ML16328A357
| ML16328A357 | |
| Person / Time | |
|---|---|
| Issue date: | 11/13/1986 |
| From: | Fraley R Advisory Committee on Reactor Safeguards |
| To: | Stello V NRC/EDO |
| References | |
| Download: ML16328A357 (9) | |
Text
D861113 MEMORANDUM FOR: Victor Stello, Jr.
Executive Director for Operations FROM: R. F. Fraley Executive Director, ACRS
SUBJECT:
ACRS COMMENTS ON THE PRIORITIZATION OF THE FOURTH GROUP OF GENERIC ISSUES During its 319th meeting, November 6-8, 1986, the ACRS reviewed the adequacy of the proposed priority rankings for a group of Generic Issues identified in the attached Table 1, and its comments are contained in the following attachments.
~ Attachment 1 lists those issues for which the ACRS agrees with the priority rankings proposed by the NRC Staff.
~ Attachment 2 includes a list of issues for which the ACRS agrees with the priority rankings proposed by the NRC Staff, but has comments.
~ Attachment 3 identifies the Generic Issue for which the ACRS disagrees with the NRC Staff's proposed priority ranking along with the reasons therefor.
Comments on Generic Issue 61, "SRV Line Break Inside the BWR Wetwell Airspace of Mark I and Mark II Containments," have been deferred pending additional review by the ACRS.
It is requested that the NRC Staff provide written responses to the ACRS comments identified in Attachments 2 and 3.
The ACRS will continue its review of the adequacy of the proposed priority rankings for additional Generic Issues when they become available.
Attachments: As Stated TABLE 1 GENERIC ISSUES REVIEWED BY THE ACRS DURING THE 319TH, NOVEMBER 6-8, 1986 MEETING
GENERIC PRIORITY RANKINGS REFERENCE ISSUE TITLE PROPOSED BY THE STAFF DOCUMENT NUMBER 21 Vibration Qualification of DROP Memorandum from Equipment Denton, dated June 23, 1986 61 SRV Line Break Inside the RESOLVED Memorandum from BWR Wetwell Airspace of Denton, dated Mark I and Mark II Con-August 8, 1986 tainments 74 Reactor Coolant Activity DROP Memorandum from Limits for Operating Reactors Denton, dated May 30, 1986 103 Design for Probable Maximum NEARLY RESOLVED Memorandum from Precipitation Denton, dated September 4, 1985 111 Stress Corrosion Cracking of LICENSING ISSUE Memorandum from Pressure Boundary Ferritic Denton, dated Steels in Selected Environ-November 22, 1985 ments 114 Seismic-Induced Relay Chatter Covered in USI A-46, Memorandum from "Seismic Qualifica-Denton, dated tion of Equipment in June 25, 1986 Operating Plants" 115 Enhancement of the Reliability HIGH Memorandum from of Westinghouse Solid State Denton, dated Protection System July 7, 1986 TABLE 1 (Cont'd)
GENERIC PRIORITY RANKINGS REFERENCE ISSUE TITLE PROPOSED BY THE STAFF DOCUMENT NUMBER 122 Davis-Besse Loss of All Memorandum from
Feedwater Event - Short-Term Denton, dated Actions January 28, 1986 122.1.a Common Mode Failure of HIGH Isolation Valves in Closed Position 122.1.b Recovery of Auxiliary Feed-MEDIUM water 122.1.c Interruption of Auxiliary HIGH Feedwater Flow 122.2 Initiating Feed-and-Bleed HIGH 122.3 Physical Security System LOW Constraints 125 Davis-Besse Loss of All Memorandum from Feedwater Event - Long-Term Denton, dated Actions June 30, 1986 125.I.2.a Need for a Test Program Covered in Generic to Establish Reliability Issue 70, "PORV and of the PORV Block Valve Relia-bility" 125.I.2.b Need for PORV Surveillance Covered in Generic Tests to Confirm Issue 70, "PORV and Operational Readiness Block Valve Relia-bility" 125.I.2.c Need for Additional Protection RESOLVED Against PORV Failure 125.I.2.d Capability of the PORV to Covered in USI A-45, Support Feed-and-Bleed "Shutdown Decay Heat Removal Requirements" TABLE 1 (Cont'd)
GENERIC PRIORITY RANKINGS REFERENCE ISSUE TITLE PROPOSED BY THE STAFF DOCUMENT NUMBER 125.II.3 Review Steam/Feedline Break DROP Memorandum from Mitigation Systems for (Safety Concerns Denton, dated Single Failure of this Issue have August 27, 1986
been addressed in Generic Issues 125.II.1.b and 125.II.7) 125.II.4 Thermal Stress of Once-Through DROP Memorandum from Steam Generator Components Denton, dated Sept. 10, 1986 125.II.7 Reevaluate Provisions to Auto-HIGH Memorandum from matically Isolate Feedwater from Denton, dated Steam Generator During a Line Sept. 10, 1986 Break 125.II.9 Enhanced Feed-and-Bleed Covered in USI A-45, Memorandum from Capability "Shutdown Decay Heat Denton, dated Removal Requirements" August 27, 1986 125.II.14 Remote Operation of Equipment LOW Memorandum from Which Must Now be Operated Denton, dated Locally August 27, 1986 C-4 Statistical Methods for REGULATORY IMPACT Memorandum from ECCS Analysis ISSUE (RESOLVED) Denton, dated June 23, 1986 C-5 Decay Heat Update REGULATORY IMPACT Memorandum from ISSUE (RESOLVED) Denton, dated June 23, 1986 C-6 LOCA Heat Sources REGULATORY IMPACT Memorandum from ISSUE (RESOLVED) Denton, dated June 24, 1986 ATTACHMENT 1 LIST OF GENERIC ISSUES FOR WHICH THE ACRS AGREES WITH THE PRIORITY RANKINGS PROPOSED BY THE NRC STAFF GENERIC ISSUE NO. TITLE 21 Vibration Qualification of Equipment 111 Stress Corrosion Cracking of Pressure Boundary Ferritic Steels in Selected Environments 122.1.a Common Mode Failure of Isolation Valves in Closed Position 122.1.b Recovery of Auxiliary Feedwater 122.1.c Interruption of Auxiliary Feedwater Flow
