ML20217J045
| ML20217J045 | |
| Person / Time | |
|---|---|
| Issue date: | 07/11/1997 |
| From: | Seale R Advisory Committee on Reactor Safeguards |
| To: | Shirley Ann Jackson, The Chairman NRC COMMISSION (OCM) |
| References | |
| ACRS-SL-0451, ACRS-SL-451, NUDOCS 9708130410 | |
| Download: ML20217J045 (9) | |
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NUCLEAR MEGULATORY COMMISSION SL-0451 6"
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ADVitORY COMMITTEE ON RE ACTOR SAFE 00ARDS PDR M///97 ADVitORY COMMITTEE ON NUCLEAR WABT' I
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July 11, 1997 of PKI OF ACht/ACNW The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555-0001
Dear Chairman Jackson:
SUBJECT:
SUMMARY
REPORT - FOUR HUNDRED FORTY-SECOND MEETING OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, JUNE 11-14, 1997, AND OTHER RELATED ACTIVITIES OF THE COMMITTEE Dur.tng its 442nd meeting, June 11-14, 1997, the Advisory Committee on Reactor Safeguards (ACRS) discussed several matters and completed the following reports and letters.
In addition, the Committee authorized Dr. Larkins, Executive Director, to transmit the memoranda noted below:
REPORTS e
ProDosed Staf f Position Recardina " Inclusion of a Containment Sorav fivst em in the AP600 Desian (Report to Shirley Ann Jackson, Chairman, NRC, from R.
L.
Seale, Chairman, ACRS, dated June 17, 1997.)
e Procosed Reaulatory Aceroach AsJociated with Steam Generator Inteority (Report to Shirley Ann Jackson, Chairman, NRC, from R. L. Seale, Chairman, ACRS, dated June 20, 1997.)
LETTERS e
ProDosed Final Generic Letter. " Assurance of Sufficient Net Positive Suction Head for Emeroency Core Coolina and Contain-ment Heat Removal Pumos" (Letter to L. Joseph Callan, Execu-tive Director for Operations, NRC, from R. L. Seale, Chairman, O\\
ACRS, dated June 17, 1997) e Proposed Generic Letter.
" Potential for Dearadation of the Emircency Core Coolina System and the Containment Sorav System After a Loss-of-Coolant Accident Because of Construction and g
Protective Coatina Deficiencies and Foreion Material in the
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Containment" (Lettar to L. Joseph Callan, Executive Director i
for Operations, NRC, from R. L. Seale, Chairman, ACRS, dated June 18, 1997) e Reculatorv Guidance for Imolementation of Dioital Instrumenta-tion and Control Systems (Letter to L. Joseph Callan, Execu-
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tive Director for Operations, NRC, f rom R. L. Seale, Chairman, ACRS, dated June 23, 1997)
MEMOPld7DA 9
Besolution of the Multiole System Resoonses Procram Issues (Memorandum to L. Joseph Callan, Executive Director for Opera-tions, NRC, from John T.
Larkins, Executive Director, ACRS, dated June 17, 1997) e Establishina a Denchmark on Risk Durina Low-Power and Shutdown Doerations (Memorandum to L. Joseph Callan, Executive Director for Operations, NRC, from John T. Larkins, Executive Director, ACRS, dated June 19, 1997) e Pronosed Reaulatorv Guidance Related to Imolementation of 10 CFR 50.59 (Chances. Testo and Exneriments) (Memorandum to L.
Joseph Callan, Executive Director for Operations, NRC, from John T.
Larkins, Executive Director, ACRS, dated June 19, 1997) e Procosed Rule. "Precuency of Reviews and Audits for Emeroency Precaredness Procrams.
Safeauards Continaency Plans.
and Security Procrams for Nuclear Power Reactors." (10 CFR Parts 50 and 73.
PRM 50-59 and 50-60r ' (Memorandum to L.
Joseph Callan, Executive Director for Operations, NRC, from John T.
Larkins, Executive Director, ACRS, dated June 30, 1997)
HIGHLIGHTS OF KEY ISSUES CONSIDERED BY THE COMMITTEE 1.
Dioital Instrumentation and Control Systems The Committee heard presentations by and held discussions with representatives of the NRC staff and its contractor, the Lawrence Livermore National Laboratory, regarding the proposed final update of Standard Review Plan (SRP),
Chapter 7,
" Instrumentation and Controls," including Branch Technical Positions, and associated Regulatory Guides related to digital instrumentation and control systems.
