ML20206S265
| ML20206S265 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 02/23/1998 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-3092, NUDOCS 9905210083 | |
| Download: ML20206S265 (10) | |
Text
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February 23, 1998 d
CERTIFIED:
March 2. 1998 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE MEETING MINUTES ADVANCED REACTOR DESIGNS FEBRUARY 3-4. 1998 ROCKVILLE MARYLAND The ACRS Subcommittee on Advanced Reactor Designs met on February 3-4. 1998, at 11545 Rockville Pike. Rockville Maryland, in Room T-2B3.
The purpose of the meeting was to review chapters 1. 4. 5. 7. 8, 11. 13. and 18 of the AP600 Standard Safety Analysis Report (SSAR) and the AP600 test and analysis program. The entire meeting was open to public attendance. Mr. Noel Dudley was the cognizant ACRS staff engineer for this meeting. The meeting was convened at 1:20 p.m. on February 3. 1998, and adjourned at 5:00 p.m. on February 4, 1998.
ATTENDEES ACffi J.
Barton. Chairman R.
Seale. Member G.
Apostolakis. Member R.
Uhrig. Member T.
Kress. Member G.
Wallis. Member D.
Miller. Member J.
Carroll, Consultant D.
Powers. Member NRC STAFF T.
Kenyon NRR J.
Wilson, NRR W.
Huffman. NRR R.
Landry. NRR A.
Levin. NRR WESTINGHOUSE ELECTRIC CORPORATION B.
McIntyre M.
Corletti TD R.
Vijuk T.
Hayes f]
,T S.
Kerch K.
Deutsch gl0g ?k 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.
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' Advanced Reactor Designs Subcommittee 2
February 3-4, 1998 SUBCOMMITTEE CHAIRMAN INTRODUCTION Mr. John Barton, Subcommittee Chairman, convened the meeting at 1:20 p.m. and noted that the ACRS had agreed to expedite its review of the AP600.
He explained that for the ACRS to meet the expedited schedule, the NRC staff must provide a complete and high-quality final safety evaluation report (FSER) with no safety-significant open items.
Mr. Barton pointed out that draft FSER chapters. were not available for the SSAR chapters that would be discussed during this meeting.
Mr. Barton read a prepared statement by Dr. Robert Seale ACRS Chairman. The statement noted that the ACRS had a statutory requirement for reviewing the AP600 design application.
Unlike previous designs, the AP600 design relies on the adequacy of the basic natural processes to drive passive heat removal systems without operator intervention. The statement concluded that the focus of the ACRS review had evolved but that the bar had not been raised with respect to the acceptance criteria as had been suggested by Westinghouse.
WESTINGHOUSE INTRODUCTION: Mr. Brian McIntyre, Westinghouse Electric Company Mr. McIntyre presented the background and status of the AP600 design review process. He noted that there were over 400 open items on which the staff and Westinghouse had not reached resolution.
He stated that the test and analysis program was completed and that documentation had been submitted. Mr. McIntyre expected that the open technical issues related to the qualification of containment coatings, the fire protection analysis, and the inspection, tests, analyses, and acceptance criteria (ITAAC) would be resolved.
STAFF INTRODUCTION:
Mr. Thomas Kenyon, NRR Mr. Kenyon presented the background and status of the staff's review of the AP600 design. The Subcommittee Members and the staff discussed the quality and completeness of Westinghouse documentation, the criteria the staff uses to assess the adequacy of documents, and the availability of the Westinghouse security design report.
WESTINGHOUSE PRESENTATION SSAR Chapter 1 - Introduction and General Description of Plant:
Mr. Ron Vijuk, Westinghouse Mr. Vijuk presented the key design features of the AP600 and described the passive safety systems, non-safety systems, and the layout of plant equipment.
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February 3-4. 1998 The Sybcommittee Members and Mr. Vijuk discussed the reason for the extensive reliance on electronic work stations, possible human actions that could prevent the operation of passive systems, deviations from the Electric Power Research Institute Utility Requirements Document (URD). and why diesel J
generators are non-safety related. The Subcommittee noted that the severe accident mitigation design alternatives (SAMDA) developed by Westinghouse use
$1,000 per person rem instead of the $2.000 per person rem, which has been adopted as NRC policy.
