ML19249F057
| ML19249F057 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/21/1979 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Mittl R Public Service Enterprise Group |
| References | |
| NUDOCS 7910030890 | |
| Download: ML19249F057 (6) | |
Text
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SEP 211979 Docket No. 50-311 M.~. R. L. Mitti, General Manager Licensing and Environment Engineering aad Construction Department Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101
Dear Mr. Mitti:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW 0F THE SALEM UNIT 2 FINAL SAFETY ANALYSIS REPORT (FSAR)
As a result of our continuing review of the Salem FSAR, we find that we need additional informction to complete our evaluation. The specific information required is listed in the enclosure.
Our review schedule is based on the assumption that this additional information will be available for our review by October 5,1979.
If you can-not meet this date. please inform us within seven days after receipt of this letter so that we may revise our schedule accordingly.
Please contact us if you desire any discussion or clarif. cation of the enclosed request.
Sincerely,
%an 19. L
. Parr, Chief Light Water Reactors, Branch No. 3 Division of Project Management
Enclosure:
As Stated cc: See Next Page 10/4 18/
7910080 h O g
I.
Mr. R. L. Mitti, Gen 2ral Manager cc:
Richard Fryling, Jr., Esq.
Assistant General Counsel Public Service Electric & Gas Company 80 Park Place Newark, New Jersey 07100 Mark Wetterhahn, Esq.
Conner, Moore & Cober 1747 Pennsylvania Avenue, N.W.
Suite 1050 Washington, D.C.
20006 Mr. Leif J. Norrholm V. S. Nuclear Regulatory Commission Region I Drawer I Hancocks Bridge, New Jersey 08038 h0lb h00 r
b ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION SALEM NUCLEAR GENER_ATING STATION, UNIT 2 DOCKET N0.__50-311 5.0 Containment Systems 5.106 In a letter dated September 10, 1979, the NRC was informed by Virginia Electric and Power Company that overoressurization of the containment at North Anna 3 and 4 could occur as a result of a main steam line break inside containment. This overpressurization resulted when auxiliary feeuater flow was included in the analysis.
NRC is currently assessing the generic implications of this letter.
To assist us in determining if a similar circumstance could occur at your facility, you should take the following actions.
1)
Review your original analysis of this event, and provide NRC with the assumptions used during this analysis. Particular emphasis should be placed un describing how auxiliary feedwater flow (AFF) was accounted for in your original analysis.
(Refer-ence to previously submitted information is acceptable if identified as to page number and date.) Any changes in your design which would impact the conclusions of your original analysis snuuld be discussed. We are particularly concerned with design changes that could lead to an underestimation of the containment pressure following a MSLB inside containment.
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i 2)
Specifically, provide the following information for the analyses performed to determine the maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels:
a.
Specify the auxiliary feedwater flow rate th't was used in your original containment pressurization an ayses.
Provide the basis for this assumed flow rate.
b.
Provide the auxiliary feedwater rated flow rate, the run out flow rate, and the pump head capacity curve of your current design.
c.
Provide schematic drawings to show the auxiliary feedwater system arrangement in your current design.
d.
Provide the time span over which it was assumed in your original analysis that AFF was added to the affected steam generator following a MSLB inside containment.
Discuss the design provisions in the auxiliary feedwater e.
system used to terminate the auxiliary feedwater flow to the affected steam generator.
If operator action is required to perform this function, discuss the information that will be available to the operator to alert him of the need to isolate the auxiliary feedwater to the affected steam generator, the time when this information would be-come available, and the time it would take the operator 1.0/4 190
i to complete this action.
If termination of auxiliary feed-water flow is dependent on automatic action, describe the basic operation of the auto-isolation system.
Describe the failure modes of the system.
Describe any annunciation devices associated with the system.
f.
Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and the contain-ment pressure response. The single failure analysis should include, but not necessarily be limited to:
partial loss of containment cooling systems and failure of the auxiliary feedwater isolation valve to close.
g.
For the single active failure case which results in the maximum containment atmosphere pressure, provide a chronology of events.
Graphically, show the containment atmosphere pressure as a function of time for at least 30 minutes fo: lowing the accident.
For this case, assume the auxiliary feedwater flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current design) for the entire transient if no automatic isolation to auxiliary feedwater is part of the current design.
h.
For the case identified in (g) above, provide the mass and energy release data in tabular form.
Discuss and justify 1074 191
s.
the assumptions made regarding the time at which active containment heat removal systems become effective.
10/4 192