ML20094B747
| ML20094B747 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-06.2.1, NUREG-75-87, NUREG-75-87-6.2.1, SRP-06.02.01, SRP-6.02.01, NUDOCS 9511010244 | |
| Download: ML20094B747 (7) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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1 STANDARD REVIEW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION
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SECTION 6.2.1 CONTAINMENT FUNCTIONAL DESIGN REVIEW RESPONSIBILITIES Primary-ContairmentSystemsBranch(CSB)
Secondary - Accident Analysis Branch (AAB)
Mechanical Engineering Branch (MEB)
Structural Engineering Branch (SEB)
Reactor Systems Branch (RSB)
Core Performance Branch (CPB)
Electrical, Instrumentation and Control Systems Branch (EICSB)
Auxiliary and Power Conversion Systems Branch (APCSB) 4 INTRODUCTION The CSB reviews information regarding the functional capability of the reactor containment pre 4 j
sented in Section 6.2.1 of the applicant's safety analysis report (SAR). The containment en-closes the reactor system and is the final barrier against the release of significant amounts of radioactive fission products in the event of an accident. The containment structure must be capable of withstanding, without loss of function, the pressure and temperature conditions resulting from postulated loss-of-coolant and steam or feedwater line break accidents. The containment structure must also maintain functional integrity in the long term following a postulated accident; i.e., it must remain a low leakage barrier against the release of fission products for as long as postulated accident conditions require.
The design and sizing of containment systems are largely based on the pressure and temperature conditions which result from release of the reactor coolant in the event of a loss-of-coolant i
accident (LOCA). The containment design basis includes the effects of stored energy in the reactor coolant system, decay energy. and energy from other sources such as the secondary system, and metal-water reactions including the recombination of hydrogen and oxygen, The containment system is not required to be a complete and independent safeguard against a LOCA by itself, but functions to contain any fission products released while the emergency core cooling system cools the reactor core to prevent any extensive fuel melting.
The evaluation of a containment functional design includes calculation of the progress of a LOCA event after an instantaneous rupture is assumed to occur in some section of the primary USNRC STANDARD REVIEW PLAN W
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coolant system piping. The subssquint thermodynamic effects in the containment resulting from the release of the coolant mass and energy in the primary system are detennined from a solution of the incremental space and time-dependent energy, mass, and momentum equations. The basic functional design requirements for containment are given in General Design Criteria 16 and 50 in Appendix A to 10 CFR Part 50. General Design Criterion 50, among other things, requires that consideration be given to the potential consequences of degraded engineered safety featurf.5,
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such as the containment heat removal system and the emergency core cooling system, the limitations in defining accident phenomena, and the conservatism of calculational models and input parameters, in assessing containment design margins.
There are a number of different containment types and designs, and several aspects of contain-ment functional design that are within the scope of SAR Section 6.2.1.
The various containment types and aspects to be reviewed under this plan have been separated and assigned to a set of "subplans" as follows:
a.
Pressurized water reactor (PWR) dry containments, including subatmospheric containments l
(SRP6.2.1.1.A).
b.
Ice condenser containments (SRP 6.2.1.1.B).
c.
Mark I, !!, and III boiling water reactor (BWR) pressure-supression type containments (SRP6.2.1.1.C).
d.
Subcompartmentanalysis(SRP6.2.1.2).
e.
Mass and energy release analysis for postulated loss-of-coolant accidents (SRP 6.2.1.3).
f.
Mass and energy release analysis for postulated secondary system pipe ruptures (SRP 6.2.1.4).
g.
Minimum containment pressure analysis for emergency core cooling system (ECCS) perform-ance capability studies (SRP 6.2.1.5).
A separate standard review plan (SRP) has been prepared for each of these areas.
Areas related to the evaluation of the containment functional capability are treated in other standard review plans; e.g., containment heat removal (SRP 6.2.2), combustible gas control (SRP6.2.5),andcontainmentleakagetesting(SRP6.2.6).
I.
AREAS OF REVIEW The items reviewed are described in the " Areas of Review" sections of the seven "subplans" listed above.
II.
