ML20062J497
| ML20062J497 | |
| Person / Time | |
|---|---|
| Issue date: | 10/26/1993 |
| From: | Throm E Office of Nuclear Reactor Regulation |
| To: | Miraglia F, Murley T, Russell W NRC |
| References | |
| CON-NRC-03-93-032, CON-NRC-3-93-32, PROJECT-679A NUDOCS 9311080107 | |
| Download: ML20062J497 (8) | |
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October 26, 1993 Project No. 679 MEMORANDUM FOR:
T. Murley G. Lainas C. Rossi F. Miraglia J. Roe R. Zimmerman W. Russell J. Zwolinski B. Boger J. Partlow E. Adensam C. Thomas D. Crutchfield M. Virgilio F. Congel W. Travers B. Grimes E. Butcher F. Gillespie J. Wiggins W. Bateman, EDO S. Varga B. Liaw M. Slosson J. Calvo A.. Thadani Operations Center THRU:
Edward D. Throm, Acting Director Advanced Reactors Project Directorate Associate Diret'3 rate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation FROM:
Dino C. Scaletti, Sr. Project Manager Advanced Reactors Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
SUBJECT:
DAILY HIGHLIGHT - FORTHCOMING MEETING WIT;l AECL TECHNOLOGIES (AECLT)
DATE &
9t:
October 28, 1993 2 p.m. - 5 p.m.
LOCATION:
U.S. NRC Headquarters 11555 Rockville Pike, Room 1/F/19 Rockville, Maryland PURPOSE:
Discuss the request for additional information related to.
staff contract NRC-03-93-032, "CANDU 3 Containment Perforruance and Consequences" (see enclosure).
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' Multiple Addressees _0ctober 26, 1993 j
PARTICIPANTS:*
!!HB AECLT ELA R. Landry V. Snell F. Sciacca j
D. Scaletti L. Rib, et al.
Original signed by:
Dino C. Scaletti, Sr. Project Manager Advanced Reactors Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation l
Enclosure:
Items of Concern cc w/ enclosure:
See next page
- Meetings between NRC technical staff and applicants or licensees are open for interested members of the public, petitioners, intervenors, or other parties-to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy," 43 Federal Reaister 28058,06/28/78.
Distribution:
0 Central F11_e; 1
PDR PDAR R/F EDThrom DCScaletti OGC 15/B/18 OPA 2/G/5 ELJordan MNBB 3701 Receptionist OWFN RLandry WUpshaw ACRS (10)
P 315 NRR Mailroom, PMAS 12/G/18 PShea CANDU R/F PDAR Meeting File t A:PDAR: ADAR PM PDAR:ADAR PD:PDAR:ADAR PShea Q M DScalet EDThrom M IOb3
[0 fI/93
/J /2h93 OFPCIAl. RECORD COPY Document Name:_ CANMTG10.28
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Mus ciple Addressees October 26, 1993 CANDU Project No. 679 cc: Louis N. Rib, Licensing Consultant l
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 KlP SS9 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 5S9 Project Director - CANDU-3 AECL CANDU 2251 Speakman Drive Mississaugua, Ontario, Canada L5K 182 t
L. Hanning Muntzing Newman & Holtzinger, P.C.
i 1615 L Street, N.W., Suite 1000 Washington, D.C.
20036 Steve Goldberg, Budget Examiner Office of Management and Budget 725 17th Street, NW.
Washington, D.C.
20503 A.D. Hink Vice President / General Manager AECL Technologies 9210 Corporate Boulevard, Suite 410 i
Rockville, Maryland 20850 t
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1.0 INTRODUCTION
1
- The contract statement-of-work tasks SEA with reviewing CANDU 3 documents submitted to NRC by
.l AECL Technologies with regard to the major areas of the containment _ system design including
-l containment system and structural response to both design basis and severe accidents, source terms t
and release rates, and containment performance analysis methods and acceptance criteria. SEA is to
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identify the preapplicant's criteria in each area and evaluate the technical and experimental bases supponing this criteria with particular regard to areas where design, materials, or acceptance criteria differ from accepted U.S. practice. The NRC criteria are found in NUREG-0800,'" Standard Review Plan, Sections 3.8.1,6.1.2,6.2.1,6.2.2, and 62.4, through 6.2.7" (SRP). Particular attention is to be l
paid to issue, unique to the CANDU 3 design.
