboratory LANL Los Alamos National Laboratory INPO Institute of Nuclear Power Operations NEI Nuclear Energy Institute A complete list of meeting attendees is in the ACRS. Office File,-
and will be made available upon request. The presentation slides andihandouts used during the meeting are attached to the office copy of these minutes.
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Aucust 28, 1997 Introduction Dr. George Apostolakis, Chairman of the Subcommittee on PRA convened the meeting at 8:30 a.m. and introduced the Co-Chairmen of the Subcommittees on Plant Operations and Fire Protection, ACRS Members, and others in attendance. He stated that the purpose of this meeting was to continue the Subcommittees review of matters included in the Staff Requirements Memorandum dated May 27, 1997:
- 1) acceptance criteria for plant-specific safety goals and deriving lower-tier acceptance criteria; and 2) the use of uncertainty versus point values in the PRA-related decisionmaking process. The Subcommittees discussed the proposed action plan to improve the Senior Management Meeting (SMM) process and the voluntary approach proposed by the industry for reporting reliability and availability information for risk-significant systems and equipment. The Subcommittees also reviewed the NRC programs ~for risk-based analysis of reactor operating experience during the August 29, 1997, portion of the meeting.
Dr. Apostolakis stated that the Subcommittees had received no requests for time to make oral statements by members of the public.
However, Mr. Stanley Levinson of the Babcox and Wilcox Owners Group has requested that a written statement be read into the public record and made available during the meeting. Dr. Apostolakis read the statement and informed the audience that copies were available in the back of the room.
NRC Staff Presentation Treatment of Uncertainty Messrs. Thomas King, RES, and Gary Holahan, NRR, provided a discussion on the treatment of uncertainties in the risk-informea decisionmaking process. Mr. Garreth Parry, NRR, provided j supporting discussion. They discussed desirable characteristics of the decisionmaking process with emphasis on aspects that relate to uncertainty, representation of uncertainty in PRAs, the nature of acceptance criteria, and comparison of PRA results with acceptance criteria. Significant points made during the discussion include:
e Desirable characteristics include: explicit use of calculated I uncertainties, simplicity in application and interpretation, defensibility of risk analysis, flexibility to deal with i I
uncalculated uncertainties, changes in state of knowledge, and consistency in application and regulatory decisions.
I e Sources of uncertainty include: 1) parameter uncertainty -
initiating event frequency, component failure rate and ,
unavailability, and human error rate, 2) model uncertainties - !
success criteria (1 or 2 pumps), analytical models and 3
dependency (seal LOCA), and 3) incompleteness -
missing initiating events, modes of operation (low power and
. shutdown), errors of commission, and influence of organizational factors, and 4) modeling ' approximations and simplifications. 1 l
e Uncertainties explicitly- characterized in the model by probability distributions can be propagated to generate probability distributions on risk metrics (e.g., CDF and LERF).
e Acceptance criteria requires both the specification of numerical guidelines and a method of comparison of the analysis results to assess acceptability. In draf t Regulatory Guide DG-1061, acceptance guidelines were established to be consistent with the Commission's quantitative health objectives (QHOs) and subsidiary objectives. The Safety Goal Policy Statement recommended that mean values be compared with the QHOs with due consideration of uncertainty. ,
l e Methods for comparing acceptance criteria include- 1 Method 1: Accept if the mean (expected) value lies below the guideline.
Method 2: Accept if specified (e.g., 95th) percentile of the distribution lies below the guideline.
Method 3: Accept if the mean value lies below one guideline, and the 95th percentile lies below another.
e Any measure of assurance from probability distributions are conditional upon how the distributions are generated. The :
distribution must ~ represent all significant sources of uncertainty. Alternatively, sensitivity studies or other arguments are required to address unquantified uncertainties and to demonstrate that the degree of assurance is not likely to be significantly changed when unquantified uncertainties concerning the completeness are taken into account.
e Comments from public workshop on the Standard Review Plan and Regulatory Guides on risk-informed, performance-based regulation include:
Concern over the definitions ' of " risk neutrality" and "small" with respect to risk increases that may be considered negligible.
Concern over the incompleteness and possible literal interpretation of acceptance guidelines in DG-1061 with respect to extensive analysis that might be required for very small increases in risk.