ACCEPTANCE CRITERIA The acceptance criteria are given in the " Acceptance Criteria" sections of the seven "sub-plans" listed above.
III.
REVIEW PROCEDURES Review procedures are given in " Review Procedures" sections of the seven "subplans" listed above.
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IV. EVALUATION FINDINGS Th2 results of the reviews under the seven "subplans" listed above are consolidated into a single set of findings. The reviewer verifies that sufficient information has been provided 4
and that his evaluation is adequate to support conclusions of the following type, to be i
included in the staff's safety evaluation report:
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" Containment Functional Design j
.The scope of review of the functional design of the containment for the ABC nuclear power plant has included a review of plant arrangement drawings, system drawings, and descriptive information for the containment building, subcompartments, and associated systems, components, and structures that are essential to the functional capability and integrity of the containment. The review has included the applicant's proposed
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design bases for the containment building and internal structures, and associated structures and systems upon which the containment function depends, and the applicant's
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analysis of postulated accidents and operational occurrences which support the adequacy of the design bases.
"The basis for the staff's acceptance has been conformance of designs and design bases for the containment building, internal structures, and associated systems, j
components, and structures to the Commission's regulations as set forth in the general i
design criteria, and to applicable regulatory guides, brarch technical positions, and industry codes and standards. (Special problems or exceptions that the staff takes
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to the design or functional capability of containment structures, systems, and components should be discussed.)
4 "The staff concludes that the containment functional design confoms to applicable regulations, guides, staff positions, and industry standards, and is acceptable." -
V.
REFERENCES 1.
10 CFR Part 50 Appendix A, General Design Criterion 16. " Containment Design;" Cri-terion 39, " Inspection of Containment Heat Removal System;" Criterion 40, " Testing of Containment Heat Removal System;" Criterion 50, " Containment Design Basis;" Criterion 54, " Systems Penetrating Containment;" and Criterion 56, " Primary Containment Isolation."
2.
10 CFR 550.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" and 10 CFR Part 50, Appendix K, "ECCS Evaluation Models."
3.
ASME Boiler and Pressure Vessel Code,Section II, Division 1. Subsection NE, " Class MC Components," American Society of Mechanical Engineers.
4.
Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors."
5.
Regulatory Guide 1.26, " Quality Group Classifications and Standards."
6.
Regulatory Guide 1.29, " Seismic Design Classifications and Standards."
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7.
C. F. Camichael and S. A. Marks, " CONTEMPT-PS, A Digital Computer Code for Predicting ths Pressure-Ttmperature History Within a Pressurs-Suppression Containmet.t Vessel in Response to a Loss-of-Coolant Accident," 100-17252 Phillips Petroleum Company, April 1969.
8.
L. C. Richardson, L. J. Finnegan, R. J. Wagner, and J. M. Waage, " CONTEMPT, A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss-of-Coolant Accident," 100-17220, Phillips Petroleum Company, June 1967.
9.
R. J. Wagner and L. L. West, " CONTEMPT-LT Users Manual," Interim Report I-214-74-12.1, Aero,let Nuclear Company, August 1973.
10.
R. I Miller, " Evaluation of the Predictive Capabilities of the CONTEMPT-PSComputer Code by Comparison of Calculated Results with the Humboldt Bay and Bodega Bay Pressure Suppression Tests," Interim Report 4.2.1.1, Idaho Nuclear Corporation, September 1970.
11.
D. C. Slaughterbeck. " Comparison of Analytical Techniques Used to Determine Distribu-tion of Maas and Energy in the Liquid and Vapor Regions of a PWR Containment Following a Loss-of-Coolant Accident," Special Interim Report Idaho Nuclear Corporation, January 1970.
12.
R. C. Schmitt, G. E. Bingham, and J. A. Norberg, " Simulated Design Basis Accident Test
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for the Carolina Virginia Tube Reactor Containment - Final Report " IN-1403, Idaho Nuclear Corporation, December 1970.
13.
D. C. Slaughterbeck " Review of Heat Transfer Coefficients for Condensing Steam in a Containment Building Following a Loss-of-Coolant Accident," IN-1388, Idaho Nuclear Corporation, September 1970.
