ML20052H484
| ML20052H484 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/31/1982 |
| From: | Ransom C EG&G, INC. |
| To: | Lantz E Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-6436 EGG-EA-5826, TAC-6714, NUDOCS 8205210126 | |
| Download: ML20052H484 (12) | |
Text
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, . t.GG-EA-5826 5 March 1982 TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE gg PROTECTION SYSTEM FOR BEAVER VALLEY POWER STATION, UNIT NO. 1 8 pge geseo,<el adu 71sbid b* 'I"" f#
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i This is an informal report intended for use as a preliminary or working document i Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-761001570 FIN No. A6436 g
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<a o3 ea INTERIM REPORT Access;on No.
Report No. EGG-EA-5826
, Contract Program or Project
Title:
Steam Generator Transients and Operating Reactors Evaluation for Reactor Systems Branch Subject of this Document:
1 Technical Evaluation Report of the Overpressure Protection System for Beaver Valley Power Station, Unit No. 1 Type of Document:
Informal Report Author (s):
C. B. Ransom Date of Document:
March 1982 Responsible NRC Individual and NRC Office or Division:
.! E. Lantz, Division of Systems Integration This document was prepared primarily for preliminary or internal use. it has not received full review and approval. Since there may be substantive changes, this docume-t should not be considered final.
EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Under DOE Contract No. DE-AC07-761001570 NRC FIN No. A6436 INTERIM REPORT i
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, TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE PROTECTION SYSTEM FOR BEAVER VALLEY POWER STATION, UNIT NO. 1 March 1982 C. B. Ransom Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.
o Draft 3-24-82 TAC NO.
5 ABSTRACT .
This report documents the technical evaluation of the low temperature overpressure protection system of the Beaver Valley Nuclear Power Station, Unit 1. The criteria used to evaluate the acceptability of the 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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CONTENT S
1.0 INTRODUCTION
..................................................... I 2.0 DESIGN CRITERIA.................................................. 1 3.0 SYSTEM DESCRIPTION AND EVALUATION................................ 1 3.1 Air Supply................................................. 2 3.2 Electrical Controls........................................ 2 e
3.3 Testability................................................ 2 3.4 S i n g l e F a i l ure C r i te r i a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.5 Seismic Design............................................. 3 3.6 Analysis Results........................................... 4 3.6.1 Mass Input Case.................................... 4 3.6.2 HeatInputCase.................................... 4 4.0 ADMINISTRATIVE CONTR0LS.......................................... 5
5.0 CONCLUSION
S...................................................... 6
6.0 REFERENCES
....................................................... 7 4
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TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE PROTECTION SYSTEM FOR BEAVER VALLEY POWER STATION, UNIT NO. I
1.0 INTRODUCTION
Several instances of reactor vessel overpressurization have occurred in pressurized water reactors in which the technical specifications imple-menting 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 Duquesne Light Company (DLC),
owner and operator of the Beaver Valley Power Station, dated August 11, 1976 (Ref. 1), the U.S. Nuclear Regulatory Commission (NRC) requested an evaluation of Beaver Valley Unit 1 to determine susceptibilty to overpres-surization events and an analysis of these possible events, and required DLC to propose interim and permanent modifications to the systems and procedures to reduce the likelihood and consequences of such events.
DLC participated as a member of a Westinghouse owner's group which provided a reference mitigating system and analyses to verify the adequacy of the system (Ref. 9). DLC modified the reference mitigating system and proposed their Reactor Vessel Overpressure Protection System (RV0PS) along with administrative proceoure modifications and operator training (Ref 6, 7, 10 and 12). The RV0PS 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 probabil-ity of the occurrence of an overpressurization event.
This is a report of the evaluation of the compliance of the licensee's Overpressure Protection System with the design criteria establisned 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 Tehnical 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 Beaver Valley RV0PS consists of two separate trains, each contain-ing a power-operated relief valve (PORV), an isolation valve and associated circuitry. When in the low pressure mode the system provides a low pressure setpoint of 350 psig for both PORV trains. When the system is enabled, .t will terminate all analyzed pressure transients below the Appendix G limit 1
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 operator 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- g required-to-open valves, which are normally supplied by the containment instrument air system. To assure operability of the valves upon loss of containment instrument air, a backup nitrogen supply is provided. Nitrogen is supplied by the nitrogen manifold that provides nitrogen to the safety injection accumulators. The backup system is seismically designed and sup-ported, and includes an accumulator for each PORV, with sufficient accumu-lated nitrogen to permit ten minutes of operation without any additional operator action. The accumulator is charged to 675 psig and at 600 psig a pressure switch will initiate an alarm in the control room.
