ML13317A146
| ML13317A146 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/28/1983 |
| From: | Paulson W Office of Nuclear Reactor Regulation |
| To: | Dietch R Southern California Edison Co |
| References | |
| TAC-62079, TASK-06-10.A, TASK-6-10.A, TASK-RR LSO5-83-04-070, LSO5-83-4-70, NUDOCS 8305030414 | |
| Download: ML13317A146 (5) | |
Text
April 28, 1983 Docket No. 50-206 LS05-83-04-070 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:
SEP TOPIC VI-lO.A, TESTING OF REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES, INCLUDING RESPONSE-TIME TESTING, FINAL SAFETY EVALUATION REPORT FOR THE SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 By letter dated August 28, 1981, the staff issued a contractor Technical Evaluation Report and a draft safety evaluation report on this topic.
The safety evaluation report has been revised to reflect the comments provided by letters from R. W. Krieger to D. M. Crutchfield dated February 28, 1983, March 18, 1983 and April 6, 1983. The staff final safety evaluation report is enclosed.
This evaluation recommends modifications to the Technical Specifications and any necessary equipment to implement response-time testing for all reactor protection systems (including engineered safety features).
The need to actually implement these changes will be determined'during 5
the integrated safety,ssessment. -This topic ;Assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is )DS&
usk't completed.
Sincerely, 0:~jjjsgnea
~by~w54 Walter A. Paulson, Project Manager 8305030414 830428 Operating Reactors Branch No. 5 PDR ADOCK 05000206 Division of Licensing PDR
Enclosure:
As stated cc w/enclosure:
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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960
Mr. R. Dietch, Vice Pres t
Docket No.
50-206 Nuclear Engineering and Operations San Onofre 1 Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770 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 Post Office Box 1831 San Diego, California 92112 Resident Inspector/ San Onofre NPS c/o U.S. Nuclear Regulatory Commission Post Office Box 4329 San Clemente, California 92672 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 Robert H. Engelken, Regional Administrator U.S. Nuclear Regulatory Commission, Region V 1450 Maria Lane Walnut Creek, California 94596
SYSTEMATIC EVALUATION PROGRAM TOPIC VI-10.A SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 TOPIC: VI-10.A, Testing of Reactor Trip System and Engineered Safety Features, Including Response-Time Testing I.
INTRODUCTION The purpose of this topic is to review the reactor trip system (RTS) and engineered safety features (ESF) test program for verification of RTS and ESF operability on a periodic basis and to verify RTS and ESF response time in order to assure the operability of the RTS and ESF.
Response times should not exceed those assumed in the plant accident analyses. Accordingly, the test program of the RTS and ESF was reviewed in accordance with the Standard Review Plan, including applicable Branch Technical Positions.
II. REVIEW CRITERIA The review criteria are presented in Section 2 of EG&G Report 0487J, "Testing of Reactor Trip System and Engineered Safety Features, Including Response-Time Testing" (Enclosure 1 to Reference 1).
III.
RELATED SAFETY TOPICS AND INTERFACES Topic VI-7.A.3 discusses the question of testing protection systems under conditions as close to design conditions as practical. There are no topics that are dependent on the present topic information for their completion.
IV. REVIEW GUIDELINES Review guidelines are presented in Sections 3 and 4 of Report 0487J.
V.
EVALUATION The Technical Specifications for San Onofre Unit No. 1 were compared with the Standard Technical Specifications for current pressurized water reactor licensing. It was found that, for the reactor trip system, one signal (loss of reactor coolant flow) is not subjected to a channel functional test as frequently as required in the Standard Technical Specifications; seven channels are not checked, tested or calibrated as required as in the Standard Technical Specifications; and several channels that are part of the RTS for the Standard Technical Specifications are not part of the San Onofre Unit 1 RTS. Additionally, the channel response time between channel trip and the operation of the reactor trip relay is not required to be tested.
-2 For the Containment Spray System, selected as typical of the ESF systems, surveillance requirements were non-existent for systems that are required to operate in support of the containment spray system (such as component cooling water).
In References 2 through 4, the licensee provided information on surveil lance and testing that is governed by plant procedures but not necessarily in the Technical Specifications. These procedures establish testing requirements at frequencies consistent with Standard Technical Specifica tion requirements. Support systems (such as component cooling and salt water cooling) are subject to periodic surveillance and inservice testing by plant procedures. The test frequency for the reactor coolant flow signal was approved by the staff in a licensee amendment dated October 20, 1971, based on plant specific operating experience.
Control rod operability, control rod drop time and operability of reactor trip system components is verified periodically by plant procedures.
Control rod drop time and operability requirements are also in the Technical Specifications. However, response time of individual components is not measured. Response-time testing of other engineered safety systems actuation (diesel generator start, sequencer operation, containment isolation) is not required to be measured in the Technical Specifications, although operability surveillance and functional testing requirements are established there.
VI.
CONCLUSION The staff recommends that the design of systems which are required for safety should include provisions for periodic verification that the minimum performance of instruments and controls is not less than that which was assumed in the safety analyses and should be incorporated in the plant Technical Specifications. The bases for this conclusion are the provisions of General Design Criterion 21, Section 3.9 of IEEE Std. 279-1971 and IEEE Std. 338-1977. Consequently, the licensee should amend the Technical Specifications to correct for the other omissions in our contractor's report by including the test frequencies now specified by plant procedures. In addition, the licensee should implement a program for response-time testing of all reactor protection systems (including engineered safety features systems such as containment isolation).
-3 VII.
REFERENCES
- 1. Letter from D. M. Crutchfield (NRC) to R. Dietch (SCE),
dated August 28, 1981.
- 2.
Letter from R. W. Kreiger (SCE) to D. M. Crutchfield (NRC),
dated February 28, 1983.
- 3. Letter from R. W. Kreiger (SCE) to D. M. Crutchfield (NRC),
dated March 18, 1983.
- 4. Letter from R. W. Kreiger (SCE) to D. M. Crutchfield (NRC),
dated April 6, 1983.