ML20203D452
| ML20203D452 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 02/19/1998 |
| From: | Seale R Advisory Committee on Reactor Safeguards |
| To: | Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| References | |
| ACRS-R-1744, FACA, NUDOCS 9802260064 | |
| Download: ML20203D452 (8) | |
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UNITED STATES 8
NUCLEAR REEULATORY COMMISSION ACRSR-1744 o
ADVISORY COMMITTEE ON REACTCH SAFE ^UARDS pgg 0,,
CA$HINGTON. C. C. 20S55 February 19, 1998 72 0 Mr.
L.
Joseph Callan Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C.
20555-0001
Dear Mr. Callan:
SUBJECT:
INTERIM LETTER ON THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 P' \\NT DESIGN Ouring the 448th meeting of the Advisory Comittee on Reactor Safeguards, February S 7,1998, we reviewed the AP600 test and analysis program and various chapters of the AP600 Standard Safety Analysis Report.
Our Subcomittees on Advanced Reactor Designs and on Thermal Hydraulic and Severe Accident Phenomena have reviewed these matters previously, as listed in Attachment 1.
During these reviews, we had the benefit of discussions with representatives of the NRC staff and the Westinghouse Electric Company (Westinghouse) and of the occuments referenced.
TEST AND ANALYSIS PROGRAM The central goals of the Westinghouse Test and Analysis Program (TAP) are to confirm the design basis for the nuclear power plant components and systems unique to the AP600 design and to provide test data to support validation of relevant plant system codes. Westinghouse has concluded its testing programs, and its current focus is on the verification of the pertinent analytical tools.
Our Thermal Hydraulic and Severe Accident Phenomena Subcomittee began its review of the Westinghouse TAP in December 1991 and several meetings of the Subcomittee have been held in the interim.
'The Subcomittee last met to review the status
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of the key elements of the Westinghouse TAP on December 9-12, 1997.
l Over the course of these reviews, the Thermal Hydraulic and Severe Accident 1
Phenomena Subcomittee has raised a number of issues that have been documented j
only in Subcomittee minutes and transcripts, Subcomittee Chairman's reports, and ACRS consultants' reports.
In the interest of documenting these issues in a single report, a listing is provided below.
We recomend that Westinghouse 98o2260064 98o219 l
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o Mr. L., Joseph Callan 4 provide responses on these issues to the NRC staff for review. Those issues that we consider to be of higher priority are marked with an asterisk.
ReactorCon.lgtt System issues:
The basis for not including the momentum fluxes in the NOTRUMP code, particularly during the blowdown phases of the accident analyses Explanation of the coplicability of Equation 3-63 of Reference 3 to the critical flow of a single component two phase fluid Validation basis for the drift-flux modeling of horizontal flow Explanation of why the bloadown flows out of the automatic depressurization system (ADS) valves 1. 2. and 3 and out of the break itself are not predicteo well by the NOTRUMP code and what will be done to assure a conservative prediction of AP600 behavior (we are particularly concerned about using modeling deficiencies as compensating ef fects)
A more complete demonstration that the proposed penalty on fluid level in the in-containment refueling water storage tank (IRWST) provides sufficient conservatism to offset the uncertainties in the calculated pressurizer level holdup and resulting minimum core level The basis for validation for the liquid entrainment model used for the ADS-4 line Justification for the absence of (or completion of) a ' multi-loop" scaling analysis during the IRW3T cooling phase when the vessel inventory approaches a minimum Description of the pressurizer flooding model and its validation basis (treatment of the surge line from the hot leg to the pressurizer)
Explanation of how upstream flow effects were trected in reducing the data in the ADS separate effects tests and in the NOTRUMP code The basis for the inconsistencies between the NOTRUMP code noding used for the integral system test configurations and that used for the AP600 plant model j
Mr. L Joseph Callan Containment Issues:
Justification for the use of an incorrect expression in the rate-of-pressure-change equation (Equation 34 in Reference 6)
Justification for the inappropriate cancellation of the partial derivative of internal energy at constant pressure by the partial derivative of internal energy at constant volume to arrive at Equation 34 of Reference 6
Re-evaluation of the derivation and quantification of the scaling p1 groups resulting from a correction of Equation 34 of Reference 6 Justification for using the WGOTHIC lumped parameter model well-mixed assumption for calculating the AP600 containment behavior Justification for the use of steady-state testing in the Passive Containment System Large Scale Test facility to validate transient heat transfer correlations in the WG0THIC code Justification for the normalization of the rate-of-pressure-change term in Equation 34 in Reference 6 Technical basis for the treatment of the cooled containment boundary laminar sublayer in the WG0THIC code Validation basis for assuming a low elevation for the main steam line break Justification that the calculated peak containment pressure has appropriate margin in view of the observation that all three of the containment cooling system mechanisms (i.e.. the passive cooling water system, heat transfer to the containment shell, and heat transfer to the internal structures) are required to turn the pressure over just as it reaches the design value Quantification of the impact of incorrect (with respect to AP600) relative magnitudes of energy and mass addition and energy removal during the Large Scale Tests on the usefulness of the data for WGOTHIC code validation for use on AP600
.. =_..
Mr. L Joseph Cahn -...
