ML20054M866
| ML20054M866 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/31/1982 |
| From: | Ransom C EG&G, INC. |
| To: | Lantz E Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6436 EGG-EA-5864, TAC-06836, TAC-6836, NUDOCS 8207150064 | |
| Download: ML20054M866 (14) | |
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a u f u e-c h d a Ti L.c a ifm s k n epn' May 1982 TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE pris PROTECTION SYSTEli FOR SAN ON0FRE NUCLEAR GENERATING ggfg PAI STATION, UNIT NO. 1 6fd ef
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C. B. Ransom U.S. Department of Energy idaho Operations Office
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Report No E.GG.EA-5864
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Contract Program or Project
Title:
Steam Generator Transients and Operating Reactors Evaluation for Reac. tor Systems Branch Subject of this Documents
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Technical Evaluation Report of the Overpressure Protection System for San Onofre Nuclear Generating Station, Unit No. 1 Type of Document:
Informal Report Author (s):
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C. B. Ransom Date of Document:
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May 1982 y
Responsible NRC Individual and NRC Of' ice or Division:
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E. Lantz, Divisie. of Systems Integration
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This docurnent was prepared primarily for preliminary or internal use. it has nn', received
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full review and approval. Since there may be substantive changes, this document should not be considered final.
s EG&G Idaho. Inc, Idaho Falls, Idaho 83415 7
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U.S. Nudear Regulatory Commission ~
4 Washington, D.C.
Under DOE Contract No. DC AC07-78tD01570 NRC FIN N'o. ' Ab4J6
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TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE PROTECTION SYSTEM i e FOR SAN ONOCRE NUCLEAR GENERATING STATION, UNIT NO. 1 May 1982 C. B. Ransom Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.
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ABSTRACT This report documents the technical evaluation of the low temperature overpressure protection system of the San Onofre Nuclear Generating Station, Unit 1.
The criteria used to evaluate the acceptability of the l
system are those criteria contained in NUREG-0224 as appended by the Branch Technical Position (RSB 5-2).
FOREWORD This report is supplied as part of the " Steam Generator Transients and Operating Reactors Evaluation for Reactor Systems Branch" being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Systems Integration, by EG&G Idaho, Inc.,
Reliability and Statistics Branch.
The U.S. Nuclear Regulatory Commission funded the work under the authorization, B&R 20-19-01-22, FIN A6436.
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s CONTENTS
1.0 INTRODUCTION
1 2.0 DESIGN CRITERIA.................................................
1 3.0 SYSTEM DESCRIPTION AND EVALUATION...............................
1 e
3.1 Air Supply................................................
2 3.2 Electrical Controls.......................................
2 3.3 Testability...............................................
2 3.4 Single Failure Criteria...................................
3 3.5 Seismic Design............................................
3 3.6 Analysis Results..........................................
4 3.6.1 Mass Input Case...................................
4 3.6.2 Heat Input Case...........
6 4.0 ADMINISTRATIVE CONTROLS.........................................
7
5.0 CONCLUSION
S.....................................................
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6.0 REFERENCES
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TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE PROTECTION SYSTEM FOR SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. I
1.0 INTRODUCTION
Several instances of reactor vessel overpressurization have occurred in pressurized water reactors in which the technical specifications implement-ing Appendix G to 10 CFR Part 50 have been exceeded.
The majority of cases have occurred during cold shutdown while the primary system was in a water-solid condition.
By letter to the Southern California Edison Company (SCE),
owner and operator of the San Onofre Nuclear Generating Station, dated August 11,1976 (Ref.1), the U.S. Nuclear Regulatory Commission (NRC) requested an evaluation of San Onofre Unit 1 to determine susceptibility to overpressurization events and an analysis of these possible events, and required SCE to propose interim and permanent modifications to the systems and procedures to reduce the likelihood and consequences of such events.
SCE participated as a member of a Westinghouse owner's group which pro-vided a reference mitigating system and analyses to verify the adequacy of the system (Ref. 10).
SCE modified the reference mitigating system and pro-posed their Overpressure Mitigating System (OMS) along with administrative procedure modifications and operator training (Ref. 8, 9,11 and 15).
The OMS is designed to mitigate the consequences of an overpressurization event and the additional operator training and the administrative procedure modifications are intended to reduce the probability of the occurrence of an overpressurization event.
This is a report of the evaluation of the compliance of the licensee's Overpressure Mitigating System with the design criteria established by the NRC.