SSAR Chapter 4 Reactor Core and Fuel Design: Mr. Ron Vijuk. Westinghouse Mr. Vijuk presented a comparison of core parameters between the AP600 and a typical 2-loop Westinghouse plant. He described the core layout, fuel assembly design, fuel rod design, and the reactivity control features.
He explained the nuclear and thermal hydraulic design. The Subcommittee Members and Mr. Vijuk discussed the number of control rod drives, the type of material used in grey rods, and the reactivity of grey and black rods. They also discussed issues related to high burnup fuel such as gas release at high temperatures, cladding integrity, and consideration of power insertion ramps.
SSAR Chapter 5 Reactor Coolant System and Connected Systems:
Mr. Ron Vijuk. Westinghouse Mr. Vijuk described the layout of the reactor coolant system (RCS), the operation of the automatic depressuization system, and the design of the normal residual heat removal system. He explained the design enhancements that are intended to prevent inter-system loss of coolant accidents, mid-loop events. and shutdown incidents.
I In response to previous ACRS questions. Mr. Vijuk explained the reactor coolant leak detection system design and the justification for the acceptability of a safe shutdown condition of 420*F or below.
The Subcommittee Members and representatives of Westinghouse discussed:
the reason for eliminating loop seals.
motor operated valves repositioning against high differential pressures.
design of pump seals for normal operating RCS pressures.
mid-loop operations.
locatien of pressurizer level indicator taps.
the leak-before-break design of carbon steel pipes, and disadvantages of the RCS pump design.
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Advanced Reactor Designs Subcomittee -
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SSAR Chapter 11 - Radioactive Waste Management: Mr. Ron Vijuk, Westinghouse Mr. Vijuk presented the two source term bases Westinghouse used in evaluating the AP600 design of radioactive wastes management systems.
He explained the technical specification requirements associated with primary coolant activity.
Mr. Vijuk described the liquid, gaseous, and solid radioactive waste system designs and the associated safety-related radiation monitoring systems. The Subcommittee Members and the Mr. Vijuk discussed the hydrogen monitors in the containment, RCS iodine _ spiking, and the reason for using mobile systems for chemical and detergent wastes.
SSAR Chapter 8 Electrical Power: Mr. Tom Hayes, Westinghouse Mr. Hayes explained that the electrical power systems were designed to criteria for three different tiers of loads (i.e., normal unit operating loads, nonsafety-related loads required at all times, and safety-related loads). He presented the offsite power system and t M onsite AC and DC power systems designs. Mr, Hayes described the 25 kw ancillary diesel generators that are designed to provide power to safety-related loads 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after an
- accident, He Hayes explained the technical specification requirements for the safety-related and nonsafety-related electrical systems.
The Subcommittee Members Mr. Hayes, and the staff discussed:
design provisions _ for a second'offsite power supply.
grid reliability concerns, exemptions to General Design Criterion 17, " Electrical power systems,"
e missile protection for the diesel generator radiators, i
actions that could prevent automatic loading of diesel generators.
the function of safety-related batteries.
main control room habitability, qualification of the ancillary diesel-generator, hookiilg up loads to the ancillary diesel generator.
lightning and grounding protection.
sizing of electrical cabling for motor operated valves, and electromagnetic and radio-frequency interference (EMI/RFI) generated by l
inverters.
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Advanced Reactor Designs Subcommittee 5
February 3-4, 1998 SSAR Ch, apter 13 - Conduct of Operations: Mr. Steven Kerch Westinghouse Mr. Kerch presented the design certification requirements for organizational structure, training, emergency planning, operational review, and plant procedures. The Subcommittee Members and Mr. Kerch discussed the following:
applicability of the Institute of Nuclear Power Operations (INPO) requirements for training programs, differences between WCAP-14655, " Designer's Input for the Training of the Human Factors Engineering Verification and Validation Personnel."
and training programs at operating plants, inspections, tests, analyses, and acceptance criteria (ITAAC) process, and similarities between the AP600 design and previously approved evolutionary reactor designs.
SSAR Chapter 18 - Human Factors Engineering: Mr. Steven Kerch, Westinghouse Mr. Kerch explained that Westinghouse used NUREG-0711, " Human Factors Engineering Program Review Model," to design the AP600 Human Factors Engineering (HFE) Program.
He presented the program review model and elaborated on the ten program elements. The Subcommittee Members and Mr.