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In our initial review, SEA has identified in several areas where additional infonnation is needed to complete our task and has prepared a request for this information, as directed by the statement-of-j work. The following request consists of a discussion of the documentation initially transmitted to l
SEA and a summary of the apes of additional information needed, as well as lists of reports referenced within the initial documents which will at least panially fulfill those infonnation needs.
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2.0 STATUS OF INITIAL DOCUMENTATION This section is a brief discussion of the completeness and applicability of the initially received documents relative to the " areas of review" specified in the SRP.
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1 2.1 Prevailing Design The Technical Description and Conceptual Safety Report are both 1989 l
documents but the design has apparently continued to evolve since then. For example, the 1989 Technical Description shows 2 main reactor cooDnt circulation pumps, whereas, the 1992 Technical Outline shows 4. Our review based on the documents currently available for our review will not l
necessarily represent the current design.
i 2.2 Concrete Containment The current documents lack much of the detail needed to complete the review as outlined in the SRP. The containment building description consists of: a few sketches (one j
cross-section, two elevations, and an intemal structural module), the basic dimensions (height, width, l
and wall thickness), the design pressure, and qualitative discussions regarding materials, access, i
construction, etc. The SRP areas of review include such items as main reinforcement and prestressing l
tendons, the anchorage of the liner, loads, and loading combinations which cannot be reviewed with the current level of technical description. Information, such as containment ultimate failure pressure, j
subcompanment failure pressure differentials, and containment structural design details is missing.
l 2L3 Containment Source Terms and Containment Resnonse The safety analysis presented in the Conceptual Safety Repon (CSR) is described as design-assist safety analysis functioning primarily to provide assistance and input to the design and is incomplete for license application, as well as not
-l reflecting the latest design modifications. The CSR presents only minimal results for selected _
calculations such as a large LOCA, presented as a' bounding LOCA. The SRP " areas of review"'
include the containment pressure and temperature due to a spectmm of LOCAs (break size and location) for both the primary and secondary systems. The verification that the LOCA source tenns r
Encl osure -
'to the containment are bounding as stated in the CSR cannot be determined from the 1989 CSR.
Other important analytical details such as the containment temperature are not presented. Sensitivity
-and uncertainty analyses are not presented.
The CSR lacks detailed analytical results needed to evaluate the validity of the source terms to the containment and containment response, such as the containment analysis nodalization and basis for that nodalization, the impact on containment response of possible combustion (with or without igniters), containment response with or without air coolers, and the possible decomposition of protective coatings.
The Technical Description lacks detailed specification of the location of piping, valves, etc. within the containment, which piping penetrates the containment walls, and safety valve discharge location and I
capacity. During our initial review, for instance, we were unable as ascertain whether the main steam safety valves were located inside or outside the containment building or whether or not main steam lines had isolation valves.
The trip specification of a LOCA signal was not found. For instance, can a steam generator tube l
rupture signal a LOCA, thereby initiating secondary crash cooling to the atmosphere (assuming safety i
valves are located outside containment) of an non-isolated steam line (assuming these lines indeed do not have isolation valves), thereby, bypassing the containment?
i 2.4 Radionuclide Source Terms and Release Rates The soume term analysis document, TTR-384, contains analysis specific to CANDU 6. While this analysis will contribute to overall understanding of CANDU souwe term methodology, the results from the analysis in TTR-384 are not directly applicable to CANDU 3.
There are major differences between the CANDU 3 and CANDU 6 containment designs. Perhaps the most significant difference influencing the radionuclide soume term to the environment is that the CANDU 6 containment employs sprays for pressure suppression and radionuclide contml, whereas, the CANDU 3 design does not include sprays. 'Ihe TTR-384 analysis involved sprays.