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5 e The major impediment to rigorous treatment of uncertainty is the issue of completeness, particularly in relation to the
. evaluation of total CDF and LERF. Modeling uncertainties are considered to have the next greatest potential for impacting decisions. Parameter uncertainties are adequately treated using mean values, if calculated correctly. I Dr. Harry Martz, an NRC contractor at the Los Alamos National Laboratory (LANL) , provided a discussion on assessing conformance to safety goals using PRA results. Significant points made during the discussion include:
e The proper way to assess conformance to the safety goal is to determine an upper critical limit using a prior distribution.
e Two classes of prior distributions need to be considered: 1) loanormal -
either degree-of-belief (Bayes - epistemic or state of knowledge uncertainty) or empirically fitted )
(empirical Bayes - aleatory or random uncertainty) plant to plant variability, and 2) Noncarametric probability density function estimate (empirically fitted, emjirical Bayes).
e PRA results can be used to compute a figure-of-merit to assess '
compatibility of the plant with the goal.
e Bayes or empirical Bayes methods are appropriate for developing criteria for use in assessing compatibility with safety goals with specified confidence.
e Two sources of uncertainty are properly accounted for when using these methods: 1) uncertainty in the PRA estimate (best estimate of the parameter of interest), and 2) uncertainty in j the parameter of interest itself for a given facility. 4 l
Improvements to the Senior Management Meeting Process Messrs. Michael Johnson, NRR, and Richard Barrett, AEOD, led the i discussions for the staf f regarding proposed changes to the Senior Management Meeting (SMM) process and information base. Messrs. Tim Martin, Alan Madison, Steven Mays, and Ronald Lloyd of AEOD i provided supporting discussion. j l
Mr. Johnson reviewed the history of the SMM process, noted other l programs to assess plant / licensee performance (SECY-97-122), and discussed the purpose of the improvements being made to the SMM I process. In particular, he noted that the Commission had directed the staff to reevaluate the SMM process to make it more scrutable to the industry and public. He also stated that some incremental changes were being implemented with each SMM meeting while major improvements are pursued in parallel and will be discussed by Mr.
' Barrett.
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f Mr. Barrett reviewed the findings and recommendations provided in the study of the SMM process completed by the accounting firm of Arthur Andersen, Inc. , discussed the SMM Implementation Plan (SECY-97-072), and milestones and schedule for completion. Mr. Madison discussed the plant performance template. Mr. Mays discussed the performance trending methodology and Mr. Lloyd discussed the use of economic indicators. Significant points made during the discussion include:
- The Arthur Andersen study concluded that the SMM process is highly subjective, that objective information was not valued, and that there is no clear criteria regarding the information base. The study recommended a shift to more objective performance measures, a shift from event reaction to predictive performance, a movement toward scrutable and enforceable decision criteria, and the use of economic indicators to assess economic stress.
e Guidelines for the template includes the use of risk information related to operational performance, human performance, engineering and design, self-assessment including problem identification and resolution, and plant material condition.
- Guiding assumptions in developing the methodology were that it use data readily available to the NRC, use valid face and statistical inputs and models, relate to past SMM decisions, consider multiple models rather than a single model, and exclude direct economic factors. Analyses would include use of a trend model utilizing existing performance indicators, regression model methods based on past SMM decisions, and supporting analyses utilizing cluster and variability analysis methods.
e Potential financial analysis models include those related to the site (e.g., revenue factor, contribution, non-fuel cost per Megawatt, and non-fuel operations and maintenance, etc.)
as well as those related to corporate performance (e.g. , debt-to-equity ratio, fixed charge coverage, net income, etc.).
Reliability and Availability Information Mr. Patrick Baranowsky, AEOD, led the' discussions for the NRC staff. Mr. Ernie Rossi, AEOD, also participated. Mr. Baranowsky discussed the need for reliability and availability data, reviewed past discussions regarding a possible Reliability Data Rule, and summarized the types of information which are expected to be available according to the industry voluntary approach.