122.2 Initiating Feed-and-Bleed 122.3 Physical Security System Constraints 125.I.2.a Need for a Test Program to Establish Reliablity of the PORV 125.I.2.b Need for PORV Surveillance Tests to Confirm Operational Readiness 125.I.2.c Need for Additional Protection Against PORV Failure 125.I.2.d Capability of the PORV to support Feed-and-Bleed 125.II.3 Review Steam/Feedline Break Mitigation Systems for Single Failure 125.II.4 Thermal Stress of Once-Through Steam Generator Components 125.II.7 Reevaluate Provisions to Automatically Isolate Feedwater from Steam Generator During a Line Break 125.II.9 Enhanced Feed-and-Bleed Capability 125.II.14 Remote Operation of Equipment Which Must Now be Operated Locally C-4 Statistical Methods for ECCS Analysis C-5 Decay Heat Update C-6 LOCA Heat Sources ATTACHMENT 2 LIST OF ISSUES FOR WHICH THE ACRS AGREES WITH THE PROPOSED PRIORITY RANKINGS, BUT WITH COMMENTS Generic Issue No: 74
Title:
Reactor Coolant Activity Limits for Operating Reactors Priority Ranking Proposed By The NRC Staff: DROP
ACRS Comments: The ACRS agrees with the proposed priority ranking for this issue. However, it offers the following comment:
Although the ACRS concurs, in general, in the NRC Staff's assessment of the expected savings in occupational radiation expo-sures at Boiling Water Reactors (BWRs) owing to the assumed implementation of more stringent controls, consideration might be given to refining these calcu-lations when data for BWRs, similar to those published in NUREG/CR-4485, "The Impact of Fuel Cladding Failure Events on Occupational Radiation Exposures at Nuclear Power Plants," for pressurized water reactors, become available.
ATTACHMENT 2 (Cont'd)
Generic Issue No: 103
Title:
Design for Probable Maximum Precipitation Priority Ranking Proposed By The NRC Staff: Nearly Resolved ACRS Comments: Both the description and the resolution of this issue are unclear.
The technical issue of what probable maximum precipitation (PMP) values should be used to determine flood levels at reactor sites would appear to be moot since the Staff has traditionally relied on the expertise and recommendations of NOAA for such values.
The regulatory issue of whether current NOAA values of PMP should be used to determine flood levels at existing plants has not been resolved. A decision has been made by the Staff that a request for new calculations of flood levels at NTOL plants is a backfit* and thus is not required; such calculations are being made by the Staff rather than being requested of the applicant. Presumably, no such calculations have been requested or are being made by the Staff for existing plants.
The ACRS believes that the safety issue of
possible flooding at plant sites has not been resolved for either existing or NTOL plants. Apparently, it has been resolved for future plants by changes in the Standard Review Plan.
- The ACRS has difficulty in understanding how the backfit rule can be applied because of the absence of any accepted means of quantifying the risk.
ATTACHMENT 2 (Cont'd)
Generic Issue No: 115
Title:
Enhancement of the Reliability of Westinghouse Solid State Protection System Priority Ranking Proposed By The NRC Staff: HIGH ACRS Comments: The ACRS has concerns about some parts of the Staff's analysis in connection with this Generic Issue:
~ Data used in the analysis appear to result from the treatment of testing-induced failures as random failures.
Given this approach, if one chose to increase the testing rate in an effort to increase reliability, the failure rate would probably increase as well.
~ The analysis of risk reduction appears to put principal emphasis on the operation (or lack of operation) of the undervoltage trip system.
Since experience indicates that this trip system is less reliable than the shunt trip system, the approach seems incomplete. The ACRS recognizes that the original intent of an undervolt-age trip system was to satisfy a
"fail safe" criterion. However, notable failures have occurred in this system that were not "fail safe."
ATTACHMENT 3 GENERIC ISSUE FOR WHICH THE ACRS DISAGREES WITH THE PRIORITY RANKING PROPOSED BY THE NRC STAFF Generic Issue No: 114
Title:
Seismic-Induced Relay Chatter Priority Ranking Proposed by the NRC Staff: Covered in USI A-46, "Seismic Qualification of Equipment in Operating Plants" ACRS Comments: The Staff has concluded that the work being done in connection with resolution of USI A-46 and the Seismic Margins program as well as related programs cover the intent of Generic Issue 114 and therefore it need not be pursued as a separate issue. The ACRS does not agree and finds the proposed resolu-tion to Generic Issue 114 unacceptable, for the following reasons:
~ Generic Issue 114 was to address the effects of seismic-induced relay chatter upon the safety and safety-related electrical and control systems as applied to all plants. USI A-46 applies to and is limited to operating plants which were docketed prior to 1972. There is no indication that USI A-46 will be expanded to cover plants docketed since 1972. Therefore, the ACRS concludes that these plants will not have been adequately reviewed for the effects of seismic-induced relay chatter.
~ There is no indication that the Seismic Margins Program and related programs will adequately address seismic-induced relay chatter for earthquakes above the SSE.