ML13331A897
| ML13331A897 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/12/1986 |
| From: | Baskin K, Cotton G San Diego Gas & Electric Co, Southern California Edison Co |
| To: | |
| Shared Package | |
| ML13331A896 | List: |
| References | |
| TAC-62164 NUDOCS 8611140379 | |
| Download: ML13331A897 (27) | |
Text
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON
)
COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY )
for a Class 104(b) License to Acquire,
) DOCKET NO. 50-206 Possess, and Use a Utilization Facility as
)
Part of Unit No. 1 of the San Onofre Nuclear ) Amendment No. 138 Generating Station
)
SOUTHERN CALIFORNIA EDISON COMPANY and SAN DIEGO GAS & ELECTRIC COMPANY, pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 138.
This amendment consists of Proposed Change No. 165 to Provisional Operating License No. DPR-13. Proposed Change No. 165 modifies the Technical Specifications incorporated into Provisional Operating License No. DPR-13 as Appendix A.
Proposed Change No. 165 is a request to revise Technical Specification Section 2.1 Reactor Core -
Limiting Combination of Power, Pressure, and Temperature of Appendix A.
In the event of conflict, the information in Amendment Application No. 138 supersedes the information previously submitted.
36P1140379D
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-2 Based on the safety analysis provided in the Description of Proposed Change and Safety Analysis, it is concluded that (1) the proposed change does not involve an unreviewed safety question as defined in 10 CFR 50.59, nor does it present significant hazards considerations not described or implicit in the Final Safety Analysis, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
Pursuant to 10 CFR 170.12 the fee of $150 is herewith remitted.
I
-3 Subscribed on this day of
/_/_'.
Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPANY By__
Subscrj bed and sworn to before me this
/A L
day of
//4 9
OFFICIAL SEAL AGNES CRABTREE Notary Public-Caifornia LOS ANGELES COUNTY My Comm. Exp. Sep. 14, 19D0 Nogr Pulic in and for the County of Los Angeles, State of California My Commission Expires: 7////0 Charles R. Kocher James A. Beoletto Attorneys for Southern California Edison Company By
-4 Subscribed on this 6th_ day of November 1986 Respectfully submitted, SAN DIEGO GAS & ELECTRIC COMPANY By G--.' Cotton Senior Vice President Electric Operations Subscribed and sworn to before me this day of
/5 t
4ary Publ,' in a fo the County of San Diego, 'tate of Ca j ornia My Commission Expires:
/
OFRL SEM JILLQUIGLEY David R. Pigott NOTARY PUBUC-CAWORNIA Samuel B. Casey PRINIPALOFFIE
- INOrrick, Herrington & Sutcliffe SON DIEGO COUMY My Commission EL March 7,98 Attorneys for San Diego Gas & Electric Company By
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN
)
CALIFORNIA EDISON COMPANY
)
and SAN DIEGO GAS & ELECTRIC
)
Docket No. 50-206 COMPANY (San Onofre Nuclear
)
Generating Station Unit No. 1
)
CERTIFICATE OF SERVICE I hereby certify that a copy of Amendment No. 138 was served on the following by deposit in the United States Mail, postage prepaid, on the 12th day of November
, 1986.
Henry 3. McGurren, Esq.
Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C. 20545 David R. Pigott, Esq.
Samuel B. Casey, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P. 0. Box 60860, Terminal Annex Los Angeles, California 90060 Michael L. Mellor, Esq.
Thelen, Marrin, Johnson & Bridges Two Embarcadero Center San Francisco, California 94111 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102
James McGuffin Western Regional Manager Westinghouse Electric Corporation Post Office Box 2728 Pittsburgh, Pennsylvania 15230 A. I. Gaede 23222 Cheswald Laguna Nigel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555
DESCRIPTION OF PROPOSED CHANGE NO. 165 AND SAFETY EVALUATION This is a request to revise Appendix A, Technical Specification 2.1 Reactor Core - Limiting Combination of Power, Pressure, and Temperature Description The failure of pressure transmitter PT-459 on July 30, 1986 resulted in the identification of a single failure which could render the Steam/Feedwater Flow Mismatch Reactor Trip inoperable. Since this trip is credited for the limiting transients used to establish minimum auxiliary feedwater flow requirements, it became necessary to reanalyze the affected events to determine if the acceptance criteria could still be met assuming that the Steam/Feedwater Flow Mismatch Reactor Trip was not available due to a single failure.
