ML20059C845
| ML20059C845 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/04/1993 |
| From: | Michelson C Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2883, NUDOCS 9311020034 | |
| Download: ML20059C845 (100) | |
Text
- i CERTIFI5'DBY: DATE ISSUED: 8/2/93
' C.<Michelson - 8/4/93 ,
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON THE ADVANCED BOILING WATER REACTORS (GE)
SITE VISIT AND MEETING JUNE 15-17, 1993 SAN JOSE, CALIFORNIA PURPOSE The ABWR Subcommittee conducted a site visit to the GE Nuclear Energy facility in San Jose, California, on June 15 and 16 1993.
The purpose of this visit was to gather information associated with the review of the GE standard Safety Analysis Report (SSAR). In addition, the Subcommittee held a meeting on June 17, 1993, at the Holiday Inn Park Central Plaza, San Jose, CA, to discuss matters related to the ABWR review. The meeting was held entirely in open session. A copy of the meeting agenda is attached. Dr. El-Zeftawy was the designated federal official for the meeting. No written comments or requests for time to make oral statements were received from members of the public.
ATTENDEES Principal meeting attendees included:
ACRS NRC Michelson, Chairman J. Wilson, NRR J. Carroll, Member C. Poslusny, NRR I. Catton, Member S. Koenick, NRR P. Davis, Member T. Kress, Member '
W. Lindblad, Member R. Seale, Member C. Wylie, Member M. Stella, Fellow M. El-Zeftawy, Staff OTHERS GE N. Fletcher, Doe C. Quirk B. Simon C. Sawyer R. Strong J. Power C. Buchholz G. Ehlert A. Beard J. Fox J. Duncan C. Oza U. Saxena A. James R. Kumar B. Genetti 02007.--O DF~ M ADO 2 01?!U, 30
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ABWR/GE Subete. 6/15-17/93 Minutes CHAIRMAN'S OPENING REMARKS ,
In his opening remarks, Mr. Michelson stated that the ABWR Subcommittee members have visited the GE Nuclear Energy facility on June 15 and 16, 1993, at San Jose, California. The Subcommittee gathered information associated with the review of the GE Standard Safety Analysis Report (SSAR). Attachment I presents a _ brief summary of the items that were pursued at the site visit. The -
handouts from the site visit are included in (Attachment II).
I. Summarv of the Items from the GE Site Visit (6/15-16/93)
- 1. Coolina of solid state electronic components.
The Subcommittee pursued the issue of how GE qualifies the equipment and cabinets to provide adequate cooling for the solid state electronic components. GE represen-tatives stated that the cabinets will be designed without any requirements for forced-convection cooling.
The Subcommittee pursued the issue of location of electronic components in the electrical equipment rooms outside secondary containment to assure that there are no sources of high- or moderate-energy fluids that would potentially jeopardize such equipment.
GE representatives stated that there are no such pipes in the electrical equipment rooms.
GE representatives also stated that the electrical equipment rooms are protected from fire by designing the floor, ceiling, and divisional barriers as fire barriers.
Mr. Michelson commented that such statements from GE representatives should be incorporated into the SSAR. GE agreed.
- 2. Location of Diesel Enaines Mr. Michelson expressed concern regarding the location of diesel generators (D.G.) with respect to the electrical equipment rooms. Currently the D.G.'s are located immediately above the electrical equipment rooms posing ,
a fire hazard situation. This item will be discussed at the July 28, 1993, ACRS Subcommittee meeting regarding fire protection.
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Carlyle Michelson - 1/4/93
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON THE ADVANCED BOILING WATER REACTORS (GE)
NOVEMBER 19-20, 1992 BETHESDA, MARYLAND ,
PURPOSE The purpose of this meeting was to continue the review of the draft final safety evaluation report (DFSER) for the GE/ABWR design. The meeting began at 8:30 a.m. on November 19, IS92 and adjourned at 5:45 p.m. The meeting was reconvened at 8:30 a.m. on November 20, 1992 and adjourned at 3:30 p.m. The meeting was held entirely in open session. No written comments or requests for time to make oral statements were received from members of the public. Dr. El-Zeftawy was the cognizant staff engineer for this meeting. The principal attendees were as follows:
ATTENDEES Principal meeting attendees included:
ACRS NRC C. Michelson, Chairman C. Poslusny, NRR I. Catton, Member H. Pastis, NRR P. Davis, Member J. Wilson, NRR P. Shewmon, Member W. Burton, NRR C. Wylie, Member H. Walker, NRR e R. Costner, Consultant C. Li, NRR M. El-Zeftawy, Staff T. Cheng, NRR -
S. Lee, NRR GE D. Terao, NRR H. Brammer, NRR J. Fox J. Lee, NRR J. Chambers J. Raval, NRR M. Munson J. Stewart, NRR B. Simon J. Guo, NRR K. Gregorie A. Howe, NRR M. Nikahd J. Lyons, NRR O. Saxena G. Georgiev, NRR F. Paradiso A. Mendiola, NRR G. Miller T. Kim, NRR J. Power T. Polich, NRR M. Herzog
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Iy Nuclear Reactors Branch
SUBJECT:
WORKING COPY OF THE
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED BOILING WATER REACTORS (GE), NOVEMBER 19-20, 1992 -
BETHESDA, MARYLAND A working copy of the Summary / Minutes for the subject meeting is attached for your review. I would appreciate your review and comment as soon as possible. Copies are being sent to the ACRS Members and Consultant, who attended the meeting for information and/or review.
Attachment:
p1Eama m As stated '
cc: I. Catton, Member P. Davis, Member , [ #
P. Shewmon, Member !
C. Wylie, Member R. Costner, Consultant S. Duraiswamy R. Savio '
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'S o J MEMORANDUM FOR: Medhat El-Zeftawy, Senior Staff Engineer Nuclear Reactors Branch FROM: Carlyle Michelson, Chairman Advanced Boiling Water Reactors Subcommittee
SUBJECT:
CERTIFICATION OF THE
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED BOILING WATER REACTORS (GE), OCTOBER 21, 1992 -
BETHESDA, MARYLAND I hereby certify that, to the best of my knowledge and belief, the Minutes of the subject meeting issued November 18, 1992, are an accurate record of the proceedings for that meeting.
d Carly14 Miche4 eon, Chairman fh 20, l9V*?_
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MEMORANDUM FOR: Carlyle Michelson, Chairman .
Advanced Boiling Water Reactors Subcommittee.
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FROM: Medhat'El-Zeftawy, Senior Staff Engineer -F . ;
Nuclear Reactors Branch 1
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SUBJECT:
WORKING COPY . 0F THE
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON-ADVANCED BOILING :
WATER REACTORS (GE) , OCTOBER 21, 1992 -
BETHESDA, MARYLAND 1
A working copy of the Summary / Minutes for the subject meeting is
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attached for your review. I would appreciate your review- and' ;
comment as soon as-possible. Copies are being sent,to_the ACRS .
4 Members and Consultant, who attended the meeting for information and/or review. s i
Attachment:
As stated ,
cc: J. 1 >}. yyjj gj g p .F P. Davis ;
R. Costner S. Duraiswamy '
R..Savio G. Quittschreiber ;
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ACRS Fellows i
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LATF. 155 2L: 11/16i92 Carlyle Michelson 11/20/92
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON THE '
ADVANCED BOILING WATER REACTORS (GE)
OCTOBER 21, 1992 BETHESDA, MARYLAND PURPOSE The purpose of this meeting was to begin the review of the draft final safety evaluation report (DFSER) for the GE/ABWR design. The meeting began at 8:30 a.m., adjourned at 5:00 p.m. and was held entirely in open session. No written comments or requests for time to make oral statements were received from members of the public.
The principal attendees were as follows:
ATTENDEES:
ACRS HEC C. Michelson, Chairman R. Nease, NRR J. Carroll, Member J. Wilson, NRR I. Catton, Member C. Poslusny, NRR P. Davis, imminent Member D. Terao, NRR R. Costner, Consultant G. Georgiev, NRR '
M. El-Zeftawy, Cognizant Staff Engineer H. Richings, NRR M. Hum, NRR H. Pastis, NRR G_E M. Rubin, NRR J. Fox J. Sharkey, NRR J. Chambers B. Mendelsohn, NRR S. Boynton, NRR Others J. Lee, NRR W. Burton, NRR Y. Kim, NUS A. Mendiola, NRR V. San Angel, Bechtel J. Lyons J. Raval, NRR F. Young, RGI CHAIRMAN'S OPENING REMARKS In his opening remarks, Mr. Michelson stated that the purpose of this meeting is to examine the resolution of open items and to discuss with the NRC staff and GE representatives the approach been taken to close these open items. Items that are still open need to be discussed. Mr. Michelson solicited any feedback from the subcommittee members and reminded them that the inspections, tests, analyses, and acceptance criteria (ITAAC) associated with certain chapters will not be part of the discussion for this meeting. This is due to the incompleteness of the ITAAC and was based on the staff's request to postpone such discussion at a later meeting.
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. , I Summary / Minutes ABWR October 21, 1992 -l NRC Staff Presentation Mr. C. Poslusny, NRR, stated that the staff is reviewing GE's '
application for design certification in accordance with the=
applicable regulatory standards of NUREG-0800, " Standard Review I Plan" (SRP), and in accordance with Regulatory Guide 1.70,
" Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition)." ,
The staff is also using the guidance from the staff requirements 4 memoranda (SRM) for each of the Commission's policy papers listed in Chapter 1 of the DFSER. In these Commission papers, the staff ,
addressed the policy issues that, for evolutionary light water reactors, modify SRP review criteria or address issues, such as severe accidents, that are beyond the scope of the SRP.
In each section of the DFSER in which the staff found the ABWR to be acceptable, according to the review criteria and guidance,.the staff indicates that, except as stated otherwise, GE has submitted a sufficient amount of design detail for the NRC to make its safety finding. In areas such as radiation protection and airborne concentrations, digital computer instrumentation and control systems, detailed piping design, and control room human factors engineering, the staf f has made its finding by reviewing the design acceptance criteria (DAC) for acceptance, as described in SECY 196, " Development of Design Acceptance Criteria for the Advanced Boiling Water Reactor (ABWR)," and SECY-92-299, " Development of Design Acceptance Criteria for the Advanced Boiling Water Reactor (ABWR) in the ~ Areas of Instrumentation and Centrols (I&C) and Control Room Design."
The staff concludes that the DFSER contains no new policy issues.
The staff will issue the DFSER to GE to inform it of the staff's current findings, including outstanding technical issues that should be resolved. In the FSER for the ABWR, the staff will indicate the resolutions of issues included in the DFSER.
Desion Control Document Mr. J. Wilson, NRR, described the form and content of the design control document (DCD) in the design certification rule. He stated that the DCD is a two-tiered document (Tier 1 and Tier 2). Tier 1 consisting of scope, design descriptions, ITAAC, site parameters, and interface requirements, and Tier 2 consisting of supporting information.
For Tier 1, the applicant for design certification will extract the information from its application, the SSAR and submit it to NRR for ;
review. The staff will provide an evaluation of the proposed DCD in a final safety evaluation report (FSER), and the applicant would then prepare a revised DCD. This later version of the DCD will be l
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CERTIFIED EY: ) DATE ISSUED: 11/17/92 Carlyle lhchelson - 11/22/92
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED BOILING WATER REACTORS AUGUST 19, 1992 BETHESDA, MARYLAND INTRODUCTION The ACRS Subcommittees on Advanced Boiling Water Reactors (ABWR).
held a meeting on August 19, 1992, in Room P-110, 7920 Norfolk Avenue, Bethesda, Maryland, to review the General Electric Nuclear Energy's (GE'S) and the NRC staffs responses to the issues included in the April 13, 1992 ACRS Letter to the NRC Executive Director for operations regarding the Draft safety Evaluation Report (DSER) for the GE ABWR design, and the ABWR passive vent actuation and its consequences. The entire meeting was open to_public attendance.
Mr. E. Igne was the cognizant ACRS staff engineer for this meeting.
The presentation schedule for the meeting is included in the Attachment. The meeting was convened at 8:30 a.m. and adjourned at 5:20 p.m.
ATTENDEES ACPS C. Michelson, Chairman D. Ward, Member C. Wylie, Member I. Catton, Member R. Costner, Consultant NRR J. Wilson M. Case M. Rubin J. Burns J. Knox P. Campbell !
W. Farmer T. D'Angelo W. Beckner W. Burton R. Palla G. Bagchi J. Stewart J. Lyons D. Terao D. Notley H. Brammer l
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Advanced Boiling 2 October 16', 1992 Water Reactors i General Electric Nuclear Enerav l C. Sawyer J. Chambers -i J. Fox C. Buchholz B. Simon J. Quirk U. Saxena M. Ross G. Ehlert '
Other V. San Angelo, SERCH Licensing /Bechtel Power K. Graney, SERCH Licensing /Bechtel Power i N. Fletcher, DOE /ALWR -
i OPENING REMARKS Mr. Michelson, Chairman of the Subcommittee, convened the meeting -
at 8:30 a.m., and stated that GE has provided written responses to ;
the questions delineated in the April 13, 1992 letter. GE's reply will form the basis for discussion at today's meeting. He stated ,
that Mr. Costner, ACRS consultant, has provided a preliminary report on the Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) process. ,
Introductory Comments by GE - Mr. J. Ouirk. GE Mr. Quirk said that Mr. G. Ehlert, GE, will discuss the' items on control of building flooding and adequacy of physical separation.
4
. 1 Control of Buildina Floodina - Mr. G. Ehlert. GE Mr. Ehlert discussed the concern on control of building flooding.
He stated that the main issue is centered on the service water system in the basement and the high energy lines in the steam-tunnel. In reply to a question by Mr. Michelson, Mr. Ehlert stated that the Standard Safety Analysis Report (SSAR) would include requirements for all welded pipes (i.e., no expansion joints or bellows) for the reactor service water (RSW) and reactor cleanup water (RCW) systems piping inside the control room. Further, Mr. ,
Ehlert stated in reply to Mr. Michelson's comments, that leak-
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%...../ November'17, 1992 MEMORANDUM FOR: Carlyle Michelson, Chairman Advanced Boiling Water Reactors Subcommittee
'f FROM: Elp;mio Igne, Senior Staff Engineer Nuclear Reactors Branch
SUBJECT:
WORKING COPY OF THE
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED BOILING WATER REACTORS, AUGUST 19, 1992 -
BETHESDA, MARYLAND A working copy of the Summary / Minutes for the subject meeting is attached for your review. I would appreciate your review and comment as soon as possible. Copies are being sent to the ACRS '
Members and Consultant, who attended the meeting for information '
and/or review.
Attachment:
As stated cc: D. Ward C. Wylie . .
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R. Costner S. Duraiswamy R. Savio G. Quittschreiber '
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MEMORANDUM FOR: Elpidio Igne, Senior Staff Engineer Nuclear Reactors Branch FROM: Carlyle Michelson, Chairman Advanced Boiling Water Reactors Subcommittee
SUBJECT:
CERTIFICATION OF THE
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED BOILING WATER REACTORS, AUGUST 19, 1992 -
BETHESDA, MARYLAND I hereby certify that, to the best of my knowledge and belief, the Minutes of the subject meeting issued November 17, 1992, are an accurate record of the proceedings for that meeting.
Carlyle*Micheliifon, Chairman W Sk,l$fk Date I
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December 1, 1992 MEMORANDUM FOR: ACRS Members FROM: Elpidio Igne, Senior Staff Engineer Nuclear Reactors Branch
SUBJECT:
CERTIFICATION OF THE MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON ADVANCED BOILING WATER REACTORS, AUGUST 19, 1992 - BETHESDA, MARYLAND The minutes for the subject meeting, issued November 17, 1992, have been certified as the official record of the proceedings of that meeting. A copy of the certified minutes is attached.
Attachment:
As stated ,
cc: R. Costner R. Fraley R. Savio S. Duraiswamy G. Quittschreiber ACRS Technical Staff ACRS Fellows i
Mr. James M. Taylor 9 April 13, 1992 We do not expect to receive a separate reply to the above items if f they are covered appropriately in the final SER. We will keep you informed of any additional concerns as our review proceeds.
Sincerely, l .
David A. Ward Chairman
References:
- 1. GE Nuclear Energy, Standard Safety Analysis Report, " Advanced Boiling Water Reactor," Chapters 1 through 20 (Amendments 1 through 18)
- 2. SECY-91-153, dated May 24, 1991, for the Commissioners from James M. Taylor, Executive Director for Operations, NRC,
Subject:
Draf t Safety Evaluation Report (DSER) on the General Electric Company Advanced Boiling Water Reactor Design Covering Chapters 1, 2, 3, 4, 5, 6, and 17 of the Standard Safety Analysis Report (SSAR)
- 3. SECY-91-235, dated August 2, 1991, for the Commissioners from James M. Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapters 1, 3, 9, 10, 11, and 13 of the SSAR
- 4. SECY-91-294, dated September 18, 1991, for the Commissioners from James M. Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapter 7 of the SSAR
- 5. SECY-91-309, dated October 1,1991, for the Commissioners from James M. Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapter 19 of the SSAR,
" Response to Severe Accident Policy Statement"
- 6. SECY-91-320, dated October 15, 1991, for the Commissioners from James M. Taylor, EDO, NRC,
Subject:
DSER on the GE Advanced Boiling Water Reactor Design Covering Chapter 18 of the SSAR ,
- 7. SECY-91-355, dated October 31, 1991, for th,e Commissioners from James M. Taylor, EDO, NRC,
Subject:
DSER on the GE Bolling Weter Reactor Design Covering Chapters 1, 2, 3, 5, 6, 8, 9, 10, 12, 13, 14, and 15 of the SSAR
- 8. Electric Power Research Institute, " Advanced Light Water Reactor Utility Requirements Document" (Volume II)/ALWR Evolutionary Plant, Revision 3, Issued November 1991 3 l
4
s Mr. James M. Taylor 7 April 13, 1992 containment. The exclusion of these breaks was based '
erroneously on an analysis of the effects of suppression pool bypass events on overall risk. However, the analysis f ailed to take into account that the bypass path (e.g. , RWCU System pipe br(ak) could be the initiator for the core-damage event.
e The PRA analysts took credit for the RWCU System as a heat removal system in all sequences where reactor pressure is assumed to remain high. The analysts assumed that the capacity of the non-regenerative heat exchanger (NRHX) capacity is adequate to remove the decay heat. The appears to be adequate; however, our calculations indicate that the outlet temperatures on the RWCU System side and cooling water side of the NRHX would exceed the design limits for the piping. Furthermore, a temperature sensor between the NRHX and the RWCU. System pumps in the present design would automatically isolate the NRHX on high temperature, making it unavailable.
The items mentioned above are among a number of issues that were identified. It is impertant for the staff to ensure that the shortcomings of the RWCU System and PRA related portions of the SSAR are not indicative of problems in the remainder of that report.
- 11. i Plant Desian Life and Acina Manacement We recommend that the SSAR clearly define the scope of the 60-year design life for the ABWR and describe a program plan for achieving it. This program should include those aging management measures which are necessary to maintain the plant within its design basis throughout its design life. This program should specify the original design and application criteria and, where required, the projected refurbishment or replacement requirements with appropriate rationale. To the extent applicable, the lessons learned from the NRC's Nuclear Plant Aging Research Program as well as other agin projects should be incorporated into this program.g research We note that the EPRI URD (Volume II, Chapter 1, Paragraph 3.3) includes a requirement for a plant design life of "60 years without necessity for an extended refurbishment outage and discusses a ")
the requirements Paragraph 11.3.
for its achievement in
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Mr. James M. Taylor 5 April 13, 1992 accidents. The policy statement published in the Federal Recister of August 8, 1985, also states that "Accordingly, within 18 months of the publication of this Severe Accident Policy Statement, the staff will issue guidance on the form, purpose and role that PRAs are to play in severe accident analysis and decision making for both existing and future plant designs...." The Statement says further, "The PRA guidance will describe the appropriate combination of deterministic and probabilistic considerations as a basis for severe accident decisions."
The staff has yet to produce the promised guidance. We urge that the staff formulate a set of criteria that it plans to use in making severe accident decisions. This should include the way in which the results of a PRA are to be used in the process (not just whether the PRA has been done properly).
- 9. Containment Hydrodynamic Loads Air-clearing loads on containment structures are the result of a complex process resulting from the drywell air being forced into the wetwell by the primary system blowdown. The water in the vent system is pushed down and out until the horizontal vents are cleared. The water-clearing process produces a jet of water into the suppression pool which causes a load on the outer part of the wetwell wall. ThiL water clearing is followed by an air-steam mixture which creates a large bubble as it exits into the pool. The steam condenses but the air expands forcing the water above it up into the wetwell air space. The wetwell air space is compressed due to the momentum of the water in the layer above the bubble.