3.2 Electrical Controls The electrical, instrumentation, and control system aspects of the Beaver Valley Unit I low temperature overpressure protection system (RV0PS) have been reviewed and reported in a separate technical evaluation (Ref 15).
3.3 Testability Testability is provided for the Beaver Valley RV0PS. DLC has stated that a channel functional test will be performed prior to enabling the system during plant cooldown and monthly thereafter when the system is required. The functional test will include actual valve stroking at the desired setpoint. Testing requirements are incorporated in the Technical Specifications.
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 transient.
The Beaver '/ alley RV0PS meets this criteria for all cases reviewed except for the case where the initiating event is a loss of power from one 125V DC bus. This loss of power would result in isolation of the letdown line and one PORV failing to open upon request. Because the other PORV is powered i
from the other DC bus, it will remain functional. However, when a single failure is postulated in the remaining PORV, no low-temperature overpressure protection is afforded the plant.
4 Beaver Valley Unit 1 is susceptible to this scenario when the letdown path is the normal letdown vic the Chemical and Volume Control System (CVCS). Letdown via the excess letdown heat exchanger and via the Residual 2
Heat Removal System (RHR) are not affected because of the use of motor-operated valves that fail in the "as is" position upon loss of control power.
DLC maintains a vapor space in the pressurizer during plant cooldown by establishing a nitrogen volume as the steam bubble is being collapsed.
The nitrogen volume reduces the amount of time that the plant is in a water-solid condition, which provides a buffer against overpressurization of the reactor coolant system (RCS). This buffer allows the operator time to take i
' corrective action to prevent exceeding the Appendix G limits. However, there are periods of time when no steam or nitrogen volume exist in the pressurizer and the plant is in a water-solid condition (i.e., during the l ,
e fill and vent procedure). There could also be periods of time when the gas or vapor volume is of insufficient size to allow the required 10 minutes after the operator is alerted to a pressure transient before he must take action to preclude exceeding the Appendix G limits. During these periods there would be no protection for an overpressure transient at Beaver Valley Unit 1.
There is a safety valve in Beaver Valley's RHR system with sufficient capacity to mitigate this scenario, however, it is set to lift at 600 psig, which is above the Appendix G limits for low temperatures, and the RHR system is automatically isolated at a pressure of 630 psig. Therefore no credit can be taken for this system in mitigating overpressure transients.
We conclude that the Beaver Valley RV0PS meets the single failure criteria except for those times when an insufficient steam or nitrogen volume exists in the pressurizer to allow for the 10 minute delay between when the operator is alerted and when he must take action to mitigate a pressure transient. DLC is presently formulating a response to this issue.
3.5 Seismic Design The specified seismic criteria is that the overpressure protection system should be designed to function during an Operating Basis Earthquake.
The RV0PS installed at Beaver Valley Unit 1 is Seismic Category I with the a exception of the PORV operatces which have not been specifically qualified for operation through an OBE. The PORVs were purchased to a Westinghouse E Spec. which required the valve operators and valve assemblies to withstand seismic loadings equivalent to 3.0g in the horizontal direction and 2.0 9 in the vertical direction and to be capable of performing all intended functions, but they were not procured nor analyzed as Seismic I components.
We conclude that the Beaver Valley RV0PS meets the seismic criteria with the exception of the PORV operators. OLC states that an OBE for Beaver Valley is defined as a normalized acceleration of 0.06g and a DBE is defined as a normalized acceleration of 0.1259 . This issue is under consideration by the NRC.
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3.6 Analysis Results 3.6.1 Mass Input Case The inadvertent start of a safety injection pump with the plant in a cold shutdown condition was selected as the limiting mass input case.
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 i opening time and relief valve setpoint. These sensitivity analyses were then applied to the Beaver Valley plant parameters to obtain a conservative estimate of the PORV setpoint overshoot. The following assumptions were <
made when running 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 l RCS volume 9,716 ft3 RCS temperature 100 F PORV relief setpoint 350 psig PORV opening time 2.5 seconds l SI pump delivery 560 gpm Mass input rate 77 lb. per second The PORV setpoint overshoot was determined to be 89 psi. With a relief valve setpoint of 350 psig a final pressure of 439 psig is reached for the worst case mass input transient. The 10 CFR 50 Appendix G curve limit for Beaver Valley is 440 psig for temperatures of 100 F and above. It is con-cluded that the Beaver Valley RV0PS performance is acceptable for mass input transients.