The reanalyzed events were the Loss of Normal Feedwater (LONF) and Feedwater Line Break (FLB).
The preliminary results of analyses were provided to the NRC by letter dated August 21, 1986. The final results are provided as.
The analyses of Attachment 1 indicated that for the FLB, the acceptance criteria previously used for this event continue to be met since the peak pressure is maintained within acceptable limits and the core remains coolable. For the LONF transient however, the Pressurizer High Level Trip setpoint must be reduced below the normal 70% level setpoint to 50% in order to meet the acceptance criterion which prohibits the pressurizer filling with water for this event. The maximum setpoint allowed by Technical Specification 2.1 must therefore be reduce to 50% while the Steam/Feedwater Flow Mismatch Trip is susceptible to a single failure.
In order to provide for plant operation with normal setpoints once modifications are completed to eliminate the single failure problem with the mismatch trip, the proposed change includes a conditional statement for the maximum Pressurizer High Level Reactor Trip setpoint and allows the restoration of the 70% setpoint. A description of these modifications will be provided to the NRC by February 27, 1987 as indicated in the SCE letter dated October 31, 1986 regarding this subject.
Technical Specification 2.1 is revised to indicate the 50% (20.8 ft.) level setpoint when the Steam/Feedwater Flow Mismatch Reactor cannot be credited and 70% (27.3 ft.) when the trip is restored to conform with its design basis.
The asterisked footnotes are also reorganized as an editorial change.
Existing Technical Specifications Refer to Attachment 2
Proposed Technical Specifications Refer to Attachment 3 Safety Evaluation Proposed Change No. 165 is determined not to constitute a significant hazards consideration based on the following review questions and responses.
- 1. Ouestion Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No. This proposed change does not significantly increase the probability or consequences of an accident previously evaluated. As discussed in, the reanalysis of the events affected by the failure of PT-459, demonstrates that for the LONF event reducing the Pressurizer High Level Reactor Trip setpoint to 50% will result in a system response which is within the previously established acceptance criteria. This change does not affect the system response to a Feedwater Line Break event. In addition, the change does not impact the probability of occurrence of the affected transients since the reduction of the trip setpoint is not related to any of the event initiators such as loss of offsite power, feedwater system equipment failures and pipe break.
- 2. Ouestion Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident?
Response
No.
The proposed change does not create the possibility of a new or different kind of accident. The change will only lower the initiation setpoint of a safety trip system to provide for earlier initiation of the trip during a transient. The change is a move in the conservative direction with respect to this safety function.
- 3. Ouestion Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response
No. Operation of the facility in accordance with this proposed change does not involve a significant reduction in a margin of safety. The reduction in the Pressurizer High Level Trip setpoint to 50% results in a transient system response for the affected event (LONF) which is within the previously established acceptance criteria for the event and therefore equivalent to the previously analyzed margin of safety.
This change is similar to example (ii) of the "Examples of Amendments that are Not Likely to Involve Significant Hazards Considerations" as published in 48 FR 14864 dated April 6, 1983.
Example (ii) states: a change that constitutes an additional limitation, restriction or control not presently included in the technical specifications. The proposed specification is similar to the example because it provides a more conservative limitation on a reactor safety system setpoint.
Safety and Significant Hazards Determination Based on the safety analysis, it is concluded that:
- 1. The proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92;
- 2. There is reasonable assurance that the health and safety of the public will not been endangered by the proposed change; and
- 3. This action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Environmental Statement 00 ATTACHMENT 1 Loss of Normal Feedwater And Feedwater Line Break Reanalysis San Onofre Unit 1 BACKGROUND The loss of Normal Feedwater (LONF) event and the Feedwater Line Rupture (FLB) event were recently analyzed in the Spring of 1986 for Southern California Edison's (SCE) SONGS 1, Reference 1. This analysis showed that for the LONF with loss of reactor coolant pumps (RCPs), an auxiliary feedwater (AFW) flow of 165 gpm to 3 steam generators initiated 30 minutes after reactor trip provides acceptable results. An AFW flow of 250 gpm to 2 steam generators initiated 20 minutes after reactor trip is adequate to remove decay heat for the FLB with loss of RCPs event. The above conclusions were based on the assumption that reactor trip was provided by the steam flow/feed flow mismatch signal.
