ML20052H490
| ML20052H490 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| 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-5822, NUDOCS 8205210130 | |
| Download: ML20052H490 (12) | |
Text
_ _ _ _
EGG-EA-5822 March 1982 TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE h
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@2N es k4 7h:h This is an informal report intended for use as a preliminary or working document Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6436 g
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FORM EG4G 396 (Rev 03 82)
INTERIM REPORT Accession No.
Report No.
EGG-EA-5822 Contract Program or Project
Title:
Steam Generator Transients and Operating Reactors Evaluation for Reactor Systems Branch Subject of this Document:
Technical Evaluation Report of the Overpressure Mitigating System for Trojan Nuclear Plant 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 document should not be considered final.
EG&G Idaho, Inc.
9 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
Linder DOE Contract No. DE AC07 761D01570 NRC FIN No.
A6436 INTERIM REPORT t
0108j 6
TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE MITIGATING SYSTEM FOR TROJAN NUCLEAR PLANT March 24, 1982 C. B. Ransom Reliability and Statistics Branch Engineering Analysis Division EG&G Idaho, Inc.
9 e
Draft 3-24-82
ABSTRACT This report documents the technical evaluation of the low temperature overpressure protection system of the Trojan Nuclear Plant. 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).
FORWDRD 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 No. A6436.
e e
ii
CONTENTS
1.0 INTRODUCTION
1 2.0 DESIGN CRITERIA.................................................
1 3.0 SYSTEM DESCRIPTION AND EVALUATION...............................
1 3.1 Air Supply................................................
2 3.2 Electrical Controls.......................................
2 o
3.3 Te s t ab i l i ty...............................................
2 3.4 Single Failure Criteria...................................
2 3.5 Seismic Design............................................
3 3.6 An al ys i s Re s ul t s..........................................
3 3.6.1 Mass Input Case...................................
3 3.6.2 Heat Input Case...................................
5 4.0 A DM I N I S TR AT I VE CO NTRO L S.........................................
6
5.0 CONCLUSION
S.....................................................
6
6.0 REFERENCES
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1 9
1 0
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TECHNICAL EVALUATION REPORT OF THE OVERPRESSURE MITIGATING SYSTEM FOR TROJAN NUCLEAR PLANT
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 e
have occurred during cold shutdown while the primary system was in a water-solid condition.
By letter to the Portland General Electric Ccmpany (PGE),
c, owner and operator of the Trojan nuclear Plant, dated August 11, 1976 (Ref.1), the U.S. Nuclear Regulatory Commission (NRC) requested an evalua-tion of the Trojan Nuclear Plant to determine susceptibility to overpres-surization events and an analysis of these possible events, and required PGE to propose interim and permanent modifications to the systems and procedures to reduce the likelihood and consequences of such events.
PGE 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. 13). PGE modified the reference mitigating system and proposed their Overpressure Mitigating System (OMS) along with adminstra-tive procedure modifications and operator training (Ref. 7, 8, 9, 10, and 12). The OMS is designed to mitigate the consequences of an overpres-surization event and the additional operator training and the administra-tive 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 Tecnnical 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 discus-sions pertaining to the vessel stresses and other aspects of vessel over-1 pressurization.
3.0 SYSTEM DESCRIPTION AND EVALUATION e
The Trojan OMS consists of two separate trains, each containing a power-operated relief valve (PORV), an isolation valve and associated cir-cuitry. When in the low pressure mode the system provides low pressure setpoints of 440 psig and 490 psig for the respective trains. When the systen is enabled, it will terminate all analyzed pressure transients below l
the Appendix G limit by automatically opening the PORVs. A manual switch is used to enable und disable the low pressure 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 sys-tem when system pressure drops to a predetermined point (375 psig).
In dddition, an alarm is provided in tne control room to indicate wnen 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 supplied by a control air source.
To assure operability of the valves upon loss of control air, a backup air o
supply is provided. The backup air supply consists of a seismically qualified passive air accumulator for each PORV.
Each accumulator contains enough air to assure that it will provide 32 cycles of the PORV which will be sufficient to mitigate an overpressure event for the 10 minutes during which no credit can be taken for operator action.