2.0 DESIGN CRITERIA The NRC formally addressed reactor vessel overpressurization in August 1976, and requested that the utilities provide a solution to the problem.
The design criteria were subsequently identified through meetings and correspondence with utility representatives.
NUREG-0224, " Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors" with appended Branch Technical Position (RSB 5-2) formalizes the staff require-ments for the overpressure mitigating system.
This NUREG also includes a thorough discussion of the background of this problem and technical dis-cussions pertaining to vessel stresses and other aspects of vessel overpressurization.
3.0 SYSTEM DESCRIPTION AND EVALUATION The San Onofre Unit 1 OMS consists of two separate trains, each con-taining a power-operated relief valve (PORV), an isolation valve and associ-ated circuitry. When in the low pressure mode the system provides a low 1
pressure setpoint of 522 psig for both PORV trains. When the system is enabled, it will terminate all analyzed pressure transients below the Appen-dix G limit by automatically opening the PORVs. A manual switch is used to enable and disable the low setpoint of each relief valve. An enabling alarm which monitors system pressure is provided to alert the control room opera-tor to enable the overpressure mitigating system when system pressure drops to a predetermined point.
In addition, an alarm is provided in the control room to indicate when an overpressure transient is occurring.
3.1 Air Supply The power-operated relief valves (PORVs) are spring-loaded-closed, air required-to-open valves, which are normally supplied by plant instrument air.
To assure operability of the valves upon loss of plant instrument air a redundant pneumatic source is automatically provided.
The backup source is the Station Nitrogen System which is independent such that a failure of the Instrument Air System will not result in a loss of nitrogen to the PORVs.
Similarly, a failure in the Station Nitrogen System will not affect the air supply to the valves.
Either system provides sufficient pneumatic capacity to handle PORV cycling for the duration of a pressure transient with the number of relief cycles conservatively estimated at 45.
Both sys-tems have a pressure switch monitoring header pressures which provide alarms in the control room to notify the operator if instrument air or station nitrogen supply is unavailable to the PORVs.
3.2 Electrical Controls The electrical, instrumentation, and control system aspects of the San Onofre Nuclear Generating Station, Ur.it I low temperature overpressure pro-tection system have been reviewed and reported in a separate technical evaluation (Ref. 19).
3.3 Testability The staff position requires that a test be performed to assure oper-ability of the system electronics prior to each shutdown and that a test for valve operability, as a minimum, be conducted as specified in the ASME Code Section XI.
The San Onofre Unit 1 OMS will be tested by the operator adjusting the pressure control bistable setpoint which would result in PORV and annunciator actuation.
Satisfactory completion of this end-to end sys-tem test would be ascertained by actual PORV opening as determined by proper valve position indication and system annunciation in the control room.
SCE states that these tests are performed prior to returning to a water-solid condition following a cold shutdown with the RCS depressurized.
SCE chose not to test the San Onofre Unit 1 OMS prior to each plant shutdown because testing at that time would increase the probability of a loss-of-coolant due to PORV malfunction or a loss of OM9 capability due to PORV isolation valve malfunction (failing in the closed position).
We conclude that the testing proposed by SCE for the San Onofre Unit 1 OMS does not meet the staff position.
Not testing the OMS prior to each shutdown would result in a decrease in the system reliability, especially 2
when it is placed in service during plant cooldown following long periods of operation.
SCE is presently evaluating the possibility of testing the electronic portion of the San Onofre Unit 1 OMS during plant shutdown.
The staff will evaluate SCE's proposal when it is submitted.
3.4 Single Failure Criteria The specified single failure criteria for the overpressure mitigating system is that it should be designed to protect the vessel given a single failure in addition to the failure that initiated the pressure trantient.
SCE was ask to address this concern and responded that there is no credible single failure which will both initiate a pressure transient and disable one of the redundant PORV trains (Reference 21).
Therefore, the San Onofre Unit 1 OMS meets the single failure criteria.
In addition to the OMS, SCE indicated that, during all of the normally encountered plant cooldown and heatup conditions, there would either be a vapor space in the pressurizer or the RHR system would be in service.
The vapor space in the pressurizer would provide a buffer against overpressurization of the RCS, which would allow the operators time to take corrective action to prevent exceeding the Appendix G limits.