Significant points made during the discussion include:
- The staff recommended (SECY-97-101) accepting the voluntary approach proposed by the industry and the Commission approved 6
this approach in a staff Requirements Memorandum dated June
- 13, 1997. The Commission also requested periodic updates to
. assure that the voluntary approach was viable for continued use, o Reliability and availability information is needed to support risk-informed, performance-based regulation. It is also used to support changes to the current licensing basis, decisionmaking at the SMM, and for the identification and resolution of generic safety issues.
e The proposed Equipment Performance and Information Exchange
'(EPIX) evaluates equipment unavailability per train on a monthly basis. Functional failures, unplanned unavailable hours, estimated demands, and estimated operating hours are measured. Input data is derived from systems and components monitored under the Maintenance Rule, licensee event reports, and monthly' operating reports.
Presentation by Industry Reoresentatives Messrs. Tony Pietrangelo of the Nuclear Energy Institute (NEI) and Thomas McHenry of the Institute of Nuclear Power Operations (INPO) provided a presentations to the Subcommittees. Mr. Pietrangelo discussed industry perspectives on the SMM process. Mr. McHenry discussed the voluntary approach proposed by the industry for reporting reliability and availability data. Significant points made during the discussion include:
SMN e The SMM "watchlist" or designation as a problem plant has serious consequences for licensees. On average, it costs about 200 million dollars for a plant placed on the watchlist to placate NRC concerns.
e Evaluating plants compared to industry averages is inappropriate because the average performance of plants today is much better than it was 5 or 10 years ago. Consequently, a plant whose performance is below industry average today would likely have been considered a good performing plant in past years.
e The SMM process has been very subjective in the past. Plants with performance issues or concerns were often improving by the time the NRC placed them on the watchlist. Decision criteria is needed to make the process more objective and scrutable.
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, J e It is inappropriate to use economic indicators to assess plant performance because there is not a good correlation between
, the ' information used by financial markets and resource allocation at the plant and corporate levels.
Reliability and Availability Information e The Nuclear Plant Reliability Data System (NPRDS) is being replaced with the Equipment Performance and Information Exchange (EPIX) system. This newer system is expected to provide reliability information to the NRC as well as a database for licensee compliance with the Maintenance Rule.
o EPIX is a focused database that provides information on components important to nuclear plant safety and reliability.
It is an automated system that integrates input data for root cause analysis, provides for exchange of operating experience,-
allows benchmarking of performance, and facilitates risk-informed decisionmaking.
- EPIX uses modules to archive, retrieve, and analyze data along four safety-scope levels.
Subcc 4ttee Questions and Cc--ents Treatment of Uncertainty Dr. Apostolakis questioned the assertion that all uncertainties l need to be quantified. He stated that the review of proposed !
changes should consider the potential for unquantifiable benefits. 1 He added that the major problem lies in the issue of I incompleteness. Dr. Apostolakis stated incompleteness is always present. Dr. Kress noted that uncertainty could be bounded by good analysis. Dr. Apostolakis stated that plant-to-plant variability l needs to be bounded to isolate the variability in the methods used !
by various analysts. Dr. Seale questioned the extent to which bias exists on the part of the analyst. Dr. Apostolakis added that the greatest problem is coming up with the " likelihood" function in Bayes' Theorem. Dr. Apostolakis summarized that these were difficult issues to address. ,
Dr. Powers noted that uncertainty analysis was not typically done in most Individual Plant Examinations (IPEs) because it was not j requested in Generic Letter 88-20. He noted that the uncertainties i in the results of Level 2 PRA are very large because the uncertainties concerning the many severe accident phenomena are very large. Dr. Garrick stated that the distribution could present problems for the analyst because the "true value" of an individual
-measure could be outside the distribution.
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Mr. Barton expressed agreement with the workshop comments that the quality and scope of PRA that the NRC will accept needs further clarification. Dr. Kress stated that it is hard to do analysis appropriate to the issue without a Level 3 PRA.
At the conclusion of this session, Dr. Apostolakis requested the staff address the issues of completeness and unquantifiable benefits at the full ACRS meeting, September 3-5, 1997.
Improvements to the Senior Management Meeting Process Mr. Barton questioned whether the decisions made at the SMM could be made via other program elements such as Systematic Assessment of Licensee Performance (SALP). The staff stated that there was some overlap in assessment programs but indicated that the SMM was intended to be a top-level, agency-wide communication between the senior NRC managers and licensee executives (i.e., SALP is a communication from the NRC regional office).