4 The wetwell air space will be subjected to an energetic two-ph.sse eruption as a result of the air-clearing process. The vacuum breakers which are in the vicinity will be exposed to this environment unless protected. The SSAR should describe what the environment will be and what protective measures, if any, are needed to ensure survival of the vacuum breakers. If a vacuum breaker does not close, the suppression pool is bypassed and the wetwell/drywell pressures will rise at a rate dictated by the capability of some means other than the suppression process (e.g., containment sprays) to remove heat and condense steam. The SSAR should contain an analysis of such a situation.
The early work to address problems arising from analyses of the Mark I, II, and III containments is not sufficient to address similar processes that will occur following a LOCA in an ABWR containment. The ABWR is different for two reasons:
(a) the volume of the wetwell air space in the ABWR is approximately that of a Mark II, and (b) the impact of the
. i Mr. James M. Taylor 3 April 13, 1992 $
We believe that the SSAR should describe and the staff should evaluate the adequacy of proposed separation barriers for the !
full range of events and conditions for which separation must .
be ensured. We continue to recommend that systems required j for safe shutdown not share a common Heating, Ventilating and :
Air Conditioning (HVAC) System during normal plant operation. ,
The secondary containment HVAC System for the ABWR is such a shared system.
- 3. Protection of Environmentally Sensitive Eauinment The ABWR makes extensive use of environmentally sensitive equipment (including solid-state electronic components) for essential protection, control, and data transmission i functions. Such components are known to be susceptible to ;
adverse environmental changes, particularly temperature extremes. We are concerned that a number of these components ;
may be located in plant areas where postulated events such as .
pipe breaks, fire, internal flooding, or loss of room cooling ,
may create an adverse environment. Such environments need to '
be identified in the SSAR to ensure appropriate environmental <
qualification of the equipment. l
- 4. Review of Chilled-Water Systems The ABWR uses large chilled-water systems to provide essential environmental cooling, which in turn includes cooling of the solid-state electronic components. Because there was no SRP for chilled-water systems, the staff used other guidance such as SRP Section 9.2.2 (Reactor Auxiliary Cooling Water Systems) ,
when the safety evaluation was performed. However, this t !
guidance is not appropriate for the evaluation of refrigeration systems.
The NRC staff needs to evaluate the performance of chilled-water systems under varying accident heat loads and during loss-of-offsite-power events, and to consider their ability to restart and function after a prolonged station blackout. The DSER sections which should evaluate the performance of large chiller packages do not address these issues. We believe they should.
- 5. Use of Leak-Before-Break Methodoloav It is our understanding that GE will not propose the use of leak-before-break methodology for the ABWR standard plant.
Thus, the DSER should be revised to ensure that consideration is given to pipe break effects for all systans and locations.
This may introduce additional structural protection and environmental qualification requirements in the SSAR.
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April 13, 1992 Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Taylor:
SUBJECT:
REVIEW OF THE DRAFT SAFETY EVALUATION REPORTS ON THE GE ADVANCED BOILING WATER REACTOR DESIGN During the 383rd and 384th meetings of the Advisory Committee on Reactor Safeguards, March 5-7 and April 2-4, 1992, we discussed the Draft Safety Evaluation Reports (DSERs) on the Advanced Boiling Water Reactor (ABWR) design which is described by GE Nuclear Energy (GE) in its Standard Safety Analysis Report (SSAR), as amended, and for which GE has applied for design certification in accordance with 10 CPR Part 50, Appendix 0. The DSERs which are the basis for this report were sent to the Commissioners for information as six-These SECY . papers (SECY-91-153, 235, 294, 309, 320, and 355).
generally cover the SSAR and its first eighteen amendments. Our Subcommittee on Advanced Boiling Water Reactors discussed these papers with representatives of GE and the NRC staff during its meetings on September 18 a ' " October 23, 1991 and January 23-24 and February 20-21, 1992. ...i also had the benefit of the documents referenced.
Our first report to you concerning the DSER for this project was dated November 24, 1989. That report conveyed our commente, on Module 1 of the design (former GE designation). We also sent a report to you on July 18, 1991, outlining several ABWR design concerns that developed during subsequent review.
We note a marked improvement in the quality of the staf f's DSER evaluations since our November 24, 1989 report. The staff reviewers appear to be following the guidance outlined in the applicable Standard Review Plans (SRPs) to the extent possible, and they are asking good in-depth questions in most areas. ;
The SECY-91-161 schedule indicates that the Final Design Approval If we l (FDA) is to be issued before the end of Calendar Year 1992.
are to provide our final report on this subject in December 1992, it will be necessary that we receive a complete and final SER no later than early September 1992. There are now more than three hundred open items in the DSERs, many of which are major. In
l e t Mr. James M. Taylor 5 July 18,1991 but also to the design of all Advanced Light Water Reactor designs.
Sincerely, David A. Ward Chairman
References:
- 1. Letter dated August 17, 1989 from Charles L. Miller, Office of Nuclear Reactor Regulation, NRC, to Patrick W. Marriott, General Electric Company, enclosing Draf t Safety Evaluation Report Related to the Final Design Approval and Design Certification of the Advanced Boiling Water Reactor, dated August 1989.
- 2. Letter dated August 7, 1987 from Thomas E. Murley, Office of Nuclear Reactor Regulation, NRC, to Ricardo Artigas, General '
Electric Company, enclosing GE Advanced Boiling Water Reactor, Licensing Review Bases, dated August.1987.
- 3. GE Nuclear Energy, Standard Safety Analysis Report, Advanced Boiling Water Reactor, Chapters 1 through 20.
Mr. James M. Taylor 3 July 18,1991 l l
i
- 3. Environmental Protection for Solid-State Electronics l The ABWR makes extensive use of solid-state electronic '
components for essential protection, control, e data transmission functions. Such components are known to be susceptible to adverse environmental changes, particularly .
temperature extremes. We are concerned that a number of these components may be located in plant areas where postulated events such as pipe rupture, fire, internal flooding, or loss ,
of room cooling may create an adverse environment. The response of such components to the environmental change may be unpredictable and lead to unacceptable system interactions or responses. The behavior of solid state electronic com-ponents in environments created by off-normal or accident situations needs to be considered before the adequacy of any physical separation and environmental control measures can be evaluated.
- 4. Review of Chilled-Water Systems The ABWR makes extensive use of large chilled-water systems to provide essential environmental cooling functions including those for the solid-state electronics. Since there is no SRP '
for chilled-water systems, the staff uses other guidance such as SRP Section 9.2.2 (Reactor Auxiliary Cooling Water Systems) when performing its safety evaluation. This guidance does not include evaluation of the large refrigeration equipment that is required for chilling the closed-cycle cooling water.
The IGC staff and GE need to evaluate the safety implications of chilled-water systems, including performance under varying accident heat loads, loss-of-offsite-power loading charac-teristics, and ability to restart and function after a prolonged station blackout. The NRC staff should develop appropriate guidance for such reviews by preparing a suitable SRP now.
- 5. Use of Leak-Before-Break Methodoloav Outside of Primary Containment In our report of March 14, 1989 to then NRC Chairman Zech on
" Additional Applications of Leak-Before-Break Technology," we expressed our belief that an avenue for consideration of further extension of the leak-before-break (LBB) concept should exist. This is still our position. We are concerned that the NRC staff is not giving serious consideration to GE proposals to extend the concept to systems outside of the l primary containment because the staff feels constrained by General Design Criterion 4 which does not propose review of I methodology.
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[gouc c; UNITED STATES y ,( ) NUCLEAR REGULATORY COMMISSION j - :. t ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o, j wAsmwcTon, o. c. rosss
%,, f July 18, 1991 Mr. James M. Taylor Executive Director for Operations '
U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Taylor:
SUBJECT:
CONCERNS RELATED TO THE GENERAL ELECTRIC ADVANCED BOILING WATER REACTOR DESIGN During the 375th meeting of the Advisory Committee on Reactor Safeguards, July 11-13, 1991, we discussed the status of the Advanced Boiling Water Reactor (ABWR) design, described in the Standard Safety Analysis Report (SSAR), for which the General Electric Company (GE) has applied for design certification in accordance with 10 CFR Part 50, Appendix 0. Our Subcommittee on Advanced Boiling Water Reactors also discussed this matter during its meetings on October 31, 1990, and May 30, 1991, with represen-tatives of GE and the NRC staff. We also had the benefit o.? the documents referenced.
1 Our previous letter to you concerning the ABWR design was dated November 24, 1989, and conveyed our comments on Module 1 of the Draft Safety Evaluation Report (DSER). Since this letter, we have been kept apprised of the design and the status of the review while awaiting receipt of additional DSERs. The staff now says that DSER preparation by modules will be discontinued in favor of prepara-tion by SSAR chapters and Standard Review Plan (SRP) sections.
To ensure the completeness of our review, it will be necessary to account for any additions or revisions to each DSER as forwarded by a SECY subsequent to issuance of our respective comment letter.
An arrangement acceptable to us is needed to ensure the identifica-tion of any additions or revisions, and we should agree on an appropriate time for their review. Our comments will not be complete, however., until we have submitted a report to the Commission concerning the final SER on which we expect to comment by mid-November 1992.
Our activities subsequent to the completion of our November 1989 letter have focused on several design concerns that were discussed with GE and the NRC staff in an effort to ensure an early awareness and understanding. We believe that it is appropriate to document them here for timely consideration and resolution in appropriate DSER sections. We expect to have additional items later. We do
e- ') i f
REFERENCES:
- 1. Report from David A. Ward (Chairman, ACRS) to James M. Taylor (EDO), " Concerns Related to the General Electric Advanced Boiling Water Reactor Design," July 18, 1991.
- 2. Report from David A. Ward (Chairman, ACRS) to James M. Taylor (EDO), " Review of the Draft Safety Evaluation Report on the GE Advanced Boiling Water Reactor Design," April 13, 1992.
- 3. Steven E. Mays and Mark E. Stella, "ABWR Reactor Water Cleanup System Review," A Report Prepared for the Advisory Committee on '
Reactor Safeguards, July 30, 1992.
" Certification of the Minutes of the ACRS Subcommittee Meeting on Advanced Boiling Water Reactors, August 19, 1992 - Bethesda, Maryland," December 1, 1992.
" Certification of the Minutes of the ACRS Subcommittee Meeting on Advanced Boiling Water Reactors (GE), October 21, 1992 -
Bethesda, Maryland," January 8, 1993. -
" Certification of the Minutes of the ACRS Subcommittee Meeting on Advanced Boiling Water Reactors (GE) , November 19-20 -
Bethesda, Maryland," January 8, 1993.
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b PROPOSED SCHEDULE FOR ACRS ABWR SUBCOMMITTEE MEETING.
SAN JOSE, CALIFORNIA JUNE 17, 1993 The ABWR subcommittee meeting will review matters related to the GE ABWR Standard Safety Analysis Report (SSAR), including the following:
- 1. The subcommittee chairman and other ACRS members will summarize observations and ask any additional questions concerning each agenda item covered during the visit to the GE facility and establish for LLe official record any important considerations relating to the ACRS review of the SSAR.
- 2. The subcommittee would like to discuss the following items to the extent defined:
o Plant Design Life and Aging Management.
Discuss the design life of the plant and the program for achieving it, including:
- a. Design and application criteria.
- b. Refurbishment or replacement requirements.
- c. Aging management measures, e Station Grounding and Surge Protuction Discuss the station grounding and surge protection measures including:
- a. Overall plant station grounding and connections to switchyards,
- b. Lightning protection measures.
- c. Grounding system and protection of sensitive solid-state electronic components.
- d. Basis for the requirement of resistance for grounding systems to absolute earth of 0.05 ohms.
- e. Design requirements for the main, auxiliary, reserve transformers, including ratings, design loads, temperature ratings, insulation class, basic insulation levels, etc.
. Corrosion Control for Structures Discuss corrosion control measures to achieve design life considering the environmental ef fects, including protection of:
- 5. Representative Conceptual Design of Ultimate Heat Bink and Reactor Service Water System Including:
- a. Safety evaluation (e.g., PRA) that verifies acceptability of proposed concept under various postulated accident or event conditions,
- b. Adequacy of the interface requirements.
- c. Verification that a control building flooding problem does not exist for the representative conceptual design, including consideration of isolation valve arrangements and ,
operability, feasibility and time available for needed :
response, and the assumption of a single active component !
failure during the response.
(This item is a follovup to item 1 in Ref.1 and 2) .
- 6. Design and Safety Evaluation of RWCS h
This agenda item should include a detailed dialogue on the ACRS report, "ABWR Reactor Water Cleanup System RevicW", Ref. 3.
The discussion should be with those GE personnel who have examined the report.
- 7. Computer Aided Design Demonstration Plant areas of special interest to the subcommittee are the ,
plant control building, the RWCS layout and compartments, the Reactor Building Common Ventilation System, and the main steam and feedwater compartments outside of the primary containment.
I d
l PROPOSED SCHEDULE FOR ACRS ABWR SUBCOMMITTEE VISIT TO GENERAL ELECTRIC NUCLEAR ENERGY (GE) FACILITY SAN JOSE, CALIFORNIA JUNE 15-16, 1993 The purpose of the subcommittee visit is to gather information associated with the ACRS review of the GE ABWR Standard Safety I Analysis Report (SSAR). This is neither a review nor a Subcommittee meeting. Information gathered during this visit will be used by the full Committee and Subcommittee in their deliberations of this matter during future open meetings.
The subcommittee is interested in (1) talking to the people who actually performed the design and safety evaluation work for the i SSAR, and in (2) looking at the design work which is under way or !
completed, or (3) looking at the documents which ensure that the following concerns will be accommodated by other suitable means:
Design of Physical Feparation Barriers and Associated Penetrations l
- a. Determination of the events which may challenge the l physical separation barriers and their penetrations (e.g. ,
i fire, internal floods, and pipe ruptures.)
- b. Determination of the environmental disruptions associated with such events (e.g. , pressure and temperature increases, flooding, spraying, smoke, and pipe whip and jet ;
impingement). j
- c. Specification of the design requirements for those !
components which make up the separation barriers (e.g., !
requirements for walls, doors, pipe and electrical penetrations, and HVAC isolation dampers and penetrations) .
Of particular interest are (1) calculation of subcompartment j pressures for postulated failures in the main steam and )
feedwater system compartments outside of the primary ;
containment and in the Reactor Water Cleanup System (RWCS) compartments, and (2) the design requirements for physical separation boundary components which confine these pressures.
(This item is a follovup to item 2 in Ref. 1 and 2. Prepare for an indepth dialogue).
d Cooling of Bolid State Electronic Components
- a. What are the bases for determining the capability of such '
components to function acceptability at the maximum specified room temperature (i.e. , tests, analyses, or other means)?
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. q'c, UNITED STATES y' gf 'e NUCLEAR REGULATORY COMMISSION S- ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
$ ~/[I
%, ' % ;j WASHINGTON, D. C. 20555 ,
- "* May 18, 1993 MEMORANDUM TOR: Dennis M. Crutchfield Associate Director for Advanced Reactors and License Renewal FROM: John T. Larkins, Executive Director Advisory Committee for Reactor Safeguards
SUBJECT:
ABWR SUBCOMMITTEE VISIT AND MEETING The ACRS Subcommittee on Advanced Boiling Water Reactors is planning to visit the GE facility in San Jose, California, on June 15 and 16, 1993. The purpose of this visit is to gather information associated with the review of the GE Standard Safety Analysis Report (SSAR). A proposed schedule for this visit is attached.
In addition, the Subcommittee is scheduled to meet with the representatives of GE and the NRC staff, as appropriate, on June 17, 1993, in an open session at Holiday Inn Park Central Plaza,- San Jose, CA, to discuss matters related to the ABWR review. Attached is a proposed schedule for this meeting. Please send copies of the '
attached to appropriate GE personnel.
.A- ,
6:' ,
John T. Larkins Executive Director, ACRS
Attachment:
As stated .i
)
cc: J. Wilson l C. Poslusny l l
Summary / Minutes ABWR -3 -
October 21, 1992 referred to as the Master DCD in the proposed standard design certification rulemaking. The Master DCD, without the proprietary information or secondary references, will be referenced in the rule that is published in the Federal Register. The legal details for withholding the proprietary portions are still under development by the staff. After the rulemaking hearings, the commission will direct the staff to make any required changes to the Master DCD for publication in the final standard design certification rule.
Mr. Wilson discussed the change process for Tier 1 and Tier 2 information. For Tier 1, the changes are limited to rulemaking, plant-specific order, or exemptions as given in various sections of 10 CFR 52.63. These changes must be necessary either to bring the certification into compliance with the Commission's regulations applicable at the time the certification was issued, or to assure adequate protection of the public health and safety. For Tier 2, a lower threshold for change is proposed in order to accommodate changes in technology, to incorporate lessons learned from construction and operating experience, and to accommodate necessary changes to a f acility or application for a facility. These changes are to be governed by the backfit standard of 10 CFR 50.109. Also, the staff proposes a change process for Tier 2 based on 10 CFR 50.59 for processing of changes for plants currently in operation.
Mr. Wilson pointed out that the staff proposal differed with Commission guidance in two aspects. One difference was in regard to the change standard for Tier 2 information. The Commission indicated that this standard should be based on adequate protection while the staff proposed that the standard be based on substantial increase in protection. The other difference was in regard to when the 10 CFR 50.59 process could be used. The Commission indicated )
that this process could only take effect after the issuance of a combined operating license (COL) while the staff proposes that the process can be used either before or after the COL.
l Mrs. R. Nease, NRR, stated that the DFSER has approximately 80 new open items in addition to the original 300 open items in the DSER.
The new open items are related to ITAAC, completed review, and updating the SSAR. There are approximately 100 open items that l were resolved from the DSER.
Mr. Michelson agreed to the staff's request of just answering
, questions, without specific presentation, regarding the open items associated with the subject chapters of the meeting.
Chapter 4 - Reactor Mr. Carroll stated that the ABWR design (as initially designed) did not include a loose parts monitoring system (LPMS). However, in response to the staff's position that an LPMS is required, GE has provided an LPMS general description. Mr. Carroll asked the staff 1
.a ,--ut A2L__ z.-. - - 4 e. >e = A Summary / Minutes ABWR -4 -
October 21, 1992 to provide discussion of LPMS requirements and the basis for such requirements at the November '92 subcommittee meeting.
Mr. Carroll requested from GE representatives to clarify operation with less th 10 reactor internal pumps (RIP). In addition, the staff requested GE to provide existing flow test results for operation with fewer than 10 RIPS. The staf f has not received such information.
Mr. Carroll asked the staff and GE to provide information on the potential for electrical faults in RIP motors, and what clears faults and how long it takes to clear such faults for rrmps with M/G sets and without M/G sets.
Chapter 5 - Reactor Coolant System and Connected Systems Mr. Poslusny, NRR, stated that the staff reviewed the measures used to provide and maintain the integrity of the reactor coolant pressure boundary (RCPB) and other pressure-retaining components and their supports that are important to safety for the plant design lifetime.
According to 10 CFR 50.55a components important to safety are subject to the following: i e RCPB components must meet the requirements for Class 1 l (Quality Group A) components in American Society of Mechanical Engineer ( ASME) Code,Section III, except for those components that meet the exclusion requirements of 10 CFR 50.55a(c) (2) .
e Components classified as Quality Groups B and C must meet the requirements for Class 2 and 3 components, respectively, in ASME Code,Section III.
l Mr. Michelson asked the staff to clarify the discussion in the .
DFSER section 5.2, regarding the overpressure prot;ection and also !
to add discussion of RCIC.
Mr. Michelson expressed concern regarding the statement made by the staff regarding the intersystem leakage to be " highly unlikely" because it would have to occur through closed check valves and/or closed containment isolation valves.
OPEN ITEMS Per the staff's request and agreement with the subcommittee members, no presentation was requested regarding Chapter 10, " Steam and Power Conversion System," Chapter 11, " Radioactive Waste Management," Chapter 12, " Radiation Protection," Chapter 13,
" Conduct of Operations," Chapter 15, " Transient and Accident I
4' .4 ABWR/GE Subcte. -3 -
11/19-20/92 '
Minutes e No expansion joints or bellows assemblies 'will be utilized in piping within the control building SSAR Revision e A statement tJ the commitments cited above~will be included in revised SSAR Section 9'2.