During cooldown, the RHR system is normally placed in service prior to collapsing the pressurizer bubble and is not normally removed from service during heatup until after a steam bubble has been established in the pres-surizer. The RHR system provides a second letdown flow path and safety Valve RV-206, which has a setpoint of 500 psig and a relief capacity of 470 gpm at 10% accumulation (550 psig). This safety valve is capable of mitigating a pressure transient resulting from any of the initiating events analyzed for San Onofre Unit 1.
The RHR system is not automatically isolated at high pressures, therefore, this safety valve should remain available.
We conclude that the San Onofre Unit 1 OMS meets the single failure criteria.
3.5 S_eismic Design The specified seismic criteria is that the Overpressure Mitigating System should be designed to function during an Operating Basis Earthquake (OBE).
The OMS installed at San Onofre Unit I does not meet this criteria.
Those portions of the OMS which interface with or could affect existing Seismic Category A equipment are designed as Seismic Category A.
The PORVs and their operators are Seismic Category A as is the backup nitrogen system except for sections that remain as original plant equipment. All other por-tions of the system are designed at least to Seismic Category B.
Seismic Category B components and supports are designed for a maximum ground accele ation of 0.2 g at San Onofre Unit 1.
A Safe Shutdown Earth-quake (SSE) for San Onofre is defined as a normalized acceleration of 0.67 g 3
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and an Operating Basis Earthquake (0BE) is defined as a normalized accelera-tion of 0.335 g.
Those portions of the San Onofre Unit 1 OMS which are i
Nonseismic Category A could not be expected to remain operable during an OBE and the OMS would therefore be disabled.
We conclude that the San Onofre Unit 1 OMS does not meet the seismic criteria. SCE is investigating resolution of this issue and it is there-fore still under consideration by the staff.
3.6 Analysis Results 3.6.1 Mass Input Case i
The Westinghouse bounding analysis considered two mass input cases:
l inadvertent safety injection, and inadvertent isolation of letdown while charging to a solid system.
SCE chose not to consider inadvertent safety injection at San Onofre Unit 1, even though it would be the most limiting mass input case.
SCE states that inadvertent safety injection is precluded by the establishment of two positive barriers between the Safety Injection System (SIS) and the Reactor Coolant System (RCS) during all cold shutdown conditions. These positive barriers are established during plant cooldown prior to reaching a Reactor Coolant System pressure of 500 psig and are returned to a normal status when the RCS pressure is 1400 psig during plant startup. Once the positive barriers are established on the Safety Injection System, they are not removed until plant startup except for the performance f
4 of the "no-flow" test.
.CE stated that the "no-flow" test is never con-i ducted with the RCS water-solid.
The two positive barriers are established by the plant operating instructions and are required by the plant technical specifications.
i A positive barrier is defined as any one of the following:
(1) a motor operated valve, when closed and tagged with the safety switch open, (2) a pneumatic hydraulic valve, when closed with the hydraulic oil isolation valve closed, or (3) a manually operated valve, when locked closed and tagged.
Two alternative valve arrangements are provided for in the San Onofre Unit 1 operating instructions.
The primary arrangement requires:
(1) the following valves are closed with a yellow " caution" tag affixed to the remote manual switch:
MOVs 850A, 850B, and 850C and HVs 851A and 851B, (2) the safety switches (breakers) for MOVs 850A and 850B are opened, the fuse blocks removed, and a yellow tag affixed to the breaker, (3) the out-put breaker from the Uninterruptable Power Supply (UPS) inverter to MOV i
850C is opened with the manual transfer switch aligned to the inverter l
output breaker; a yellow tag is aftixed to the output breaker and the transfer switch, and (4) the hydraulic oil block valves on the operators for HVs 851A and 851B are closed and tagged.
i The alternative arrangement requires:
(1) closing, locking, and tag-ging the safety injection pump suction Valves 861A and 8618, (2) closing, locking, and tagging the feedwater safety injection bypass Valves 856A and 4
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8568, (3) closing HVs 853A and 853B and (4) closing and tagging the hydrau-lic oil valves on the operators for HVs 853A and 853B.
All plant operating conditions were examined to determine the potential for an inadvertent safety injection leading to RCS overpressurization at San Onofre Unit 1.
All potential scenarios which could lead to inadvertent safety injection were identified and evaluated. These scenarios included consideration of operator errors, equipment failures, and combinations thereof.
In each case a minimum of three independent operator errors and/or equipment failures was required to create an inadvertent safety injection.
Therefore, the probability of such an event occurring is very low.
