ML18095A253
| ML18095A253 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/29/1990 |
| From: | Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCR-90-01, LCR-90-1, NLR-N90107, NUDOCS 9006060247 | |
| Download: ML18095A253 (2) | |
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Public Service Electric and Gas Company Stanley LaBruna Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President - Nuclear Operations May 29, 1990 NLR-N90107 LCR 90-01 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR ADDITION INFORMATION END OF LIFE MODERATOR TEMPERATURE COEFFICIENT LIMIT SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSE NOS. DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 Public Service Electric and Gas (PSE&G) hereby provides the additional information requested in your letter of April 20, 1990.
The relationship of moderator feedwater used in the main steam line break analyses to the most negative Moderator Temperature Coefficient (MTC) Limiting Condition for Operation (LCO) (generated for hot, full power conditions with all control rods out) or to the Moderator Density Coefficient (MDC) is discussed below.
The most negative MTC LCO is based on a constant value of MDC which is used in the Westinghouse safety analysis of the events described in the License Change Request.
This MDC value is based on hot full power conditions.
In general, for transients which result in an increase in heat removal from the core, this constant MDC is assumed.
The intent is to maximize the overpower transient which leads to a decrease in DNB margin.
The steam line break event is an exception to this generalization.
From a DNB margin standpoint, the limiting steam line break occurs at hot zero power.
Therefore, credit is taken in the analysis for smaller moderator feedbacks due to reduced moderator temperature and pressure, and boron from the safety injection, which occurs after the reactor trip.
Figure 15.4-48 in the Salem FSAR shows the effect of moderator temperature on the feedback only at one state point of 1000 psia.
The safety analysis would include many such curves, covering a range of system pressures and boron concentrations.
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Document Control Desk NLR-90107 5-29-90 Therefore, the feedback used in the steam line break has no relationship to the hot full power MDC.
Since credit is taken in the reactivity feedback, the steam line break is evaluated for each fuel cycle.
Additionally, analysis is performed for containment mass and energy release during the steam line break, prior to the reactor trip.
From the containment standpoint, the limiting steam line break occurs at partial power.
Therefore, this analysis assumes the constant hot full power MDC value.
Should you have any questions regarding this transmittal, please feel free to contact us.
c Mr. J. c. Stone Licensing Project Manager Mr. T. Johnson Senior Resident Inspector Sincerely, Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625