ML13309A235
| ML13309A235 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/19/1992 |
| From: | Rosenblum R Southern California Edison Co |
| To: | NRC/IRM |
| References | |
| NUDOCS 9205200277 | |
| Download: ML13309A235 (8) | |
Text
Southern Calfornia Eason Company 23 PARKER STREET IRVINE, CALIFORNIA 92718 R. M. ROSENBLUM May 19, 1992 TELEPHONE MANAGER OF (714) 454-4505 NUCLEAR REGULATORY AFFAIRS U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Plant Operation Above Cycle 11 92% Power Limit San Onofre Nuclear Generating Station, Unit 1
References:
- 1. Letter, R. M. Rosenblum (SCE) to NRClDocument Control Desk, "Main Feedwater Line Break Analysis/Reduced Tavg Operation,"
February 15, 1991.
- 2. Licensee Event Report 91-004, "Calculational Error Affecting the Time Interval for Initiation of Hot Leg Recirculation,"
March 18, 1991.
The purpose of this letter is to inform the NRC that on occasion San Onofre Unit 1 (SONGS 1) has been operated slightly above the power limit assumed in a Cycle 11 post-accident environmental assessment. This occurred because we did not proceduralize the power limit assumed in the environmental assessment. We have concluded that the assumed power limit was overly conservative and the periods of plant operation above 92% posed no adverse safety consequences.
Review of Operation at 94.5% Power On February 11, 1992, the plant power level increased to 94.5% reactor thermal power (RTP).
The power increase resulted from increased heat transfer in the steam generators due to the removal of a feedwater heater from service for maintenance purposes. Engineers in our nuclear engineering organization questioned if such operation was consistent with our commitment (Reference 1) to operate with a reduced Reactor Coolant System average coolant temperature, Tavg (which had typically resulted in a reactor power level of approximately 92%); and initiated a review of our Cycle 11 analyses, design calculations, previous LERs, and other licensing documents to confirm the acceptability of operating at 94.5% RTP. Our review focused on documents for which reactor power was an input assumption that could be potentially important to safety.
The documents that were reviewed are listed in Attachment 1.
The review determined that one of our post-accident environmental assessments for Cycle 11 included an analytical limit on reactor power which was 9205200277 920519 PDR ADOCK 05000206 PDR
Document Control Desk
-2 inconsistent with operation at 94.5% RTP.
Specifically, the Long-Term Post Loss of Coolant Accident Containment Pressure/Temperature COTRAN Analysis (SCE Calculation No. DC-3538, Revision 0) assumed the plant would not be operated above 92% RTP during Cycle 11.
This assumption was made in the mistaken belief that there was a procedural limit on reactor power due to steam generator tube plugging. Subsequent investigation determined neither the SONGS 1 operations staff nor our licensing organization were notified about the need to restrict reactor power to not exceed 92%. This review also identified that the plant had operated slightly above 92% power on other occasions.
We have not found any other analyses, design calculations, LERs, or other licensing documents which imposed a reactor power limit or whose results or conclusions would be adversely affected by operation at 94.5% RTP. However, Reference 2 discusses a calculation that was based on actual plant conditions rather than design basis limits which could incorrectly be interpreted as implementing a power restriction.
The calculation assumed a 92% reactor power level and other actual plant conditions to demonstrate that boron precipitation limits were not exceeded under realistic-post loss of coolant accident conditions.
Unfortunately, the wording in the LER concerning the power level assumption could incorrectly be interpreted as implementing an operational restriction on reactor power. This was not the intent of the LER. Rather, the LER should have stated 92% is the nominal power level which results from our administrative control on Tavg and that the power level can vary depending on other plant parameters. However, given the minor nature of this matter, we do not plan on revising LER 91-004 to clarify that no power limit was intended.
Immediate Action to Rectify Operation Above 92% Power Limit Upon confirmation that one of our Cycle 11 analyses had assumed a 92% power limit, we reduced the reactor power below that level (to approximately 90%) on February 13, 1992.