The mass input case that SCE chose to analyze is an. inadvertent isola-tion of letdown while charging to a water-solid RCS. Westinghouse provided the licensee with a series of curves based on the LOFTRAN analysis of a generic plant design which indicates PORV setpoint overshoot for this transient as a function of system volume, relief valve opening time and relief valve setpoint.
These sensitivity analyses were then applied to the San Onofre Unit 1 plant parameters to obtain a conservative estimate of the PORV setpoint overshoot.
The following assumptions were made when perform-ing the analysis:
1.
One PORV was assumed to fail.
2.
The RCS was assumed to be rigid with respect to expansion.
3.
Conservative heat transfer coefficients were assumed for the steam generator.
Parameter Value and/or Reference Initial RCS pressure 50 psig 3
RCS volume 6752 ft RCS temperature 100 F PORV relief setpoint 500 psig Charging pump delivery 110 gpm PORV relief flow per valve C = 31 gpm/ psi y
PORV opening time 2 seconds PORV closing time 1.5 seconds The PORV setpoint overshoot was determined to be 21 psi. With a relief valve setpoint of 500 psig, a final pressure of 521 psig is reached for the worst case mass input transient.
The 10 CFR 50 Appendix G curve limit for San Onofre Unit 1 is 555 psig for a temperature of 100 F.
It is concluded that the San Onofre Unit 1 OMS performance is acceptable for mass input transients.
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3.6.2 Heat Input Case An inadvertent startup of a reactor coolant pump with a primary to secondary temperature differential across the steam generator of 50 F, and with the plant in a water-solid condition, was selected as the limiting heat input case.
For the heat input case, Westinghouse provided the licensee with a series of curves based on the LOFTRAN analysis of a generic plant design to determine the PORV setpoint overshoot as a function of RCS volume, steam generator area, and initial RCS temperature.
For this transient, the following assumptions were used in the analysis:
Parameter Value and/or Reference Initial RCS pressure 300 psig 3
RCS volume 6752 ft Initial RCS temperature 100, 140, 180 and 250 F RCS/ Steam generator AT 50 F 2
Steam generator heat transfer area 27,700 ft PORV relief setpoint 500 psig PORV opening time 3 seconds PORV flow capacity per valve C = 31 gpm/ psi y
The analyses results for the heat input transient depend on the initial RCS temperature; the results for the various initial temperatures are given below:
-P a
RCS Temperature max setpoint Maximum RCS Pressure RCS Pressure Limit l
( F)
(psi)
(psig)
(psig) 100 18 518 555 140 34 534 584 180 58 558 634 250 99 509 834 e
a.
Determined from 0 F/hr cooldown rate pressure-temperature limits curve from the San Onofre Unit 1 Technical Specification 3.1.3.
t The above analysis utilized a differential temperature of 50 F between the Reactor Coolant System and the secondary side of the steam generator.
The transient has not been analyzed for any differential temperature greater than 50 F.
Therefore, a positive means of determining the differential 6
temperature and administrative controls that maintain it within the bounds that have been analyzed, should be provided in order to insure that the San Onofre OMS can mitigate any pressure transient resulting from heat input.
SCE has not provided any means of directly measuring the steam gener-ator secondary temperature'or any indirect method of determining the temper-ature for all situations that may be encountered.
SCE stated that the only means available to determine this temperature at San Onofre Unit 1 is to use the steam generator secondary pressure indication in conjunction with steam tables, which results in the saturation temperature of the steam generator.
This sytem will not provide adequate temperature indication as temperatures approach 212 F.
SCE stated that San Onofre Unit 1 should never get into a situation where any significant differential temperature exists between the RCS and the steam generator. There are no RCS loop isolation valves and at least one reactor coolant pump is kept running during plant cooldown until the final cold shutdown reactor coolant system temperature (approximately 100 -120 F) is achieved.
The San Onofre Unit 1 technical specifications do not specify a maxi-mum differential temperature when starting a reactor coolant pump. They require that the potential for having developed coolant system temperature gradients be evaluated prior to starting a pump only when pressurizer water level is greater than 80%.
This appears to allow several scenarios that could possibly result in a pressure transient of greater magnitude than the bounding analysis.
In the heat input analysis, for the given RCS temperature, the Appen-dix G limits are not exceeded, therefore the performance of the San Orofre Unit 1 OMS is judged to be adequate for heat induced transients providing
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SCE can assure that the bounding conditions assumed in the analysis are not exceeded.