The Committee also discussed the staff's integration of insights from the National Research Council's Phase 2 Study Final Report and a draft safety evaluation report on an Electric Power Research Institute (EPRI)
Topical Report, TR-106439,
" Guideline on Evaluation and Acceptance of Commercial Grade Digital Equip-ment for Nuclear Safety Applications."
Conclusion The ACRS issued a letter to the Executive Director for Operations, L. Joseph Callan on June 23, 1997, on this matter,
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4 2.
Consecuences of Reactor Water Cleanun System Line Break Outside the Containment During the review of the Advanced Boiling Water Reactor (ABWR)
- design, the ACRS identified an event that might cause a serious environmental disruption through the secondary containment before it could be isolated.
In its letters of July 13,1994 and February 15, 1995, the ACRS pointed out that an added third valve inside the primary containment would provide long-term post-accident isolation for the ABWR design.
General Electric Nuclear Energy agreed and added the third valve to the ABWR design.
The ACRS recommended that this issue be investigated for operating boiling water reactors (BWRs).
The NRC staff performed a study regarding the consequences of reactor water cleanup (RWCU) system line break using three representative BWR plants. These are Browns Ferry 2 (BWR4, Mark I), Susquehanna 1 (BWR4, Mark II), and Grand Gulf 1 (BWR6, Mark III).
The 9t'ff's overall conclusion is that to ensure with certainty Mb 1e health and safety under RWCU 3
system line break acciegnt./w ditions, it is necessary that the RWCU containment isolni s valyes close upon demand.
The committee in general expres2ed concern about the limited scope of the study.
Other items of concern to some members of the Committee include lack of staff efforts to assess the effects of break location and size, water cascades and water droplets, alternate plant configurations, the reliability of the containment isolation volves under blowdown conditions, and the ef fects of aging on these values.
The staff agreed to address these concerns at a future date.
The staff described an RWCU break condition at the Monticello plant.
As part of its power uprate, Monticello dotsrmined that the energy and mass release assumed for a high energy line break in the RWCU system was incorrect.
Conclusion The committee decided to postpone its comments on this matter until the staf f provides additional information regarding the concerns cited by the committee members.
3.
PRA Imolementation Plan The committee heard presentations by and held discussions with representatives of the NRC staff regarding the PRA Implementa-tion Plan (SECY-97-076) with emphasis on risk-informed initiatives in the areas of training and inspection.
The
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Committee also heard an overview briefing on the proposed SRP Section and associated Regulatory Guide for risk-informed, performance-based inservice inspection (ISI).
Conclusion This briefing was for information only.
The Committee plans to continue its discussion of the proposed SRP Section and Regulatory Guide for risk-informed, performance-based ISI during the July 9-11, 1997 ACRS meeting.
4.
Procomed Staff Position on the severe Accident Rulernakino The Committee heard presentations by and held discussions with representatives of the NRC staf f concerning the proposed staf f decision not to proceed with a rulemaking on severe accident performance design criteria for future plants.
The staff discussed the stability of the regulaton process, the need for design features to prevent containment
- bypass, the benefits of codifying present knowicdge with a rule, and the adequacy of the present policy statement regarding severe accidents.
conclusion This briefing war for information only.
No Committee action was required.
5.
honroach and Preliminary Evaluation of Site-Soecific DeviatioD of Subsidiary Risk Accentance Criteria The Committee discussed the Commission's request included in the Staff Requirements Memorandum (SRM) dated May 27, 1997, that the ACRS determine the change in core damage frequency and large, early release f requency f rom site to site, when lower-tier criteria are derived from the prompt facility quantitative health objective.
Mr. Rick Sherry, Senior ACRS Fellow, presented the statun of the analysis being performed by him to address this issue.
He presented the model being used to calculate site-spacific
- large, early release frequencies, and provided a partial listing of plant-specific exposure indices.
The Committee discussed the uncertainties associated with the results of gaussian distribution plume models and the methodology used to calculate the site-specific results.
Conclusion The Committee decided to continue discussion of this matter at the July ' -11, 1997 ACRS meeting.
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6.
Pronosed Reculatorv Acoroach Associated with Steam Generator Tube Intecrity The Committee heard presentations by and held discussions with representatives of the NRC staff and the Nuclear Energy Institute (NEI) concerning the proposed regulatory approach for ensuring steam generator tube integrity.