Safety Consequences of Operation Above 92% Power Concurrent with the power reduction, we began a COTRAN reanalysis of the long term containment pressure/temperature response to a LOCA to evaluate the potential consequences of operation above 92% power. The results of the reanalysis demonstrate that the 92% limit assumed in the analysis is overly conservative and that operation up to 95.5% RTP (which corresponds to a theoretical analytical limit of 97.5% RTP to allow for a 2% uncertainty) is acceptable during Cycle 11.
That is, for indicated reactor power levels at or below 95.5%, the calculated containment pressures and temperatures remain within the environmental qualification limits of all equipment and systems located inside the containment. Therefore, past Cycle 11 operation:posed no threat to the health and safety of the public since reactor power never exceeded 95.5% RTP.
Document Control Desk
-3 Reactor Power Administrative Control/License Condition For conservatism, we have reserved an additional 2.5% power margin and, therefore, have now implemented an administrative control on reactor power to not exceed 93% (i.e., 95.5% minus 2.5%).
Our administrative control will remain in force throughout the current fuel cycle. An amendment application to place this power limit in the plant's license will be submitted if plant operation is extended beyond the current fuel cycle.
Root Cause Evaluation As a result of this incident, we have initiated a root cause evaluation to determine why a plant operational restriction was not correctly implemented.
Once completed, this root cause evaluation will be available for the NRC's inspection in our offices.
If you have any questions on this matter, please let me know.
Very truly yours, cc:
J. B. Martin, Regional Administrator, NRC Region V George Kalman, NRC Senior Project Manager, San Onofre Unit 1 J. 0. Bradfute, NRC Project Manager, San Onofre Unit 1 C. W. Caldwell, NRC Senior Resident Inspector, San Onofre Units 1, 2&3
ATTACHMENT 1 Analyses, Design Calculations, Licensee Event Reports, and Other Licensing Documents Reviewed for Reactor Power Limit Applicability Cycle 11 Analyses 1..
SONGS 1 UFSAR Chapter 15, "Accident Analyses," and Chapter 6.2, "Containment Systems."
- 2.
Calculation SEP-19, Revision 8, Supplement A, "Containment Pressure Temperature Transient Analysis," January 12, 1991.
- 3.
Calculation DC-3538, "Long-Term Post-LOCA Containment Pressure Temperature COTRAN Analysis," January 21, 1991.
- 4.
Calculation MSL-NC-02.1, "Containment Analysis for Main Steamline Break," Revision 3, February 28, 1990.
- 5.
Calculation MSL-NC-02.5, "Component Thermal Analysis for Main Steamline Break," Revision 5, January 9, 1991.
- 6.
Letter from H. C. Carlton (Westinghouse) to B. Carlisle (SCE),
"Feedwater Line Break Analysis for Pressure Safety Valve Water Discharge/NUREG-0737 Concerns," October 19, 1990.
- 7.
SCE Report, "Evaluation of Safety Analysis Assumptions for Initial Upper Head Temperature, SONGS 1," D. C. Wood, July 19, 1991.
- 8.
SCE Report, Document 90050, "Station Blackout Analysis, SONGS 1,"
Revision 1.
- 9.
Calculation FCE 1-90-001, "Unit 1 Cycle 11 Core Reload," February, 1991.
Design Calculations
- 1.
Calculation DC-197, "SONGS 1 Long-Term Core Cooling," October 1, 1975.
- 2.
Calculation DC-73, "Hydraulic Transient Flow Analysis,"
December 3, 1975.
- 3.
Calculation SEP-14, "Post Loss of Coolant Accident Sump pH,"
January 22, 1976.
- 4.
Calculation DC-182, "SONGS 1 Post Loss of Coolant Accident Boron Depletion," December 21, 1976.
- 5.
Calculation DC-2, "Spent Fuel Pool Thermal Inertia," August 10, 1976.
- 6.
Calculation DC-177, "Boron Concentration vs. Time for SONGS 1,"
August 4, 1976.
1.1
Design Calculations (continued)
- 7.
Calculation PS035, "Spray and Recirculation System Modification Drawing 714500," November 30, 1975.
- 8.
Calculation 181, "SONGS 1 Boron Inventory," October 15, 1976.