In addition to the above, we are disturbed by the poor status of documentation related to information needed to certify the AD600 design. We believe that any certification should be contingent upon documentation of sufficient quality to ~
provide a traceable and well-archived licensing basis.
SAFETY ANALYSIS REPORT Our Advanced Reactor Designs Subcommittee began its review of the AP600 design t
- in January 1995. Since then, we have issued two reports to the Commission: one report concerned policy and key technical issues, and the second supported the requirement for a containment spray system.
We have reviewed the following Standard Safety Analysis Report chapters and have no comments at this time:
Chapter 1 - Introduction Chapter 4 - Reactor Chapter 5 - Reactor Coolant and Connected Systems Chapter 7 - Instrumentation and Controls Chapter 8 - Electrical Power Chapter 11 - Radioactive Waste Management Chapter 13 - Plant Operations' (excluding security)
Chapter 18 - Human Factors Engineering
SUMMARY
We have identified a number of issues associated with the Westinghouse Test and Analysis Program that should be resolved during the staff review. Our assessment of the adequacy of the Standard Safety Analysis Report chapters discussed to date is ir, complete. Completion of our review is contingent on the timely receipt of draft Final Safety Evaluation Report chapters.
Sincerely, Robert L. Seale Chairman
4 Mr. L. Joseph Callan
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References:
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-F-A "NOTRUMP - A Nodal Transient Small Break and General Netwcrk Code,"
August 1985 (Proprietary).
. 5.
Westinghouse Electric Corporation. WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code."
August 1985 i
- (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.
"WG0THIC 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 WG0THIC 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).
Attachment:
1.
Chronology of the ACRS Review of the Westinghouse Application for AP600 Standard Design Certification 1
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ATTACHMENT 1 CHRONOLOGY OF THE ACRS REVIEW 0F THE WESTINGH0lJSE APPLICATION FOR AF500 STANDARD DESIGN CERTIFICATION SUBCOMMITTEE DAIL SUBJECT Thermal Hydraulic Pheromena Review of Proposed Comission Paper on Need for 12/17/91 Full-Height. Full-Pressure Integral System Testing of AP600 Design Thermal Hydraulic Phenomena Continue Review of Integral System Testing 3/3/92 Requirements for AP600 Passive Plant Design Thermal Hydraulic Phenomena Continue Review of Integral System Testing 6/23-24/92 Requirements for AP600 Passive Plant Design Thermal Hydraulic Phenomms Continue Review of Westinghouse Test and 7/22-23/93 Analysis Program for AP600 Design Thermal Hydraulic Phenomena Continue Review of Westinghouse Test and 9/21/93 Analysis Program - Oregon State University Test Facility Thermal Hydraulic Phenomena Continue Review of Westinghouse Test and 3/15 16/94 Analysis Program - Core Makeup Tank Test Facility W Standard Plants Designs Overview Chap, 1: Introduction and 1/11/95 General Description of Plant Thermal Hydraulic Phenomena Review of COBRA / TRAC codes for W AP600 2/15-16/95 Thermal Hydraulic Phenomena Review test and analysis programs for AP600 3/29-30/95 Passive Containment Cooling System W Standard Plant Designs Review Staff Commission Paper on Status of Ten 5/31/95 Key Technical and Policy issues
e Thermal Hydraulic Pnenomena Staff review of Qualification Documentation for 7/26 27/95 the W COBRA / TRAC code Thermal Hydraulic Phanomena Staf f review of Qualification Documentation for 1/18 19/96 the W COBRA / TRAC code W Standard Plant Designs SECY-96 128. " Policy and Key Technical Issues 7/19/96 Pertaining to the AP600 Design" 433rd ACRS Meeting SECY 96128, " Policy and Key Technical Issues" 8/8/96 ACRS Report Issued 8/15/96 W Standard Plant Designs Chap. 4: Reactor 12/4/96 Chap. 5: Reactor Coolant System and Connected Systems Chap. 9: Auxiliary Systems Chap. 11: Radioactive Waste Managemeat Thermal Hydraulic Phenomena Scaling and PIRT Closure Report 12/18-19/96 Thermal Hydraulic Phenomena Test and Analysis Program: Long-Term Core 3/28/97 Cooling With W COBRA / TRAC Code 442nd ACRS Meeting AP600 Containment Spray Design 6/13/97 ACRS Report issued 6/17/97 Thermal Hydraulic Phenomena NOTRUMP Small-Break LOCA Code 7/29-30/97 Thermal Hydraulic Phenomena Passive Containment Systen Test and Analysis 9/29 30/97 Program Thermal Hydraulic Phenomena PIRT: Scaling of RCS: NOTRUMP Small Break LOCA 12/9-10/97 Code Thermal Hydraulic Phenomena WG0THIC Containment System Code 12/11-12/97 Advanced Reactor Designs Chap.
1: Introduction 2/4/98 Chap.
4: Reactor
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1 Chap. 5: Reactor Coolant System and Connected Systems Chap. 7:_ Jnstrtmentation and Controls Chap. 8: Electrical Power Chap.11: Radioactive Waste Management-Chap.13: Conduct _of 0perations l
Chap.18: Human Factors Engineeric.g 448th ACRS Meeting TAP and SSAR Chapters 1, 4, 5, 7, 8, 11, 13, and 18 Interim Letter issued February 19, 1998 g
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