Dr. Apostolakis questioned how the staff determined the specific performance characteristics to be the most important. He asked why top-level criteria like core damage frequency (CDF) and large, early release frequency (LERF) were not listed as critical measures. He also questioned the modeling that was used to support the proposed template and suggested that an influence diagram might be useful. Dr. Apostolakis stated that decision criteria were needed to make the process scrutable. The staff stated that the listed characteristics represent categories of information used in decisionmaking. The staf f stated that the Commission had requested the attributes of the current SMM process be maintained but wanted more balance in the evaluation and decisionmaking processes. Dr. {
Apostolakis acknowledged that the Commission's Staff Requirements !
Memoranda (SRMs) say keep the process and improve on it.
Dr. Powers expressed concern over the use of industry averages to assess performance and suggested that the proposed template would 1 allow for possible " creep" or " rising" NRC expectations for '
licensees performance when overall industry performance is improving. Dr. Powers also suggested that the NRC indicators do not sufficiently consider the potential for false positives or negatives in the data. Dr. Seale questioned the validity of data counting used to assess licensee performance (e.g., missed counts, j double counts, etc.). Dr. Powers also suggested that there was limited benefit in the ACRS evaluating financial indicators and suggested that the Subcommittees focus on issues subject to the Commission's SRMs.
Dr. Miller questioned whether the NRC's performance measures were the same or similar to those used by the Institute of Nuclear Powhr Operations. The ataff stated that they did not know the specific measures used by INPO although they believed that many were the .
same as those used by the NRC. l I
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Dr. Fontana questioned whether the process was effective in identifying the right plants to discuss at the SMM or whether the SMM did not place the appropriate plants on the watchlist. The staff stated that the proper plants were identified for discussion at the SMM but that action to place them on the watchlist was, at times, slow.
At the conclusion of this session, Mr. Barton requested the staff to discuss the integration of the various plant performance assessment programs (e . g . , inspection, plant performance reviews (PPRs), plant issues matrix (PIM), SALP and the SMM processes, etc.) at the September 1997 ACRS meeting. The Subcommittees requested the staff to define the need for each of these programs and to discuss how this reassessment would satisfy these needs.
Reliability and Availability Information Dr. Apostolakis questioned how human error would be considered in the assessment of reliability data. The staff stated that human error would be considered as it is associated with offnormal equipment performance and failures.
Dr. Powers questioned whether the INPO database was considered proprietary and how the public could evaluate the validity on which the conclusions of the studies were based. He also questioned whether public access was sacrificed in accepting the approach proposed by the industry rather than the proposed Reliability Data Rule. The staff stated that the studies would be available to the public but acknowledged that the raw data was considered proprietary and was not publicly available. The staff did, however, state that the data could be made available if the individual plants were not identified in the subject database available to the public.
The Subcommittees noted that the decision taken by the Commission and staff to accept the voluntary approach proposed by the industry was at odds with the position recommended in the ACRS letter dated April 12, 1995. Consequently, Dr. Apostolakis requested the staff to address the public accessibility of reliability and availability data during the September 1997 ACRS meeting.
August 28, 1997 Introduction Dr. George Apostolakis, Chairman of the Subcommittee on PRA convened the meeting at 8:30 a.m. and introduced the Co-Chairmen of the Subcommittees on Plant Operations and Fire Protection, ACRS Members, and others in attendance. He stated that the purpose of this meeting was to review the NRC programs for risk-based analysis of reactor operating experience. Dr. Apostolakis stated that the 10
Subcommittee had received no written comments or requests for time to make oral statements from members of the public for this meeting.*
MRC Staff Presentation Mr. Patrick Baranowsky, AEOD, led the discussions for the NRC staff. Messrs. Steven Mays and Patrick O'Reilly, AEOD, and Martin sattison, Idaho National Engineering and Environmental Laboratory (INEEL) provided presentations to the Subcommittees. Mr.
Baranowsky provided an overview of AEOD's risk-based programs. Mr.
Mays discussed the system reliability and special studies. Messrs.