4.0 ADMINISTRATIVE CONTROLS To supplement the hardware modifications and to limit the magnitude of postulated pressure transients to within the bounds of the analysis provided by the licensee, a defense in-depth approach is adopted using procedural and administrative controls. These specific conditions required to assure that the plant is operated within the bounds of the analysis are spelled out in the Technical Specifications.
t A number of provisions for prevention of pressure transients are con-tained in the San Onofre Unit 1 Operating Instructions.
These instructions prohibit startup of a reactor coolant pump at RCS pressures less than or equal to 400 psig unless the pressurizer water level is less than 80% or 7
the potential for having developed coolant system temperature gradients has been evaluated.
Reactor coolant system temperature gradients are precluded by continuous operation of a reactor coolant pump until the final cold shut-down reactor coolant system temperature is achieved. An overpressure trans-ient due to an inadvertent initiation of safety injection is precluded by establishment of two positive barriers between the safety injection system and the reactor coolant system during cold shutdown conditions.
A charging /
letdown mismatch is precluded by maintaining multiple relief paths open through the normal letdown lines and the residual heat removal inlets.
When the RCS pressure is s400 psig and the pressurizer water level is greater than 50%, a maximum of one of the two centrifugal charging pumps shall be operable.
The inoperable centrifugal charging pump shall have the motor circuit breaker removed from the electrical power supply circuit and shall be condition tagged.
5.0 CONCLUSION
S The administrative controls and plant modifications proposed by Southern California Edison Company provide protection for the San Onofre Nuclear Generating Station, Unit 1 from pressure transients at low tempera-tures by reducing the probability of initiation of a transient and by limit-ing the pressure of such a transient to below the limits set by 10 CFR 50 Appendix G.
We find that the San Onofre Unit 1 Overpressure Mitigating System meets GDC 15 and 31 and that SCE has implemented the guidelines of NUREG-0224 except as noted in Sections 3.3, 3.5, and 3.6.2 of this report.
Pending resolution of these items, the San Onofre Unit 1 Overpressure Mitigating System is judged as an adequate solution to the problem of low temperature overpressure transients.
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6.0 REFERENCES
1.
NRC letter (Schwencer) to Southern California Edison Company (SCEC),
dated August 11, 1976.
2.
SCEC letter (Haynes) to NRC (Schwencer), dated September 2,1976.
3.
SCEC letter (Baskin) to NRC (Schwencer), dated October 29, 1976.
4.
SCEC letter (Baskin) to NRC (Schwencer), dated December 30, 1976.
5.
NRC letter (Schwencer) to SCEC (Moore), dated January 10, 1977.
6.
NRC letter (Schwencer) to SCEC (Moore), dated February 14, 1977.
7.
NRC letter (Schwencer) to SCEC (Moore), dated April 1,1977.
8.
SCEC letter (Baskin) to NRC (Schwencer), dated April 22, 1977.
9.
SCEC letter (Baskin) to NRC (Schwencer), dated May 2,1977.
- 10. " Pressure Mitigating System Transient Analysis Results," prepared by Westinghouse for the Westinghouse user's group on reactor coolant system overpressurization, dated July 1977.
11.
SCEC letter (Baskin) to NRC (Schwencer), dated October 12, 1977.
- 12. SCEC letter (Baskin) to NRC (Schwencer), dated December 2,1977.
- 13. NRC letter (Schwencer) to SCEC (Drake), dated December 7,1977.
14.
SCEC letter (Baskin) to NRC (Schwencer), dated January 27, 1978.
15.
SCEC letter (Baskin) to NRC (Ziemann), dated hrch 17, 1978.
16.
SCEC letter (Baskin) to NRC (Ziemann), dated April 13, 1978.
17.
SCEC letter (Baskin) to NRC (Ziemann), dated May 3,1978.
18.
" Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors," NUREG-0224, September 1978.
19.
" Technical Evaluation of the Electrical, Instrumentation, and Control Design Aspects of the Low Temperature Overpressure Protection System for the San Onofre Nuclear Power Station, Unit 1," prepared by Lawrence Livermore National Laboratory for the NRC, dated June 1980.
20.
SCEC letter (Baskin) to NRC (Crutchfield), dated February 23, 1982.
- 21. Telecon J. Hammond, SCEC, E. Lantz, NRC and C. Ransom, EG&G Idaho, Inc., April 27, 1982.
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