3.2 Electrical Controls The electrical, instrumentation, and control system aspects of the Trojan low temperature overpressure mitigating system (OMS) have been reviewed and reported in a separate technical evaluation (R,f. 15).
3.3 Testability Testability is provided for the Trojan OMS. PGE has stated that a channel functional test will be performed prior to enabling the system during plant cooldown. 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 Trojan OMS 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 from the other DC bus, it will remain functional. However, when a single fail-ure is postulated in the remaining PORV, no low-temperature overpressure protection is afforded the plant.
In their analysis of this scenario, PGE indicated that during all of the normally encountered plant cooldown and heatup conditions that there would either be a vapor space in the pressurizer or the RHR system would be in service.
T.a vapor space in the pressurizer would provide a buffer against over-pressurization of the RCS, which would allow the operators time to take 2
corrective action to prevent exceeding the Appendix G limits. The amount of time provided to the operator is dependent on the water level in the pressurizer. At the normal no-load level of 30% there would be 1260 ft3 of vapor space, which would result in an 18% pressure increase 10 minutes after the operator was alerted to the transient by various alarms. How-ever, the Trojan Technical Specifications allow plant operation in Modes 1 through 3 with a pressurizer level of 92%, which results in a vapor space of only 306 ft.0 This smaller vapor space would not provide the required ten minute o
buffer for operator action prior to exceeding the Appendix G limits. PGE stated that the level should always be maintained at the 30% level in Modes 1, 2, and 3 except for abnormal transient conditions.
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 protection from this scenario in two ways; first by providing a second letdown flow path and second by providing safety valve PSV-8708, whicn has a setpoint of 450 psig and a relief capa-city of 900 gpa. This safety valve is capable of mitigating a pressure transient resulting from isolation of letdown at Trojan. The RHR system is not automatically isolated at high pressures due to the isolation valve circuit breakers being racked out to prevent this automatic closure; there-fore this safety valve should remain available.
We conclude that the Trojan OMS meets the single failure criteria except for those times when pressurizer level is increased above the 30%
no-load level and the RHR system is not in service. PGE assures that this would occur only during abnormal transients, but this point is still under consideration by the staff.
3.5 Seismic Design The specified seismic criteria is that the overpressure protection system should be designed to function during an Operating Basis Earthouake.
The OMS installed in the Trojan Nuclear Plant is Seismic Category I with the exception of two components. The PORV operators and their associated air regulators have not been specifically qualified for operation through an OBE. The PORVs are designed to withstand seismic loddings equivalent to 3.09 in the horizontal direction and 2.0g in the vertical direction, but the PORV operators were not procured nor analyzed as Seismic I components.
We conclude that the Trojan OMS meets the seismic criteria with the exception of the PORV operators and regulators. PGE is currently investi-gating the requirements for qualification of these Components.
4 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.
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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 Trojan plant parameters tc obtain a conservative esti-13dte of the PORV setpoint overshoot.
The following assumptions were made when performing 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 e
generator.
PGE performed an independent mass input pressure transient analysis using the RETRAN computer code. The assumptions used in the RETRAN analyses are listed below:
Parameter Value and/or Reference Initial RCS pressure 65 psia RCS volume 12,500 ft3 RCS temperature ICO F PORV relief setpoint 425 psia (410.3 psig)
PORV reseat setpoint 405 psia SI pump delivery Trojan FSAR Fig. 15.2-446 PORV relief flow Ref. 13, Fig. 2.2.1 PORV opening characteristics 0.32 sec. delay time 0.28 sec stroke time PORV closing characteristics 1.56 sec. delay time 1.88 sec. stroke time Using the RETRAN code, the Trojan PORV setpoint overshoot was determined to be 9.2 psi. With a relief valve setpoint of 425 psia, a final pressure of 434.2 psia is reached for the worst case mass input transient. Subse-quent to performing the analysis, PGE increased the PORV relief setpoints to 440 psig (PCV 455A, slower acting PORV) and 490 psig (PCV 456, faster acting PORV). These relief setpoints are not conservative in comparison to the 410 psig value used in the analysis. PGE states that the setpoints were increased for operational convenience and that, even with the increase, these setpoints assure successful mitigation of the worst-case postulated transient. Even though the overshoot may increase if the PORV setpoint were increased to a more conservative value; we conclude that the 9.2 psi 4
overshoot value would not increase sufficiently to cause the resultant pressure to exceed the Appendix G limit and that the system performance is acceptable.
