ML20206S567
| ML20206S567 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/30/1998 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-3108, NUDOCS 9905210159 | |
| Download: ML20206S567 (13) | |
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l-Da,te Issued: May 30, 1998 CERTIFIED:
June 16. 1998 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l
ADVANCED REACTOR DESIGNS SUBCOMMITTEE MEETING MINUTES MAY 13-15, 1998 ROCKVILLE, MARYLAND The ACRS Subcommittee on Advanced Reactor Designs met on May 13-15, 1998, at-11545 Rockville Pike. Rockville. Maryland, in Room T-2B3. The purpose of the meeting was to review the AP600 Standard Safety Analysis Report (SSAR) and the associated advanced Final Safety Evaluation Report (FSER) chapters 3, 6.14.
- 16. and 17; regulatory treatment of non-safety systems: the Level 2 and 3 AP600 probabilistic risk assessments; and severe accidents including external reactor vessel cooling.
Mr. Noel Dudley was the cognizant ACRS staff engineer for this meeting. The meeting was convened at 8:30 a.m. and recessed at 5:43 p.m. on May 13. 1998: reconvened at 8:30 a.m. and recessed at 2:12 p.m. on May 14, 1998; and reconvened at 8:30 a.m. and adjourned at 11:50 a.m. on May 15, 1998.
ATTEND 3ES ACBS J. Barton Chairman W. Shack, Member M. Fontana, i>mber R. Uhrig Member T. Kress, Member.
C. Chen Consultant D. Miller, Member J. Carroll, Consultant D. Powers, Member I. Catton, Consultant R. Seale, Member N. Dudley. Staff Engineer NRC STAFF T. Quay. NRR A. Levin, NRR T. Kenyon, NRR J. Wilson, NRR W. Huffman, NRR A. Chu, NRR J. Lyons NRR G. Bagchi, NRR M. Snodderly. NRR S. Hou. NRR R. Palla,'NRR T. Cheng. NRR D i R. Gramm. NRR R. Pettis. Jr., NRR
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.s WESTINGHOUSE ELECTRIC COMPANY g~~ p v 'q B. McIntyre, Westinghouse R. Orr Westinghouse D. Lindgren, Westinghouse T. Schulz, Westinghouse J. Scobel. Westinghouse R. Lutz, Jr., Westinghouse I
~ p;3 E. Piplica Westinghouse DESIGNMED ORIGINAL N 10159 9805M N-x
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Advanced. Reactor Designs Subcomittee 2
May 13 15, 1998
.There were no written comments or requests for time to make oral statements received from members of the public. An attendance list of the NRC staff and public is available in the ACRS office files and will be made available upon request.
SUBCOMMITTEE CHAIRMAN INTRODUCTION Mr. John Barton Subcommittee Chairman, convened the meeting at 8:30 a.m. and noted that this was a three day meeting that would cover a wide range of topics related to the AP600 design.
NRC STAFF PRESENTATION Mr. Thomas Kenyon. NRR summarized the past AP600 review schedule and the planned meetings with the ACRS.
He noted that there were no significant open items related to the AP600 design and listed those topics that the staff planned to specifically address.
Mr. Kenyon noted that the Director of the Office of Nuclear Reactor Regulation decides on issuing the AP600 Final Design Approval with negative consent from the Commission. The Subcommittee members and the NRC staff discussed the eight open items in the FSER, and the number and types of exemptions granted by the staff WESTINGHOUSE PRESENTATION Mr. Brian McIntyre. Westinghouse, noted that information concerning the containment analysis would be presented during an upcoming meeting of the Thermal Hydraulic Phenomena Subcommittee.
SSAR CHAPTER 6: ENGINEERED SAFETY FEATURES - Mr. Terry Schulz. Westinghouse Mr. Terry Schulz. Westinghouse, presented information on the passive containment cooling system (PCCS) safety-related functions and design.
He explained that the PCCS provides long term cooling, a backup water supply to the fire protection system, and air only heat removal capability to prevent failure of the containment due to over pressurization.
