ML20035C345
| ML20035C345 | |
| Person / Time | |
|---|---|
| Issue date: | 03/30/1993 |
| From: | Donohew J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-679A NUDOCS 9304070090 | |
| Download: ML20035C345 (5) | |
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UNITED STATES P
NUCLEAR REGULATORY COMMISSIOM i
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WASHINGTON, D. C. 20555 e
t k... e,e8 March 30, 1993 Project No. 679 i
ORGANIZATION: Atomic Energy of Canada, Ltd., Technologies (AECLT)
SUBJECT:
MEETING WITH AECLT TO DISCUSS ACCIDENT ANALYSIS DOCUMENTATION ON THE CANADIAN DEUTERIUM URANIUM 3 (CANDU 3) REACTOR DESIGN On the afternoon of Thursday, February 11, 1993, a meeting was held between members of the Advanced Reactors Project Directorate (PDAR) of the Office of Nuclear Reactor Regulation (NRR) in the United States (U.S.) Nuclear i
Regulatory Commission (NRC) and representatives of AECL Technologies (AECLT).
Representatives of AECL and of the Atomic Energy Control Board (AECB) of Canada also attended. The AECLT represents AECL, the designer of CANDU 3 in Canada, within the U.S. and before the NRC, and the AECB is the federal government regulator for nuclear power reactors in Canada.
This meeting was requested by PDAR/NRC to discuss the accident analysis documentation on the CANDU 3 reactor design that has been submitted to the NRC and the additional information needed for NRC to continue its reviews of the CANDU 3 during this current preapplication review. This meeting follows (1) the letters of September 23, 1992, and January 15, 1993, to AECLT in which the NRC requested accident evaluation data and (2) the meeting of February 2,
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1993, in which, as part of the meeting, PDAR NRC requested the consequence f
analysis for a large-break loss-c?-coolant accident (LOCA) with a failure to shutdown. The summary for the February 2, 1993, meeting was issued on February 18, 1993.
This report is a summary of the important issues discussed and the decisions, if any, made in the meeting.
Enclosed is the list of individuals and their affiliation who attended the meeting. There were no materials handed out during the meeting.
The meeting started with a presentation by PDAR on (1) the accident analysis j
information that is given in the CANDU 3 Conceptual Safety Report (CSR) and (2) the scope and depth of the information needed by NRC, for the CANDU 3 design. The information in the CSR is essentially the result of three different analyses:
peak containment pressure from the steamline break, minimum emergency core cooling system (ECCS) flow for the large-break LOCA, t
and time needed to scram for the large-break LOCA. The discussion on these analyses does not give NRC the consequences occurring in the core during the specific event, the pressure, temperature, and fission product transport (if any) in the containment during the event, and the fission product release from the containment and done consequences to the public for the event.
PDAR stated that from the review of the CSR there appear to be several design-M i
%y. Md basis events that are being analyzed by AECL for the CANDU 3 design, but none ofthedetailoftheconse0uencesoftheseaccidentshasbeensubmittedto 02 110
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AECL Technologies March 30,1993 submitted to the NRC. This information is needed for the NRC to maintain the schedule for the CANDU 3 preapplication review.
PDAR stated that it needed to understand what events have completed consequence analyses for the CANDU 3 design and what events do not.
PDAR explained that the NRC focuses on peak clad temperature (2200 *F),
maximum clad oxidation (0.17 times clad thickness), maximum hydrogen generation (0.01 of total clad-water reaction), coolable core geometry, and long-term fuel cooling during the most severe design-basis event for the core which, for light-water reactors, is contained within 10 CFR 50.46(b) and Appendix K to Part 50. The details of this type of information on fuel and channel integrity during design-basis events have not been presented to the NRC.
AECLT stated that Technology Transfer Reports (TTR) -291, -409, and -276 provide information on fuel channel integrity, on ECCS performance, and on the LOCA analysis, and have been submitted to the NRC for the CANDU 3 design.
These three TTRs have been reviewed by the NRC and are discussed below:
TTR-409, "CANDU 3 and U.S. NRC Requirements, Equivalent Safety Issues: Emergency Core Cooling," dated July 1992, deals with the reliability of the ECC function and the ECCS criteria of 1200 *C maximum fuel clad temperature. The report has a brief summary-type discussion of the results of an evaluation of the CANDU 3 design against the NRC requirement on maximum fuel clad temperature in which AECL concludes that fuel cladding temperatures remain below 2200 *F for all postulated breaks in the inlet and outlet headers; however, this discussion does not present any details of whatever calculations were made or any consequences in the core for the postulated breaks. This TTR does not present any of the information needed by the NRC statf to review the CANDU 3 design against ECCS requirements. Also, there is no discussion to relate the break sizes mentioned in TTR-409 to the design-basis events discussed in the CSR.