- 2. Provide additional information relative to (a) the main steam tunnel potential flooding level and its ability to accommodate and contain the resultant Water; (b) water .
tightness of the structure under flooded- conditions including the pressurization effects; and (c) discuss the potential flooding effecte and impacts on the reactor building and the control building GE Response e Some confusion has existed relative to the current SSAR documentation in the areas of: main steam tunnel configuration, main steam (MS) and feedwater line (FWL) configuration, main steam tunnel blowvent panels, and main steam tunnel performance characteristics SSAR Revision a
e Plant building drawings will be corrected
- MSL and FWL drawings will be corrected-e Location of blowout panels will be shown e A new MS tunnel write-up may be added e A flooding analysis will be added
- 3. Provide additional information relative to the reactor
ABWR/GE Subcte. _4 -
11/19-20/92 P.inutes service water (RSW) system-isolation capability to-preclude excessive control building and/or pump house flooding for a wide variety of RSW pipe breaks,fsizes and !
location l GE Respongg A new safety evaluation was performed for:- '
- RSWS configurations, e RSWS break spectrum,
- RSWS isolation capabilities, .
e flooding inside control building-basement, e
flooding inside pump house building, t flooding outside both building in. pipe chase, and .
e alternative enhancements.
SSAR Revision e
Additional requirements will be added to SSAR-RSWS write-up ... air break considerations A change will be necessary relative to loss of two RSWS1 divisions -
e Flooding analysis writing will be added -
- 4. Confirm that the' RSW system will not utilize " lined. ,
piping." That is piping with an inside protective liner. '
1 GE Response e
ABWR Standard design vs. COL design requirements
ABWR/GE Subete. 6/15-17/93 Minutes
- 3. Chilled Water Systems GE representatives indicated that'the Chillers for the-next generation of chilled water systems will be computer-controlled or at least microprocessor-controlled. The chiller systems will have built-in protective features with an anticipation that it takes 10 minutes for restart from a sudden stop.
Currently the standard review plan (SRP) section for the Chillers has not been completed.
Mr. Carroll questioned how the NRC staff .would _be performing its reviews regarding this issue and if they are planning to-include an outside consultant. The NRC staff replied that they will brief the ACRS _later regarding this issue.
- 4. Coolina of Instrumentation Rooms (Inside Secondary Containment)
Cooling is performed by normal ventilation. Upon secondary containment isolation, the cooling is cut off.
- 5. Auto-Start of Chillers Mr. Michelson expressed concern regarding the scenario following loss of power (e.g., station blackout). The Chillor compressors will proceed to cool off and the oil in them will proceed to absorb the refrigerant over several hours. For the case of station blackout, can the compressors be restarted without outgassing the oil?
GE representatives responded by stating that this does not represent a problem especially with the requirement that combustion-turbine generator (CTG) to provide power to ev.ergency buses within 10 minutes.during blackout.
However, GE is committed to respond to the concern about refrigerant-oil interaction and degassing requirements.
- 6. Combustion-Turbine Generator (CTG)
The Subcommittee discussed the intended use of the CTG and raised questions regarding the various modes of operation and the use of CTG as a backup for emergency diesels and compliance with the station blackout- (SBO) rule. GE is planning to cover all the CTG requirements in Appendix 1C of the SSAR. The CTG is not seismically qualified. However, it is protected against tornado and hurricane winds, but not associated missiles.
l l
ABWR/GE Subcte. 6/15-17/93 l Minutes ,
The CTG has a 0.95 target reliability (i.e., a 5 percent probability of failure) to start and carry load.
Mr. Carroll expressed concern regarding the non-protected status of the CTG, especially if GE is planning on taking credit for the CTG to meet the SBO rule. GE representa- '
tives stated that the ABWR design will meet all the SBO requirements without the CTG. ,
- 7. Pressurization Analysis Dr. Catton stated that he has reviewed the calculations for the repressurization analysis and concluded that GE's calculations are " grossly conservative." For example, ,
the room walls are assumed to be pressurized on one side only, no heat transfer is allowed and no equilibrium is assumed. Dr. Catton commented that this is an example of over design based on an overly submissive model.
Mr. Michelson stated that the Committae is concerned as to the pressures that might be reached in the reactor .
water clean-up (RWCU) compartment in the event of a pipe l break in the compartment. The in!tial calculations resulted in pressures in the order of 5 psi. Previously, ,
the Committee has commented on such calculations as being unrealistic with extremely short time for isolation valve closure.
GE currently performed a new set of calculations in which they introduced a 45 second time delay and a maximum >
pressure of 15 psi and approximate temperature of 230 F. -
As a result, the design basis will be modified to allow l for the RWCU compartment to vent into secondary contain-ment. This would require the equipment within secondary containment to be qualified for 15 psi and the corre-I sponding saturation temperature (s) . GE is proposing such ,
approach.
! Mr. Michelson expressed concern that the equipment be able to function properly under such pressure and r temperature conditions. For example, the room coolers f for the RHR pumps would be heavily loaded. It is not i just the environmental qualification but rather the need :
for equipment to perform its functions in such environ-ment.
- 8. Containment Performance Dr. Catton described GE's basis for containment perfor-mance loading. GE is using the MAAP 3.0 and BW-SAR I
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Advanced Boiling 3 October 16, 1992 Water Reactors before-break (LBB) provisions are not being specified for the ABWR design, and that all reference to the use of the LBB criteria in the SSAR is being removed.
Mr. Ehlert, stated that compartment pressurization analysis has been performed for assumed high energy line breaks. For the steam tunnel, the governing line breaks are from the main steam and feedwater piping systems. Assuming a double-ended guillotinc break in main steamline results in maximum compartment pressure of 10.9-pounds per square inch. Mr. Michelson stated that the SSAR should include requirements for drainline systems to prevent pressurization or flooding by interconnected drain line systems to vital areas; Mr. Ehlert agreed.
Mr. Michelson indicated that according to the drawings in the SSAR, a blowout panel is located between the reactor building and the control building. Mr. Ehlert stated that drawings are in error.
The indicated blowout panel is actually a seismic restraint. Mr.
Ehlert stated that there are only two blowout panels; one from the reactor water cleanup and the other from the reactor core isolation coolant (RCIC). Both blowout panels empties into the steam tunnel.
Mr. Ehlert stated that what is shown in the drawing is a blowout panel for a Japanese nuclear power plant. For the U.S. plant, the ,
drawings will be corrected to show a seismic restraint, not a l blowout panel. In reply to Mr. Michelson's question whether the staff is aware of this change, Mr. Burton, NRR, said-that this '
matter is on the staff's list of items that needs-to be clarified with GE. Mr. Ehlert explained that they are revising these drawings for a Japanese plant to U.S. criteria. Mr. Michelson stated that the revised drawings should also show that the feedwater line and steamline should be on the same elevation, as the Japanese plant apparently runs the feedwater lines above the steamlines. Mr. Michelson stated that the quality control problem, e.g., drawings, writeups, etc., is becoming exasperating.
Mr. Ehlert stated that water from an assumed feedwater line break will accumulate in the steam tunnel. Since the control building
- , -- 4 , -- -,
Advanced Boiling 4 October 16, 1992 -l Water Reactors 4
area is the highest portion of the steam tunnel, water will flow to the turbine or reactor building areas of the steam tunnel. The i resulting water levels in either the reactor or turbine buildings will accommodate the assumed feedwater line break water volume. In reply to a question by Mr. Michelson, Mr. Ehlert stated that ;
waterproof requirements for the water pits and description of the drain system for the steam tunnel will be specified in the SSAR.
Further, Mr. Ehlert stated that these service water lines will not be lined.
In reply to a question by Mr. Michelson, Mr. Ehlert stated that provision for a redundant isolation capability outside of the control building on the RSW supply and ~ return lines will be included in the SSAR. Also, a description / requirements of the pump house design and buried yard piping and the penetration of the reactor service water pipe through the control building wall will be included in the SSAR.
idecuacy of Divisional Separation - Mr. G. Ehlert. GE Mr. Ehlert discussed the following with respect to the adequacy of ,
divisional separation:
e Combined ef fects of fire, smoke and pressure from a fire.
- Flooding from the ef fects of fire suppression activities.
,
- Effects of pipe breaks.
e Secondary containment common HVAC System.
In reply to questions by the subcommittee, Mr. Ehlert said that the requirements for the following items will be included in the SSAR:
ABWR/GE Subete. 11/19-20/92 .
Minutes
- Outside of GE-ABWR standard design
- No current specif.c inclusion or exclusion ,
Current RSWS drawing indicates liner Required liner protection indirectly cited e Operating experience recommends internal pipe f protection SSAR Revision e Correct RSWS Figure e Additional Discussion in COL Section 9.2.15.2.1 e Additional COL Requirements e Accommodate Liner Degradation with Safety Evaluation
- 5. Provide additional information relative to the design basis of the reactor service water system pump house and the buried yard piping network between the pump house and the control building GE Response e ABWR Standard design vs. COL design requirements Outside ABWR standard design -
- GE scope is within the control building All portions of RSWS outside control building are l COL-site specific aspects
- Pump house building and yard-buried. piping chase f structure are not covered in SSAR 2
.6 Y
r ASWR/GE Subcte. 11/19-20/92 ;
Minutes !
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SSAR Revisign '
S
- A new brief but comprehensive SSAR section will be developed and issued
- 6. Provide the design basis and safety evaluation requirements relative to the RSW system piping penetration 1 into the control building i
GE Response e Current Standard-COL Interface Requirements-in SSAR Current RSWS standard requires that . inside and outside equipment be designed to seismic, ASME, >
quality and safety standards Current control building standard is designed against external flood damage by wall thickness, water stops at construction joints, water tight' door and piping penetrations, etc.-
SSAR Revision e A paragraph relative the subject piping penetration '
will be added to SSAR Section 9.2.15
- 7. Provide additional information relative to plant equipment rooms structural / barrier walls capabilities with specific emphasis on the reactor building, the secondary containment, the control building and the divisional separation compartments under pipe break, flooding and fire conditions. The information should address delta P integrity retention capabilities, compartment
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. .- . l ABWR/GE Subete. 6/15-17/93 Minutes ,
codes. The ACRS did not review the MAAP code. Some aspects of concern are, for example, the concrete attack basis. Dr. Catton indicated that the assumptions need to ,
be examined for such items as radial attack used to evaluate the pedestal attack, the lateral relative heat transfer, and the suppression pool carry-over. Another area of uncertainty in the calculations is the amount of radionuclides coming out of solution as'a result of 10 percent core vaporization. ;
A Subcommittee meeting will be scheduled in September 1993, to discuss the above matters.
- 9. PRA and Severe Accident Issues Dr. Kress and Mr. Davis summarized the issues that were discussed with GE representatives.
Mr. Davis stated that the PRA issues will be discussed at the September 1993, Subcommittee meeting to be held at UCLA (now rescheduled for Portland, Oregon, and probably :
it will become a meeting of 3 days duration). !
Dr. Kress described the impact of the fine-motion control rod drive platform grating underneath the core in case of a core melt accident. GE designed this grating to be !
made of steel and it is open. It is unlikely to create a hold-up and then suddenly drop molten core into the ;
water. l l
GE has performed the analysis of core / concrete interac- l tions and the coolability of it. GE claims that their j analysis shows a substantial floor area (greater than 0.02 m2 /Mwt). GE is also using a special concrete to reduce the amount of gases generated. GE claims that in most cases, the cavity will be dry and there will be top flooding after the core melts, either by a passive i flutter or by the fire water system to mitigate the ;
core / concrete interactions. I The containment shell is protected by the pedestal. The pedestal analysis indicates that it will not erode away.
For long-term mitigation of the core / concrete gases, GE could use the containment overpressure system for relief.
Another issue is the hydrogen explosion--GE has contain-ment vent design features that take care of this issue, e.g., two relief valves in series with nitrogen in between.
l
ABWR/GE Subete. 6/15-17/93 Minutes The Subcommittee also discussed the issue of aerosol plugging of a stuck-open vacuum relief valve. GE is assuming a nonporous plug. GE performed a sensitivity J analysis of the consequences of the radiological release I that showed assuming a nonporous plug-is of no concern. l Dr. Kress commented that in past practices, credit was i not given for aerosol plugging due to uncertainty. GE '
claims that this is not a design basis issue.
In response to Mr. Carroll's question of how the Commit- I tee feels regarding the " flood after philosophy?" Dr.
Kress stated that this is the best way to finesse the question of fuel / coolant interactions and steam explosion '
problems. However, this is not the best way to cool the debris.
- 10. Suppression Pool DH Control
, Dr. Kress stated that research performed at Oak Ridge indicates that a substantial increase in airborne iodine can occur if the suppression pool becomes acidic. The radiolytic formation of nitric acid has been identified as a mechanism for the pH of the suppression pool to decrease.
GE has performed a calculation of pool pH. The results ,
indicate that the pool will not become acidic during the i first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the transient. GE representatives stated that a section will be added to the USAR to address this issue. ,
- 11. Suppression Pool-Swell ModeliDS i Dr. Catton stated that he has looked at the calculation and concluded that GE's modeling is too simplistic and arbitrary and it should be more concise. Dr. Catton expressed concern regarding " fallback" of pool swell on vacuum breakers located inside the wet well. .
- 12. Plant Physical Senaration Barriers. l t
Mr. Michelson summarized GE's new arrangement for confining the energetic releases from reactor water clean-up system pipe breaks into the secondary'contain-ment.. Mr. Michelson indicated that with the present l arrangement, the effluent from the break would enter all I
- three divisions. No special provisions have been made in i the common normal ventilation system to prevent it.
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1 Advanced Boiling 5 October 16, 1992 Water Reactors e Dif ferential pressure retention capability for secondary containment and divisional separation on walls. [
t
. Barrier penetrations i.e., door, pipes, cable trap, and HVAC ducts. ,
I e Double wide barrier doors.
s e Floor leak-tightness provisions, o HVAC isolation valves and back-flow dampers, f i
e Blowout panels. l
- Floor drain system design and isolation arrangements.
- No safety grade remote multiplexing units (RMUs) inside I the secondary containment.
- Maximum rate of rcom heat-up for each equipment area following total loss of room HVAC. -
- The determination of environmental conditions following various accidents.
In addition to adding the above requirements in the SSAR, GE promises to provide the RWCU piping drawings to the ACRS. -
Protection of Environmentally Sensitive Eculoment - Mr. B. Simon. >
Mr. Simon stated that the issues are that 1) solid-slate electronic components are susceptible to adverse environmental changes, ,
particularly temperature extremes and 2) response of solid-state components to fire, flood, or loss of HVAC may be unpredictable, j thus, causing adverse system response. In response to a question l by Mr. Michelson, Mr. Simon stated that the design maximum ambient
.&, y
- 9. ,- -
Advanced Boiling 6 October 16, 1992 Water Reactors room temperature is 104*F and that the equipment is qualified to a maximum ambient temperature of 122 *F. (There seems to be confusion in the staff's position--component temperature of 122*F or the '
cabinets that house the components to be at an ambient temperature of 122*F--the staff person cognizant in this area was not present ;
and needs to be consulted.) Mr. Michelson stated that all Class IE solid-state components procured to MIL-STD-883C should function 4 satisfactorily to 230*F--these components are available, but expensive.
Mr. Sinon stated that GE's resolution of this issue is defense-in-depth built into the design. This means robust component design--
components meet MIL-STD 883C (125* test; 110 operating) ; separation and distribution of components--four redundant I&C divisions that i are separated and independent, and all data transmirsion via fire-resistant fiber optic cables; and data error detection circuits.
In reply to a question by Mr. Michelson, Mr. Simon stated there are no essential safety-related RMUs in the secondary containment. In reply to a question by Mr. Wylie, Mr. Simon stated that the Japanese are developing solid-state electronic equipment and system for their plants--in fact, there is some operating solid-state electronic equipment now available for industry use. Further, the ,
NRC is not planning on building and testing a prototype system--but '
the Japanese are performing environmental qualification testing of -
an integrated system.
i Mr. Simon stated that the solid-state electronic. components will ,
also be tested in an electromagnetic environment. The electromagnetic capability (EMC) commitments in ITAAC require that the design, installation, and test be in accordance . with the applicable electromagnetic interference (EMI), radiofrequency interference (RFI), electrostatic discharge (ESD), and electrical I surge (surge withstand capability (SWC) codes and standards. In reply to a question by Mr. Michelson regarding the equipment qualification list for ITAAC, Mr. Lyons, NRR, stated that it is required by 10 CFR 50.49, " Equipment qualification of electrical equipment important to safety for nuclear power plant," and will be
i Summary / Minutes ABWR October 21, 1992 Analysis " and Chapter 17, " Quality Assurance." However, review of the resolution of selective open items regarding these chapters was discussed.
1 Mr. Poslusny stated that GE has submitted ITAAC for the RCIC l system. The staff will review the GE submittal and provide an R evaluation in the FSER.
i FOLLOW-UP ITEMS As a result of the subcomraittee discussion, the following items were discussed, with the staff and GE representatives agreeing to provide the necessary information at a later ABWR subcommittee meeting:
- 1. Loose Parts Monitorina Clear up historical discussion in DFSER Section 4.4, page 4-9.
(staff)
Clarify operation with <10 reactor internal pumps operable, ,
page 4-9. (GE)
I Revise FSER writeup. (staff)
Provide discussion of LPMS requirements at November meeting. '
(staff) r
- 2. Electrical Faults in RIP's -
Provide information on the !
potential for faulting in RIP motors, what clears faults, and how long it takes irr pumps with M/G sets and w/o MG sets.
(staff, GE)
- 3. Add reference to Chapter 14 in Chapter 5 discussion of Code
~
editions, page 5-3. (staff) ,
- 4. Over pressure protection -
Clarify operation of system in discussion on page 5-6 DFSER Section 5.2, and add discussion of RCIC. (staff)
- 5. RCP Boundary Leakage . Detection, DFSER page 5-19 -
Delete sentence on intersystem leakage. (staff)
- 6. RCIC - Clarification of COL Action Item 5.4.6 RCIC valve closure verification ITAAC vs COL Action Item. (staff)
Summary / Minutes ABWR October 21, 1992
- 8. Revise DFSER as requested to address SSAR Sections 5.4.3, 5.4.4, 5.4.5, 5.4.9, 5.4.12, 5.4.13, 5.4.14 (staff)
- 11. Discussion of uninterruptable power source for protected lighting area. (GE)
- 12. Decrease in core coolant temperature, DFSER Section 15.1, page 15-4.
Remove redundant sentence. (staff)
- 13. Clarification of credit for nonsafety equipment, DFSER Section +
15.1 page 15-7 -
Address basis for conclusion and include extent of redundancy in power and water supplies. Include table in the SSAR. (staff)
- 14. Trip of all RIPS -
Describe basis for and better define "special case" on page 15-9 of the DFSER. (staff)
- 15. Verify that CRAC weather input error identified in Chapter 19 ~
was not used in Chapter 154 analysis. (staff)
- 16. DFSER Section 15.4.3, MSL -
On page 15-12 of the DFSER, include cross reference to Chapter 3 discussion of environment effects. Provide discussion about steamline break being limiting accident for offsite doses. (staff, GE)
- 17. DFSER Section 15.4.4.1, Containment Leakage Contribution -
Discuss why sizing of SATS is adequate to address potential ex-leakage with extremely low outside pressure. Add discussion to page 15-15 of the DFSER. (staff)
- 18. Discuss contribution to offsite release from gland seal system leakage via steam packing exhauster. (GE, staff)
- 19. DFSER Section 15.4.4.3 -
Discuss post LOCA radiological analysis to address the assumption of 25 gpm identified leakage versus 1 gpm. (staff)
- 20. Discuss why the design does not include continuous drywell to wetvell leakage monitoring system. (GE)
- 21. Discuss details of heavy object drop accident and control rod damage in accident scenario. (staff)
- 22. Reconsider writeup in DFSER Section 10.2.2, page 10-4, to better describe staf f position regarding SRP. Revise wriceup for COL Action Item 10.2.2-1 on turbine orientation. (staff)
s s
ABWR/GE Subete. -7 -
11/19-20/92- .
Minutes pressurization flooding and fire analysis and compartment and quadrant venting / relief techniques.
GE Response e The secondary containment divisional separation zone ,
loads are confined to two high energy pipe breaks.
Conservative pressurization analysis and technique are used. In addition, reliable venting equipment is used.
The control building barriers are leak-tightness ,
oriented.
SSAR Revision l e A brief paragraph relative to the structural barrier dC ca P design basis requirements will be added to appropriate SSAR ,
- 8. Provide additional information relative to the use of stacked block wall within the secondary containment i structures and compartments. Provide the design basis and ;
safety evaluation of these temporary structures including affects resulting from pipe breaks within these compartnents. -
GE Response o Most structural barriers-compartments and equipment room are permanent r
e Some semi-permanent structures are used: Removable l l
shielding walls, equipment hatches, blowout panels j l
1
e ABWR/GE Subcte. -8 -
11/19-20/92 Minutes e Temporary barriers may consist of a variety of designs -bricks and blocks e subject temporary barriers will be designed to stay in place under pressurization events SSAR Revision e A paragraph reflecting the above will be added to the appropriate SSAR section (e.g., compartment pressurization (6.2.3) or missile potentials (3.5.1).