- 9.
Calculation DC-1, "Main Steam Line Break Flooding Level," August 9, 1976.
- 10.
Calculation 180, "Boil-Off Mass Rate at Dual Recirculation Initiation,"
April 14, 1977.
- 11.
Calculation DC-175, "Boron Accumulation Equations," June 2, 1977.
- 12.
Calculation SEP-32, "Containment Loss of Coolant Accident Pressures CONTEMPT," April 11, 1977.
- 13.
Calculation 202, "Hot Leg Recirculation Failure Modes and Effects Analysis," September 29, 1977.
- 14.
Calculation DC-803, "Main Steam Line Break in Containment Secondary System Mass/Energy Releases," February 5, 1981.
- 15.
Calculation DC-912, "Uncertainty Analysis, Reactor Thermal Calibration,"
January 1, 1981.
- 16.
Calculation DC-1211, "Boron Dilution Accident," May 3, 1982.
- 17.
Calculation DC-1275, "MTC Calculations for an Inactive Reactor Coolant Loop," September 22, 1982.
- 18.
Calculation MSL-NC-02.3, "Limited Area Main Steam Line Breaks Pressure/Temperature Analysis," March 18, 1982.
- 19.
Calculation MSL-NC-02.6, "Main Steam Line Break with PORV Blowdown,"
February 8, 1983.
- 20.
Calculation.DC-1365, "Auxiliary Feedwater Tank Volume Requirement,"
March 31, 1983.
- 21.
Calculation MC-422-1, "Hydraulic Transient Analysis of Salt Water Cooling System," May 31, 1984.
- 22.
Calculation DC-1834, "Pressurizer Empty Time," December 23, 1985.
- 23.
Calculation 0310-082-A.3, "SONGS 1 Design Criteria for Environmental Qualification," September 20, 1985.
- 24.
Calculation EQ-NC-03, "Pressure/Temperature Analysis Modeling and Results Outside Containment," November 19, 1985.
- 25.
Calculation EQ-NC-04, "Pressure/Temperature Analysis, Blowdown Outside Containment," November 19, 1985.
1.2
Design Calculations (continued)
- 26.
Calculation EQ-NC-05, "Pressure/Temperature Analysis, Steam Line Breaks Outside Containment," November 19, 1985.
- 27.
Calculation 0310-082-A4, "Environmental Zone Maps Reference Calculation," December 13, 1985.
- 28.
Calculation DC-2484, "Saltwater Cooling Flow Requirements for Loss of Coolant Accident," November 26, 1986.
- 29.
Calculation DC-2483, "Recirculation Heat Exchanger Performance,"
November 21, 1986.
- 30.
Calculation DC-2475, Suppl. A, "Feedwater System Check Valve Leakage Criteria," August 15, 1986.
- 31.
Calculation DC-1318, "Calculation of Safety Injection Termination,"
September 27, 1986.
- 32.
Calculation 310-036-1356, "Recirculation Heat Exchanger, E-11,"
July 30, 1986.
- 33.
Calculation DC-859, "Three Mile Island Review," September 22, 1986.
- 34.
Calculation 01-0310-1321, "High Energy Line Break Environmental Analysis, SONGS 1," June 22, 1987.
- 35.
Calculation DC-2888-1, "Post Safety Injection Cooling," April 18, 1988.
- 36.
Calculation DC-2847, "Charging Pump Horsepower vs. Flow for Safety Injection Recirculation Phase," July 22, 1988.
- 37.
Calculation DC-3088, "Main Feedwater System Hydraulic Calculation,"
December 20, 1988.
- 38.
Calculation DC-2999, "Charging Pump Break Horsepower at Maximum Flow During Injection Phase of Safety Injection," October 4, 1989.
- 39.
Calculation DC-3089, "Recirculation Alignment Criteria," July 10, 1989.
- 40.
Calculation DC-3233, "Updated Main Feedwater Addition for Large Main Steam Line Breaks and Impact on Pipe Break Blowdown," December 19, 1989.
- 41.
Calculation DC-2990, "Loss of Coolant Accident Injection Mode Pump Loads," February 28, 1989.