Kerch discussed:
updating of NUREG-0711, compliance of the HFE Program with the Utility Requirements Document, a
staffing requirements for the AP600 human factors organization, e
utilization of shift operators, e
the database for activities in the main control room during accidents, integration of human reliability assessment activities in the HFE
- Program, compliance with NUREG-0700, Revision 1 guidance for digital equipment.
e approval process for rapidly changing technologies, justification for designs that use digital software and PC terminals, use of symptom based emergency operating procedures (EOPs),
possible use of computer based E0Ps, and procedural requirements associated with containment hydrogen monitors.
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Advanced Reactor. Designs Subcomittee 6
February 3-4, 1998 SSAR Chapter 7 - Instrumentation and Control: Mr. Ken Deutsch, Westinghouse Dr. Miller introduced the Subcommittee review of the instrumentation and control (I&C) design by noting his concerns related to the design of the nuclear instrumentation instruments. He stated that the use of fission chambers is always superior to the use of boron-type instruments. He also commented that the wide use of fiber optics throughout the I&C system was a significant advancement in the design.
Mr. Deutsch, Westinghouse, stated that the use of digital, microprocessor-based technology interconnected by multiplexed communications provides significant benefits over analog control technology.
He explained the I&C system architecture and described the monitor bus, the protection and safety monitoring system, the reactor trip switchgear, and the diverse actuation system. Mr. Deutsch presented the design of the incore and excore nuclear instrumentation, as well as the plant control system architecture.
The Subcommittee Members, Mr. Deutsch, and the staff discussed:
the disadvantages of digital equipment:
qualification of the digital systems for smoke:
the fiber optic cable qualification program:
disadvantages of decentralizing systems and tasks:
i implications of system reliability for PRAs:
regulatory guide for producing, verifying and validating software:
diverse use of hardwired I&C systems:
types of incore detectors; and
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concerns related to EMI/RFI effects on digital equipment.
AP600 Test and Analysis Program:.Mr. Brian McIntyre. Westinghouse Mr. McIntyre provided an overview of the test and analysis process and the
-test program objectives. He described the component design verification tests, the passive containment _ cooling system tests, and the passive core cooling system tests. Mr. McIntyre identified the computer codes used in the test and analysis program and explained how the codes were verified and validated. He presented the key open issues related to passive containment
- cooling. The key issues included the modeling of circulation and and stratification of the containment atmosphere, the revised scaling report, the adequacy of computer codes. Mr. McIntyre concluded that the AP600 testing program was adequate to support the design certification application.
a Advanced Reactor Designs Subcommittee 7
February 3-4. 1998 The Subcommittee Members, Mr. McIntyre, and the staff discussed steam condifions inside the containment during a severe accident, how adequate water distribution on the outside of the containment would be assured, and the variations between test data and the results of computer codes. They also
' discussed the process for staff review of the application for design certification.
STAFF PRESENTATION ON TEST AND ANALYSIS PROGRAM:
Mr. William Huffman, Jr.. NRR Mr. Huffman presented the staff's key open issues related to its review of the NOTRUMP and WGOTHIC codes.
He describes the staff's key issues related to the automatic depressurization system test program, scaling and quantifying the uncertainty of the Oregon State University test data, and the Phenomena Identification and Ranking Table. The Subcommittee Members indicated that they agreed with the staff's list of open items.
SUBCOMMITTEE COMMENTS. CONCERNS. AND RECOMMENDATIONS Dr Miller stated that the two significant differences between the operating reactors and AP600 I&C system designs are the use of digital microprocessor-based systems with multiplexed data links and the use of fiber optics for data links and signal isolation.
Dr. Miller recommends that-reference to neutron detector types in SSAR sections 7.1.2.8.2 and 4.4.6.3 be deleted.
He stated that in other sections of the SSAR where sensors and their performance requirements are discussed no sensor types are specified.
Dr. Miller suggests that if low level neutron detectors are specified, the type referred to as " pulse fission chamber" should be referred to as " fission chamber using the pulse and Campbelling (mean square voltage) modes of detector operations."
STAFF AND INDUSTRY COMMITMENTS Westinghouse agreed to discuss during its presentation of Chapter 15." Accident Analysis." the suitability and results of the reactivity code used to predict the effect of a grey rod ejection.