The staff's revised approach involved issuing a generic lettur requesting licensees to amend steam generator tube integrity technical specifications and developing a Regulatory Guide related to steam generator tube integrity.
The staff discussed the nondestructive examination techniques, performance monitoring criteria, the uncertainties associated with the draft risk assessment, and the application of the defensedin-depth philosophy in developing steam generator tube performance criteria.
NEI discussed the possibility of referencing the guidelines prepared by the Electric Power Research Institute in the proposed Regulatory Guide.
Conclusion The Committee issued a report to Chairman Jackson, dated June 20, 1997, on this matter.
7.
Procosed Final Generic Letter on Assurance of Net Positive Suction Head for Emeroency Core Coolina and Containment Heat Removal Pumos The Committee reviewed the proposed final version of the subject Generic Letter.
In the past few years, a number of nuclear power plant licensees have discovered (or been found to have) concerns relative to assuring that adequate net positive suction head (NPSH) existed for the emergency core cooling or containment cooling pumps during accident condi-tions.
The NRC staff is particularly concerned that some licensees were taking credit for, or found the need to take credit for, containment overpressure to ensure sufficient NPSH.
The Generic Letter requests that all licensees review their current design-basis analysis to determine available
- NPSH, and provide a specified set of detailed technical information for NRC staff review.
Conclusion The ACRS issued a letter to the Executive Director for Operations, dated June 17, 1997, on this matter.
8.
Pronosed Generic Letter on Potential for Deoradation of Emeroency Core Coolina and Containment Sorav Systems followina a Loss-of-Coolant Accident Due to Construction and Protective coatina Deficiencies and Foreien Material in the Containment
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The committee heard presentations by and held discussions with representatives of the staff regarding the subject Generic Letter. The staf f stated that the foreign materials, degraded coatings inside the containment that detach from their substrate, along with debris generated by a loss-of-coolant accident, are potential common cause failure mechanisms which may clog the suction strainers and sump screens.
The NRC has issued a number of generic communications on various aspects of the potential for the loss of the Emergency Core Cooling System and safety-related Containment Spray System as a result of strainer clogging and debris blockage.
The information requested in the proposed Generic Letter will enable the staff to determine whether the licensees' use of protective coatings inside the containment is in compliar.ca with the current licensing basis.
Conclusion The Committee issued a letter to the Executive Director for Operations, dated June 18, 1997, on this issue.
9.
Poliev Issue Pertainina to AP600 Desian The Committee heard presentations bp and held discussions with representatives of the NRC staff and Westinghouse Electric Corporation concerning the staff position that the AP600 design should include a Containment Spray System for accident management following a severe accident.
The staff discussed the effectiveness of the natural aerosol removal mechanisms, the uncertainties associated with aerosol removal models, other contributors to AP600 design uncertainties, and how a spray system would reduce these uncertainties.
Westinghouse discussed the uncertainties and assumptions associated with the AP600 design calculations.
The ACRS discussed the need for an uncertainty analysis associated with the PRA.
Conclusion The Committee issued a report to Chairman Jackson, dated June 17, 1997, on this matter.
RECONCILIATION OF ACRS COMMENTS AND RECOMMENDATIONS The Committee discussed the response from the NRC Executive Director for Operations dated May 5, 1997, responding to ACRS comments and recommendations included in the ACRS report dated April 8, l',7, concerning proposed regulatory guidance related to implementation of 10 CFR 50.59
(" Changes, Tests and Experiments").
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The Committee decided to continue its review of this matter i
after reconciliation of public comments, t
e The committee discussed the response from the NRC Executive l
Director for operations dated May 29, 1997, responding to ACRS I
comments and recommendations included in the ACRS report dated June 3, 1996, concerning resolution of the Multiple System i
Responses Program issues.
The committee remains interested in the full resolution of
- these issues and looks forward to a briefing on the staff's
- summary report when it is completed, o
The Committee discussed the response from the NRC Executive Director for Operations dated May 28, 1997, responding to ACRS comments and recommendations included in the ACRS report dated April 18,1997, concerning the establishment of a benchmark on risk during low power and shutdown operations.
The Committee looks forward to working with the staff on this issue and would like to discuss the staf f's goals, objectives, r
strategy, and schedules for completing this activity during a
- future meeting.