Mr. Schulz described the containment hydrogen control system including the passive autocatalytic hydrogen recombiners (PAR) hydrogen igniters, and hydrogen monitoring sub-systems. He presented information on the non-safety systems used for primary coolant makeup. decay heat removal, and main control room emergency habitability control.
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Advanced. Reactor Designs Subcommittee 3
May 13-15. 1998 The Subcommittee members, the NRC staff, and Westinghouse discussed:
the 2-dimensional model used to calculate peak containment pressure.
the time to reach half of peak containment pressure.
volume of the containment.
freeze protection for components external to the containment.
corrosion protection and inspection of the exterior of the containment.
spent fuel off-load capacity of the spent fuel pool.
purging of the containment through the spent fuel pool.
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. environmental qualification testing of the PARS.
J location of hydrogen igniters.
PCCS system design features that minimize water hammer.
inservice inspection of PCCS check valves, and operating experience and reliability of squib valves.
Mr. William Huffman. NRR stated that the staff considered all open items i
related to this chapter to be technically resolves.
He noted that the issue of heat sinks in the containment will be a Combined License (COL) confirmatory item.
SSAR CHAPTER 14: INITIAL TEST PROGRAM - Mr. Eugene Piplica. Westinghouse Mr. Eugene Piplica. Westinghouse, presented a summary of the initial test program. He described the purpose of the program and the test objectives of the construction and installation, preoperational, and startup tests.
Mr.
Piplica explained the organization of the preoperational and startup test abstracts. He differentiated between the first plant only special tests and the first three plant special tests. Mr. Piplica described the natural circulation tests for the steam generator and the passive residual heat removal heat exchanger.
He listed the responsibilities of the COL applicant.
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The Subcommittee members, the NRC staff, and Mr. Piplica discussed:
completing preoperational tests before fuel load.
definition of defense-in-depth systems.
reason for first three plant special tests.
similarity among different plants' piping designs.
types of initial tests for the instrument air system, and control rod configuration.
l Mr. Jerry Wilson. NRR. stated that the staff found the initial test program acceptable.
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l Ad'vanced Reactor Designs Subcomittee 4
May 13-15, 1998 LEVEL 2 AND 3.PROBABILISTIC RISK ASSESSMENT - Mr. James Scobel. Westinghouse i
i Mr. James Scobel, Westinghouse, identified the interface between the Level 1 and the Level 2 AP600 probabilistic risk assessments (PRAs) and the associated AP600 accident classes.
He described the severe accident phenomena, j
simplifying assumptions, and the top eleven events in the containment event tree.
Mr. Scobel presented the containment event tree structure and the results of the hydrogen, core damage frequency (CDF), and large, early release frequency (LERF) analyses.
Mr. Scobel identified the interface between the Level 2 and the Level 3 PRAs.
He presented the release categories and the release category contribution to LERF. He explained the fission product source terms and the model used in the consequence analysis. Mr. Scobel summarized the at-power, shutdown risk, and j
total PRA results for CDF, LERF, and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> off-site population dose _ risk.
The Subcommittee members, Westinghouse, and the NRC staff discussed:
validation and verification of the LOFTRAN code:
hydrogen diffusion, deflagration', and detonation:
containment fragility:
risk contribution of steam generator tube rupture events:
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consequences of in-vessel steam explosion:
probability of reactor vessel failure:
use of probabilistic fracture mechanics:
proper balance between prevention and mitigation of risk:
1 basis for containment leak rate limits; and experimental base for assuming adequate water chemistry controls.
i IN VESSEL RETENTION OF MOLTEN CORE DEBRIS - Mr. James Scobel, Westinghouse i
Mr. James Scobel, Westinghouse, explained why the AP600 design is uniquely suited for in-vessel retention of the reactor core during severe accidents.
He presented the conclusion of DOE /ID Report 10460. "In-Vessel Coolability and Retention of a Core Melt." that the AP600 reactor vessel will not fail.
He summarized the principle assumptions and scenario used in DOE /ID Report 10460.
Mr. Scobel described the thermal loads on the vessel, the critical heat flux determined by experiments, and the results of structural analyses. He noted l
that the'INEEL analysis INEEL/ EXT-97-00779, Potential for AP600 In-Vessel Retention through Ex-Vessel Flooding," identified core melt configurations l
that were not bounded by DOE /ID Report 10460 and could result in failure of l
the reactor vesse?.