- 9. Provide additional specific design basis and evaluation information relative to the reactor building, the secondary containment (SC) , the control building (CB) , and the divisional separation barrier penetrations. . Special-emphasis should be directed at the wide variety of barrier penetration types (including piping, cable trays, and FNAC ducts) used in the ABWR standard design under pipe' break; pressurization, flooding and fire conditions. Information should cite integrity assurances and pressure and temperature accommodations.
GE Response <
e ABWR Standard design is unique relative to past plant designs.
. Secondary containment - divicional separation quadrant barrier penetration are of most important and of greatest interest.
. - - = _ ~ - . . - - =. . -
4 4
ABWR/GE Subcte. 6/15-17/93 !
Minutes GE representatives were uncertain as to how to treat the ,
pressure loading of 15 psi on secondary containment and t divisional walls and will include additional drawings in .
the SSAR to define the pressurized boundary.
- 13. Reactor Water Clean-Up System (RWCU) ;
Mr. Stella summarized the issues that were satisfactorily i addressed. These issues were raised originally in the -
ACRS/RWCU report issued in July 1992. The issues deal i
with the process design of the system, the configuration '
of the system, and the use of the system as a high-pressure decay heat removal system in the PRA. Other !
issues that still remain open are: {
e Type and extent of design information required to be included in SSAR for RWCU system containment isola-tion valves, e overpressure protection for isolatable system compo- ,
nents, e Final configuration selected for instantaneous reverse flow isolation for RWCU system return to NDS/ Main feedwater headers, e Identification of design transients applicable to ;
ASME class III-3 components (Loading combinations ;
I and stress limits) - SRP 3.9.3, e Risk significance of RWCU system pipe breaks outside primary containment, and ;
- 14. Ultimate Heat Sink (UHS) Arrancement Mr. Michelson summarized his understanding of the UHS arrangement. He stated that the concern is control l building flooding. GE representatives stated that the i spray header return to the UHS will be divisionalized so l there will be no potential for flow from other divisions l back through the common spray header. In addition, all l safety-related piping will be in chases and not buried below grade.
. - . - . - ~ ..
ABWR/GE Subcte. 6/15-17/93 Minutes
- 15. Computer Aided Desian (CAD) and Conficuration Manacement 1 The Subcommittee was provided with a brief presentation and demonstration of certain key elements of the GE/ CAD / database and plant modeling systems. GE and its partners have used the CAD extensively for the ABWR design. These systems are being integrated into a single data network with advanced computerized design and configuration management capabilities. Currently, other advanced reactor designers are also-introducing similar CAD systems in their Projects.
The Subcommittee members were pleased and impressed by ,
the CAD system and GE's efforts to utilize the state-of-the art technology in their design of the ABWR.
II.
SUMMARY
OF THE ITEMS FROM THE SUBCOMMITTEE MEETING (6/17/93)
- 1. Plant Desian Life and Acina Manaaement - Mr. Genetti (GE) .
The ABWR design accommodates component refurbishment and replacement for early in life failures and obsolescence. ,
Breakthrough type technical developments are not needed ;
to achieve a 60-year life. .
The COL applicant is expected to study.the design to evaluate the longevity of structures, systems and components and develop a design . life understanding, management and classification system in conjunction with +
the procurement and maintenance programs. The classifi-cation system will categorize items (i.e., structures, systems, subsystems and components) according to design life capability for developing the strategy to be employed to support the design life requirement.
For those components which are expected to be replaced :
for obsolescence or early failure, the COL applicant will provide a plan for replacement, refurbishment, and repair >
activities, as appropriate, to assure the design life of ,
the overall plant. Such activities will be' scheduled to demonstrate that the plant availability requirements are satisfied. The plan shall include a comprehensive program for obtaining behavior data and record keeping for evaluating life capability of long life components based upon their operating history and measurement of their life limiting characteristics. A program- is-required for surveillance specimens, instrumentation, ;
material condition monitoring, environmental monitoring, etc.
i
Summary / Minutes ABWR- -7 - October 21, 1992
- 24. DFSER, page 10-9, Section 10. 4.1 - Discussion of circulating water system and power heat sink as an interface. SSAR and SAR revisions required. (GE, staff)
- 25. DFSER Sections 10.4.2 and 10.4.5, Main Condenser Evacuation System -
Provide cross reference to turbine building compartment exhaust system page 10-10, and defined medium efficiency filter. Clarify Statement about intercondenser.
(staff)
- 26. Turbine Gland Sealing System, DFSER Section 10.4.3 - Better define " fairly clean" steam on DFSER pages 10-11 and 10-13.
(staff)
- 27. DFSER section 10.4.4, Turbine Bypass -
Identify industry standards on page 10-13. (staff)
- 28. DFSER Section 10.4.5, CWS -
Discuss staff evaluation of GE calculators of complete rupture of single expansion of joint on CWS and provide additional discussion in DFSER. (staff) ,
- 29. SSAR Section 10.4.5 - Discuss flooding protection for at-grade ;
loads. (GE) ,
- 30. Discuss whether strainers in RW, CWS and condensate cleanup system are designed to full shut off head of pumps. (GE)
resolution the generic issue related to Reactor Vessel Overfill Protection. (GE)
- 32. DFSER Section 11.2.1, Li' quid Waste Management System - Reword passage on page 11-2 concerning demineralizers and resin cleaning. (staff)
- 33. DFSER Table 11.1, page 11-3, Discuss meaning of steam / water concentration, reactor vessel data. Are these partition valves? (staff)
- 34. Section 12.2.2 of the DCM - Clarify use of the word 'DAC' and include cross references to generic ITAAC. Include in FSER pertinent references to GE's specific ITAAC documents (GE, staff)
- 35. DFSER Chapter 17 - Section 17.1 needs to address role of COL applicant during design and construction. Need to reflect how a QA program comparable to GE's and which meets NRC 1 requirements will be implemented. (staff)
a 1 s Summary / Minutes ABWR -8 -
October 21, 1992
- 36. Add RG titles to DFSER Table 17.1 (staff)
- 37. Provide the data of the audit of GE's QA program in text in DFSER and provide details of review of GE's audit of Japanese QA programs. Add appropriate information in conclusion in DFSER 17.1 (staff) '
- 38. Discuss means to prevent interconnections between floor drains systems.
- 39. Discuss the need for vent and fill system design features.
(GE, staff)
- 40. Provide to ACRS a set of submittals concerning PRA which will support FSER development by staff. (GE)
FUTURE ACTIONS The ABWR subcommittee will meet on November 19 and 20, 1992 to '
continue its review and discussion of the DFSER. Pending the results of such discussion the Subcommittee Chairman may recommend certain course of action to the full Committee.
BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE FOR THIS MEETING
- 1. DFSER - dated October 5, 1992 (Predecisional)
- 2. Note from C. Michelson to Igne - " Items to include in future SSAR revision" - dated August 28, 1992 ;
NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006, (202) 634-3273, or can be purchased from Ann Riley and Associates, Ltd., 1612 K Street, NW, Suite 300, Washington, DC 20006, (202) 293-3950.
1
- . . = _ . - - . -. -- .- -. _. - . - .
Advanced Boiling 7 October 16, 1992 Water Reactors developed by the combined operating license (COL) applicant. In response to Mr. Michelson's concern on the criteria used to determine the list of electrical equipment that needs to be qualified, Mr. Fox stated that the SSAR will contain the requirements for qualification, not a specific list of equipment, but what will constitute the list. Mr. Michelson stated that his main concern is not so much a list of components that needs to be qualified, but the requirement of the qualification criteria of the components, i.e., water spray, temperature, electromagnetic compatibility, etc. Mr. Lyons, NRR, stated that water spray is covered under harsh environment found in Section 3.11 of the SSAR. ,
Mr. Michelson will revisit this section. Mr. Michelson asked if ,
the staff has reviewed and accepted the American Society of Heating, Refrigeration, and Air-Conditioning Engineers, Inc.
(ASHRAE) Standard 30 for the evaluation of refrigeration systems for safety-related equipment. Mr. Lyons stated that he will check on it. ,
Cavity Floor Area Beneath the Reactor Vessel - Ms. C. Buchholz, GE Ms. Buchholz discussed severe accident issues related to the lower drywell. She identified five major items:
e Flooding of the lower drywell.
- Passive flooder.
e Core-concrete interaction.
e Lower drywell sumps.
e Fuel-coolent interactions.
Ms. Buchholz identified three mechanisms for flooding the lower ,
drywell: !
e Low pressure injection systems which begins water injection when the vessel pressure drops.
l l
- AC Independent firewater addition system that requires !
manual actuation. l J
- -1
Advanced Boiling 8 October 16, 1992 Water Reactors e Lower drywell flooder wPich is a passive system actuated by the melting of a fusible link.
In response to subcommittee concerns on containment leakage at the lower and upper drywell seal due to high temperatures caused by corium at the lower cavity, Ms. Buchholz stated that results of the analysis indicate that the fusible link is expected to actuate the ten valves flooder system, about.10 minutes after the temperature in the drywell reaches 500*F, which should prevent the seals from degradation. In response to Dr. Catton's question on the MAAP
- code, Ms. Buchholz stated that the code is used to calculate the thermal-hydraulic characteristics, the fission product behavior and the core melt process. Dr. Catton stated that he is not familiar with the current version of the MAAP code and would like to hear '
from the staff on its review of the code and calculations using MAAP. Mr. D'Angelo, NRR, stated that the staff is currently reviewing the MAAP code and calculations. Further, the staff's present position is not to either accept or reject the MAAP code.
For severe accident calculations, the staff accepts the use of the MAAP code, if uncertainty analysis is performed. Dr. Catton questioned the validity of the use of the MAAP code for severe accident analysis. He stated that other severe accident investigators concluded that twice the lower cavity floor area is 2
needed to quench the molten core. Further, questions about prediction of steam explosions depend significantly on the initial conditions of a severe accident, which relies on the MAAP code.
With respect to molten core coolability issue, Mr. D'Angelo stated that the staff, in its FSER, has not accepted a steady state heat flux of 300 KWt per square meter, but will probably recommend a time-dependent heat flux. Dr. Catton stated that for core debris coolability, the Germans use a value of .04 square meter per megawatt thermal or twice the value that GE recommends (.02) . Mr.
Beckner, NRR, stated that with respect to the severe accident issues, the staff has asked GE to perform analyses to determine the consequences of the molten core not being quenched on the structural integrity of the reactor vessel pedestal and
Advanced Boiling 9 October 16, 1992 ;
Water Reactors containment. In addition to these analyses, the staff has asked GE to perform impact analyses of an assumed steam explosion. These analyses, according to Mr. Beckner, should be bounding analyses.
Preliminary results of GE's analyses indicate that the pedestal and containment integrity will be within the design basis. These analyses are being reviewed by the staff.
2 Dr. Catton stated that the technical basis for the .02m /MWt value :
f or core debris coolability and steam explosion should be addressed !
in the SSAR. '
Dr. Catton noted that SECY-90-016 requires that the core debris be quenchable. In or:*er to achieve this requirement based on today's knowledge, Dr. Catton believes that a much larger cavity floor area .
is needed than now provided by GE. He then asked the staff about what its plans are if the core debris is not to be shown :
quenchable. Mr. Beckn'er stated that he believes that SECY-90-016 does not require the core debris to be quenchable. Further, Mr.
Beckner believes that SECY-90-016 states that the core debris only '
be spreadable and coolable and that the impact of an unquenchable l core debris be determined.
i Ms. Buchholz stated that she believes that SECY-09-016 does ntt require that the core debris be quenchable. The staff stated that it will provide the Subcommittee with the latest staff position on '
the requirements for core debris coolability delineated in SECY 016. Further, Mr. Wilson stated that the FSER is-scheduled to be issued in early October 1992, which will have an evaluation of the core debris coolability matter. The staff will also discuss this matter with the Subcommittee during its next meeting, scheduled for October 21-22, 1992.
In reply to a question by Mr. Ward, Mr. D'Angelo stated that the staff is not planning to verify the MAAP code. The staff plans to confirm the data obtained by MAAP using the MELCOR code, although Mr. D'Angelo stated that, quite honestly, he does not think that MELCOR is any better. Dr. Catton agrees with Mr. D'Angelo's i
1
8 4 Advanced Boiling 10 October 16, 1992 Water Reactors observation. Further, Mr. D'Angelo stated that if the results obtained by MELCOR be about the same magnitude as those obtained by MAAP, we would be happy.
Ms. Buchholz discussed the lower drywell sump. She stated that the ,
floor sump is required in the lower dryvell for normal operation.
The flow f core debris into the sump may. limit coolability of the debris. She stated that a corium shield is designed to prevent the debris from flowing into the sump.
Dr. Catton requested that the staff provide him with a copy of the RSK document on the BETA core concrete interaction when it becomes publicly available.
LOCA Pool Swell Loads - Mr. N. Saxena. GE Mr. Saxena briefly' described the ABWR containment hydrodynamic loads. Mr. Saxena stated that the ABWR containment pool swell phenomena features are expected to be similar to those observed in
- earlier BWRs. The ABWR design includes design features from the MARK II (confirmed wet well air space) and the MARK III (horizontal vent system) containments. The MARK II containment pool swell phenomena were based on a single analytical model delineated in NEDE-21544-P/NUREG-0808, " MARK II Containment Program Food Evaluation and Acceptance Criteria," while the MARK III pool swell phenomena were evaluated based on test data.
Dr. Catton stated that the staff has this issue well in hand.
Issue 4-Review of the Chilled Water System - Mr. J. Lyons. NRR Mr. Lyons responded the ACRS concern that the staff's review of the Chilled Water System based on SRP Section 9.2.2, is not appropriate for the review of refrigeration systems. The staff should revise the SRP to account for the performance of the Chilled Water System under varying accident heat loads and during loss of offsite power events, including the ability of the system to restart and function
ABWR/GE Subcte. 11/19-20/92 Minutes ,
e Divisional separation quadrant, house the critical mitigation equipment (ECCS).
l t
- Divisional separation also provide protective housing for piping penetrations, instrumentation, control and power penetrations, HVAC penetrations and auxiliary service penetrations (e.g., cooling water)..
SSAR Revision e A set of paragraphs will be added to the appropriate SSAR sections relative to the above SC and CB specific penetration design.
- 10. Provide additional specific design basis and safety evaluation information relative to the reactor building, the secondary containment, the control building and the divisional separation barrier doors. Special emphasis should be directed at the wide variety of barrier door closures types (including single and double wide doors, motor-operated doors, sliding doors, equipment hatches / doors, large vehicle entry doors, elevator doors and stairwell doors) used in the ABWR standard design under pipe break conditions: pressurizations, flooding and fire. Information should cite delta P capabilities, air and water leak-tightness, fire resistance, and closure reliability.
gE Response
- ABWR Standard design -
unique compartmentization &
freely uses barrier door '
1 l
_ _ _ _ _ _ _ _ _ _ ____ _ _ _ _ J
ABWR/GE Subcte. 11/19-20/92 !
Minutes
~
e Doors provide ready yet controhled access to modular equipment cubicles / areas ,
o Ready access enhances plant surveillance, test, inspection and maintenance operations !
e Minimizes personnel and equipment dose; maximizes ALARA initiatives
- SSAR Plant layout drawings identify most doors e ABWR will heavily utilize past reactor door operating experience insights - reliability, operability, etc.
SSAR Revision e A table will be added to appropriate SSAR section reflecting the above. A new paragraph will be added to appropriate SSAR section on barrier door design basis. !
- 11. Provide additional specific design basis and evaluation information relative to the reactor building, the {
secondary containment, the control building and the -
divisional separation compartments floors and ceiling leaktightness. Special emphasis-should be directed at i
I floor and ceiling barrier integrity under flood water !
retention, fire containment, pressurization (delta P), and accident effects (pipe whip, jet impingement, etc.).
l Floor equipment and divisional drain system l
l interconnections should be considered. I l
l r
=
h' l ABWR/GE Subete. 6/15-17/93 Minutes The reactor vessel replacement is not anticipated during !
plant lifetime and need not be considered within the ,
availability requirements. The reactor vessel is designed for replaceability of internal parts; therefore, !
replaceability of the vessel itself is not considered necessary. However, the ABWR design has considered reactor vessel replacement in the development of plant layout and structural design such that, if required, reactor vessel replacement would be possible and would be accomplished without inadvertent obstacles.
Mr. Wylie asked if GE is planning to identify items to be !
designed for 60-years. GE representatives stated no.
Mr. Carroll commented that the NRC practice of issuing a 40-year license to a 60-year design life plant is inconsistent.
i
- 2. Station Groundina and Surae Protection - Mr. R. Strong (GE)
The electrical grounding system is comprised of:
e An instrument grounding network, >
e An equipment grounding network for grounding elec- [
trical equipment (e.g., switchgear, motors, distri-bution panels, cables, etc.) and selected mechanical componenis (e.g., fuel tanks, chemical tanks, etc.) ,
e A plant grounding grid, and e A lightning protection network for protection of structures, transformers and equipment located j outside buildings.
?
The plant instrumentation is grounded through a separate .
insulated radial grounding system comprised of buses and insulated cables. The instrumentation grounding systems are connected to the station grounding grid at only one point and are insulated from all other grounding cir-cuits. Separate instrumentation grounding systems are provided f or plant analog (i.e. , relays, solenoids, etc. )
and digital instrumentation systems.
Each building is equipped with grounding systems connect-ed to the station grounding grid. As a minimum, every other steel column of the building perimeter will connect directly to the grounding grid.
_.~ _ ,._ _ _ _. . _ _ . _ . . . _ ._ ._
l ABWR/GE Subcte. 6/15-17/93 Minutes i
The plant's main generator is grounded with a' neutral 1 grounding device. The impedance of that device will !
limit the maximum phase current under short-circuit l conditions to a value not greater than that'for a three- i phase fault at its terminals.
l The target value of ground resistance is 0.05 ohms or less for the reactor, turbine, control, service and ;
radwaste buildings. If-the target grounding resistance '
is not achieved by the ground grid, auxiliary ground grids, shallow buried ground rods, or deep buried ground rods will be used in combination as necessary to meet the ;
target ground resistance value.
The lightning protection system covers all major plant '
structures and is designed to prevent direct lightning strikes to the buildings, electric power equipment and instruments.
l Currently, there are no SRP or regulatory guidance for the grounding and lightning protection system. It is designed and required to be installed to such codes and-standards as IEEE Std. 80, 81, 665 and ' National- Fire Protection Association (NFPA-78).
It is the responsibility of the COL applicant to perform-ground resistance measurements to determine that the required value of 0.05 ohms or less has been met and to make additions to the system,--if necessary, to meet.the target resistance. This information will be documented in the SSAR.
- 3. Erosion / Corrosion Protection - Mr. G. Ehlert (GE)
The Category 1 structures are designed to have an epoxy based water proof coating on exterior walls to prevent water intrusion. There are water stops at construction joints below grade. There is a water proof lining under the basemat. Below grade yard piping and tanks such as the diesel fuel oil storage tanks, are in concrete chases or bunkers to avoid exposure to the soil or ground water.
GE specifies the use of waterproof coatings for the exterior of below grade structures to prevent water intrusion. Consideration is also given to the use of coated rebar.
Mr. Lindblad questioned if the use of coated rebar would interfere with the steel-concrete bond. GE representa-tives indicated that it may result in an increase in the slip length by 20 to 30 percent.
.. . v ABWR/GE Subcte. 11/19-20/92 Minutes GE Resoonse i
e ABWR Standard is a unique design-horizontal..
compartmentization and vertical open communication e Primary containment (PC) is a leaktight structure with internal leaktight compartments '
e Secondary containment (SC) is a 'leaktight structure for inside PC DBA breaks !
e Secondary containment (SC) is not a leaktight structure for external DBA breaks e Reactor Building (RB) --in and of itself--is not a leaktight structure e RB-Clean rooms, major equipment rooms, support equipment rooms are not leaktight to environs e SC-Divisional separation quadrants ~are inter-leaktight relative to one another SSAR Regponse -
I e A brief paragraph citing selective ceiling and floor leaktightness requirements and considerations will be added to appropriate SSAR sections
- 12. Provide additional specific design basis and evaluation information relative to the reactor building, the secondary containment, the control building and the divisional separation compartments EVAC system. Special
- :m . . . e - _mn
ABWR/GE Subcte. 11/19-20/92 Minutes emphasis should be directed at HVAC isolation valves and' ,
backflow dampers under a variety of plant disturbance '
conditions (including DBAs, internal floods, internal -
fires, radiological releases, environmental pertubations,- !
etc.).