1 OTHER RELATED ACTIVITIES OF THE COMMITTtE During the period from May 4 through June 10, 1997, the following Subcommittee meetings were held:
e Instrumentation and control Systems and comeuters - May 28-29, 1997 The Subcommittee on Instrumentation and control Systems and computers met with the.NRC staf f to continue the review of the proposed final Standard Review Plan sections, Regulatory Guides, and Branch Technical Positions associated with digital instrumentation and control systems.
6 Human Factorm - June 3, 1997 The Subcommittee on Human Factors met with the NRC staff to discuss the latest draf t of the NRC Human Performance Program Plan and related matters.
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-e Plannina and Procedures. June-10, 1997 The Planning and Procedures Subcommittee discussed proposed ACRS activities, practices, -and procedures for conducting Committee business and organizational and personnel matters relating to ACRS and its staff.
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LIST OF FOLLOW-UP MATTERS FOR THE EXECUTIVE DIRECTOR FOR OPERATIONS e
The Committee remains interested in the full resolution of the Multiple System Responses Program issues and looks forward to a briefing on the staf f's summary report when it is completed.
e The Committee looks forward to working with the staff on the establishment of a benchmark on risk during low-power and shutdown operations and would like to discuss the staff's goals, objectives, strategy, and schedules for completing this activity during a future meeting.
e The Committee plans to review additional information, which the staf f has committed to provide to address the concerns of the ACRS members regarding the staff study on Reactor Water Cleanup System line break outside the containment, o
The Committee plans to review the proposed Generic Letter, Regulatory Guide, and related documents, including staf f's response to previous ACRS concerns and resolution of issues raised in the Differing Professional opinion, associated with steam generator tube integrity when they become available.
The Committee decided to continue its review of the proposed regulatory guidance related to impfementation of 10 CFR 50.59 (Changes, Tests and Experiments) after reconciliation of public comments.
PROPOSED SCHEDULE FOR THE 443RD ACRS MEETING The Committee agreed to consider the following during the 443rd ACRS Meeting, July 9-11, 1997:
Meetino with the Director of the NRC Office of Nuclear Reactor Reculation (NRR)
The Committee will hear presentations by and hold discussions with the Director,
- NRR, on items of mutual interest, including low-power and shutdown operations risk, NRR research needs, status of the fire inspection program, use of e risk-informed, performance-based process for prioritizing comp 11-ance issues, and coordination of ACRS reviews of NRR activities.
Accentance Criteria for Plant-Soecific Safety Goals and Derivina Lower-Tier Acceotance Criteria The Committee will hear a
presentation by and hold discussions with the ACRS Senior Fellow regarding the results of his analysis to determine the site-specific changes in core damage frequency and large, early release frequency, when these lower-tier criteria are derived from the prompt f atality quantitative health objectives. Representatives of the NRC staff will participate.
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Prooosed f,tandard Review Plan Section and Reculatory Guide for Risk-Informed.
Performance-Based Inservice InsDection The Committee will hear presentations by and hold discussions with rep-resentatives of the NRC staff regarding the proposed Standard Re-view Plan Section and Regulatory Guide for risk-informed, perfor-mance-based inservice inspection.
Meetina with NRC Commissioner McGaf fica.D - The Committee will meet i
with NRC Commissioner McGaffigan to discuss items of mutual in-terest, including ACRS activities and NRC research.
ProDosed Final Modifications to 10 CFR Part 26, Fitness-For-Duty Procram Recuirements - The Committee will hear presentations by and hold discussions with representatives of the NRC staff regarding the proposed final modifications to fitness-for-duty program requirements.
Halden Reactor Proiect - The Committee will hear presentations by and hold discussions with representatives of the NRC staff and its contractors regarding the ongoing and planned work at the OECD Halden Reactor Project in the areas of human factors, instrumenta-tion and control r.ystems, software quality, and reactor fuels.
The Committee will Use of RASCAL Code Durina Incident Resoonee hear presentations by and hold discussit/ns with representatives of the NRC staff regarding the use of the RASCAL code to calculace off-site doses.
NUREG/CR-6372. Recommendations for Probabilistic Seismic Hazard Analysier Guidance on Uncertainty and Use of Exoerts The Committee will hear presentations by and hold discussions with rep-resentatives of the NRC staff and its contractors regarding NUREG/CR-6372.
Sincerely,
/1 R. L. Seale Chairman
.