TTR-291, "The Technology of CANDU Fuel Channels," dated January 1991, deals with the design, technology, and performance of generic CANDU fuel channels. Chapter 14 on consequences of pressure tube rupture provides a qualitative description of the event sequence for a pressure tube rupture accident, the safety acceptance criteria, the analysis methodology, and a summary. The summary states that following a postulated fuel channel failure (1) there is no systematic fuel failure predicted to occur in the remaining fuel channels and (2) the remaining fuel channels remain intact. There are no details on what the predicted fuel and channel temperatures are in the intact channels and no discussion of the fuel failure in the ruptured fuel channel.
TTR-276, "The Technology of CANDU Loss of Coolant Analysis," dated
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February 1991. The purpose of this document is to introduce readers
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AECL Technologies
.3-March 30, 1993 j
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familiar with the U.S. accident modelling to the' Canadian modeling.
methods and thus, provides an overview of Canadian accident analysis j
practice. This document does not present-any consequences of i
accidents calculated for the CANDU 3 design.
PDAR stated that these three TTRs'do not provide ~the detailed accident d
analysis information discussed above and needed to review the CANDU 3 design.
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AECLT also stated that the power pulse break survey' analysis was submitted in the letter of January 20, 1993, from AECL and provided accident-evaluation data for a break in the new inlet header design.
PDAR stated that this-i analysis has not been reviewed yet.-
r After further discussion, AECLT stated that it agreed that NRC did need more j
information for.its review. AECL and AECLT agreed to meet to determine what-additional accident information was available on the CANDU 3 design and what could be submitted to the NRC because of proprietary information consider-
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a*.i on s. AECLT stated that it would get back to NRC with this information.
This ended the meeting.
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n_ph s,s,
N. Donohew,, SeMar. Project Manager dvanced Reactors Project Directorate' Associate Directorate for Advanced Reactors-and License Renewal
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Office of Nuclear Reactor' Regulation 1
Enclosure.
j List of Attendees cc w/ enclosure:
i See next pageh LA ABA PM:l W :ADAR Distribution:
t PShe JNDonohew:sa Central File JLKennedy 93 y/g93 PDR JNDonohew 1
TEMurley WUpshaw SC:PDAR:ADAR D:PD ADAR FJMiraglia PShea
'THCox %
RCPi s'en DMCrutchfield ACRS (10)
--3/Lf/93
$ /4 93 RCPierson OGC
' l THCox EJordan EDThrom PDAR R/F i
RMeyer, RES CANDU R/F j
i 0FFICIAL' RECORD COPY Document Name: SUMRY2.11
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AECL Technologies March 30, 1993-CANDU Project No. 679 cc: Louis N. Rib, Licensing Consultant AECL Technologies 9210 Corporate Boulevard, Suite 410 Rockville, Maryland 20850 Bernie Ewing, Manager Studies and Codification Division Atomic Energy' Control Board P.O. Box 1046, Station B 270 Albert Street Ottawa, Ontario, Canada KIP 559 A.M. Mortada Aly, Senior Project Officer Advanced Projects Licensing Group Studies and Codification Division Atomic Energy Control Board P.O. Box 1046, Station B 270 Albert Street Ottawa, Ontario, Canada KlP SS9 Project Director - CANDU-3 AECL CANDU 2251 Speakman Drive Mississaugua, Ontario, Canada L5K IB2 L. Manning Muntzing Newman & Holtzinger, P.C.
1615 L Street, N.W., Suite 1000 Washington, DC '20036 Steve Goldberg, Budget Examiner Office of Management and Budget 725 17th Street, NW.
Washington, DC 20503 Mr. A.D. Hink Vice President / General Manager AECL Technologies 9210 Corporate Boulevard, Suite 410 Rockville, Maryland 20850
...,1 i.
ENCLOSURE MEETING BETWEEN PDAR/NRC AND'AECL TECHNOLOGIES TO DISCUSS ACCIDENT ANALYSIS DOCUMENTATION ON THE CANDU 3 DESIGN FEBRUARY 11. 1593 ATTD. DEES HAME AFFILIATION Janet Kennedy NRC/NRR/PDAR Robert Pierson NRC/NRR/PDAR Tom Cox NRC/NRR/PDAR Edward Throm NRC/NRR/PDA1.
Jack Donohew NRC/NRR/PDAR Louis Rib AECLT Michael fletcher AECLT Robert Ferguson AECLT A.C. David Wright AECL - CANDU Duane Pendergast AECL - CANDU A.M. Mortada Aly AECB John Tong AECB AECL - Atomic Energy of Canada Limited AECLT - AECL Technologies AECb - Atomic Energy Control Board PDAR - Advanced Reactor Project Directorate of NRR/NRC
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