GE Response e
ABWR Standard design utilizes a unique network of HVAC subsystems for various plant conditions i e Reactor building / secondary containment -HVAC is different from past plant designs e RBVs provides no HVAC tc SC Under DBA conditions i
e -
RBVs is isolated and safety equipment get individual / local HVAC subsystem service e
Selective reactor building safety related rooms receive their normal and accident HVAC service from individual divisional HVAC subsystems e For breaks inside SC, RBVs is isolated and local HVAC subsystems operate as in DBAs e SC - is not required !
I SSAR Revision Additional information will be incorporated in the current plant ventilation system descriptions.
?
t ABWR/GE Subcte. 6/15-17/93 Minutes The containment is made of reinforced concrete. The ,
containment top cap and drywell cylinder are carbon steel. The exposed surfaces are coated with an epoxy paint. The wetwell floor is constructed of either stainless steel or carbon steel with stainless steel i cladding on the wetted surface. GE does have an in-service inspection (ISI) program for the upper portion of the pool liner to monitor corrosion and pitting. The
~
containment internal structures such as the . reactor shield wall and pedestal are steel structures filled with concrete. The exposed surfaces are coated with an epoxy paint.
GE used the K-7 RWCU layout and EPRI's CHEC-MATE program to assess the acceptability of carbon steel piping for the RWCU. Dr. Seale indicated that this issue has to be revisited.
GE requires that cathodic protection be provided. !
However, its design is plant unique as it must be tailored to the site conditions. It is the COL applicant responsibility to provide a cathodic protection system.
- 4. Combustion-Turbine Generator (CTG) - Mr. R. Strong (GE)
Mr. Strong presented a single-line diagram for the ABWR loading groups A, B, and C and the use of the CTG as a backup for the emergency diesel generators. This will be added to the SSAR as Appendix 1C. The normal design ,
function of the CTG will be to act as a standby, nonsafety-related power source for the plant investment protection (PIP) nonsafety related loads during loss of preferred power (LOPP) events. The CTG will be able to provide power within 10 minutes during SB0 events. The CTG will automatically start and will be ready to accept PIP loads within 2 minutes of the receipt of its start signal. The target reliability of the CTG will be greater than 0.95. The CTG is not seismically qualified and it is not protected against wind induced missiles.
Mr. Carroll expressed concern regarding the lack of tech.
specs. for CE/ System 80+ and GE/ABWR's CTG. The NRC staff responded by stating that the design has to be >
finalized first. Currently, the Tech. Spec. schedule has been removed from the ABWR schedule.
- 5. Reactor Internal Pumps (RIPS) - Mr. Kumar (GE) !
Mr. Kumar described the RIP design requirements and maintenance. The reactor recirculation system (RRS) j e
ABWR/GE Subete. 6/15-17/93 Minutes features an arrangement of 10 RIPS. Mr. Kumar presented '
a cross-section view of a RIP. Collectively the. RIPS provide forced circulation of the reactor coolant through the lower plenum of the reactor and up through the-lower grid, the reactor core, steam separators, and back down the downcomer annulus. The recirculation flow rate is variable over a range (termed the flow control range) from minimum flow established by certain pump performance characteristics. Regulation of reactor power output over an approximate power range (70 percent s reactor power ,
output $100 percent rated output) without need for moving '
control rods is made possible by varying the recircula-tion flow rate over the flow control range.
The motor casing has a closure assembly at its bottom most end, termed a motor cover. The motor cover, in addition to its reactor pressure-boundary closure ,
function, provides a foundation for the bearing assembly which holds the non-rotating bearing elements of the thrust bearings. The motor cover is sealed to the motor l casing with a Flexitallic-type gasket and an 0-ring. The ,
recire. motor (RM) region surrounded by the inner surface t of the motor casing and the inner surface of the motor ;
cover, is termed the motor cavity.
The principal element of the stretch tube section is a thin-walled Inconel tube configured as a hollow bolt fitting around the pump shaf t and within the pump nozzle.
It has an external lip (bolt head) at its upper'end and an external threaded section at its lower end. The stretch tube function is to achieve tight clamping of the i internal pump diffuser to the gasketed, internal-mount end of the RPV pump nozzle, at the extremes of thermal-transients and pump operating conditions. Clamping action is achieved by capturing, with the stretch tube upper lip, a mating lip on the diffuser, and a stretch -
tube nut threaded onto the stretch tube lower end where
, it projects into the upper region of the motor cavity. .
When the stretch tube is hydraulically pretensioned, the prescribed preload is exerted on the diffuser.
The RM internals, including the water present in the motor cavity, are cooled by a circulating water process I which cycles the water in the motor cavity out through the RMC subsystem to a recirculation motor heat exchange (RMHX) and through return piping connections back to the RM. There is one RMHX per RIP located near the RM and within the reactor support pedestal. While the RIP is operating, flow circulation is powered principally by the a
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Advanced Boiling 11 October 16, 1992 Water Reactors 1
after a prolonged station blackout. Mr. Lyons stated that the need for an SRP section for use by the staff in reviewing the Chilled Water System is being considered as part of the SRP update program, which is still ongoing. In the meantime, the ABWR chilled water system is being reviewed in accordance with the current staff guidance. Mr. Michelson stated that he is not satisfied with the staff's evaluation, and will await the staff's evaluation when the FSER is completed.
Issue 8-ABWR PRA - Mr. W. Beckner, NRR Mr. Beckner, summarized the ACRS concerns as 1) how to determine when the PRA is adequate, and 2) where is the guidance on PRA that the staff has promised. In reply to the latter concern, Mr. l Beckner stated that the staff has presented to the commission a l
paper on PRA guidance. Mr. Beckner stated that he will send a copy of the PRA guidance paper to the ACRS. This guidance will be transformed into a regulatory guide that is scheduled to be completed by the end of the year or early in 1993. l This regulatory guide is intended to provide information on the PRA form and purpose as well as role of the PRAs in severe accident analysis and decision-making process for both existing and future plants.
Further, Mr. Beckner stated that the ABWR SSAR should reflect the contents of the proposed regulatory guide.
Adeauacy of SSAR Treatment of the Reactor Watgr Cleanuo (RWCU1 System - Mr. C. Sawyer, GE Mr. Sawyer discussed the following issues raised in the ACRS April 13, 1992 ACRS letter,1) interpretation of the Japanese codes and standards, 2) capability of the isolation valves, l
- 3) description of leak detection system, 4) inclusion of the RWCU )
I line break as core damage initiator, and 5) credit for the RWCU for backup heat removal. In addition, Mr. Sawyer, briefly responded to the findings of the study documented in a report by M. Stella and S. Mays, ACRS Fellows, dated July 30, 1992. Mr. Sawyer stated that GE will respond in writing to the findings of the study.
Advanced Boiling 12 October 16, 1992 Water Reactors With respect to the interpretation of Japanese codes, Mr. Sawyer stated that each Piping & Instrumentation Diagram (P&ID) will now explicitly show the required safety description and seismic class.
The RWCU (outboard of the isolation valve) is SC-3, Quality Group' C and non-seismic. Further, the transition from SC-1 to SC-3 is not treated explicitly in the ANSI /ANS 52.1 standard; however, GE believes the criteria of Section 3.3.2.1 of the standard are met.
Note: NRC has not endorsed the ANSI /ANS standard. (A copy of the ABWR Design Certification, Use of Metric Units was presented to the Subcommittee by Mr. J. Fox, GE.)
Mr. Sawyer stated that GE is aware of recent tests on the RWCU i isolation valves showing that closure of the valve under LOCA conditions is 2 to 3 times that originally assumed during procurement. The results of these operability tests will be factored into the SSAR.
Mr. Quirk, stated that the P& ids are available in San Jose, California. The P& ids for the RWCU are also available and he stated that a layout drawing will be provided to the ACRS. The staff and the ACRS have been invited to look at the piping drawings at San Jose in the near future.
Plant Desion Life and Aoina Manacement - Mr. C. Sawyer, GE Mr. Sawyer stated that he agrees with the ACRS concern that the i SSAR Section 1.2.1.3 should be rewritten to include some l
commitments for planned replacement interval of equipment, e.g.,
l motors, not planned to last the full 60-year life. The rewrite will basically follow the EPRI Utility Requirements Document, j Station Groundina, Lichtnino Protection and Corrosion Control -
Mr. M. Ross, GE Mr. Ross stated that with respect to the station grounding and surge protection, test has been added to the SSAR Appendix 8A, which includes references to existing codes that are a basis for 1
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ABWR/GE Subete. - 13 -
11/19-20/92 Minutes
- 13. Provide additional information relative to the use of, the location of, the design of and the performance analysis of blowout panel cited throughout the SSAR in safety related structures / compartments.
GE Response e Considerable amount of confusion exist relative to SSAR blowout panels i e Blowout panels are being used in a number of safety-related structure / compartment pressure relief functions e Basic operating principles of pressure blowout panels is rather simple l e Current analysis models are f airly straight forward and accurate e A variety of physical blowout panel hardwares is available SSAR Revision e Plant layout drawings will be adjusted for design refinements e Pressurization tables and figure will adjust for design refinements e Add a new paragraph on main steam tunnel blowdown-panels aspects i
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4 s ABWR/GE Subete. Minutes 11/19-20/92 i
14.
Describe in general where remote multiplexing units (RMUs) are utilized and located throughout the plant and {
specifically determine whether safety essential RMUs are located within the secondary containment structure.
GE Resoonse e
Make extensive use of proven and reliable digital components i
e Recognizes that similar components in past have been environmentally sensitive Will continue to use conventional qualified hardwired electronics in harsh environmental areas Design will cautiously locate safety-related system RMUs in environmentally protected, or e Will use equipment for data transmission, and activating trip signals in sensitive areas e
Also can use RMUs with non-safety related systems (e.g., feedwater control) -
e Current SSAR multiplex system writeup identify critical signal and control equipment locations e
Safety-related RMUs are located outside the primary and secondary containment boundaries.
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ABWR/GE Subcte. 6/15-17/93 Minutes RM auxiliary impeller. The RMHXs are positioned verti-cally such that should the RM stop during reactor operation, natural circulation through the RMC Subsystem piping will occur at flow rates sufficient to limit the RM temperature to acceptable values. ;
The maintenance requirements for the RIPS are as follows:
e 2 RIPS are removed at each outage and fully inspect-ed. The remaining RIPS are routinely checked in ,
place.
e Motor removal achieved without draining the reactor f vessel. There is a primary seal on the pump shaft collar, and a secondary seal. activated by using pressurized water. Minimum design life is 5 years.
- GE requires special tools for handling, installa-tion, removal, and inspection of the impeller, diffuser, pump shaft, stretch tube, etc.
Mr. McGrady (GE) described the RIP power supply protec-tion devices. The major protection devices are vacuum a
circuit breakers (open on fault current detection),
motor / generator set protection trips, and an output over- i current trip protection sometimes referred to as the adjustable speed drive (ASD) trip. Postulated major RIP failures will lead directly to ASD trip.
The GE design allows for parallel operation and load for the RIP power supply. During normal plant operation, all 10 RIPS are operated at approximately the same speed.
- 6. Containment Performance - Mr. Saxena (GE)
Mr. Saxena described the calculations that he performed to evaluate the ABWR suppression pool behavior during air ,
carryover (LOCA pool swell phase) . Basically, Mr. Saxena '
indicated that the ABWR LOCA pool swell response is much milder than that for earlier BWRs. The ABWR pool swell calculations are based on a simplified one-dimensional-basic pool swell analytical model used earlier for Mark II plants. The basic model, without modification, has-been shown to over predict Mark III horizontal vent test data.
The ABWR pool swell response is calculated using~ the basic model modified to yield additional conservatism in predicted response:
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-ABWR/GE Subete. 6/15-17/93 ;
Minutes '
e Peak pool swell velocity calculated using input l assumptions in a conservative manner.
1:
o Peak pool swell height taken as calculated maximum.
swell height. t Mr. Saxena stated that based on the conservatism in the models and the calculated loads, the vacuum breaker assembly is protected from likely pool swell loads by adequate structural shielding. '
Dr. Catton expressed concern regarding validity-of the assumed conservatism and arbitrary assumptions. He requested additional clarifications from GE regarding the ,
model assumptions, data base (e.g., Moody's calcula- ,
tions), and interpretation of the results. j
- 7. Electrical Insulation - Mr. Oza (GE) !
Mr. Oza described the electrical insulation for the ABWR i design. He stated that polyvinyl chloride or neoprene ;
cables are not used. Cross-thermosetting polyethylene '
insulated wires and cables are used for transmission and distribution of energy. Cable trays are made of noncom- ,
bustible material. Multi-conductor cables are specified ,
to pass vertical flame tests as part of the fire protec-tion features of the ABWR design. Each power, control and instrumentation cable is specified to pass the vertical tray flame test. The power cables are stranded conductor with flame and radiation resistant covering.
They operate at 100 percent relative humidity with '
service life expectancy of 60 years. The Class lE cables are designed and qualified for a LOCA at the end of.'60-year lifespan.
The COL applicant is required to demonstrate testing ,
methodology to assure such attributes for 60-year lifespan.
The ABWR design requires heat and smoke detectors. The l smoke detectors are of the ionization and photoelectric type. The ABWR design also utilizes other detectors such as gas sensing fire, flame, infrared flame, and ultravio-let flame detector's. The ABWR designers believe that effective detection and smoke removal systems are vital.
l Dr. Catton expressed concern regarding the electrical l insulation material that have been used, which could result in more toxicity than GE's estimates. This matter )
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9 4 Advanced Boiling 13 October 16, 1992 Water Reactors i
station grounding and lightning protection designs. Mr. Wylie -
questioned the valve of 0.05 ohms as the target value of the 3 resistance to absolute earth ground for the grid _ as to low.
Mr. Wylie stated that this valu: is very low and that many siten will not be able to meet this value. Mr. Ross stated that this is :
not a requirement, but nice to have. Mr. Wylie stated that GE should address the 0.05 ohms value at the next Subcommittee meeting and to clarify if this value is a target value or a requirement.
Mr. Wylie stated that the corresion control issue relates to the structures e.g., containment and components and should be in a l separate subsection in the structural area rather than in the electrical section. An element for corrosion control is cathodic protection. This issue should be addressed at the next '
Subcommittee meeting.
SUBCOMMITTEE ACTION / FUTURE PLANS Mr. Michelson stated that if the final SER arrives on time--the next Subcommittee meeting will be held in September 1992, if not, the next meeting will be held in October 1992.
Mr. Michelson stated that most of GE's replies to the ACRS concerns were satisfactory. A checklist of issues or questions that need further considerations will be generated based on this meeting as a starting point for the next Subcommittee meeting.
No presentation of this matter to the full ACRS is planned at this time.
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Advanced Boiling 14 October 16, 1992 Water Reactors
SUMMARY
OF AGREEMENTS. ASSIGNMENTS. REOUEST. AND FOLLOW-UP MATTERS i During this meeting, GE committed to provide / discuss the following issues in the SSAR or during the next subcommittee meeting on this matter. (This list of items include those provided by C. ;
Michelson.)
- 1. All welded pipe (i.e., no expansion joints) for the RSW and RCW systems piping inside the control building r
- 2. Water-tight main steam line tunnel in reactor and control building up to maximum interval flood line
- 3. Provision for redundant isolation capability outside of the control building on the RSW supply and return lines
- 4. No lined pipe for RSW system
- 5. Pump house design and buried yard piping
- 6. Penetration of RSW pipe through control building wall '
- 7. Differential pressure retention capability for secondary containment and divisional separation walls i
- 8. Stacked block walls subject to differential pressure due to '
pipe breaks
- 9. Barrier penetrations (i.e., doors, pipes, cable trays, and HVAC ducts)
- 10. Double wide barrier doors i
- 11. Floor leak-tightness provisions
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l ABWR/GE Subcte. 6/15-17/93 Minutes will be discussed in more detail at an ACRS Subcommittee
- meeting concerning fire in July 1993. ;
Mr. James provided a snapshot summary of the Tier 1/ITAAC status. He stated that there has been two years of '
multi-party interactions of Tier 1 form and content.
GE had a major submittal in May 1992.
Currently, GE is' planning on revising and resubmitting its final set of ABWR Tier 1/ITAAC material by end of August 1993.
The scope of Tier 1 is reserved for top-level informa- ,
tion. The ITAAC are in Tier 1. The purpose of the ITAAC are to verify conformance of as-built facility to Tier 1 design. Tier 2 is the design described in the.SSAR.
Verification of non-Tier 1 is via the Part 50 processes.
The GE approach is to structure Tier 1 on a system-by-system basis covering most plant systems, with graded treatment of systems that reflect safety significance.
The first step is to prepare the Tier 1 design descrip-tion for each system, then prepare the ITAAC table'for each system. Other Tier i entries will be prepared as needed. The Tier 1 design description derives directly from the SSAR. ,
The elements included in Tier 1 are as follows:
- System entries - Design descriptions and ITAAC
- e Non-System material - DAC and program commitments e Interface requirements i e Site parameters Currently, there are 79 Tier 1 entry systems, 4 DAC, 2 program entries, 9 interface requirements, and 1 site parameters table.
. In conclusion, Mr. James stated that Part 52 can be made .
to work using the step-by-step approach (pilot examples) .
However, it requires a core team of knowledgeable participants and lots of time. The overall .GE/NRC/
industry agreement on Tier 1 scope, form and content is '!
now in place. )
Mr. Stella requested the " lessons learned document" that t Mr. James mentioned in his presentation regarding the !
process and development of ITAAC. Mr. Quirk (GE) .,
promised to send it to the ACRS.
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ABWR/GE Subcte. 6/15-17/93 Minutes l
Mr. Carroll indicated that he will schedule a meeting of l the Ad Hoc Subcommittee to review the ITAAC document as !
soon as the final version is submitted by GE.
In closing, Mr. Michelson thanked GE for their fine effort to support the meeting and noted that additional meetings would be scheduled to complete the review of this matter.
ACTIONS. AGREEMENTS. AND COMMITMENTS
- 1. GE acknowledged that 46 open issues would still remain follow-ing the officially committed July 31, 1993 SSAR completion date. There was no commitment to a certain date for final closure on these items. Neither GE nor the staff gave the ACRS a list of these open issues.
- 2. The Tier 1/ITAAC material will not be completed until late August 1993 (GE estimate) . The ITAAC material has been changed significantly from the previous version seen by the Subcommit-tee. Mr. Michelson and Mr. Carroll stated that they would -
prefer not to review the Tier 1/ITAAC material until the staff and GE had agreed entirely on them.
- 3. The ABWR Tech. Specs. Will not be completed until late October 1993, and it is not included in the ABWR schedule. The ACRS needs to decide if the Subcommittee will be required to review the Tech. Specs. or not.
- 4. GE was requested to have test data available to verify the containment performance results of the Mark II model used or convert the model from the simplified two-dimensions to three dimensions. GE will write an evaluation to justify their use of the Mark II model. '
S. The Subcommittee requested GE to clarify what walls and penetrations would fall under the new design basis, due to the changes in GE's position for secondary containment pressure loading (from 5 psi to 15 psi).
- 6. The Subcommittee members discussed the service of Mr. E. Burns as a consultant for future Subcommittee meetings on severe accident issues.
- 7. The Subcommittee members raised the issue of the " security aspects" of the ABWR design and if the ACRS would get involved in reviewing this issue. Mr. Michelson agreed to bring the :
issue before the full Committee in July 1993.
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ABWE/GE Subete.
15 - 11/19-20/92 Minutes SSAR Revision
. PRUs will not be identified inside PC or SC on plant layout drawings e Statement to the effect that no essential RMUs are located within the PC or SC boundaries e Plant layout drawings will reflect essential and ncn-essential PEU locations e A new write-up will be provided relative to essential and non-essential RMUs
- 15. Provide additional information relative to the plant floor drain systems with specific attention to the reactor building, the secondary containment, the control building, and the divisional separation areas under normal conditions (equipment and piping leakage) and under accident -
pipe break condition (compartment flooding, pressurization and fire)
GE Response e Not part of ABWR standard
& Interface requirem;nps have been provided e Reactor building sump collection is shown on plant layout drawings e Radwaste system transfer interconnections are noted on various P & ids
ABWR/GE Subcte. 11/19-20/92 '
Minutes e i Internal flooding analysis ignores successful system operation SSAR Revision e Some additional COL Interface Information. Will be included in revised Section 9.3.12
- 16. Provide additional specific information relative to post accident (DBA) environmental conditions within the reactor building, the secondary containment, the control building and within various divisional separation compartments with special emphasis on their effects' on safety-related equipment in the affected areas GE Response e Significant amount of general information is available throughout the SSAR
- Additionally rather comprehensive / specific discussions are also provided i
Design basis and protective measures for safe i shutdown capabilities (SSCs) under DBA's Worte environs for SSCs at their locations Break effects both inside/outside containment including system interactions l e Internal event evaluation -
breaks, floods, fire, adverse environs i
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Advanced Boiling 15 October 16, 1992 Water Reactors
- 13. Blow-out panels.