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Advanced ' Reactor Designs Subcommittee 5
May 13-15, 1998 The Subcommittee members, the NRC staff, and Westinghouse discussed:
the inability to accurately model core melt accidents, uncertainties' calculated by the AP600 PRA, e
exothermic reactions between molten uranium and the reactor vessel, e
the adequacy of critical heat flux experiments at Penn State, and e
the critical heat flux that would result in failure of a reactor vessel.
EX VESSEL SEVERE ACCIDENT TOPICS - Mr. Robert Lutz, Jr., Westinghouse Mr. Robert Lutz, Jr., Westinghouse, explained that the AP600 design assumes that external vessel cooling prevents failure of the reactor vessel under all core damage accidents, however, limited analyses of ex-vessel phenomena were conducted.
He described the two reactor vessel failure modes that were assumed in the ex-vessel severe accident analyses and the characteristics of the associated core debris transport mechanisms. Mr. Lutz presented the results of the analyses for direct containment heating, ex-vessel steam explosions, and core-concrete interaction.
He concluded that based on the limited set of deterministic analyses the goal of protecting the containment fission product boundaries during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a severe accident was met.
The Subcommittee members, the NRC staff, and Westinghouse discussed the adequacy of the TEXAS code to model ex-vessel steam explosion events, consequences of core melt penetrating the basemat, and the effects of the hydrogen generated in the containment.
EXTERNAL REACTOR VESSEL COOLING - Mr. Robert Palla, NRR Mr. Robert Palla. NRR, explained that the staff used a balanced approach to address the adequacy of the AP600 design for ex-vessel events by relying on the high probability of in-vessel retention of the core and on limited j
analytical evaluations of ex-vessel phenomena.
He listed the reports
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submitted by Westinghouse and the technical reviews performed by the NRC staff.
Mr. Palla concluded that reactor pressure vessel failure cannot be l
ruled out for all possible core melt scenarios. However, deterministic analyses indicate reactor pressure vessel (RPV) failure will not result in early containment' failure and probabilistic analyses indicate that containment
' failure frequency will remain below the large, early release goal.
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' Advanced Reactor Designs Subcomittee 6
May 13-15, 1998 The Subcommittee members and Mr. Palla discussed:
experimental RPV heat fluxes compared to the critical heat.
modeling uncertainties associated with chemical interactions, use of defense-in-depth approach in the design review, and operating plants adopting the external reactor vessel cooling approach.
I SSAR CHAPTER 3: DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS -
Mr. Donald Lindgren. Westinghouse Mr. Donald Lindgren, Westinghouse, explained that Westinghouse requested two exemptions from the general design criteria (GDC) for the AP600 design.
One exemption was for one offsite power circuit instead of two. The other exemption was to GDC-19 for measuring the main control room operator dose limit in units of total effective dose equivalent (TEDE) instead of REM.
Mr. Lindgren presented the different safety and equipment classifications. 'He
/ described the design considerations for wind and tornado loads, flooding, missile protection, and pipe rupture.
He explained the basis for the original feedwater line leak-before-break (LBB) design, which was unacceptable to the i
staff. Mr. Lindgren present design features that were different than the design features for operating plants. He noted that documenting seismic and environmental qualification tests is the responsibility of the COL applicant.
The Subcommittee members. Mr. Lindgren, and the NRC staff discussed:
design of the containment baffle.
design of the switchgear floor drains.
design features that minimize water hammer.
advantage of LBB design for the feedwater lines.
effects of high burnup fuel on the reactor design, and whether GDC 19 will be amended.
LEAK BEFORE BREAK PIPING DESIGN - Mr. Shou-Nien Hou NRR 4
Mr. Shou-Nien Hou. NRR. listed the requirements against which the staff reviewed the feedwater line LBB design and described the scope of the review.
