ML13331A886

Forwards Proposed Change 165 to Tech Specs,Requiring That Lower Pressurizer Level Trip Setpoint Be Maintained Until Mods to Mismatch Trip Completed,Per NRC 860813 Request. Formal Amend Application Will Be Submitted Later

ML13331A886
Person / Time
Site:	San Onofre
Issue date:	08/21/1986
From:	Medford M Southern California Edison Co
To:	Lear G Office of Nuclear Reactor Regulation
Shared Package
ML13331A887	List: ML13331A886 ML13331A888
References
TAC-44652, TAC-62164 NUDOCS 8608250043
Download: ML13331A886 (5)

Text

REGULATORNFORMATION DISTRIBUTION

• TEM (RIDS)

ACCESSION NBR:8608250043 DOC.DATE: 86/08/21 NOTARIZED: NO DOCKET #

FACIL:50-206 San Onofre Nuclear Station, Unit 1, Southern Californ 05000206 AUTH.NAME AUTHOR AFFILIATION MEDFORD,M.O.

Southern California Edison Co.

RECIP.NAME RECIPIENT AFFILIATION LEARG.E.

PWR Project Directorate 1

SUBJECT:

Forwards Proposed Change 165 to Tech Specsurequiring that lower pressurizer level trip setpoint be maintained until mods to mismatch trip completedper NRC 860813 request.

Formal amend application will be submitted later.

DISTRIBUTION CODE: AOO1D COPIES RECEIVED:LTR I ENCL 1 SIZE:

TITLE: OR Submittal: General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109.

05000206 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PWR--A EB 1

1 PWR-A EICSB 2

2 PWR-A FOB 1

1 PWR-A PD1 LA 1

0 PWR-A PD1 PD 01 5

5 DUDLEYR 1

1 PWR-A PSB 1

1 PWR-A RSB 1

1 INTERNAL: ADMILFMB 1

0 1

0 NRR/ORAS 1

0 04 1

1 RGN5 1

1 EXTERNAL: EG&G BRUSKE,S 1

1 LPDR 03 1

1 NRC PDR 02 1

1 NSIC 05 1

1 TOTAL NUMBER OF COPIES REGUIRED: LTTR 22 ENCL 18

Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 M. 0.

MEDFORD TELEPHONE MANAGER. NUCLEAR LICENSING (618) 302-1749 August 21, 1986 Director, Office of Nuclear Reactor Regulation Attention: G. E. Lear, Director PWR Project Directorate No., 1 U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

Subject:

Docket No. 50-206 Proposed Change No. 165 San Onofre Nuclear Generating Station Unit 1 The failure of pressure transmitter PT-459 was reported to the NRC, (by red phone) pursuant to 10 CFR 50.72 on July 29, 1986. The SCE review"of this failure and its consequences has identified a deficiency in the design, of the Steam/Feedwater Flow Mismatch Reactor Trip System. This deficiency has been the subject of several telephone discussions with the NRC staff during the past two weeks. The identified design deficiency can, due to a single failure, prevent the mismatch trip from occurring when required. This single failure therefore impacts the safety analyses for the Loss of Main Feedwater and Feedline Break transients since credit is taken for this trip in these two events.

As previously discussed with the NRC staff, immediate compensatory measures were instituted following identification of the problem. In addition, the affected safety analyses were reanalyzed without credit for the mismatch trip. The results of these analyses required that the Pressurizer High Level Trip setpoint be lowered to 50% (for Loss of Main Feedwater only) from the then existing 70% level in order to provide a transient response within the acceptance criteria previously committed. These analyses are provided as Enclosure 1. The lower pressurizer level trip is at the present time required by administrative controls, to be maintained until corrective actions are completed to assure that the mismatch trip is single failure proof.

During the telephone discussion of August 13, 1986, the NRC staff requested that the administrative controls be formalized by submitting a revision to Technical Specification 2.1 which required that the lower,

pressurizer level trip setpoint be maintained until modifications to the mismatch trip are completed. Accordingly, Enclosure 2 providis A preliminary copy of Proposed Change No. 165 to the SONGS 1 Technical Specifications.

PPDR 60LOB250043 86082 PDR

e 0

Mr. G.

August 21, 1986 This preliminary copy is provided in conjunction with the safety analyses of in order to assist the NRC staff in their request to expedite the issuance of this proposed change. The formal submittal in the form of a License Amendment Application will be provided shortly, following the required safety review and approval.

If you have any questions or desire additional information, please contact me.

Very trul yo rs, cc:

Mr. A. E. Chaffee, Regulatory Projects Branch, Region V Mr. F. R. Huey, NRC Senior Resident Inspector, SONGS 1, 2 and 3

ENCLOSURE 1 Loss of Normal Feedwater And Main Feedline Rupture Reanalysis San Onofre Unit 1 BACKGROUND The loss of Normal Feedwater (LONF) event and the Main Feedline 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.

Feedline Break 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 0018P