2.5 Design Bases and Severe Accidents The analyses in the documents provided to SEA are all based on design basis accidents. TTR-429 (June,1992) states that information regarding containment response to severe core damage accidents will be pmvided to the NRC when this response is assessed.
2.6 Analysis Methods The SRP states that the applicant should use calculational methods that have been previously reviewed by the staff and found acceptable.
Most of the computer codes used by the Canadians are CANDU specific and probably have not been evaluated by the NRC. Much of the j-technical basis for such review areas as the LOCA mass and energy soume tenns to the containment l
and the containtnem rnponse to these sources is contained within the models of these Canadian codes.
and within the input and results of those codes. While a complete review of all of these codes and their associated experimental verification and validation to ensure conservative results is beyond the scope of this task, some preliminary understanding of these codes is necessary in order to make arm assessment of the containment soume tenns. Descriptions of the models employed by these codes are almost completely missing in the initial documents. Further, the Canadians apparently use several separate codes in perfonning an accident analysis, as opposed to a fully integrated systems code, and their methods of integrating the code results are not discussed.
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w M. A. Cormier, " Containment Node Link Model for the CANDU 3 Design", AR-74-68400 002.
- M. S. Quraishi, et al., "Assessmeret af Higher Leakage from Containment Following Postulated y
Accidents", TTR-168, Volume 2,1986 September.
+ J. M. Hopwood, R. S. Poner, S. Pang, B. A. Shalaby, S. D. Grant, E. Kohn, A. Lai, V, K.
Molindra, "Large LOCA Power Transient Assessment for CANDU 3", AR-7443500-016,1989 January.
- S. D. Grant, V. I. Nath, "A Study of Pressure Tube Heat-Up Following Postulated Large Breaks in an Inlet Header", AR-74-03500-022,1989 January.
M. A. Wright and M. S. Quraishi, " Analysis of the Consequences of an End Fitting Failure", AECL Repon TTR-153,1985 May 3.4 Radionuclide FalTf Tenns and Release Rates We request CANDU 3 specific radionuclide source tenn and release analysis repons,if available and the following repons.
- G. I. Hadaller, G. H. Archinoff, and E. Kohn, "CANDU Fuel Bundle Behavior During Degraded Cooling Conditiorf,4th Annual Conference of the Canadian Nuclear Society, Montmal,1983 June.
+ E. Kohn, G. I. Hadaller, R. M. Sawala, G. I. Archinoff and S. L. Wadswonh, "CANDU Fuel Deformation During Degraded Cooling - Experimental Results", Canadian Nuclear Society Confemnce,1985 June.
3.5 Desien Basis and Severe Accidents We request repons addmssing CANDU 3 severe accident j
analysis, if available, including probabilistic analysis of beyond design basis events.
3.6 Analysis Methods We request repons discussing Canadian accident analysis codes and their technical and experimental basis in addition to the following efemnced repons.
- M. R. Lin, S. Prawirosoehardjo, " FIREBIRD-III Mod 1 Program Description", TDAI-373,1984 j
November.
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. A. Lai, " FIREBIRD Model for CANDU 3 232-Channel,4-Header Primary Heat Transpon System Design, AR-74-03500-032,1988 August.
. W. M. Collins, "PRESCON2 Pmgnm Description", AECL Repon TDAI-292 Volume 1,1982 September.
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~ W. M. Collins,"MICROPRESCON 2 Program Description", TDAI-368, Vol.1,1985 February.
. P. Muzundar, R. L. Sakaguchi, J. K. Presley, "HOTSPOT-II Fuel Bundle Thermal Response Code",
Ontario Hydro Report NSSD 83058,1983 April.
- H. Keil, MODHT: A Computer Pmgmm to Predict Tmnsient Moderator Temperatures -Model Description", TDAl-183, Volume 1,1980 March.