Due to concerns raised by SCE regarding the availability of the steam flow/feed flow mismatch reactor trip, SCE has requested Westinghouse to reanalyze the LONF and the FLB events assuming the steam flow/feed flow mismatch reactor trip is not available.
ANALYSIS As in the 1986 LONF and FLB analysis, the LOFTRAN code is used to simulate the transients. The reanalysis consisted of two cases to determine the impact of no steam flow/feed flow mismatch reactor trip for the LONF and FLB events.
Protection is expected to be provided by the high pressurizer pressure, the high pressurizer water level, or the variable low pressurizer pressure reactor trip.
The assumptions modelled in this reanalysis are the same assumptions used in the Spring 1986 analysis except as noted below.
- 1. A high pressurizer water level reactor trip setpoint of 50% NRS for the LONF and 70% NRS for the FLB are assumed with a delay time of 2 seconds.
- 2. A high pressurizer pressure reactor trip setpoint of 2260 psia (including uncertainties) is assumed with a delay time of 2 seconds.
- 3. The delay time of the variable low pressurizer pressure reactor trip is assumed to be 2.32 seconds.
- 4. An initial pressurizer water level of 37.5% NRS is modelled which corresponds to the programmed pressurizer level at full power and reduced average temperature.
- 5. For this reanalysis, the unavailability of the steam flow/feed flow reactor trip is assumed to be the single failure. As such, the turbine driven AFW pump is available. The turbine driven AFW pump, which delivers a total of 165 gpm to 3 steam generators, is modelled only for the LONF event. The steam pressure during the FLB event is not expected to be sufficient for the turbine driven AFW pump. A delay of 3 minutes following a low steam generator water level signal is assumed for initiation of the turbine driven AFW pump.
For the FLB event, a delay of 15 minutes following a low steam generator water level is assumed for initiation of the motor driven AFW pump followed by a line fill time of approximately 5 minutes. The motor driven AFW pump delivers a total of 250 gpm to 2 steam generators.
RESULTS The results of the transient analysis are shown in the attached figures.
Loss of Normal Feedwater The results show that lowering the high pressurizer water level reactor trip setpoint to 50% prevents the pressurizer from filling. Reactor trip occurs on pressurizer high level at 65.3 seconds. Peak pressurizer volume of 1092 cubic feet occurs at 1729 seconds.
Feedwater Line Break In order to determine the worst case peak pressure and core cooling response, the event is modelled both with and without steam dump valve actuation. The plots provided are for the case with steam dump.
The results show that the pressurizer fills and hot leg boiling occurs.
Reactor trip occurs on the existing variable low pressure trip at 24.4 seconds. Peak pressurizer pressure of 2204 psia occurs at 1151 seconds.
Peak pressurizer volume of 1316 cubic feet occurs at 1145 seconds. Detailed calculations showed that the mass relieved through the PORV's between the time of bulk boiling in the primary system and the time that the heat removal capability of the AFW exceeds the core decay heat (turnaround) was not sufficient to uncover the core. As such, the ultimate acceptance criteria for a FLB that the core remains coolable during the transient is shown to be met.
CONCLUSIONS The results of the analysis show that for the LONF with loss of RCPs event, an AFW flow of 165 gpm to 3 steam generators with a 3 minute delay in initiating the turbine driven AFW pump and a pressurizer high water level reactor trip setpoint of 50% NRS provide adequate protection to prevent the pressurizer from filling.
For the FLB with loss of RCPs event, an AFW flow of 250 gpm to 2 steam generators with a 15 minute delay in initiating the motor driven AFW pump and the existing variable low pressure trip are adequate to remove decay heat and keep the core coolable. Hence, the change in the high pressurizer water level trip is required only for the LONF event.
REFERENCE M. 0. Medford (SCE) to G. E. Lear (NRC) letter dated May 1, 1986 SCE STATION LACOUT ANALYSIS
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ATTACHMENT 2 Existing Technical Specifications Affected by Proposed Change No. 165 San Onofre Nuclear Generating Station Unit 1