ML19282B957
| ML19282B957 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 02/27/1979 |
| From: | Baer R Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 7903190156 | |
| Download: ML19282B957 (5) | |
Text
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d, UtdTED STATES
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NUCLEAR REGULATORY COT.*T."IsSION n ASHINGTON, D. C. 20555 hh f
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\\, U Y'l Mcket N:s. 50-369 and 50-370 Mr. William 0. Parker, Jr.
Vice President, Steam Production Duke Power Company P. O. Box 2178 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Parker:
SUBJECT:
REANALYSIS OF CONTAINMENT RESPONSE TO POSTULATED MAIN STEAM LINE BREAK ACCIDENT (McGuire Nuclear Station, Units 1 and 2)
This is to confirm our discussions with your staff regarding our require-ment for a reanalysis of the McGuire containment temperature and pressure response to postulatej main steam line breaks in the containment lower compartment using the revised Westinghouse Clectric Corporation LOTIC-3 ice condenser containment code. Our request for this information is described in the Enclosure.
We request that this information be provided by March 23, 1979.
Our letter to you dated February 8,1979 included the above as an outstanding review matter.
Sincerely,
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Robert L. Baer, Chief Light Water Reactors Branch No. 2 Division of Project Managemen:
Enclosure:
Request for Additional Information ces w/ enclosure:
See next page 7903190MG
. <. l l: 5r C. Da rker, Jr.
"' : _ i c r i dc r.:, Ste ct Fr:dscticn E.ce Faus-r.o r.: 3 ny L. '. Ecx 217i 4.22 Lu.th Cnuren St reet Cr.ar ic::e,.cr:h Carolina 252 2 cc: :~ r.... L. P o rte r Cse,e Power Ccapany P. O. Eox 2178 422 South Church Street Charlotte, North Carolina 28242 t.r. R. S. Ho wa rd Poter Systems Civisicn Westinchcuse Electric Corporation P. O. Eox 355 Pittsburgh, Pennsylvania 15230 Mr. E. J. Keith EDS Nuclear Incorporated.
220 Montgomery Street San Francisco, California 94104 Mr. J. E. Houghtaling f.US Corporation 2536 Ccuntryside Boulevard Cleerwater, Florida 33515 Fr. Jesse L. Riley, President The Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207 J. Michael PcGarry, III, Esq.
Debevoise a Liberman 1200 Seventeenth Street, N. W.
Washington, D. C.
20036 Shelley Bl un, Esq..
418 Law Euilding 730 East Trade Street Charictte, North Carolina 28202
illier D. Parker, Jr.
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- c c e r'. ';. L a:c, E sc., C nai rman
' cr.ic Safe ;. and Licensing Board U. S. 'bcleSr Regula: cry Commission
'..ds hi n gton, D. C.
205E5 Dr. Emmeth A. Luebke Atonic Safety and Licensing Board U. S. l'uclear Regulatory Commission Washington, D. C.
2C555 Dr. Cadet H. Hand, Jr., Director Bodega f'arine Lab of California P. O. Box 247
, Bodega Bay, California 94923 e
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ENCLOSURE Request for Additional Information Containment Systems Branch M Guire Nuclear Station, Units 1 & 2 Docket Nos. 50-359/370 042.73 The analysis of the containment response to postulated main steam line (6.2.1.3) breaks in the containment lower compartment were performed using the Westinghouse Electric Corporation LOTIC-3 code prior to completion of the NRC staff's review of the code. During the course of our generic review of the LOTIC-3 code, modifications were made to the code by the Westinghouse Electric Corporation. Thus, the main steam line break analysis currently described in the McGuire FSAR (Section 6.2.1.3.1.2) was not performed in accordance with the provisions of the NRC staff approval of the LOTIC-3 code.
In our safety evaluation report for the LOTIC-3 topical report (WCAP 8354-Supplement 2), we concluded that the.LOTIC-3 code is acceptable for calculating the temperature and pressure response of ice condenser containments following certain postulated secondary system pipe breaks provided the following items are satisfied:
1.
The mass and energy release data, which are input to the code should be cal 1:ulated using a model approved by the NRC.
2.
The model (Option 2) which assumes steam to be condensed and added to the sump (due to heat transfer with structural elements) should be used for break sizes that produce no liquid entrainment and for all other break sizes until models with liquid entrainment are approved.
et
- 3.
Analyses using the LOTIC-3 code should include a break spectrum analysis to demonstrate that the most severe containment conditicr.s have been identified, since small breaks have been shown to produce elevated temperatures within the containment for periods longer than those for double-ended breaks.
Accordingly, to complete our review of the containment transient response to a postulated main steam line break (MSLB) we require the following:
The results of a re-analysis of the containment response to postulated MSLB's using the approved version of the LOTIC-3 code as prescribed in the NRC's safety evaluation report on the code. The re-analysis should include the spectrum of circumferential pipe ruptures previously 2
analyzed (1.4, 0.6 and 0.4 ft ) as well as a spectrum of smaller breaks.
f This spectrum should include the largest break which will not result in a feedwater isolation signal or liquid entrainment from the break, and the largest break which would not result in automatic operation of the containment sprays and return air fans. The spectrum of breaks considered should include both circunferential pipe breaks and longitudinal 3plit pipe breaks, unless it can be shown that split breaks can not c:
occur.
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