ML20141A286
| ML20141A286 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/16/1997 |
| From: | Huffman W NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9706200207 | |
| Download: ML20141A286 (5) | |
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UNITED STATES p
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 4 o01 t
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May 16, 1997 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
SUBJECT:
AP600 CONTAINMENT PRESSURE ANALYSIS REQUESTS FOR ADDITIONAL INFORMA-TION (RAI) l
Dear Mr. Liparulo:
In support of the AP600 design certification review, the Nuclear Regulatory Commission (NRC) staff is evaluating the Westinghouse containment analysis in Section 6.2 of the AP600 Standard Safety Analysis Report.
Based on its review of the containment subcompartment pressurization analysis and minimum contain-ment backpressure analysis, the staff requests additional information. The l
RAI are provided in an enclosure to this letter.
You have requested that portions of the information submitted in the June 1992, application for design certification be exempt from mandatory public disclosure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that portion of the submitted information is being withheld from public disclosure pending the staff's final determination.
The staff concludes that these followon ques-tions do not contain those portions of the information for which exemption is sought. However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westinghouse the opportunity to verify the staff's conclusions.
If, after that time, you do not request that all or portions of the information in the enclosures be withheld from public disclosure in accordance with 10 CFR 2.790, this letter will be placed in the NRC Public Document Room, j
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NRC H E CENTIB COPY
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I 9706200207 970516 PDR ADOCK 05200003 A
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l Mr. Nicholas J. Liparulo May 16, 1997 i
If you have any questions regarding this matter, you can contact me at (301) 415-1141.
Sincerely, original signed by:
William C. Huffman, Project Manager Standardization Project Directorate l
Division of Reactor Program Management Office of Nuclear Reactor Regulation l
Docket No.52-003 i
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION:
- Enclosure to be held for 30 days
- Docket File PDST R/F MSlosson PUBLIC SWeiss TRQuay TKenyon BHuffman JSebrosky DJackson JMoore, 0-15 B18 WDean, 0-5 E23 ACRS (11)
CBerlinger, 0-8 H7 MSnodderly, 0-8 H7 JDawson, 0-8 H7 DOCUMENT NAME: A:DWN2-CNT.RAI
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(E' = Copy with attachment / enclos To aceive a copy of this decwnnat,indcote in the boa: "C" = Copy without attachmark/o.
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- N' = No copy l
OFFICE PM:PDST:DRPM D:SCSB:DSSAT. A D:PhsT!ORPM l
f NAME WCHuffman:sg//9 & CBerlinged.7rk' TRQuhk '
DATE 05/11,,/97 V
05/l[r/97 05/tk/97 0FFICIAL RECORD COPY
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,":f Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 t
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cc: Mr. B. A. McIntyre Ms. Cindy L. Haag Advanced Plant Safety & Licensing Advanced Plant Safety & Licensing 4
Westinghouse Electric Corporation Westinghouse Electric Corporation i
Energy Systems Business Unit Energy Systems Business Unit P.O. Box 355 Box 355 i
Pittsburgh, PA 15230 Pittsburgh, PA 15230 i
Mr. M. D. Beaumont Mr. S. M. Modre 4
Nuclear and Advanced Technology Division Nuclear Systeas Analysis Technologies Westinghouse Electric Corporation Lockheed Idaho Technologies Company One Montrose Metro Post Office Box 1625 11921 Rockville Pike Idaho Falls, ID 83415 Suite 350 Rockville, MD 20852 Enclosure to be distributed to the following addressees after the result of the proprietary evaluation is received from Westinghouse:
Mr. Ronald Simard, Director Ms. Lynn Connor Advanced Reactor Programs DOC-Search Associates Nuclear Energy Institute Post Office Box 31 1776 Eye Street, N.W.