He stated the LBB design met all review criteria with the exception of withstanding water hammer. Mr. Alan Levin, NRR, stated that the initial L
-conditions used to calculate the potential stress on the pipe caused by a water hammer were not based on any specific scenario nor was the probability l
l of the initial conditions considered in the staff assessment. Mr. Hou l
presented four examples where water hammer events resulted in pipe ruptures.
r Advanced. Reactor Designs Subcomittee 7
May 13-15, 1998 However, none of the ruptures cited by the staff were documented in licensee event reports.
He explained that due to the uncertainty in assessing the frequency and magnitude of water hammer events, the staff could not approve the feedwater line LBB design.
The Subcommittee members and the NRC staff discussed the low probability of water hammer events, the water ~ hammer report written by Dr. Griffith of MIT, water hammer loads postulated by the staff, and the possible scenarios leading to the calculated loads.
DESIGN OF SEISHIC CATEGORY I STRUCTURES - Mr. Richard Orr, Westinghouse Mr. Richard Orr, Westinghouse, presented the design soil profile, nuclear island seismic analyses, and the design of category I structures, He provided details of the design of the nuclear island basemat, containment vessel, and the containment internal structures.
The Subcommittee members, Mr. Orr, and j
the NRC staff discussed:
the seismic design of the water tank on top of the containment, amount of engineering judgement used in performing hand calculations, e
l type of reinforcement used in the concrete structures, failure modes of the containment vessel, e
use of cathodic protection.
methods for filtering and venting the containment, and l
seismic qualification of cranes.
a AP600 NUCLEAR ISLAND BASEMAT DESIGN - Mr. Thomas Cheng, NRR Mr. Thomas Cheng, NRR, explained that the containment shell design meets the General Design Criteria and the requirements of ASME code Section III.
He described how the shell is embedded in concrete, covered with an inorganic zinc coating, and protected by the shield building.
Mr. Cheng stated that the l
shell would be inspected in accordance with ASME code section XI to ensure l
that its integrity is maintained.
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L Mr. Cheng stated that as a result of the NRC staff review Westinghouse increased the' steel reinforcement of the basemat and added shear reinforcement stirrups to the embedded exterior walls. The staff noted that Westinghouse and the staff disagreed on the service level C limits for the containment and nuclear island basemat.
However, the design meets the more conservative service level C limits used by the staff.
Mr. Cheng concluded that the final AP600 basemat design is acceptable.
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I Advanced. Reactor Designs Subcomittee 8
May 13-15, 1998 The Subcommittee members and Mr. Cheng discussed:
the reason for the thickness of the shell:
expected corrosion rates, inservice inspections, and repair techniques:
details of water seal between the shell and concrete: and the qualification of the paint used on the containment shell.
SSAR CHAPTER 16: TECHNICAL SPECIFICATIONS - Mr. Terry Schulz. Westinghouse:
Ms. Angela Chu NRR
'Mr. Terry Schulz, Westinghouse, explained that the AP600 technical specifications are based on the improved Westinghouse Standard Technical Specifications (STS). but that-there are differences due to the AP600 passive features and extensive use of microprocessor based instrumentation and control systems. He noted that the AP600 technical specifications were expanded to cover events occurring during low-power and shutdown conditions. Mr. Schulz detailed the differences between the AP600 technical specifications and the improved Westinghouse STS.
Ms. Angela Chu NRR. stated that the difference between the AP600 technical specifications and the improved Westinghouse STS were based on design differences. She identified three confirmatory items related to containment requirements for irradiated fuel handling, isolation functions of valves in the steam supply system, and correction of general inconsistencies.
Ms. Chu noted that there were no open items related to the AP600 technical
' specifications (TS).
The Subcommittee members. Westinghouse, and NRC staff discussed:
treatment of non-safety systems.
technical specification requirement for incore instrumentation.
short term availability items.
assessment of the safety significance of equipment both for at power and shutdown modes of operation.
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extension of AP600 TS shutdown requirements to operating plants.
developing limiting conditions for operation for passive systems, and surveillance requirements for instrumentation and control systems.
SSAR_ CHAPTER 17: QUALITY ASSURANCE - Messrs. Brian McIntyre and Terry Schulz, Westinghouse: Messrs. Robert Pettis, Jr. and Robert Gramm. NRR Mr. Brian McIntyre. Westinghouse, explained that the COL applicant will be responsible for developing quality assurance programs for the design.