O'Reilly and Sattison discussed the Accident Sequence Precursor (ASP) program. Significant points made during the discussion includes e AEOD uses data from actual reactor operating experience to assess and trend risk indicators. This is accomplished, in part, through rational decomposition of risk. AEOD has a major role in the NRC's PRA Implementation Plan with regard to ,
evaluating trends and patterns of operating experience.
e Program elements include: the ASP program, initiating event / frequency evaluations, reliability studies, common-cause failure. (CCF) analysis, performance indicators, human performance analysis, and special studies, as needed, e AEOD is conducting the following special studies:
Initiating events frequency -
to compare operating experience with initiating accident event sequences Loss of offsite power - to compare operating experience with the analysis supporting the Station Blackout Rule System reliability studies -
to provide engineering insights on risk-important systems based on actual operating experience. Studies include: high pressure core spray, reactor core isolation cooling, auxiliary feedwater, and Westinghouse reactor protection system Fire events study - to characterize the frequency and nature of fire event data and examine the potential impact as compared to fire risk assessments
- The ASP program continues to evolve including the development of improved models and methods, systematic screening of actual operating experience to assist management in making risk-informed decisions, and for communicating significant event insights to the industry and the public.
INEEL has.been developing ASP models using the SAPPHIRE computer code.
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Plant-specific fault trees are used based on plant Safety Analysis Reports, PRAs, and Individual Plant Examinations
. (IPEs).
Support systems are not modeled except for emergency AC power.
Modeling does not include test and maintenance unavailabilities.
Revision 20A for Level 1 enhancements is expected to be completed in December 1997.
Models are also being developed for PRA Level 2/3.
subcommittee ouestions and cc ents Drs. Apostolakis and Powers questioned who the users were for the i AEOD studies. Dr. Apostolakis questioned the status of various AEOD programs including those for CCFs, risk-based performance indicators (PIs), and ASP programs. The staff stated that the technical staff in NRR and RES are the primary customers of AEOD's risk-based work products. The staff stated that it was requesting informal feedback at this time and was not requesting a letter from the Committee during the September ACRS meeting.
At the conclusion of the meeting, Dr. Apostolakis requested the staff to discuss the initiating events study and Bayes methods at the next Subcommittee meeting on these matters.
3 Followun Actions The Subcommittee requested the staff to provide the following documents meetings:
e Dr. Apostolakis requested copies of the AEOD risk-based analysis methodology reports.
e The Subcommittees decided to hold future meetings on October 21-22 and November 13-14, 1997, to review policy issues related to risk-informed decisionmaking and to discuss the public comments on the associated Standard Review Plan and Regulatory Guides. l 1
Representatives of the NRC staff agreed to provide these documents I subsequent to the Subcommittee meeting.
Backaround Material Provided to Subc - 1ttee for this Meetina
- 1. Staff Requirements Memorandum dated May 27, 1997
- 2. Staff Requirements Memorandum dated June 5, 1997 '
- 3. Memorandum dated July 2, 1997, from S. Jackson, NRC Chairman to L. Joseph Callan, EDO,
Subject:
"The Statement of Core 3 Damage Frequency of 10E-4 as a Fundamental Commission Goal" i 12
- 4. Issues list dated June 12, 1997, from G. Apostolakis to
, Subcommittee participants, subject: Questions to be Addressed at t.he ACRS Probabilistic Risk Assessment Subcommittee Meeting
' on July 7, 1997.
Subject:
Consultant's report for ACRS PRA Subcommittee meeting, July 7-8, 1997.
- 6. Letter dated July 23, 1997, from J. Colvin, to S. Jackson, NRC Chairman,
Subject:
Comments on NRC proposed guidance for risk-informed, performance-based regulation.
- 7. LA-UR-97-248 Revision, " Assessing Conformance to Safety Goals Using Nonparametric Empirical Bayes Methods: A Nuclear Reactor Application," by H. Martz and J. Johnson, draft publication submitted to Nuclear Safety, 1997.
- 8. Nuclear Safety, May-June 1984, " Assessing Compatibility with Reactor Safety Goals Using Uncertain Risk Analysis Results with Application to Core Melt," by H. Martz and J. Johnson.