Cabin John, MD 20818 Suite 300 Washington, DC 20006-3706 Mr. Robert H. Buchholz GE Nuclear Energy Mr. James E. Quinn, Projects Manager 175 Curtner Avenue, MC-781 LMR and SBWR Programs San Jose, CA 95125 GE Nuclear Energy 175 Curtner Avenue, M/C 165 Mr. Sterling Franks San Jose, CA 95125 U.S. Department of Energy NE-50 Barton Z. Cowan, Esq.
19901 Germantown Road Eckert Seamans Cherin & Mellott Germantown, MD 20874 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. Charles Thompsi n, Nuclear Engineer AP600 Certificat!oa Mr. Frank A. Ross NE-50 U.S. Department of Energy, NE-42 19901 Germantown load Office of LWR Safety and Technology Germantown, MD 20874 1990) Germantown Road Germantown, MD 20874 Mr. Ed Rodwell, Manager PWR Design Certification r
Electric Power Research institute 3412 Hillview Avenue Palo Alto, CA 94303 i
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i AP600 STANDARD SAFETY ANALYSIS REPORT (SSAR) SECTION 6.2 REQUESTS FOR ADDITIONAL INFORMATION Ouestions 480.1042 throuah 480.1045 aoolv to the Containment Subcompartment Pressurization Analysis 480.1042 Besides those subcompartments listed in SSAR Section 6.2.1.2, were any
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other subcompartments analyzed? In particular, the staff has identified from drawinas the reactor coolant drain tan.k room, the RHR valve room, and contaiiment isolation valve areas. Do these, or any other compart-ments, need to be analyzed? Please justify your response.
480.1043 In the equilibrium flow model used to calculate the vent flow, was an
" equilibrium" or " frozen" model used.
Please provide any clarifying details.
b 480.1044 Explain the difference between the " modified" Zaloudek coefficient and the Zaloudek coefficient the staff is familiar with. What are the modifications, and how is the coefficient different from the " unmodified" Zaloudek coefficient? Justify use of the modified coefficient, and discuss any impact of the modifications on the calculated pressures.
480.1045 Section 3.6.1.2 of SSAR Rev. 7 states that the loads for the reactor vessel annulus asymmetric compartment pressurization analysis are based on a 5 anm leakage crack in the primary system piping.
However, this analysi. is not discussed in Section 6.2.1.2 of the SSAR, " Containment Subcompartments." Please submit details on the analysis that was conducted to determine the reactor vessel cavity asymmetric pressuriza-tion loads (at a minimum, include those details considered in the analyses conducted for the other containment subcompartments).
Question 480.1046 and 480.1049 apply to the Minimum Containment Pressure Analysis for Performance Capability Studies _of ECCS 480.D46 Was the mixing of subcooled ECCS water with steam h !he containment atmosphere modeled in accordance with Branch Technicai Position (BTP)
CSB 6-17 That is, was it assumed that the subcooled ECCS water and steam in the atmosphere mixed, and thus helped minimize the containment pressure? If not, provide justification for deviation from this aspect BTP CSB 6-17 480.1047 What backpressure is credited for the ECCS/LOCA analysis? Table 15.6.5-5 of SSAR Revision 12, dated April 30, 1997, says only that a " lower bound" pressure was used.
480.1048 Is the mass and energy release used in the minimum containment backpres-sure analysis the same as that used in the WCOBRA/ TRAC ECCS LOCA analysis described in WCAP-14171?
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480.1049 Was the containecnt recirculation cooling system assumed to be operating in the containment minimum backpressure analysis conducted to assess the ECCS performance capability? If so, then provide details on the assump-tions (e.g., fan speed, number of units operating, chilled water tempera-ture and flow assumed and rationale for these values).
If the units were not assumed to be operating, then provide the rationale for this assumption. Provide justification that the assumptions used in the current SSAR minimum backpressure analysis bound any cases where the recirculation cooling system would be assumed to be operating in accor-l dance with Branch Technical Position (BTP) CSB 6-1 (i.e., with maximum flow and minimum chilled water temperature).
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