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 o
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 RCS volume 12,500 ft3 Initial RCS temperature 100, 140, 180 and 250 F RCS/ Steam Generator AT 50 F Steam generator heat transfer area 51,500 ft2 PORV relief setpoint 500 psig The analyses results for the heat input transient depend on the initial RCS temperature; the results for various initial temperatures are given below:
Maximum RCS Pressure RCS Pressure Limitd RCS Temperature Pmax -Psetfoint
( F)
(ps1 (psig)
(psig) 100 24.2 524.2 560 140 43.9 543.9 700 180 63.7 563.7 920 250 108.3 608.4 1800 In all these cases, for the given RCS temperature, the Appendix G limits are not exceeded, therefore the performance of the Trojan OMS is judged to
)
De adequate for heat induced transients.
a.
Determined from 0 F/hr cooldown rate pressure-temperature limits curve from Trojan Technical Specificatico Figure 3.4-2.
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4.0 ADMINISTRATIVE CONTROL 5 To supplement the hardware modifications and to limit the magnitude of postulated pressure transients to within the bounds of the analysis pro-vided by the licensee, a defense in-depth approach is adopted using pro-cedural 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-e tained in the Trojan operating procedures. The procedures for startup (and jogging) of a reactor coolant pump require that
" temperatures below 290*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 operated throughout a normal cooldown to 160 F to assure that the steam generators follows the RCS temperature.
Both high pressure safety injection pumps are de-energized by procedure below 200 F to prevent inadvertent starts. The accumulator isolation valves are closed and power removed from the valve actuator circuit breakers below 1000 psig to prevent inadvertent accumulator discharge. Procedures prohibit isolating the RHR inlet line from the reactor coolant loop unless there is a steam bubble in the pressurizer or the charging pumps are stopped.
5.0 CONCLUSION
S The administrative controls and plant modifications proposed by Portland General Electric Company provide protection for the Trojan Nuclear Plant from pressure transients at low temperatures by reducing the probabil-ity 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 overpressure mitigating system meets GDC 15 and 31 and that PGE nas implemented the guidelines of NUREG-0224 except as noted in Sections 3.4 and 3.5 of tnis report. Pending resolution of these items, tne Trojan 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 Portland General Electric Company dated August 11, 1976.
2.
PGE letter (Porter) to NRC (Schwencer) doted October 19, 1976.
3.
PGE letter (Porter) to NRC (Schwencer) dated December 8, 1976.
o 4.
NRC letter (Schwencer) to PGE (Goodwin) dated January 10, 1977.
5.
NRC letter (Schwencer) to PGE (Goodwin) dated February 14, 1977.
6.
NRC letter (Schwencer) to PGE (Goodwin) dated April 1, 1977.
7.
PGE letter (Goodwin) to NRC (Schwencer) dated April 8, 1977.
8.
PGE letter (Goodwin) to NRC (Schwencer) dated April 21, 1977.
9.
PGE letter (Goodwin) to NRC (Schwencer) dated July 21, 1977.
- 10. PGE letter (Goodwin) to NRC (Schwencer) dated February 28, 1978.
- 11. NRC letter (Schwencer) to PGE (Goodwin) dated January 9,1979.
- 12. PGE letter (Broehl) to NRC (Schwencer) dated May 1,1973.
- 13. " Pressure Mitigating Systems Transient Analysis Results" prepared by Westinghouse Electric Corp. for the Westinghouse Owner's Group on Reactor Coolant System Overpressurization, dated July 1977.
- 14. " Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors", NUREG-0224, September 1978.
- 15. " Technical Evaluation of the Electrical, Instrumentation, and Controi Design Aspects of the Low Temperature Overpressure Protection System of the Trojan Nuclear Power Plant," prepared by the Lawrence Livermore National Laboratory for the NRC, dated June,1980.
- 16. PGE letter (Withers) to NRC (Clark), dated February 16, 1982.
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