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Advanced Reactor Designs Subcomittee 9
May 13-15, 1998 construction, and operating phases of the AP600. He noted that the quality I
assurance programs will be developed in accordance with quality plans described in Westinghouse management system documents and topical reports.
Mr. McIntyre described how Westinghouse assured the overall quality of the f
AP600 design by evaluations, inspections, and audits of quality assurance
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programs at external organizations, which worked on the AP600 design.
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noted that the NRC staff inspected the Westinghouse quality assurance program and identified nonconformances and unresolved items. He concluded that the NRC findings had been resolved and that the Westinghouse quality assurance program meets NRC regulatory requirements.
Mr. Robert Pettis Jr.. NRR. presented details of the advanced FSER open item related to the implementation of the Westinghouse quality assurance program.
He described the NRC inspections findings and Westinghouse's corrective actions.
Mr. Pettis concluded that all NRC inspection-related issues had oeen satisfactorily addressed by Westinghouse. Mr. Robert Gramm. NRR. stated that j
the FSER open item would be closed.
Mr. Schulz presented information related to the design reliability assurance program, which assesses the risk-significant structures.. systems, and components (SSC).
The Subcommittee members. Westinghouse, and the NRC staff discussed:
how the safety-significant SSC were identified.
evaluating multiple system lineups in shutdown conditions.
establishing and measuring minimum reliability criteria for SSC.
common cause failures.
nonsafety-related SSC. and acceptance of changes that resulted in small increases in risk.
FSER CHAPTER 22: REGULATORY TREATMENT OF NON SAFETY SYSTEMS (RTNSS)
Mr. Terry Schulz, Westinghouse: Mr. Alan Levin. NRR Mr. Terry Schulz, Westinghouse, explained how Westinghouse assessed the l
importance of non-safety systems and proposed regulatory oversight for l
important non-safety systems.
He presented the results of the probabilistic l
and deterministic criteria used in the assessment. Mr. Schulz described how SSC were used to compensate for uncertainties in the probabilistic analyses.
l He explained the investment protection short-term availability controls for non-safety systems in the technical specifications.
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Advanced, Reactor Designs Subcomittee 10-May 13-15, 1998 Mr. Alan Levin. NRR. stated'that the RTNSS process is defined in SECY-94-084.
He provided details of the staff's review of RTNSS the AP600 focused PRA, and the uncertainty analyses of. thermal hydraulic models. Mr. Levin presented a list of systems that will-have RTNSS controls.
He concluded that the identified systems' and the format of RTNSS controls provide additional assurance of plant safety and address staff concerns.
The Subcommittee members. Westinghouse, and the NRC staff discussed:
use of the baseline PRA results.
containment performance.
systematic process for adverse interactions.
consideration of defense-in-depth.
- examples of thermal hydraulic uncertainties.
missile protection for diesel generator radiator, and.
significance of required actions for short-term availability controls.
i CLOSURE OF ACRS CONCERNS - Mr. Brian McIntyre. Westinghouse The Subcommittee members agreed that the following ACRS questions were answered during the Westinghouse presentations:
how.the AP600 design improved the in-service test program, f
description of vacuum breakers on the in-containment refueling water storage tank (IRWST).
freezing of the drain and/or supply lines associated with the containment passive cooling system.
instrument air system initial test program is more extensive than for previous designs.
potential for creating a vacuum in the containment.
l effects of partially filling the containment cavity on containment i
l failure probability, l
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check valve testing program.
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why hydrogen levels are displayed in the main control room but are not alarmed.-
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Adv'anced Reactor Designs Subcomittee 11 May 13-15, 1998 analysis of steam generator tube rupture during severe accident temperature and pressure conditions, and 1
are containment hydrogen level indicators diverse actuation devices.
SUBCOMMITTEE COMMENTS AND RECOMMENDATIONS Dr. Chang Chen. ACRS consultant, concluded that based on his review of the
. SSAR and advanced FSER, the AP600 seismic design of the nuclear island basemat is in compliance with nuclear industry practice and NRC regulations.