Westinghouse agreed to explain what database was used for determining the activities of operators in the main control room during plant events.
Westinghouse agreed to determine why containment hydrogen levels are displayed in the main control room but are not alarmed.
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Advanced Reactor Designs Subcomittee 8
February 3-4, 1998 Westinghouse agreed to determine whether the containment hydrogen level indicators are included as a diverse actuation device.
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Westinghouse agreed to provide the ACRS a copy of a revised passive containment cooling system scaling report that would address ACRS concerns.
l Westinghouse planned to provide the report in February 1998.
The staff agreed to provide the ACRS a copy of the AP600 security design l
' plan.
[ Received 3-3-98]
The staff agreed to provide the Subcomittee Members a letter listing the l
differences between the AP600 design and the URD. [ Received 2-4-98]
The staff agreed to provide the ACRS a copy of the Westinghouse report
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concerning circulation and stratification inside the containment.
[ Received 2-9-98].
SUBCOMMITTEE DECISIONS The Subcomittee requested that Westinghouse and the staff present a sumary of the information provided during this meeting at the March 5.1998 ACRS meeting. The Subcomittee recommended that the Comittee prepare an interim letter on the status of its review of the AP600 test and analysis program and SSAR chapters 1. 4. 5, 7. 8. 11. 13. and 18.
FOLLOW UP ACTIONS Subcomittee members identified the following items to be considered at future Subcomittee meetings:
Dr. Powers requested that Westinghouse explain how high burnup fuel was addressed in the AP600 design.
In particular. the explanation should address the lack of consideration of the power insertion ramp, the difference in reactivity between black and grey rods, and the effect on iodine spiking in the reactor coolant system.
Mr. Carroll requested that the staff explain why the use of a leak-before-break design criteria for feedwater piping was unacceptable.
Dr. Seale requested that Westinghouse discuss whether, due to the use digital I&C systems, there are too many operators required for the back i
shifts.
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a Advanced Reactor Designs Subcommittee 9
February 3-4. 1998 Dr. Powers and Mr. Carroll requested that Westinghouse explain how the
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effect of smoke on digital equipment was analyzed.
In particular, they requested information regarding which components were coated to reduce the ef fects of smoke and what tests, if any, where conducted in a smoky environment.
Mr. Carroll requested that Westinghouse provide additional information concerning the check valve testing program.
BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE 1.
Westinghouse Electric Corporation, "AP600 Standard Safety Analysis Report " updated through Revision 16 dated September 2, 1997.
2.
Letter dated January 16, 1998, from William Huffman, NRC, to Nicholas Liparulo, Westinghouse Electric Corporation,
Subject:
Open Items Associated with the AP600 Safety Evaluation Report on the AP600 Containment Design and Accident Analyses.
3.
Westinghouse Electric Corporation WCAP-14727. Revision 1, "AP600 Scaling and PIRT Closure Report," July 1997 (Proprietary).
4.
Westinghouse Electric Corporation. WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985 (Proprietary).
5.
Westinghouse Electric Corporation, WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (Proprietary).
6.
Westinghouse Electric Corporation, WCAP-14845, Revision 2: " Scaling Analysis for AP600 Containment Pressure During Design Basis Accidents,"
June 1997 (Proprietary).
7.
Westinghouse Electric Corporation, WCAP-14407. Revision 1. "WGOTHIC Application to AP600," July 1997 (Proprietary).
8.
Westinghouse Electric Corporation WCAP-14326, Revision 1. " Experimental Basis for the AP600 Containment Vessel Heat and Mass Transfer Correlations," May 1997 (Proprietary).
9.
Westinghouse Electric Corporation WCAP-14807, Revision 2. "NOTRUMP Final Validation Report for AP600," June 1997 (Proprietary).
10.
Westinghouse Electric Corporation, WCAP-14967. Revision 0, " Assessment of Effects of WGOTHIC Solver Upgrade From Version 1.2 to 4.1," September 1997 (Proprietary).
11.
Westinghouse Electric Corporation. WCAP-14135.
" Final Data Report for PCS Large-Scale Tests Phase 2 and Phase 3," July 1994 (Proprietary).
Advanced Reactor Designs Subcomittee 10 February 3-4, 1998 PRESEN,TATION 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.
NOTE: 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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