- 9. Speech dated July 21, 1997, Massachusetts Institute of Technology, Nuclear Power Reactor Safety Course, by S.
Jackson, NRC Chairman,
Subject:
Current Regulatory Issues
- 10. Memorandum dated August 13, 1997, from A. Madison, AEOD, to M.
Markley, subject: Improvements to the SMM Information Base, and briefing material attachments.
- 11. Staff Requirements Memorandum dated June 24, 1997.
- 12. Staff Requirements Memorandum dated June 30, 1997.-
- 13. Staff Requirements Memorandum dated March 14, 1997.
- 14. Memorandum dated June 11, 1997, from. D. Ross, AEOD, to A.
Thadani, RES,
Subject:
Request for Assistance from the Office of Nuclear Regulatory Research.
- 15. SECY-97-122, dated June 6, 1997,
Subject:
" Integrated Review of the NRC Assessment Process for Operating Commercial Nuclear Reactors". 1
- 16. Commission briefing handouts dated June 25, 1997,
Subject:
Periodic Briefing on Operating Reactors and Material Facilities.
- 17. SECY-97-072, dated April 2,1997, Subject : " Staff Action Plan to Improve the Senior Management Process".
- 18. GAO/ RECD-97-145, U.S. General Accounting Office Report to Congressional Requesters, dated May 30, 1997,
Subject:
NRC's Oversight of Nuclear Power Plants, Nuclear Regulation-l Preventing Problem Plants Requires More Effective NRC Action.
- 19. Staff Requirements Memorandum dated June 12, 1997.
Subject:
Future Activities Discussion:
SECY-97-101, Proposed Rule, 10 CFR 50.76, " Reporting Reliability and Availability Information for Risk-Significant Systems and Equipment".
t Report dated April 12, 1995, from T. Kress, ACRS I Chairman, to I. Selin, NRC Chairman,
Subject:
Proposed Rulemaking on Reporting Reliability and Availability Information for Risk-Significant Systems and Equipment.
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3*
Letter dated May 12, 1997, from J, Taylor, EDO, to T.
. Kress, ACRS Chairman,
Subject:
Proposed Rulemaking on
, . Reporting Reliability and Availability Information for Risk-Significant Systems and Equipment.
I
- 21. Letter dated July 25, 1997, from E. Jordan, DEDRO, to W.
Hastie, INPO,
Subject:
Commission action on Reporting of Reliability and Availability Information for Risk-Significant Systems and Equipment.
Subject:
NRC Programs for Risk-Based Analysis of Reactor Operating Experience.
- 23. Letter dated December 19, 1996, from J. Taylor, EDO, to T.
Kress, ACRS Chairman,
Subject:
NRC Programs for Risk-Based Analysis of Reactor Operating Experience.
- 24. Memorandum dated June 19, 1997, from E. Rossi, AEOD, to R.
Zimmerman, T. Martin, B. Sheron, F. Gillespie, B. Boger, M.
Slosson, and G. Holahan, NRR; L. Shao, B. Morris, and W.
Hodges, RES; J. Wiggins, RI; J. Jaudon, RII; J. Grobe, RIII; l and A. Howell, RIV;
Subject:
"Special Study: Fire Events -
Feedback of U.S. Operating Experience - Final Report," and attached proprietary materials.
- 25. Memorandum dated June 17, 1997, from E. Rossi, AEOD,*to B.
Sheron, F. Gillespie, B. Boger, M. Slosson, and G. Holahan, NRR; L. Shao, B. Morris, and W. Hodges, RES; C. Hehl and J.
Wiggins, RI; J. Jognson and J. Jaudon, RII; G. Grant and J.
Grobe, RIII; and P. Gwynn and A. Howell, RIV;
Subject:
"Special Report: -
Reactor Core Isolation Cooling System Reliability 1987-1993, AEOD/S97-02 (INEL-95-0196).
Presentation Blides The presentation slides and handouts used during this meeting are attached to the office copy of these minutes.
Note: Additional details of this meeting can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, N.W. Washington, D.C.
20006, (202) 634-3274, or can be purchased from Neal R.
Gross & Co., Inc. Court reporters and Transcribers, 1323 Rhode Island Avenue, N.W. Washington, D.C. 20005, (202) 234-4433.
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