i 3.6.2 Heat Input Case Inadvertent startup of a reactor coolant pump with a primary to secon-
- dary 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 l 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, I l
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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 RCS volume 9,716 ft3
- Initial RCS temperature 100,180 and 250 F RCS/ Steam generator AT 50 F 7
Steam generator heat transfer area 61,500 ft2 PORV relief setpoint 350 psig PORV opening time 3 seconds The analysis results for the heat input transient depend on the initial RCS temperature; the results for various initial temperatures are given below:
RCS Temperature P -P max setpoint Maximum RCS Pressure
( F) (psi) (psig) 100 23.7 373.7 180 45 395 250 64.3 4 14 .3 In all these cases, for the given RCS temperature, tne Appendix G limits are not exceeded, therefore the performance of the Beaver Valley RV0PS is judged to be adequate for heat induced transients. t 4.0 ADMINISTRATIVE CONTROLS t
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. Those specific conditions required to assure that the plant is operated within the bounds of the analysis are spelled out in the Technical Specifications.
A number of provisions for prevention of pressure transients are con-tained in the Beaver Valley operating procedures. The procedures for startup of a reactor coolant pump require that at RCS temperatures above 160 F, a steam bubble be established in the pressurizer prior to pump start. The steam generator shell-side temperature is monitored to assure that it is within 50 F of the RCS temperature. Also, at least one RCP is !
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operated throughout a normal cooldown to 160*F to assure that the steam generator follows the RCS temperature.
When RCS temperature decreases to less than 350 F, station shutdown procedures require the isolation of the boron injection tank flow path to prevent overpressurization from a safety injection signal . When RCS pres-sure decreases to less than 1000 psi the safety injection accumulator dis-charge isolation valves are closed and the integrity of their discharge check valves is verified periodically. Power is removed from the operators of these valves by disconnecting the control power to the coil of the line-starter by removing jacks in the Control Room. Procedures require placing the RV0PS in service and establishing a nitrogen bubble in the pressurizer as the steam bubble is collapsed. The standby charging pump control switch '
is placed in the pull-to-lock position at low system temperatures.
5.0 CONCLUSION
S The administrative controls and plant modifications proposed by Duquesne Light Company provide protection for the Beaver Valley Power Sta-tion, Unit 1, from pressure transients at low temperatures by reducing the probability of initiation of a transient and by limiting the pressure of such a transient to below the limits set by 10 CFR 50 Appendix G. We find that the Beaver Valley Reactor Vessel Overpressure Protection System meets GDC 15 and 31 and that 01.C has implemented the guidelines of NUREG-0224 except as noted in Sections 3.4 and 3.5 of this report. Pending resolution of these items, the Beaver Valley Overpressure Protection System is judged as an adequate solution to the' problem of low temperature overpressure trans ien ts .
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6.0 REFERENCES
- 1. NRC letter (Reid) to Duquesne Light Company (DLC), dated August 11, 1976.
- 2. DLC letter (Dunn) to NRC (Reid), dated October 12, 1976.
- 3. DLC letter (Dunn) to NRC (Reid), dated December 6,1976.
/ 4 NRC letter (Reid) to DLC (Dunn), dated January 10, 1977.
, 5. NRC letter (Reid) to DLC (Dunn), dated February 15, 1977.
- 6. DLC letter (Dunn) to NRC (Reid), dated March 2,1977.
- 7. DLC letter (Dunn) to NRC (Reid), dated April 6,1977.
- 8. " Summary of Meeting Held on May 25, 1977, with Westinghouse and the Licensees of Westinghouse Designed PWR Facilities Concerning the Status of their Efforts to Prevent Reactor Vessel Overpressurization",
Zech, NRC, dated July 11, 1977.
- 9. " Pressure Mitigating System Transient Analysis Results" prepared by Westinghouse for the Westinghouse Owner's Group on Reactor Coolant System Overpressurization, dated July 1977.
- 10. DLC letter (Dunn) to NRC (Reid), dated November 23, 1977.
- 11. " Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors", NUREG-0224, September 1978.
- 12. DLC letter (Dunn) to NRC (Schwencer), dated October 4,1978.
- 13. DLC letter (Dunn) to NRC (Schwencer), dated December 14, 1978.
- 14. DLC letter (Dunn) to NRC (Schwencer), dated December 6,1979.
- 15. " Technical Evaluation of the Electrical, Instrumentation, and Control Design Aspects of the Low Temperature Overpressure Protection System for the Beaver Valley Nuclear Power Station, Unit 1", prepared for the NRC by Lawrence Livermore National Laboratory, dated June 1980.
- 16. DLC letter (Carey) to NRC (Varga), dated February 16, 1982.
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