14.
No Safety grade RMUs will be located inside the secondary I containment, t
- 15. Floor drain system design and isolation arrangements
- 16. Determination-of environmental conditions following various .,
accidents 17.
Maximum rate of room heat-up for each equipment area following total loss-of-room HVAC
- 18. The valve of 0.05 ohms of the resistance to ground will be clarified and determine if it is a target valve or requirement.
- 19. Corrosion protection fer structures and componerts 20.
GE to provide the ACRS with RWCU piping drawings
- 21. Discuss the low pressure integral turbine design i
P NOTE: Additional meeting details can be obtained from' a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006, (202) 634-3273, or can be purchased from Ann Riley and Associates, Ltd., 1612 K Street, NW, Suite 300, Washington, DC 20006, (202) 293-3950 b
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ABWR/GE Subete. 11/19-20/92 Minutes SSAR Revision >
e RMUs will not be ident!fied inside PC or SC on plant layout drawings ,
e Statement to the effect that no essential RMUs are .
located within the PC or SC boundaries e Plant layout drawings will reflect essential and non-essential RMU locations e A new write-up will be provided relative to essential and non-essential RMUs
- 15. Provide additional information relative to the plant floor ,
drain systems with specific attention to the reactor building, the secondary containment, the control building, and the divisional separation areas under normal conditions (equipment and piping leakage) and under accident - pipe break condition (compartment flooding, pressurization and fire)
GE Response ,
. Not part of ABWR standard e Interface requirements have been provided e Reactor building sump collection is shown on plant layout drawings ,
o Radwaste system transfer interconnections are noted on various P & ids
i .
ABWR/GE Subete. 11/19-20/92 Minutes Internal flooding analysis ignores successful system operation SSAR Revision e
Some additional COL Interface Information will be included in revised Section 9.3.12
- 16. Provide additional specific information relative to post accident (DBA) environmental conditions within the reactor building, the secondary containment, the control building and within various divisional separation compartments with special emphasis on their effects on safety-related equipment in the affected areas GE Response o Significant amount of general information is available throughout the SSAR e Additionally rather comprehensive / specific discussions are also provided Design basis and protective measures for safe shutdown capabilities (SSCs) under DBA's Worse environs for SSCs at their locations Break effects both inside/outside containment including system interactions e Internal event evaluation -
breaks, floods, fire, adverse environs
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ABWR/GE Subcte. 6/15-17/93
,1 Minutes r
- 8. The Subcommittee members discussed the issue of SAMDA. Mr.
Michelson agreed to bring the issue before the full Committee in July 1993, for a final decision.
- 9. GE will issue a final written response to the issues raised in the ACRS (RWCU) report by early August 1993.
FUTURE ACRS ACTIONS Future Subcommittee meeting will be scheduled as follows:
- 1. July 27-28, 1993 (Auxiliary and Secondary Systems) Re: Fire PRA and Effects of Fire Protection System Actuation on-Safety-Related Equipment li. September 8, 1993 (ABWR Subcommittee) Re: Open Issues, and USIs/GSIs lii. September 22-24, 1993 (Severe Accident) Re: ABWR Severe Accident, MAAP code, and PRA considerations ,
iv. October 1993 (Ad Hoc Subcommittee) Re: DAC and ITAACs
{
- v. October 26-27, 1993 (ABWR S/C) Re: Review of preliminary FSER vi. November 16-17, 1993 (ABWR S/C) Re: Remaining FSER vii. January 25-26, 1994 (ABWR S/C) Re: Any remaining FSER items and ACRS report preparation viii. Unscheduled (Computers / Human Factors) ,
DOCUMENTS The review document for this Subcommittee meeting was the Advanced Boiling Water Reactor (ABWR) Standard Safety Analysis Report (SSAR) ~
up to Amendment 29 (June 1993). ,
1 NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room ,
i 2120 L Street, NW, Washington, DC 20006, (202) 634-3273, or can be purchased from Ann Riley and Associates, Ltd.,.1612 K Street, NW, Suite 300, Washington, DC 20006, (202) 293-3950. t t
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J ABWR/GE Subcte, - 17 -
11/19-20/92 Minutes e Key assumptions: Break terminations outside containment, divisional separation, and SSCs 'and requirements ,
e Safety related equipment performances under post-DBA SSAR Responsa .
e Some' additional updated information will be added to appropriate sections reflecting new insights and inquiry responses
- 17. Provide additional specific information relative to the individual equipment room heat up rates upon loss of room {'
HVAC for reactor building, secondary containment, control building and divisional separation compartments with i emphasis on equipment temperature effects, I
special l
temperature detection and monitoring, structural
{
implications, operator information and system interactions i i
GE RESPONSE i
e Basic design philosophy o Unique application -
RB/SC/ divisional separation; CB/MCR/ equip rooms / basement e
t Current available information sources - extensive SSAR sections l
e' Traditional concerns - DBA effects e Special I&C/ loss of HVAC concerns
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ABWR/GE Subete. -
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11/19-20/92 Minutes SSAR REVISION e No major changes to the current SSAR - sections are contemplated for the plant ventilation systems documentation (Section 9.4) nor for the plant safety analysis (Section 14.0) t
- 18. Provide a set of RWCUS piping / component layout drawings. '
4 9
GE RESPONSE e A preliminary set of detailed RWCUS piping and ;
component drawings have been forwarded .
SSAR Revision :
- Appropriate RWCUS SSAR sections will be amended to !
provide previous comment responses.
- The forwarded drawings will not be added as standard SSAR enclosure documents. They have been submitted as special information.
i NRC STAFF PRESENTATION Mr. C. Poslusny, NRR, responded to the subcommittee member's ;
concerns regarding the open items related to the following DFSER chapters: ;
Chapter 2 - Site Characteristics Chapter 3 - Design of Structures, Components, Equipment t and Systems 1
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e ABhP/GE Subcte. 11/19-20/92 Minutes >
I Chapter 6 - Engineered Safety Features Chapter 8 - Electric Power Systems Chapter 9 - Auxiliary Systems Chapter 14.2 - Initial Plant Test Programs Based on general discussions with the staff and GE representatives the following are concerns and requests from the subcommittee members:
l Dr. Catton
- 1. The calculational tools used-by GE and the staff (COMPARE and SCAM) for sub-compartment behavior yield volume averaged temperature and pressure. Design limits are given for temperature. If it is calculated that these limits are approached, they will certainly have been exceeded because thermal stratification is a real physical phenomena. How is this dif ference resolved? This concern applies to environmental qualification requirements as well as drywell and wetwell temperature limits.
- 2. Although on the surface the treatment of LOCA loads appears to be satisfactory, many of the important details are missing. For example, the water slug thickness is chosen to be the pool surface to top vent thickness. What is the basis for this? According to the SER,-
" adaptations" of an earlier code was made. What were the adaptations? What were the " hydrodynamic loading assumptions" and why doesn't the staff like them?
- 3. The vacuum breakers are now protected from pool swell and fall back. What does the design look like?
a 4, ABWR/GE Subete. 11/19-20/92 Minutes
- 4. GE did tiansient calculations using fl/d's for standard fittings. Were the values corrected for accelerating flows?
- 5. The general question of pool thermal mixing needs to be addressed. How much stratification is a reasonable amount? Doing low mass flow chugging, how much of the pool acts as a heat sink?
- 6. What kind of debris could result from a SLB or other high energy line break and where will it be carried?
Considertions should be given to the sheet steel insulation and the possibility that flow paths to the suppression pool may get blocked.
- 7. High energy line breaks can result in high velocity flow through doorways and other restricted flow areas. This goes well beyond " jet impingement" and " environmental qualification." Sheet metal ducting can be damaged.
Cabinets can be damaged by buffeting. To be properly treated, environmental qualification needs to include flow effects.
Mr. Michelson ,
- 1. Expressed concern regarding the misuse of COL action items. He stated that the staff refers to design to be completed in the future and it is not clear how the staf f can make a final safety determination with this type of action items left unaddressed.
- 2. Indicated that nothing in 10 CFR Part , 52, subpart B recognizes a COL. He added that the final ABWR safety i
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A 4 ABWR/GE Subcte. 11/19-20/92 Minutes determination should not be based on COL holder.
Mr. Wylie Commented that there are cases where COL action items are
" promises" of characteristics of a design that must be known to make a final safety determination, especially in Chapter 8, " Electrical Power Systems."
Mr. Michelson Stated that justification that interf ace requirements can be verified by inspection, testing, analysis, and acceptance criteria is required by 10 CFR Part 52.47.
However, such justification is not found in the staff DFSER.
Mr. Carroll Faxed his concerns (general and specific) regarding the staff DFSER. The general concerns are:
- During the 10/21/92 meeting, Mr. Carroll expressed concern that the staff was misusing the " COL Action Item" designation in the DFSER. He pointed out a number of places where it was being used by staff members as a hedge against an incomplete Tier i design description, ITAAC and/or Tier 2 material. In his view of Part 52, this is totally unacceptable. The staff must make a clean final safety determination solely on the basis of the Tier i design description, the associated ITAAC(s) and the Tier 2 material. (The only exception is the limited number of issue that are being
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j ABWR/GE Subcte. 11/19-20/92 Minutes i
i treated as DACs or as site-specific interface issues.)
With these exceptions, the " Confirmatory Action Item" designation should be used only for operational phase (post fuel load) issues. The acceptance criteria for satisfying these operational phase issues should be clearly spelled out in the FSER.
1 e There is a wide variation in format between chapters.
Mr. Carroll likes what was done in Chapters 8 and 9, where each section ends with a summary of the Design Certification Material.
. Mr. Carroll had trouble working up a lot of enthusiasm for reviewing this material given that the ACRS has to do it over again when the final FSER and the completed ITAAC will be received.
ACTIONS, AGREEMENTS, AND FOLLOW-UP ITEMS
- 1. The subcommittee members agreed that the staff was not ready to discuss three major-chapters: Chapter 7, on instrumentation and Control; Chapter 18, on Human Factors Engineering; and Chapter 19 on Severe Accidents and PRA.
These chapters are now undergoing rework by the staff and GE.
The staff has not indicated when they will be ready for ACRS review.
- 2. The ACRS stand prepared to complete its review whenever the FSER and the associated Design Descriptions, DACs, and the ITAACs are reasonably complete, up to date, and have been subjected to appropriate quality checks; and the staff and GE are ready to support final meetings. The quality checks are
, OP-'
ABWR/GE Subete. - 23 -
11/19-20/92 Minutes essential to the resolution of numerous technical conflicts between the SSAR, and FSER, and the various Design Descriptions, DACs, and ITAACs.
- 3. Certain other ACRS review work remains. The ACRS is proposing to schedule a meeting with the staff when they are ready to discuss Apendix 9A, which is the Fire Hazards Analysis, and Chapter 16 which contains the proposed Technical specifications.
- 4. Still remaining for selective review will be the final closure of over 300 open items in the draft FSER; consideration of certain Design Descriptions, DACs, and ITAACs as they reach final development; and a number of requests for additional information from previous ABWR subcommittee meetings.
- 5. Also pending is the ACRS review of the proposed technical resolution of the Unresolved Safety Issues (USIs) and the medium- and high-priority Generic Safety Issues (GSIs) which are technically relevant to the ABWR design. In addition, there are a number of miscellaneous SSAR sections which the ABWR subcommittee has not seen for the first time. ,
9 FUTURE ACTIONS The ABWR subcommittee chairman is planning to brief the full Committee at the December '9-12, 1992 ACRS meeting. Pending the results of such briefing, the subcommittee chairman may recommend a certain course of action. In addition, the subcommittee chairman I is planning to brief the Commission on December 11, 1992 regarding the status of the ACRS review of the ABWR design.
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ABWR/GE Subete. 11/19-20/92 Minutes BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE
- 1. DFSER - dated October 5, 1992.
Note from C. Michelson to E. Igne, " Items to Include in Future 3.
SSAR Revision," dated August 28, 1992
- 4. Items from October 21, 1992 Subcommittee meeting.
NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, NW, Washington, DC 20006, (202) 634-3273, or can be purchased from Ann Riley and Associates, Ltd., 1612 K Street, .NW, Suite 300, Washington, DC 20006, (202) 293-3950.
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I O ATTACHMENT II 1
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PROPOSED PRESENTATION MATERIALS (AREAS REQUIRING PREPARATION AND ATTENTION)
FOR ACRS SUBCOMMITTE MEETINGS WITH GE IN SAN JOSE JUNE 15-17,1993 0 SPECIFIC ACRS INFORMATION REQUESTS l AREAS OF PREVIOUS ACRS INTEREST / REQUESTS l
l John William Power 6/10/93 0-1
-, 1
.1 .
J UIIS AND RSWS-DESIGN AND EVALUATION ;
SUBJECT OVERVEW
- DESIGN AND EVALUATION STATUS NRC REVEW STATUS
- ACRS ISSUES STATUS O Describe Safety Evaluation (eg PRA) That Verifies Proposed Concept Under Various Postulated Accident or Event Conditions O Adequacy of Interface Requirements O Verify Control Building Can Accommodate Flooding With ...
- Degraded Isolation Operability
- Limited Feasibilityfrime Available For Needed Response
- Assumption That S ACF Occurs During Recovery l
G "4 iller l J. rower G. Elhen i
1-1 I
. V. l i
i UIIS AND RSWS- DESIGN AND EVALUATION i
Ultimate Heat Sink UHS Design and Evaluation Systems Involved System Flow Diagram, Major Components, Design Details
- System Supply Responsibility Decisions System Interface Requirements System Safety Analysis- Failure Analysis, Effect, Implications Turbine Building Flooding Recovery, Impact on Other Systems, Extemal and Intemal Sources Reactor Service Water Sysics
- RSWS Design System Design Basis a
System Arrangement - Physical, Electrical System Flow Diagram, Design Details RSWS Evaluation
+
RSWS Break Analysis- Feed, Discharge, Air / Water, Pump House / Tunnel /CB Isolation Valve Configurations - Double Valving Leak Detection and Location and Control Actions (Auto and Manual)
Operator Response Action Time Domain
- Break / Flooding Interactions Divisional Interactions RSWS Scopes of Supply
- Boundaries - CB, PCS, PHB Interface Requirements G. Miller J. Power G. Elbert 1-2
4 .. .
1 SOLID STATE ELECTRONIC COMPONENT - DESIGN AND EVALUATION (COOLING)
. SUBJECT OVERVIEW
. DESIGN AND EVALUATION STATUS
= NRC REVIEW STATUS ACRS ISSUES STATUS O What Are the Bases For Determining the Capability of SSECs To Function Acceptably at Maximum Specified Room Temperatures?
- Tests
- Analysis
- Other O Where Are SSECs Located?
O What Is Maximum Post Accident or Event Room Temperatures? (With SACF Considerations)
O What Will the SSECs Temperatures Be In a Generic Room Temperature?
O How Will the SSECs Behave if Maximum Temperatures Are Specified?
O What Room and SSECs Cabinet Cooling Will Be Needed? (Consider Need For Redundancy)
O What Will Be SSEC Response to Other Envimnmental Disruptions?
- Pipe Breaks
- Fire
- Flood O How Will the SSECs and Its Cooling Requirements Be Specified?
B. Simons B. Genetu U. Saxena 21
. t, SOLID STATE ELECTRONIC COMPONENT - DESIGN AND EVALUATION (COOLING)
Use of SSECs in Plant Systems - Design Bases a
Safety /Non-Safety Multiplex / Hardwired Software / Hardware Divisional Separation a
Redundancy / Diversity Single Line/ Box Diagram Locations / Environs Conditions / Control Use of SSECs In Plant Systems- Evaluation EQ - Testing, Surveillance, Capabilities FMEAs Break / Fire / Flood / Harsh Environ Effects Temp Rise Criteria a
Maximum Accident Aspects EMI Protection
+
Performance of SSECs Reliability Self Test Operating Experience B. Simons i B. Genetti i l U. Saxena !
2-2
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ESSENTIAL CIIILLED WATER SYSTEM - SAFETY EVALUATION SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS O Chiller Performance Under Varying Accident Heat Loads O System Behavior and Electrical Loading During LOOP /LOPP O Ability to Restart / Function After Prolonged SBO 1
G. Miller l J. Pow er l
?
l U. Sasena j 31
, t ESSENTIAL CHILLED WATER SYSTEM - SAFETY EVALUATION Chiller HVAC Systems - Design
- Chiller Applications-(MCR.DG Rooms)
- Chiller Hardware Aspects
- Chiller Performance Restarts / Sequencing Loading Aspects Component Characteristics a Standby Unit Status Switchover Plans Chiller HVAC System- Evaluation
- Component Switching
- Divisional Chiller Losses
- Breaks Outside Containment
- LOOP /LOPP Events
- Short SBOs
- Postulated Elongated SBOs
- Recovery Aspects
- Means to Circumvent Shonfalls
- Chiller Operating Experience
- Cunently Used Throughout Plants Several Years
- MCR and Equipment Rooms Applications
- Reliability Analyses Have Been Documented l
l t
G. Miller l
J. Power
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U. Saxena 3-2
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- STANDBY COMBUSTION GAS TURBINE GENERATOR- DESIGN AND EVALUATION
. SUBJECT STATUS DESIGN AND EVALUATION STATUS NRC REVIEW STATUS
- ACRS ISSUES STATUS O Discuss Design Criteria and Ratings O Intended Use O Physical Protection Against Design Basis Events l
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1 B. Strong l J. Power l 1
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l STANDBY COMBUSTION G AS TURBINE GENERATOR - DESIGN AND EVALUATION
- CTG (AAC) Design Design Basis System Equipment Description Ratings /Imads/ Capabilities Auto Operation / Controls /I&C Demonstration Testing / Surveillance / Reliability CTG (AAC) Evaluation 10CFR50.63
- RG 1.155 NUMARC 87-00 EPRI-URD DFSER CTG (AAC) Operation Use Standby PIP Source Standby LOOP /LOPP Source SBO AAC Sourre Extended Capability Risk Reduction Valve B. Strong J. Power 4-2
( 4
(
BREAKS OUTSIDE CONTAINMENT-COMPARTMENT PRESSURIZATION ANALYSES SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS O Calculations of Subcompanment Pressures For Breaks:
Main Steam and Feedwater Outside Containment RWCUS Companments O Desi n Pressures For Physical Sepamtion Boundary Components asseens I Separation Walls, Ceilings, Floors hl Separation Doors, Hatches
. W :^--I Separation Penetrations Duswaseenal Separation Drains C:...... . ..al Separation HVAC Ductwork 0;..~. ~.ml Separation Blowout Panels U. Saventa -
G. Elhert J. Power EQ Man -
5-1
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-BREAKS OUTSIDE CONTAINMENT-COMPARTMENT PRESSURIZATION ANALYSES I Pressurization Analysis- Methodology
. Basic Model :
Initial Assumptions, Conditions Break Consideration Venting Aspects Volumetric Considerations MS/FWS/RWCU/RCICS Pressurization Analysis- Results '
Peak Pressure, Temperature, RH Inter / Intra Divisional Effects Unisolated Effects Residual Effects Blowout Panet Performance Utilization of Pressurization Analyses '
Blowout Panels Turbine Building and Divisional Separation Effects - Dynamic Effects ESF Equipment - Envimnmental Qualifications ESF Equipment - Placement / Separation Radiological Consequences Isolation Valve Considerations I
- System Interactions Recovery Aspects i
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9 U. Sas ents ;
G. Elhen
- 1. Pos er EQ Man 52
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PRIMARY CONTAINMENT- PERFORMANCE j i
SUBJECT OVERVIEW I DESIGN AND EVALUATION STATUS l NRC REVIEW STATUS l
. ACRS ISSUES STATUS ,
O Discussion Suppression Pool Behavior and Carryover Following Depressurization ,
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'Ii U. Saxena C. Buchholz C. Sawyer l
6-1
. s PRIM ARY CONTAINMENT - PERFORMANCE Containment Pool Dynamic Behavior Under Accident Conditions Impact in Froth Vacuum Breaker Interaction With Water Slug / Fall Back Load Definitions Test Data Containment Intemal By-Pass Considerations Leak Paths Vacuum Breaker Functions / Failures Top Vent Unrecovery A/Vk By-Pass Testing Severe Accident Containment Aspects Containment Response to S A Events Coolables Confirm Overpressurization Protection Alternative Cooling Aspects Enhanced Capabilities U. Sasena C. Buchholz C. Saw y er 6-2
6 .
