ML13330B166
| ML13330B166 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 03/11/1987 |
| From: | Medford M, Palmer D Southern California Edison Co |
| To: | NRC/IRM |
| Shared Package | |
| ML13311A415 | List: |
| References | |
| NUDOCS 8703160001 | |
| Download: ML13330B166 (4) | |
Text
Southem California Edison Company P. 0.
BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 M. 0. MEDFORD TELEPHONE MANAGER OF NUCLEAR ENGINEERING (818) 302-1749 AND LICENSING March 11, 1987 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Pressure Transmitter 459 Failure RPS Single Failure Analysis San Onofre Nuclear Generating Station Unit 1 By letter dated September 23, 1986, the NRC requested that we perform a review of the San Onofre Unit 1 Reactor Protection System (RPS) to determine conformance to its design basis criteria related to the single failure criterion and control/protection system interactions. This NRC request resulted from the failure of a pressure transmitter (PT-459) which caused the inoperability of the Steam/Feedwater Flow Mismatch Trip (mismatch trip).
The review has been completed and a report which documents the evaluation is enclosed for your use.
The results of the evaluation determined that the RPS meets its design basis criteria with the exception of the mismatch trip as previously identified. It was also determined that the mismatch trip design contains additional shortcomings which involve components other than PT-459. These problems result from the steam and feedwater flow analog amplifier design which includes channel-common signal path and power supply configurations and from the spatial distribution of inputs to the RPS for loop specific events.
These additional design deficiencies do not impact safe operation since the affected safety analyses have already been re-evaluated and produced acceptable system response considering the mismatch trip as a potential single active failure, and with the Pressurizer Level Trip reduced to the present value of 50%.
The principal conclusion which results from the review of the RPS is that in order to provide the mismatch trip with the capability to perform its design function assuming a single failure of one of its components, significant modifications must be made which correct design problems other than those previously identified as associated with PT-459. Our design efforts to date have concentrated on the resolution of the PT-459 problem and we have therefore not identified the preferred approach for resolving the newly identified concerns.
8703160001 870311 PDR ADOCK 05000206 P
PDR._
Document Control Desk March 11, 1987 The preferred approach for resolving the single failure concerns associated with the mismatch trip cannot be developed at this time because any approach would be impacted by the ongoing redesign of the auxiliary feedwater system (AFWS).
As you are aware, the AFWS will be modified at the next refueling outage to provide for a single failure proof three pump system which will be available to respond during the transients and accidents for which the AFWS is used as the emergency recovery system and which form the design basis events. These same design basis events also determine the required reactor trips and the redundancy requirements for the mismatch trip if credit is taken for this trip as a safety function. As previously noted, the mismatch trip is required if the single active failure is the turbine driven auxiliary feedwater pump, but is backed up by the reduced set point Pressurizer Level Trip should the mismatch trip be the single failure. Modification of the mismatch trip to provide a single failure proof capability is therefore required if the setpoint is raised to the higher setting of 70% unless it can be demonstrated by safety analysis that the design basis events result in acceptable consequences without the implementation of any related modifications to the mismatch trip. With the improvements that are planned for the AFWS, it may be possible to obtain acceptable results without modifying the mismatch trip. It is therefore the AFWS redesign effort which controls the schedule for determining the preferred approach to resolving the mismatch trip concerns. It is expected that the AFWS design and the design basis event reanalyses can be completed in time to support a submittal of information by May 31, 1987. This date coincides with the date provided for submittal of the additional information requested by the NRC as described in our letter of October 31, 1986. It is our intent, therefore, to provide the NRC staff, by May 31, 1987, with the design descriptions, if they are still applicable, for the mismatch trip as well as any implementation schedules as requested in the NRC letter of September 23, 1986.
If you have any questions or desire additional information, please contact me.
Very truly yours, cc:
J. B. Martin, Regional Administrator, NRC Region V F. R. Huey, NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3 R. F. Dudley, NRC Project Manager
Southern California Edison Company SAN ONOFRE NUCLEAR GENERATING STATION P.O. BOX 128 SAN CLEMENTE, CALIFORNIA 92672 February 27, 1987 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555 Gentlemen:
Subject:
Docket Nos. 50-206, 50-361 and 50-362 Revised Emergency Plan Implementing Procedures San Onofre Nuclear Generating Station, Units 1, 2 and 3 Pursuant to 10 CFR 50, Appendix E, Section V, enclosed are two copies of revised Emergency Plan Implementing Procedures:
PROCEDURE REVISION -;TCN TITLE S0123-VIII-0.302 TCN 0-4 Annual Surveillance of the Onsite Emergency Evacuation System (OEES)
S0123-VIII-40.100 TCN 0-1 Source Term and Dose Assessment
DOCUMENT CONTROL DESK
-2 For your convenience, we have enclosed an updated index (Enclosure I) to replace the previous submittal listing the titles, latest revisions and TCN's to all of the San Onofre Nuclear Generating Station Emergency Plan Implementing Procedures. The procedures included in this transmittal are indicated on Enclosure I by a bar in the margin.
Sincerely, W. G. ZINTL
- MANAGER, COMPLIANCE Enclosures cc: F. R. Huey, (USNRC Senior Resident Inspector, Units 1, 2 and 3)
J. B. Martin, (Regional Administrator, Region V)
(2 copies of all Enclosures)