- 42.
Calculation DC-3130, "RELAP4/MOD5 Analysis, SONGS 1 Main Steam Line Break Superheat Resolution," March 5, 1989.
- 43.
Calculation DC-3127, "Auxiliary Feedwater Turbine Bypass Line Orifice Sizing," February 24, 1989.
1.3
Design Calculations (continued)
- 44.
Calculation DC-3414, "Auxiliary Feedwater System Flow Requirements,"
August 8, 1990.
- 45.
Calculation DC-3380, "Decay Heat 150 Hours After Cycle 11 Shutdown,"
May 17, 1990.
- 46.
Calculation DC-3283, "SONGS 1 PORV Opening Time Calculation,"
April 3Q, 1990.
- 47.
Calculation DC-3351, "Assessment of the Unit 1 Peak Clad Temperature on Delayed Safety Injection," July 31, 1990.
- 48.
Calculation DC-3051, "Containment Analysis, Main Steam Line Break,"
March 12, 1990.
- 49.
Calculation DC-3462, "Refueling Water Storage Tank Technical Specification Requirement," January 24, 1990.
- 50.
Calculation DC-3275, "Post Loss of Coolant Accident Recirculation System Pressure at Check Valve VCC-316," January 21, 1990.
- 51.
Calculation DC-3318, "Safety Injection Pump Adequacy for Main Steam Line Break Long-Term Cooling," May 25, 1990.
- 52.
Calculation DC-3353, "Evaluation of Piping for Anticipated Transient without Scram Pressure Surge," May 11, 1990.
53..
Calculation DC-3177, "Temperature Rise in Compartments During Station Blackout," November 6, 1990.
- 54.
Calculation DC-859, "Auxiliary Feedwater System," March 22, 1990.
- 55.
Calculation SEP-19, "Containment Pressure/Temperature Transient Analysis," January 12, 1991.
- 56.
Calculation DC-3273, "Core Boron Accumulation Control Post Loss of Coolant Accident," February 20, 1991.
- 57.
Calculation DC-3466, "Safety Injection Delay Times," February 28, 1991.
- 58.
Calculation DC-3476, "Steam Generator Blowdown Line Break Inside the Sphere Enclosure Building," February 4, 1991.
- 59.
Calculation DC-3547, "Small Break Loss of Coolant Accident NOTRUMP Input Parameters," February 1, 1991.
- 60.
Calculation DC-3601, "SONGS 1 Reactor Coolant System Inventory Analysis for Station Blackout," July 23, 1991.
- 61.
Calculation DC-3605, "Post Accident Thermal Analysis of the Amphenol Electrical Penetrations at SONGS 1," July 29, 1991.
1.4
Design Calculations (continued)
- 62.
Calculation DC-3521, "SONGS 1 NIS Overpower Trip and Rod Stop Setpoints," January 24, 1991.
- 63.
Calculation SEP-20, "pH Control," February 20, 1991.
- 64.
Calculation MSL-NC-02.1, Supplement D, "Containment Pressure/Temperature Transient Analysis (Main Steam Line Break)," January 10, 1991.
- 65.
Calculation DC-3480, "SONGS 1 Containment Spray Iodine Removal Efficiency," January 16, 1991.
License Event Reports
- 1.
All LERs issued during Fuel Cycles 9, 10, and 11.
Other Licensing Documents
- 1.
Letter from R. M. Rosenblum (SCE) to NRC, Docket 50-206, "Underestimation of Refill Volume Assumed in Large Break Loss of Coolant Accident Analysis," March 29, 1991.
- 2.
Letter from R. M. Rosenblum (SCE) to NRC, Docket 50-206, "Volume Differences Identified in Accident Analyses," May 10, 1991.
- 3.
Letter from R. M. Rosenblum (SCE) to NRC, Docket 50-206, "Evaluation of Volume Differences in Accident Analysis, SONGS 1,"
May 17, 1991.
- 4.
Letter from M. 0. Medford (SCE) to NRC, Docket 50-206, "Fire Protection Program Review," May 21, 1985.
1.5