PLANT DESIGN AND PRA -SAFETY EV ALUATIONS f
- SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS NRC REVIEW STATUS
- ACRS CONCERNS F
J. Duncan C. Buchholz i D. Knecht i Careway 7-1
.. 4 :
i PLANT DIVISIONAL SEPARATION REACTOR BUILDING / SECONDARY CONTAINMENT / CONTROL BUILDING PIIILOSOPH Y - DESIGN B ASIS '- EVALUATION SUBJECT OVERVIEW
= DESIGN AND EVALUATION STATUS
. NRC REVIEW STATUS ;
- ACRS ISSUES STATUS :
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- 1. Quirk :
H. Tow nsend C. Sawy er . 3 G. Elbert D. Mnw ell J. Pow er 8 . - . - -_. . . - .. - _ , . .
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PLANT PIIYSICAL SEPARATION B ARRIERS - DESIGN AND EVALUATION REACTOR BUILi>ING - SECONDARY CONTAINMENT - MSL TUNNEL - CONTROL '
BUILDING SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS O Identify and Evaluate Events Which May Challenge Plant Physical Separation Barriers
- Pipe Ruptures
- Fires
- Floods O Identify and Evaluate Environmental Disruptions Associated With Such Events
- Pressure and Temperature Increases
- Flooding
- Spraying
- Smoke
- Pipe Whip
- Jet Impingement O Identify and Evaluate Specifications For Those Barrier Components
- Walls
- Doors
- Pipe and Electrical Penetrations
- HVAC isolation Dampers and Penetrations i
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A. McSherry J. Quirk J. Duncan H. Townsend C. Sawyer D. Maxwell j B. Genetti G. Elbert !
C. Oza J. Power U. Saxena M. Munson 9-1 !
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PLANT PHYSICAL SEPARATION BARRIERS -DESIGN AND EVALUATION REACTOR BUILDING - SECONDARY CONTAINMENT - MSL TUNNEL - CONTROL BUILDING .
Banier - Events '
Inside Breaks .
- Outside Isolated Breaks Special Effects: Fire, Flood, Harsh Environ Outside Unisolatable Breaks Super Special Events Severe Accident Events Barrier- Design Bases Structural Walls
- Doors Penetration
- Drains HVAC Barrier- Evaluation Leak Tightness - Full Integrity
- Limited Communication Full Communication A. McSherry J. Quirk J. Duncan H. Townsend C. Sawyer D. Maxwell B. Genetti G.Elhert C. Oza J. Power U.Saxena M. Munson 9-2
i PLANT HVAC SYSTEMS - DESIGN AND EVALUATION SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS t
=
NRC REVIEW STATUS ACRS ISSUES STATUS O Arrangement of Ducts and Air Handling Units O Duct Work Design O Isolation Dampener Requirements and Design O Air Handling Unit Specification O Provisions For Redundancy s
i M. M unen I
J. Pow er U. Sasena '
EQ M.sn ,
G. Ethen 10 1
Pl. ANT HVAC SYSTEMS- DESIGN AND EVALUATION 1
HVAC Systems - Design .
Spectrum of Subsystems - RB/SC, RB/EEE, RB/DG, CB/MCR, CB/ER Divisional Separations Divisional Equipment Qualifications Compartment Heat Up Aspects STGS - Design Basis ,
HVAC Systems - General Evaluation
HVAC Systems - Specific Evaluation Loss of All HVAC -
Breaks Inside Containment Breaks Outside Containment LOOP /SBO Events
+
Fire Events Recovery Aspects i
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M. Munson J. Power U. Saxena l EQ Man G. Elhert 10-2
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l REACTOR BUILDING / SECONDARY CONTAINMENT DRAIN SYSTEMS - DESIGN AND 1 EVALUATION l l
SUBJECT OVERVIEW ;
DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS ;
5 i
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G Ehlen ll I t'
. l REACTOR WATER CLEANUP SYSTEM - DESIGN AND EVALUATION SUBJECT OVERVIEW Current Design and Evaluation is Complete. Accurate and Adequate
+ DESIGN AND EVALUATION STATUS
+ NRC REVIEW STATUS ,
+ ACRS ISSUES STATUS O Questing Response to ACRS Fellows Report J. W Power C. Sawyer D. Knecht E. Nazareno U. Saxena 12-1
e REACTOR WATER CLEANUP SYSTEM - DESIGN AND EVALUATION GE Response Report Status Ten (10) Major Issue Areas GE Initial Responses - Thirty-Two (32) Areas Fifteen (15) Specific Areas Five (5) Additional Consideration Areas Response Findings and Conclusions RWCUS Break Outside Containment Deterministic Safety Evaluations Beyond DBA Aspects Unisolatable Events Severe Accident Considerations Effects On Other Divisional Zones and Equipment RWCUS - Isolation Valving - Design and Evaluation System Valving Connections Leak / Break Detection Isolation Valving Special Isolation Valve Requirements and Considerations Demonstrated Valve Closure Capabilities RWCUS - Perspective Operating Experience Feedback Record Divisional Zone and Component Compartmentalization Coolant Boundary Integrity, Maintenance and Inspectability Evolutionary Design Enhancements Risk Aspects l
I J. W. Pow er C. Saw y er D. Knecht E. Naweno U. Saxena 12-2
- 4 &
f COMPUTER - AIDED - DESIGN DEMONSTR ATION SUBJECT OVERVIEW CURRENT USE- APPLCIATIONS NRC INTEREST a
ACRS INTEREST AREAS O Control Building O RWCS Layouts and Compartments O RB/SC-Common HVAC System O Main Stream and Feedwater Compartment (Outside Containment) i l
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1 S. Hucik .
D. Tibbels -)
- 3. Stickler 13-1 :
COMPUTER - AIDED - DESIGN DEMONSTRATION i
+
RWCUS Drawings Provided Previously
+
2 3-D Views
+
3 Plan Views Structural Cubicles Added p
l S. Hucik D. Tibbels J. Stickler 13-2
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PLANT DESIGN LIFE AND AGING MANAGEMENT
. SUBJECT OVERVIEW .
DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS O Discuss Plant Design Life Internal and Program To Achieve It -
- Design and Application Criteria
- Refurnishment or Replacement Requirements
- Aging Management Measures l
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14-1
.... i 4 8 STATION GROUNDING AND SURGE PROTECTION SUBJECT OVERVIEW
+
DESIGN AND EVALUATION STATUS NRC REVIEW STATUS
=
ACRS ISSUES STATUS O Discuss Station Grounding and Surge Protection Measures
- Overall Plant Station Grounding and Connections To Switchyards
- Lightning Protection Measures j
- Grounding System and Protection of Sensitive Solid State Electronic Components
- Basis For Requirement of Resistance For Ground System (Absolute Eanh 0.05 ohms)
- Design Requirements For Main, Auxiliary, Reserve Transformers
+ Ratings Design Loads ]
Temperature Rating ,
1 Insulation Class j Basic Insulation Level l
+ Other Applicable Data, Specification l l
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l B. Stron g 15-1
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I PL ANT STRUCTURES - CORROSION CONTROL SUBJECT OVERVIEW '
DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS O Discuss Corrosion Control Measures To Achieve Design Life
- Considering Environmental Effect and including Protection For Reactor Building Liners Substructures
+
Underground Installations G. Elhert 16-1
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. REACTOR INTERNAL PUMPS (RIPS)- DESIGN AND EVALUATION l l
SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS .
NRC REVIEW STATUS ACRS ISSUES STATUS O Discuss RIP System
- Pump / Motor Design Requirements
- Electrical Distribution Arrangement and Protective Devices Parallel Operation and Loading a
Locked Rotor and Short Circuit Protection
- Speed Control System and Equipment - Design Requirements I
D. Robertshaw i
17-1
' STANDBY COMBUSTION G AS TURBINE GENERATOR - DESIGN AND EVALUATION SUBJECT STATUS
- DESIGN AND EVALUATION STATUS
- NRC REVIEW STATUS
- ACRS ISSUES STATUS O Discuss Design Criteria and Ratings O Intended Use O Physical Protection Against Design Basis Events B. Strong J. Pow er 18 1
& B
. . p 4
PRIMARY CONTAINMENT- PERFORMANCE SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS NRC REVIEW STATUS
- ACRS ISSUES STATUS O Discussion Suppression Pool Behavior and Canyover Following Depressurization 3
b U. Saxena C. Buchholz C. Sawyer 19-1
A ELECTRICAL INSULATION-DESIGN AND EVALUATION SUBJECT OVERVIEW DESIGN AND EVALUATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS O Discuss Briefly "Supertoxicity" Smoke Combustion Products From the Ignition of Electrical Insulation O What Considerations Are Being Taken To Deal With Products? .
O What Kind of Insulation Will Be Used?
O What Pmducts Are Produced Doing Short Circuits or Shorts?
O Discuss Smoke Detector Response to Such Smoke B. S tron g 20-1
PRA STATUS- ABWR DESIGN SUBJECT OVERVEW DESIGN AND EVALUATION STATUS
=
NRC REVEW STATUS ACRS ISSUES STATUS J. Duncan C. Buchholz 21-1
.r-.
ITAAC ASPECTS
. SUBJECT OVERVIEW DOCUMENTATION STATUS NRC REVIEW STATUS ACRS ISSUES STATUS -
T. James J. Quirk 22 1
STATUS REPORT - RESPONSE TO ACRS SUBCOMMITTEE MEETING - OPEN ITEMS
. ~ SUBJECT OVERVIEW ISSUES TRACKING STATUS ;
=
NRC REVIEW STATUS :
ACRS ISSUES STATUS O ' Discuss Status and Proposed Closure Method For Items Previously Identified '
- August 19,1992 Meeting
- October 21,1992 Meeting November 19 20,1992 Meetings I
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J. Quirk ,
J. Power l J. Fox 23-1
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- - - .- . . _~ . .. .
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& GENuclearEnergy Severe Accident Containment Performance Issues Presentation to the ACRS Subcommittees on ABWR andSevere AccidentAnalysis Carol E. Buchholz, Princips! Engineer ABWR Engineering June 15,19D Containment perfonnance issues during a severe accident
- Generalcantninmentperformancetrends
- hnpoct ofIntCRDplatform grating
- Impact of core concrete interaction ,
- Aerosolplugging of bypasspathwey
- Suppression poolpit control
- Poolflashing following rupture disk opening LTC6%*32 g e t ,.- ,+my - ,, y,y
. e e e e
Containment performance methodology
- htodified version of htAAP code used as de primary nasysis tool
- Code changes reflect difference between ABM design andprevious BM designs
- A few errors in the code were corrected On& one change made to the physicalmodels to better evatuste suppression pool vapor space response
- Separate ans&ses developedto address specificphenomenn such as airect ContainmentHosting
- Some sensitivity ans&ses were performed using furner modifications of MAAP to investigere the impset of some phenomenological uncertainties, for example, Core Concrete Intsraction ca. , ,
Sample severe accident sequence - pressure
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FMCR0 platform grating What is the impact of the FhtCR0 handling machine refueling platform in the lower drywell on fuel coolant interactions and core debris coolabilityt
- De rotating pistform is circular and mounted on the rotating rail under ne reactor vessel Dere willbe en opening aren at the center and will be provided win a traveling railfor he CR0 handling devise. Gratings willbe installed on bon sides of the rails.
- ne grating consists of 1*by Mt"metalslats mounted edge-wise to form a grid with a grid size on no order of I*by 2"
- Early fragmentation of debris willtend to increase the voiding of any pre-existing water pool. This wi!Itend to reduce the loading from an FCf decreasing ne potentialfor early containment failure from this mechanism e no grating willquickly ablate due to ne flow of debris in a manner similarto the ablation of tha vesselbottom head, nerefore,6are willbe virtually so effect of &n grating after de initialdebris pour.
- ne late debris pour auld be a slow, drip-like relocation which would fallstraight through de ablatedregion of the pistforn cre ems ,
Core-concreteinteraction
- Can lead to Inte fission product rolesse or failure of the containment Structuralintegrity of 6e pedestal could be degraded sher significant erosion
- Non-condensible gas generatiera can cause containment overpresswization
. !! debris is not covered by water, high drywell gas temperatures could cause containmentleakage or failurs
- Containment liner could be compromised
- Core debris is expected to be coolable
- A smalltraction of ne debris enters the lower drywellupon vessel failure
- Firewster addition system orpassive flooder covers debris win water
- Large floor aren prornotes quenching
- Remainder of debris enters ne lower drywell very slowly allowing debris to be cooled as et drips from the vessel cra sus so
Mitigation of continued core-concrete interaction
- Debris willbe covered by water from either firennter addition system or passive flooder
- Promotes upwardhasttransfer
- Relatively deep waterpoolassures low drymiltemperatures
- AbIntion rates arelow
- Analysis indicates minimum upward heat flux is 100 k W/ra!
- Resuking long term axial ablation is on the order of I cn#r -
- Containment liner not reached for about one day
- Podestalcan withstand Intge amount of ablation
- Concentric rings with web stiffeners about 11m thick
- Outer ring plus a fe w centimeters of stiWooers sufficient to support loads
- Assuming 1:1 ratio of axial to radial ablation rates, pedestalintegrity assuredfor two doys cre ams ,,
Mitigation of continued core-concrete interaction
- Containmentpressuritation
- Basskic concrete limits production of non-condensable gasses
- Increase in partialpressure of non-condensables is partinfly offset by lowersteam generation
- Impact on containment response is isirly benign
- Consider a case with 100% of the core debris in the lower drywoll an d 10ln Whn2 upward h eat flus
- Timing of CDPS actuation About 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if firewster addition system is used About 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> if passive flooder is used
- No significant impact on fission product rolesse because of suppression poolscrubbing cumsms a
AerosolPlugging issue:
Could the plugged leak paths through the partially opened vacuum breakers be porous enough to allow significant gas flow?
Reference:
'lenksge ot Aerosols trom Containment Buildings *, H. A. Morewits Hesith Physics Vol. 41, No. 2 (February), pp 1%-207, Int.
Response
Plugs are not expected to be porous.
The impact ofplugging on the consequence anslysis is very smsII.
ces sus n PorosityofAerosolPlug
- A small, concrete, tilt up panelbuilding was tested at Atomics InternationsIin the esfly 1.960's
- Additicanttests were cited with smallbypass stess which indicate that plugs will n ot allow significant flow for low pressure differences
- Some experiments indicate a porous plug was generated for Isvge pressure differences (30-1000puis)
- Maximum pressure diMerence scross the plug in the ABWR willbe limited to the hest of water above the first row of horizontal vents (14 psid)
- Complete blockago is expected can aus u
Ifnpact of Aerosol Plugging Model on Consequence Analysis
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$ u,: -. ( l L - u. .- sl u _ l x. l i 10 Ito lt118 t.AR1Y M E11151)Y [1riE IN RIM
.Arpagapng ash Cf 8 4%9: f5 Suppression PoolpH Corrtrol Onestion:
Research parformed at Oak Ridge indicates that a substantialincrease in sirborne iodine can accer if ne suppression poolbecomes scidic.
The radiolytic formation of nitric acid has been identified as a mechsaism for tire 91 r of the suppression poolto decrease. Inst impact does this mechanism have on the fission product behsviar in 6eABWRT Refetwace:
NUREG/CR 573% 'fodine ChemicalForms in iWR Severe Accidents *.
FinalReport January t.9st,
Response
A calculstron of poolpH has been performed. The results indicate pool willnot become acidic during the firstN hours of the transient.
A section willbe added to de $$AR to address this issue.
cu sner w
- w. ..-
Suppression Pool pH - Scoping Calculation .,
- porthereference:
'11ne pH is controlled so not it stays above 1, a reasonable value for the >
traction ofI-converted to12is JE-4. Tekle A6 indicates a small :
production of volatilen for pWRs het virtusNy none forRWRs.* ;
- Reseks of preliminary analysis indicate poolpH will NOTbecome acidic .
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> forphysicsHy realistic accidents due to presence of of Cs0Hin nepeel.
pH' # #
Timelkr) pH pH 0 1 km itsE ' '
1 5.6 15 1056 10 4.6 KS3 It51 ,
24 4.3 K59 ft49 Case 1 Initially neutralpool, M of Csl and Cs0H energy not physical >
Case 2 1% initini Cs0H in pool. In of Csl and Cs0H energy .
Case 3 - M initial Cs0H in pool, M of Col and Cs0H energy craames n i
l Containment Response to Rupture Disk Opening ,
Conce:w:
Fission product relensa from he suppressinn poolat no time ne rupture disk opens Aanlysis considers: l Rete of containment dcr iressuritation l Poolswell Carryover of fission products arough ne vent due to entrainment '
h 4
b i
crewrses se -
1 4
l Blowdown of Containtnent lower Bound time constset enlculated for no blowdown of de cutsinment
- Blowdown of wetwellfrom Npsig considered (rupture disk setpoint subsequentlytnisetto S0psig)
- Flashing of suppression poolneglected
- Steam flow from drywellneglected Blowdown time constant is more than S minutes.
Non-uniform effects around the circumference of de suppression pool are negligible.
cr. .., ,,
PoolSwelldue to flashing
- Based on driftflux model
- Uniform void generation rate assumed based on conservative depressariintion rate
- Bulk poolswellis a6 meters cre ser n
Poolswelldue to flowfrom dryweII
- Analysis again based att drift flux model
- Difficuky is in determining the effected ares in the suppression pse,'
- Flow rete from de drywellbesed on the depresseritation response of the suppression pool and de drywell 90% of steam flow through ne vent will originate in ne suppressionpool
- Assume poolswellto de level of ne vents and calculated the atlected radius ifit is smaller dan the distnace from the inner to the outer well d the wetwell, poolswellis not a concern y
(
- PoolsweIIdue to dryweIIflow would have to be 3.3 m to reach level ofrupture disk
- Ettactive radius of region allected by drywellflowis as m
- Inner and outer walls of wetwsilare 1.5 m spart l Poolswellis not a threat to COPSpiping l c
as sus >,
Carryover of fission penducts due to entrainment -
- Superficial velocity from de surface of the poolis more than two orders of msgaitude below Kutateladze entrainment dreshold
- A more sophisticated estimats of entrainmentis possible using the experimental work byRozen
- Superficial velocity andpressure determine ne retio oflipid mass to vapor mass which leaves the surface d the pool
- Approximately one-tenth of the poolmass flashes during blowdown
- Fisction of suppression poollipic which might be transported through i ne rupture disk is about 4E 1
- Since fission products are in the form of a dissolved salt this is an upper bound on the fission product release due to entrainment
- Sensitivity studies indicate dat ettsits consepences not significantly -
stlected for volatile fission product releases of IE-3 Carryover of fission products will not contribute le offsite consepence as sus er i
l
b b
- Primary use ofPRA -
Improve the Design Input to Reliability Assurance Program Identification ofimportant Features -
Support improved Technical Specifications Summary ofKeyResults andinsights.
PRA Status JD Duncan CEBuchholz GENuclear June 1993 ,
- . = . . . . . . - . . . - - .
4 ABWR Design Changes Made as a Result of PRA Studies
- 3-division ECCS
- ACindependentwateraddition
- Gascombustionturbine
- RWCU filter domin bypass to permit decay heat removal at high
, pressure
,
- Containmentoverpressureprotection +
- -Automation of Suppression pool cooling ^
\
- Automation ofATWSmitigation
- ADS drywellpressure bypass timer i
- RCIC capabilityforlocalcontrol
,
- P SRVadded to Remote Shutdown Panel
- Increased rating oflowpressure piping to eliminate ISLOCA concerns
- RWCUdrainlineisolation valve
- Service Waterand Circ Waterpump trips and isolation on flood
- Improved RHR heat exchanger supports .
1
..___._._.-.m...___.____ _._________._m__ ___r2&_m.. -. . . __i,. i mr ~. _, ,, ,, 4 . ,, . . , - -- s.... -.-. . .-. - - ,w,.. - . ., # , , . .. _ m_ -c .._._ . . , _ .