Dr. Kress. ACRS member, stated that the RTNSS process is probably the best example of a risk-informed and performance-based regulatory process that he has seen.
-STAFFANDINDUSTFlyCOMMITMENTS The Subcommittee members requested and received the following documents.
F.B. Cheung and Y.'C. Liu. Pennsylvania State University. " External j
Vessel Cooling in a Flooded Cavity with the Effects of Thermal Insulations." presented at CSARP Meeting. Bethesda, Maryland. May 4-7, 1998.
Electric Power Research Institute report " Effects of Inhibitors and Poisons on the Performance of Passive Autocatalytic Recombiners (PARS) for Combustible Gas Control in ALWRs." dated May 22, 1997.
Westinghouse Electric Corporation WCAP-14477, revision 2. "The AP600 Adverse System Interactions Evaluation Report." November 1997.
Westinghouse agreed to provide Dr. Miller the response to an NRC request for additional information related to the type of instruments used for the core makeup tank level indication.
SUBCOMMITTEE DECISIONS i
The Subcommittee requested that the staff and Westinghouse present a summary i
I of. the.information provided during this meeting at the June 3,1998 ACRS meeting. The Subcommittee recommended that the Committee prepare an interim letter on the status of its review of the AP600 design.
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Advanced Reactor Designs Subcomittee 12 May 13-15. 1998 FOLLOW UP ACTIONS
. Subcommittee members identified the following items to be considered at future
. Subcommittee meetings:
the Westinghouse process for assessing adverse system interactions, and how the AP600 design minimizes the possibility of repetition of water
~ hammer operating events.
PRESENTATION SLIDES AND HANDOUTS PROVIDED DURING THE MEETING The presentation slides and handouts used during the meeting are available in the ACRS office files or as attachments to the transcript.
BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE 1.
' Westinghouse Electric Corporation AP600 Probabilistic Risk Assessment updated through revision 11 issued March 1998.
2.
Westinghouse Electric Corporation AP600 Standard Safety Analysis Report revision 22 issued on April 6, 1998.
~3, U.S. Nuclear Regulatory Commission Advance Final Safety Evaluation Report on the AP600 issued on May 2. 1998.
4.
Report dated April 29, 1998, from Richard Sherry, Senior ACRS Fellow, to ACRS-Members
Subject:
Initial Results of Review of AP600 PRA.
.[Predecisional].
5.
Report dated April 30, 1998,. from Richard Sherry. Senior ACRS Fellow, to ACRS Members,
Subject:
AP600 Explosive Squib Valves Failure Probability. [Predecisional]
6.
Memorandum dated March 11,1998,.from Brian Sharon, Chairperson, NRR DPV Review Panel, to Samuel Collins Director. NRR,
Subject:
Differing Professional. View Concerning Westinghouse AP600 Fire Pump Installation within the Turbine' Building,
[Provided for internal ACRS use only]
7.
Memorandum dated May 10, 1998, from Robert J.
Budnitz, Future Resources Associated, Inc., to Noel Dudley, ACRS,
Subject:
Review of the l
External-Events Parts of the AP600 PRA.
[Predecisional]
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. Memorandum dated April 7, 1998, from Ivan Catton.-ACRS Consultant, to John Barton, Chairman, ACRS Subcommittee on Advanced Reactor Designs.
Subject:
AP600' Advanced Reactor Design Review 31. March 1998.
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[Predecisional]
9.
Memorandum dated May 10, 1998, from Ivan Catton, ACRS Consultant, to t
Noel Dudley and Paul Boehnert, ACRS..
Subject:
Computational Tools for DNBR, THINC-IV and LOFTRAN.
[Predecisional]'
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Advanced. Reactor' Designs Subcomittee-13 May 13-15, 1998
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l NOTEi' Additional details of this meeting can be obtained from a transcript available'in the NRC Public Document Room, 2120 L Street, N.W.,
Washington, D.C. 20006..(202) 634-3274, or can be purchased from Ann Riley & Associates, LTD.. 1250 I Street,'N.W., Suite 300. Washington, D.C. 20005, (202) 842-0034.
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