14.
T. Tagami, " Interim Report on Safety Assessments and Facilities Establishment Project in Japan for Period Ending June 1965 (No.1)," prepared for the National Reactor Testing Station, February 28,1966 (unpublished work).
15.
H. Uchida, A. Oyama, and Y. Toga, " Evaluation of Post-Incident Cooling Systems'of Light-Water Power Reactors," Proc. Third International Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, United Nations, Geneva (1964).
16.
" FLOOD / MOD 002 - A Code to Detemine the Core Reflood Rate for a PWR Plant with Two Core Vessel Outlet Legs and Four Core Vessel Inlet Legs," Interim Report, Aerojet Nuclear Company, November 2, 1972.
17.
" FLOOD / MOD 001 - A Code to Determine the Core Reflood Rate for a PWR Plant with Two Core Vessel Outlet Legs and Two Core Vessel Inlet Legs, Interim Report, Aerojet Nuclear Company October 11, 1972.
18.
C. L. King and G. B. Peeler, " Moisture Carryover During an NSSS Steam Line Break Accident," CENPD-80 (Rev.1), Combustion Engineering, Inc., June 1973.
6.2.1-4 11/24/75
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19.
P. A. Lowe, J. R. Brodrick, and W. E. Burchill, " Steam-Water Mixing Test Program Task D: Fonr.31 Report fer Task A: 1/5 Scale Intact Loop," CENPD-65 (Rev.1), Combustion Enginering. Inc., March 1973 2 0.'
J. R.. Brodrick, W. E. Burchill, and P. A. Lowe, "l/5 Scale Intact Loop Post-LOCA Steam j
Relief Tests," CENPD-63 (Rev.1), Combustion Engineering, Inc., March 1973.
j 21.
W. H. Retting, G. A. Jayne, K. V. Moore C. E. Slater, and M. L. Uptmor, "RELAp3 -
A Computer Program for Reactor Blowdown Analysis," IN-1321. Idaho Nuclear Corporation, June 1970.
22.
F. J. Moody. " Maximum Flow Rate of a Single Component. Two-Phase Mixture," Jour. of Heat Transfer Trans. Am. Soc. of Mechanical Engineers Vol. 87 No.1 February 1965.
23.
" CRAFT - Description of Model for Equilibrium LOCA Analysis Program " BAW-10030, Babcock and Wilcox Company, October 1971.
i 24.
" SATAN-VI Program, A Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant,"
WCAP-8302, Westinghouse Electric Corporation, June 1974.
25.
" Description of Loss-of-Coolant Calculational Procedures," EPFD-26. Combustion Engineer-ing, Inc., August 1971.
I 26.
F. C. Cadek, et al., "PWR FLECHT (Full Length Emergency Cooling Heat Transfer), Final Report " WCAP-7665. Westinghouse Electric Corporation, April 1971.
27.
Final Safety Analysis Report for Donald C. Cook Nuclear Plant, Units 1 and 2. Appendices M and N, American Electric Power Company, and the Staff Safety Evaluation Report. AEC Docket Nos. 50-315/316.
28.
" Ice Condenser Containment Pressure Transient Analysis Methods," WCAP-8077. Westinghouse Electric Corporation, March 1973.
29.
" Mark III Analytical Investigation of Small-Scale Tests Progress Report," NEDM-10976, General Electric Company August 1973.
30.
"The General Electric Pressure Suppression Containment Analytical Model," NEDO-10320, General Electric Company, April 1971; Supplement 1, May 1971; Supplement 2, January 1973.
31.
ANC Letter, " Rationale for ANC Vent Clearing Model Nodalization " Oben-5 74, Aerojet Nuclear Company, February 14, 1974.
32.
ANC Letter, " Review of General Electric Mark III Experimental and Analytical Programs,"
Oben-28-73, Aerojet Nuclear Company, December 26, 1973.
6.2.1-5 11/24/75
- 33. Branch Technical Position CSB 6-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation," attached to Standard Review Plan 6.2.1.5.
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