Summary of InternalEventResults CDFby Type of Event Type of Event CDF % of Total
- Transients 3.7E-08 23
- L oss of offsite power 1.2E-07 76
- LOCAs 6.9E-10 <1
- ATWS 2.7E-10 <1 Total 1.6E-07 100 o
O
Primary Features contributing to Low Core Damage
. Frequency 3 ECCS Divisions, including steam driven RCIC Automatic Depresurization I
3 Divisions ofheatremoval i 1
. Combustion Turbine Generator and AC independent.
waterinjection
.DiverseReactivity Control
, i i
) .-
._,..~._,.. _...;
Summary ofSeismic Margins Study
< cc>y 9aw"abe fICLPF> 0.60g
, ar W y&Yy p- 9 Forallaccidentsequences Forallaccidentclasses e
b
.- __---..-_.--=__----_--:--------
--=x
.. ; ~
Primary Features Contributing to High Seismic Margins
- Many ABWR design features provide protection againstcore damage events
- Multiple systems for reactivity control, core cooling and -
containment cooling f
- Fire waterinjection (with valves having manual operation .
capability) e i
I 4
__ ._____ . _ m .m _- -. - - __ . - ~ ~ ,,,.-,s . . .-%.. - - +e,~-, u ~ .m,- . w>. - ._ + - - , . r -w, , -~ _ _ , , - . ,
Summary Conclusions from Flooding Analysis Total core damage frequency for internal floods is conservatively estimated at SE-9 per reactor-year for low UHS and BE-9 per reactor-year forhigh UilS Most potential floods will be automatically terminated by level sensors (turbine and controlbuildings)
- Floods in the reactor building can be contained in the ECCS rooms or the corridorof the firstfloor e
b
. . - _ - _ . _ _ _ _ . . - - - . - - - _ - - . - . - - . - _ - _ . - . _ _ . _ _ - - - - - - - _ - _ - - - _ - , - - ~.
~
9 4
^
FireAnalysis .
- Conservative Screening Methodology
~
- Fire Assumed to Result in loss of Function Area Fire Frecuency CDFis less than Divisional 0.1peryear each 2E-7 ControlRoom 0.04per year 9E-7 i Turbine Building 0.2oer voar 2E-7 0.5per year E-6
- . Resulting PlantModifications ,
- Control of additional SRV at remote shutdown panel
. - Insure RCIC operable from outside control room I-l l
_ _ _ _ _ . _ _ _ _ _ - _ ~ _ _ . . . . . _ . .. . _ - . . . _ . _ . -__ _ _ - . . _ _ . _ _ - _ _ . _ _. _ _ _ _ _ _ _ _ _ _
L Shutdown Assessment: Loss of Decay Heat Removal ;
5
Purpose:
- - Determine minimum equipment sets that meet core damage frequency criterien.
- Methodology: '
, - Operating RHR system assumed to randomly fail with a frequency of
(
0.1/yr
- Success criteria forpreventing core damage identified l - Fault trees and event trees developed similar to fullpower PRA
- Minimum sets of systems identified to not be in maintenance (all >
remaining systems assumed to be unavailable) l - Conditional core damage frequency calculated to meet criterion of ;
V 1.0E-5 which when combined with a loss of RHR frequency '
[ corresponds to the USNRC "large release" goal of 1.0E-Glyr .
l l ,~...
~
l- .. .
[
u Shutdown Assessment Many minimum sets of equipment can meet safety criterion. - This provides significant flexibility to the utility to plan maintenance activities. This is possible because of the multiple means ofproviding l makeup waterto the core.
Minimum sets identified which are comprised of systems required to be operable (ECCS, ACIWA) ornormally operable (CRD)
If credit was taken for all available systems, the core damage frequency i wouldbe 1.0E-7/yr l
'l
.. c . . . - . < - s -
y , ,. e as,%+ -.e + 9.-. , ,* .su c ' .iE.,,mm $_ .. -- - ,,-<-
i L
- Overallcontainmentperformance I a
- Conditional containment failure probability goal of 0.1 established in the .
ABWR Licensing ReviewBasis ;
- Potential for containment failure modes examined for all core damage ,
sequences :
Two definitions for containment failure considered Large release (assumed to be greater than 25 rem at the site boundary)
CCFP = 0.002 Pressure lategrity (" Integrity as a pressure boundary can no longer be
- i
- controlled"- SECY-90-016) .
CCFP = 0.005 1
ABWR meets established goal with wide margin .
g ]
CEB W15A92-2 : -
. . . . . . - _ . _ _ - . . - - _ _ --.-....._,...______..1--.
DominantRelease Sequences Sequence Frequency %
Normalcontainmentleakage 1.34E-7 85.9 %
Long term containment overpressurization 2.08E-8 13.3 %
leading to rupture disk opening Long term containment overpressurization 3.91E-10 0.3%
leading to drywellhead failure Early containment failure due to DCH or 4.05E-10 0.3%
suppression poolbypass long term station blackout leading to drywell 1.70E-10 0.1%
failure l
l l
CEB 6'15/32-3 l
ABWR design features important for accident mitigation
- Pressure suppression containment
- Containmentisolation
- Drywell-wetweII vacuum breakers
- Reliable vesseldepressurization
- Containmentinerted with nitrogen
- Lowerdrywellconfiguration Floorarea Basaltic concrete Protection forlowerdrywellsumps
- Firewateraddition system
- Lower dryweIIflooder
- Containment overpressure protection system O
CEB W1582-4 -
4 ..
- Containment Overpressure Protection i
Stack Lowrupturepressure
/ N2
'c
/
TA Reactor A0 C)G A0
~
Wetwell Building
,
- Operation Rupture disk opens at pressure above design pressure and below:
1 servicelevelC Later, operator closesisolation valves to regain control of l containmentintegrity.
l l
- Benefits
- Passive operation, high seismic capability. Very reliable
- Suppression pool scrubbing, elevated and monitored release .
CEB 6/15/92-5
- _ - _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ . . .-. . . . , , . .:..._.. ~ . . . ..~ ..,
l .
l' ABWR severe accidentmitigation l
l
- Containment overpressure I
protection i
Containment m
ABWRpassive features which mitigate Pressure severe accidents vessel
- Inertedcontainment
- Lowerdrywellfloodcapability core
- Suppressionpool-fissionproducts ac,,yy;,y ,,g ,,geggy,,
.
- Containmentoverpressureprotection 1 Fusible valve F
High degree ofpublicprotection ..
CEB 6/15192-6
~
- = , - . --- , ,w - . - -
h PRA STATUS i
- Calculations complete (except tech. spec.) .
- Final verification underway
- PRA Analysis .
- ConsistencywithDesign
- Developing last 1% forNRC
- - Early venting forloss of decay heat removalsequences
- Emergencyprocedure guideline improvements
- Improved containment overpressure relief design k t
. , +
l -
& GE Nuclear Energy CAD Systems Image Systems Data Base System PlantIAodeling M30357.1
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- Current Status
=lAultiple Systems
- Major Projects
- IAigration Strategy M30357.2
V as:er Mocel Conce:r:
Libraries and Specifications aster Configuration = Model Administrator Model n
Model Cycle Processing UO V Submodels v _ , , . -
i Lifecycle Apalica: ions '
, inee Design 1 Decomrn.issioning !
Plant Lifecycle Plant
[ Construction &
, i Configuration Maintenance & -
- "8' er t Mode (Procurement Ped Turnover &
art-u 3- '
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. =. . . . - _ , _ - _ _ - _ _
In':egratec Aaaroac1 Projec': Configuration Mocel O. Control Project Data Base Start-up - l 3 - Maintenance -
Other .
l QA/QC Human Resources ,
e
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___.____._______.___________________m_ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _~ _ _ . _ -- - _m.-m -... ~. .- .,--
e GENuclear Energy Image Systems
- Current Status
- Planned Upgrade
- Client Server System M30357.3
- + e - r m n- v ~ e -+e w e
1
@ GE Nuclear Energy Data Base System
- Current Status
- IAigration Strategy
- POWRTRAK M30357.4
.I ..
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GE Nuclear Energy PlantIAodeling System
- Need
- Applications .
- Implementation M30357.5 4
Ua .__ __ . .. . __. __
__.___.-_.__-.-_______.______.._-__..__-_____.__-_m_ - - _ . _ _ _ _ _ _; . *. --
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O GENuclear Energy Summary .
- Qualit/
- Productivitt
- System Evolution
. M30357.6
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- L
. PROPOSED AGENDA ACRS SUBCOMMITTEE MEETING .
MICHELSON BRIEFING SESSIONS
-JUNE 15, 1993 JUNE 15, 1993 - TUESDAY ROOM J-1010 o REACTOR SERVICE WATER SYSTEM (l' HOUR)
- OVERVIEW PROSPECTUS
- DESIGN AND EVALUATION STATUS
- NRC REVIEW STATUS { EACH ISSUE } 1
- ACRS CONCERNS STATUS r i
o SOLID STATE. ELECTRONIC COMPONENTS (1 HOUR) -;
o ESSENTIAL CHILLED WATER SYSTEMS -(1 HOUR) o COMBUSTION TURBINE GENERATOR - 580 (1 HOUR)
LUNCH o PRESSURIZATION ANALYSIS (1 HOUR) ,
o CONTAINMENT PERFORMANCE (1 HOUR)- I o PLANT DESIGN AND PRA - SAFETY EVALUATIONS (2 HOURS) o ULTIMATE HEAT SINK REACTOR SERVICE (I HOUR) i 4
JWP 6/10/93 ,
i
^
PROPOSED: AGENDA ACRS SUBCOMMITTEE' MEETING MICHELSON BRIEFING SESSIONS JUNE 16, 1993 JUNE 16, 1993 - WEDNESDAY ROOM J-1010 P
O PLANT DIVISIONAL SEPARATION - TUTORIAL REVIEW
- REACTOR BUILDING / SECONDARY CONTAINMENT /
CONTROL BUILDING (1 HOUR) r
- PLANT PHYSICAL SEPARATION BARRIERS -
DESIGN AND EVALUATION i (REACTOR' BUILDING / SECONDARY CONTAINMENT /
MSL TUNNEL / CONTROL BUILDING) (1 HOUR)
- PLANT HEATING, VENTILATING AND AIR CONDITIONING SYSTEMS (1 HOUR)
- REACTOR BUILDING / SECONDARY CONTAINMENT DRAIN SYSTEMS (1 HOUR)
LUNCH o REACTOR WATER CLEANUP SYSTEM (2 HOURS) o SPECIAL 0/As (I HOUR) ,
O COMPUTER-AIDED DESIGN DEMONSTRATION (1 HOUR)
JWP 6/10/93 '
.o e i 4
PROPOSED AGENDA ;
ACRS SUBCOMMITTEE MEETING FULL SUBCOMMITTEE BRIEFING SESSION JUNE 17, 1993 JUNE 17, 1993 - THURSDAY HOLIDAY INN / PARK CENTER PLAZA o PLANT DESIGN LIFE AND AGING MANAGEMENT u) 3s- '
,o STATION GROUNDING AND SURGE PROTECTION (pour) 30 ~
o CORROSION CONTROL
([ HOUR)g.# .
o REACTOR INTERNAL PUMPS IlyduR) o STANDBY COMBUSTION TURBINE GENERATOR (1 dour).
er 7
LUNCH .
3&'
o CONTAINHENT PERFORMANCE ( OUR) ,
16 ,
pp o ELECTRICAL INSULATION (1 uR)
-o-PRA4TATUS --
(1-teouRFj- ,
t$
o ITAAC STATUS (1 UR)
, I( ?
o ACRS OPEN-ISSUES STATUS (1 HOUR) s
- [
_, OST C.fi JWP 6/10/93
.* ')
l ABWR DIVISIONAL SEPARATION SECONDARY CONTAINMENT -
SAFETY-RELATED EQUIPMENT IS DIVISIONALLY SEPARATED
. PHYSICAL BARRIERS SEPARATE EACH DIVISION-EQUIPMENT QUALIFIED FOR ENVELOPING PRESSURE AND TEMPERATURE EFFECTS DUE TO AN EVENT IN ANOTHER DIVISION
. FIRE OUTSIDE CONTAINMENT 3 HOUR UL RATING ON PENETRATIONS THROUGH DIVISIONAL WALLS AND FLOORS SAFETY-RELATED EQUIPMENT IN NONFIRE DIVISION QUALIFIED FOR TEMPERATURE EFFECTS OF FIRE
. FLOOD OUTSIDE CONTAINMENT
- EQUIPMENT RAISED OFF FLOOR SLABS FOR PROTECTION
- FIRE DOORS MINIMIZE SPREAD OF WATER BETWEEN DIVISIONS
- DRAIN SYSTEM CARRIES WATER TO BASEMENT
- BASEMENT ACTS AS HOLDING POOL FOR ALL BREAKS
- WATERTIGHT DOORS PROTECT SAFETY-RELATED EQUlPMENT FROM WATER 4
- c. ..
- ~
ABWR DIVISIONAL SEPARATION SECONDARY CONTAINMENT ,
. HELB OUTSIDE CONTAINMENT
- RCIC AND RWCU ANALYZED PRESSURE DETERMINED BY MINIMIZING VOLUMES TO A GIVEN DIVISION
- ROOM TEMPERATURE BASED UPON STEAM SATURATION TEMPERATURE AT PEAK PRESSURE
- EQ VALUES BASED ON RESULTING PRESS'URE AND TEMPERATURE
. LOSS OF COOLANT ACCIDENT
- 0.5% OF VOLUME PER DAY GROSS CONTAINMENT LEAKAGE ASSUMED STANDBY GAS TREATMENT DRAWS DOWN SECONDARY CONTAINMENT TREATS LEAKAGE SGTS DESIGN ASSUMES MIXING OCCURS BETWEEN DIVISIONS ALL SAFETY-RELATED EQUIPMENT INSIDE SECONDARY CO'NTAINMENT ARE PROTECTED FROM DESIGN BASIS EVENTS USING A MlXTURE OF PHYSICAL PROTECTION FEATUIES AND EQ PROFILES
e .
~,
1 ABWR EVALUATION OF EX-CONTAINMENT LOCAs ACRS CONCERNS (JULY 1992)
EX-CONTAINMENT LOCA IS MOST SIGNIFICANT RWCU HAZARD LINE ROUTING OUTSIDE CONTAINMENT NOT DEFINED -
ISOLATION VALVES MAY NOT CLOSE WITH LOCA ,
PRA DOES NOT CONSIDER RWCU BREAK AS AN INITIATING EVENT BYPASS STUDY METHODOLOGY NOT NECESSARILY CONSERVATIVE 1
^l ACRS CONCLUSION: RWCU BREAKS SHOULD BE SPECIFICALLY CONSIDERED IN THE PRA I
i GE CONCLUSION: RWCU LINE BREAKS DO NOT POSE A SIGNIFICANT RISK j i
I l
i i
ABWR EVALUATION OF EX-CONTAINAfENT LOCAs RWCU LINE BREAKS POTENTIAL LINES LINE SIZE ISOLATION EXCLUSION BASIS RWCU Suction 8" Af0-AfD Bottom IIead Drain 4" AIANUAL BHD ROUTED TO (BHD) CLOSURE SUCTION LINE ADDED INSIDE DRYWELL RWCU Return 8" A10-Af0-CK-CK BOUNDED BY FW LINE BREAK RWCU Head Spray 6" CK-Af0-AfD (NC) NORA1 ALLY CLOSED RWCU Instrmnents .25" XFCK RESTRICTIONS
~
PREVENT SIGNIFICANT RELEASE l
i l
i l
s i
ARWR EVALUATION OF EX-CONTAINMENT LOCAs :
i 1
EVENT SCENARIO ,
.j AfEDIUM SIZE LINE BREAK IN RWCU ROOh!
310 ISOL1 TION VALVES FAIL TO CLOSE
.i NO OPERATOR ACTION TO CLOSE RPV A1AKEUP FAILURE ,
RPV LEVEL FALLS TO RWCU SUCTION N0ZZLE (> TAF)
CORE UNCOVERY AND EVENTUAL CORE DAMAGE WITHOUT '
MAKEUP OR RECOVERY h
i r
l v
s i
'l 2 1 2
- . - = = ' .- e- m.. ~ y ,j
ARWR EVALUATION OF EX-CONTAINMENT LOCAs REALISTIC CALCUIATION OF PRORARII2TY :
LINE BREAK
[2.4x10'IYR]
I ISOLATION VALVE FAILURE .
[3.1x1W] ;
.g .
NO OPER. ACTION
[lx10']
I a LOSS OF CORE COOLING
[6.2x10'] -
TOTAL CDF APPROX 4.8x10"/YR
.jl
- y 3
6 I
- . ~
.~ .
6BWR EVALUA TION OF EX-CONTAINMENT LOCAs BASIS FOR PROBABILITIES EVENT VALUE BASIS LINE BREAK 2.4E-4/YR WASH 1400 (> 3") -
= 8.6E-10 PER SEGMENT- -
IIOUR x 7000 HRS /YR x 40 SEGMENTS REDUNDANT 3.2E-4 MOV COMMON CAUSE ISOLATION VALVE FAILURE FAILURE [1.8E-3 GE PRA DATA FOR MOVs; .18 BETA]
INDUSTRY ISSUE TO BE RESOLVED NO OPERATOR 1E-2 EVENT SPECIFIC ACTION PROCEDURE - CLOSE BHD AND MAINTAIN LEVEL BELOW BREAK LOSS OF CORE 6.2E-7 CONDITIONAL COOLING PROBABILITY FOR LOCA EVENT DIV 4 AREA BLOWOUT PANELS RELIEVE PRESSURE AND EQUIPMENT DESIGNED FOR 100% STEAM ENVIRONMENT TOTAL 4.8E-16/YR l
l I
l
.A' -.e .
b . . )
H
- ABWR EVALUATION OF EX-CONTAINMENT LOCAs. ;i CONCLUSIONS ACRS CONCERNS ADDRESSED
.i IS0btTION VALVE CLOSURE IS AN INDUSTR)' CONCERN .
CORE DAMAGE FREQUENCY IS EXTREMELY LOW l ABOUT 10' % OF TOTAL CDF l MUCH LESS THAN POOL SCRUBBING FACTOR OVERALL CONCLUSION: ;
TREATMENT OF BYPASS IN PRA IS VERY CONSERVATIVE RWCU LINE BREAKS DO NOT POSE A SIGNIFICANT RISK IN ABWR ,
f
)
1
ABWR DESIGN BASIS FLOODING OVERVIEW OF KEY ASSUMPTIONS
. CONSERVATIVE ASSUMPTIONS TAKEN
- 1. THROUGH WALL CRACK FOR MODERATE ENERGY PIPES
- 2. COMPLETE SEVERANCE OF PIPE FOR HIGH ENERGY PIPES
- 3. 30 MINUTE OPERATOR RESPONSE FOR FLOODING ANALYSIS
. NO CREDIT FOR DIVISION THAT IS EXPERIENCING THE PIPE BREAK
. SINGLE ACTIVE FAILURE TAKEN OR COMPLETE DRAINING OF LEAKING SYSTEM
. ASSUME PIPE DRAINS BETWEEN BREAK AND CLOSED VALVE NO FAILURE IN A DIVISION WILL CAUSE THE LOSS OF A SECOND DIVISION
I J s4.L..., .64.g A+a. s=.. s a 4 i , 4 - .,a
- " .. 4- ,
,4 AB~WR DESIGN BASIS FLOODING REACTOR BUILDING FLOODING ANALYSIS
. DONE ON A FLOOR BY FLOOR BASIS CASES ON FLOOR B4F (-8200mm TMSL) ,
t
- 1. 450mm ECCS SUCTION LINE FAILURE
- 2. 200mm SUPPRESSION POOL COOLING LINE FAILURE
- 3. 200mm RWCU LINE FAILURE
- 4. ALL FLOORS ABOVE DRAIN TO BASEMENT LEVEL SUMPS i A MAXIMUM OF 2.44m IS EXPECTED IN THE HALLWAY PLUS AT MOST ONE SAFETY DIVISION :
. CASES ABOVE B4F AND OUTSIDE MAIN STEAM TUNNEL
- 1. 300mm REACTOR COOLING WATER LINE FAILURE j i'
- 2. 200mm FIREHOSE STAND PIPE FAILURE .
- 3. 200mm RWCU LINE FAILURE A MAXIMUM OF 150mm OF WATER IS EXPECTED
. CASES IN MAIN STEAM TUNNEL
.i
- 1. 550mm FEEDWATER LINE FAILURE ,
1 A MAXIMUM OF 4.3m OF WATER IS {
EXPECTED i
I l
~. .
ABWR DESIGN BASIS FLOODING CONTROL BUILDING FLOODING ANALYSIS
. DONE ON A FLOOR BY FLOOR BASIS
. FLOODING ON BASEMENT LEVEL
- 1. RSW LINE FAILURE A MAXIMUM OF 3.14m OF WATER IS EXPECTED IN ONE OF THE THREE SAFETY DIVISIONS
. FLOODING ABOVE B4F AND OUTSIDE MAIN STEAM TUNNEL
- 1. HECW LINE FAILURE
- 2. RCW LINE FAILURE
- 3. FIRE HOSE STANDPIPE FAILURE A MAXIMUM OF 150mm OF WATER IS EXPECTED
. FLOODING OF MAIN STEAM TUNNEL
- 1. FEEDWATER LINE FAILURE NO STANDING WATER IS EXPECTED