ML13330A388
| ML13330A388 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/26/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Dietch R SOUTHERN CALIFORNIA EDISON CO. |
| References | |
| TASK-15-03, TASK-15-19, TASK-15-3, TASK-RR LSO5-81-08-061, LSO5-81-8-61, NUDOCS 8109030057 | |
| Download: ML13330A388 (8) | |
Text
August 26, 981 Docket No. 50-206 LS05 08-061 Mr. R. Dietch, Vice President Nuclear Engineering and Operations Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770
Dear Mr. Dietch:
SUBJECT:
SAN ONOFRE 1 - SEP TOPICS XV-3, LOSS OF LOAD AND XV-19, LOSS OF COOLANT ACCIDENT (SYSTEMS)
By letter dated July 1, 1981, you submitted safety assessment reports for the above topics.
T'hE staff has reviewed these assessments and our con clusions are presented in the enclosed safety evaluation reports, which complete these topics for San Onofre 1.
These evaluations will be basic inputs to the integrated assessment for your facility. The evaluations may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosures:
As stated cc w/enclosures:
See next page
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Mr. R. Dietch cc Charles R. Kocher, Assistant General Counsel James Beoletto, Esquire Southern California Edison Company Post Office Box 800 Rosemead, California 91770 David R. Pigott Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 Harry B. Stoehr San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Resident Inspector/San Onofre NPS c/o U. S. NRC P. 0. Box 4329 San Clemente, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego San'Diego, California 92101 California Department of Health ATTN: Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 U. S. Environmental Protection Agency Region IX Office ATTN:
Regional Radiation Representative 215 Freemont Street San Francisco, California 94111
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SAN ONOFRE 1, SEP TOPIC XV-3 Loss of External Load I. INTRODUCTION Loss of electrical load can result from a number of system disturbainces, and the most likely source is a turbine generator trip. A partial loss of load may also result for the turbine governor response to a sudden increase in frequency of the distribution network. The rapid reduction in generator load will result in a significant reduction in heat removal rate from the primary coolant system and may cause an increase in reactor coolant temperature and pressure in excess of normal core operating limits.
II. EVALUATION The licensee has performed an analysis of complete loss of load with a reactor trip (Ref.1). The initial conditions include reactor power, coolant temperature and pressure at maximum values with the plant at full power (103%) which leads to maximum power difference and minimum margin to core protection limits, at the initiation of the loss of load. The analysis assumes the reactor in manual control without control rod insertion, prior to reactor trip on high pressure (2250 psia).
No credit is assumed for steam release to the atmosphere and con denser or the effects of spray and relief valves in reducing or limiting pressurizer pressure.
The direct reactor trip on turbine trip, derived from turbine autostop oil pressure was neglected.
The results of the analyses indicate that the integrity of the core is maintained by a high pressure reactor trip in seven seconds and a minimum DNB ratio of 1.55.
-2 III.
CONCLUSION As part of the SEP review for San Onofre 1, we have evaluated the licensee's analysis of loss of external load (Ref. 1), against the criteria of SRP Section 15.2.1, and have concluded that it is in conformance with these criteria.
TURBINE TRIP The turbine trip event is bounded by the analysis of loss of external load because a turbine trip will initiate an immediate reactor trip and thus limit the increase in temperature and pressure of the reactor coolant system.
LOSS OF CONDENSER VACUUM The loss of condenser vacuum event is identical to a turbine trip which is bounded by the loss of external load. Therefore, a separate analysis is not required for loss of condenser vacuum.
STEAM PRESSURE REGULATOR FAILURE Steam flow to the turbine is controlled by the turbine generator control system.
A malfunction in the control system, which results in zero demand for the turbine, would lead to closure of the turbine control valves and could initiate load rejection. This event is bounded by the analysis of loss of external load, and a separate analyses is not required.
REFERENCES
- 1. San Onofre Nuclear Generating Station Unit 1,.Part II Final Safety Analysis, 1970.
SEP TECHNICAL EVALUATION TOPIC XV-19 LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY SAN ONOFRE UNIT 1
- 1. INTRODUCTION The objective of this review is to assure that the consequences of Loss of Coolant Accidents (LOCA) are acceptable, i.e., that the requirements of the AEC Interi Policy Statement and Appendix K to 10 CFR 50 are met. Loss-of-coolant accidents are postulated accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant make-up system, from piping breaks in the reactor coolant pressure boundary.
The review consists of evaluating the licensee's analysis of the spectrum of loss-of-coolant accidents including break locations, break sizes, and initial conditions assumed, the evaluation model used, failure modes and the acceptability of auxiliary systems used.
II. EVALUATION Assuming the most pessimistic combination of circumstances which could lead to core uncovery and excessive heatup following a loss-of-coolant accident, fuel cladding integrity is ultimately maintained by successful operation of the Emergency Core Cooling System. The Safety Injection System in San Onofre Unit 1 provides the necessary protection to mitigate the consequences of a loss-of-coolant accident. The Safety Injection pumps are in redundant trains
-2 each capable of injecting 7000 gpm at 715 psig. The safety injection systems are automatically initiated on low pressurizer pressure or on high containment pressure signals.
The safety injection pumps discharge to the main feedwater pumps which provide the high volume injection flow. A charging pump is also automatically actuated on safety injection, but the analysis does not take credit for this flow.
Other aspects of post-LOCA performance are addressed under Topics VI-7.B and IX-4.
The licensee has analyzed the performance of the emergency core cooling system (ECCS) in accordance with the Interim Acceptance Criteria (IAC) for Emergency Core Cooling Systems (effective June 29, 1971).
Since San Onofre Unit I uses stainless steel clad fuel, the ECCS.performance analysis under the guidelines of the IAC is acceptable.
The limiting failure is the loss of one Safety Injection train. The break spectrum analysis performed with the Westinghouse evaluation model identified the worst break as a double-ended guillotine breakf (CD = 0.8) in the cold leg.
The highest peak clad temperature (2272 0F) is reached for this break; therefore, small breaks are bounded by the large break analysis.
- Reload Safety Evaluation, San Onofre Nuclear Generating Station Unit 1 Cycle 8 Revision 2, April, 1981.
-3 III.
CONCLUSIONS As part of the SEP review of San Onofre Unit 1, the loss-of-coolant analysis was reviewed against the acceptance criteria of SRP Section 15.6.5 and Section 6.3. The initial conditions relative to single failure, break size and loca tion, power level, and operating conditions have been reviewed and found to conform to the requirements of the SRP.