'a 'D. T. Richards, B. N. Hanna. N. Hobson, and K. H. Ardon, "CATHENA: A Two-Fluid Code for CANDU LOCA Analysis, (renamed fmm ATHENA)," Pesented at the Third Intemational Topical Meeting on Reactor Thermal Hydraulics, Newport, RI,1985 October.
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- D. J. Richards, " Validation of the CATHENA Two-Fluid Code", presented at the Third Intemational Topical Meeting on Reactor Thennal-Hydraulics, Soeul, Koma,1988 November 14-27.
- J. P. Mallory, M. A. Wright and H. Huynh, " Validation of CATHENA at High Temperature Conditions Using CHAN Thennal-Chemical Experiment Results," Presented at the 12th Annual CNS Conference, Saskatoon, SK,1991 June.
J. E. Kowolski, V. S. Krishnan, " Transient Stratified Flow Experiments in a Horizontal Channel Cor,taining Rod Bundles and Numerical Simulations Using the ATHENA Two-Fluid Code", Pesented at the Intemational ANS-ENS Topical Meeting on Thennal Reactor Safety, San Diego, Ca.,1986 Febmary 2-6.
3.7 Performance of Unique Features We request the following referenced reports.
- AECL, " Unique Aspects of the CANDU 3 Design", Atomic Energy of Canada, Limited, June 1989.
. S. D. Grant and J. M. Hopwood, Ihe Effect of Fuel Heat Transfer on Early Void Production Following a Large Pipe Break in CANDU Reactors", Canadian Nuclear Society Simulation Symposium, Winnipeg, Manitoba,1988 April.
- P. G. Gulshani, "Pediction of Pessure Tube Integrity for Large Loss-ofCoolant Accident in CANDU", American Nuclear Society,1987 Winter Meeting, Los Angles, Ca,1987 November 15-19.
- V. I. Nath and Kohn, "High Temperature Oxidation of CANDU Fuel During a LOCA", Pmceedings of the Fifth Intemational Meeting on Thennal Nuclear Reactor Safety, Karismhe,9-13 September, 1984, Kraftwerk Union Report KFK 388011,1984 December.
3.8 Reculatory and Standards Documents We request the following reports.
- Hedges, K. R., M. Bonechi, and E. M. Hinchley, "CANDU 3 Meets ALWR Requirements", Modem Power Systems,1990 December.
- AECB, " Requirements for the Safety Analysis of CANDU Nuclear Power Plants", Atomic Energy Control Board, Consultative Document C-6, Ottawa, Ontario, Canada,1980 June.
+ The following Canadian Standards. The Canadian Standards Association (CSA) Standards have apparently already been submitted to NRC.
CAN3-A23.3-M 84, Design of Concrete Structures for Buildings CAN3-N287.1-M 82 General Requirements for Concrete Containment Stmetures for CANDU '
Nuclear Power Plants 3
CAN3-N287.2-M 82 Material Requimments for Concrete Containment Structures for CANDU Nuclear Power Plants -
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'CAN31N2873-M82 Design Requirements for Concrete Containment Stmetures for CANDU i
Nuclear Power Plants CAN3-N287.4-M 83 Constmetion, Fabrication, and Installation Rcquirements for Concrete
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Containment Structures for CANDU Nuclear Power Plants CAN3-N287.5-M81 Testing and Examination Requirements for Concrete Containment Stmetures for CANDU Nuclear Power Plants UAN3-N287.6-M 80 Pre-Operational Proof and Leakage Rate Testing Requirements for Concrete Containment Structures for CANDU Nuclear Power Plants CAN3-N287.7-M80 In-Service Examination and Testing Requirements for Concrete Containment Structures for CANDU Nuclear Power Plants CAN3-N289.1-80 General Requirements for Seismic Qualification of CANDU Nuclear Power Plants CAN3-N289.2-M 81 Ground Motion Determination for Seismic Qualification of CANDU Nuclear Power Plants CAN3-N2893-M 81 Design Pmcedures for Seismic Qualification for CANDU Nuclear Power Plants CAN3-N289.4-M86 Testing Procedures for Seismic Qualification of CANDU Nuclear Power Plants i
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