ML20024B098
| ML20024B098 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/29/1983 |
| From: | Dixon O SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8307050209 | |
| Download: ML20024B098 (9) | |
Text
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SOUTH CAROLINA ELECTRIC & GAS COMPANY post orrica 7s4 CotuustA. South CAROLINA 29218 O. W. OtxoN. JR.
VICE PatslDENT June 29, 1983 NUCLes n orf matio~.
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Virgil C, Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Cold Overpressure Protection
Dear Mr. Denton:
South Carolina Electric and Gas Company (SCE&G) currently utilizes a cold Overpressurization Mitigation System (OMS) based on the protection afforded by two (2) pressurizer power operated relief valves (PORVs) augmented by administrative procedures.
A complete description of this system is available in Section 5.2 of the Virgil C. Summer Nuclear Station Final Safety Analysis Report (FSAR).
In Supplement Four of the Virgil C.
Summer Nuclear Station Safety Evaluation Report (SER), NUREG 0717, the NRC Staff concluded that no credible low-temperature overpressurization transients would
. fail the reactor vessel during first cycle operation.
- However, License Condition 2.C.10 states: " Prior to startup after the first refueling outage, SCE&G shall install an NRC staff-approved low-temperature overpressurization protection system.
The preliminary design shall be provided for NRC Staff review no later than June 30, 1983."
SCE&G, in conjunction with the work being completed to obtain NRC approval of PORVs for cold overpressure protection, initiated an engineering evaluation to determine the feasibility of utilizing the mechanical relief valves in the Residual Heat Removal System (RHRS) for mitigating overpressurization during low temperature operation.
The results of the. engineering evaluation were positive, lending substantial credibility toward making a decision to utilize RHR relief valves as the most reliable and eilficient method of mitigating cold overpressure transients.
8307050209 830629 PDR ADOCK 05000395 P
PDR l
Mr. Harold R. Denton-June 29, 1983 Page #2 The present OMS employing the PORVs to mitigate an over-pressurization transient is operational.
SCE&G plans to utilize RHR relief valves as the method of mitigating cold overpressure transients in the future, but will continue to operate'the present system until NRC Staff approval of the OMS based on'the RHR relief valves is received.
A Safety Analysis and appropriate attachments are included and constitute a preliminary design for NRC Staf f review as required by License Condition 2.C.10.
The SCE&G engineering design verification is continuing and is expected to be completed in Septenber 1983.
Any additional information resulting from the design verification will be forwarded for NRC Staff review.
The revision to the FSAR will be incorporated after NRC approval is received.
Please respond in writing with your evaluation.
If you require any additional information or have any questions,
_please let us know.-
Very truly yours, O. W. Di on, Jr.
GMF:WRM:RBC:OWD/fjc cc:
V.
C. Summer T.
C.
Nichols, Jr./O. W.
Dixon, Jr.
E.
H.
Crews, Jr.
E.
C.
Roberts H.
N. Cyrus J.
P.
O'Reilly Group / General Managers O.
S.' Bradham R.
B.
Clary C.
A.
Price A.
R.-Koon C.
L. Ligon (NSRC)
G. J. Braddick.
J. C. Miller J.
L.'Skolds J.
B.
Knotts, Jr.
NPCF.
File (Lic./Eng.)
1
1.0 SYSTEM DESCRIPTION The Reactor Coolant System (RCS) cold Overpressurization Mitigating System (OMS) utilizes the Residual Heat Removal System Relief Valves (RHRSRVs) to mitigate overpressurization transients.
The RHRSRVs are manufactured by the Crosby Valve and Gage Company (Figure 4).
These relief valves are spring-loaded, bellows-type valves, one (1) installed on each RHRS suction header.
The RHRSRVs have a setpoint of 450 psig, and are fully open at 495 psig (design accumulation of 10%).
These valves discharge to the pressurizer relief tank which provides a conservatively assumed maximum back pressure of 100 psig.
There are two (2) motor operated RHRS isolation valves between each of the RHRSRVs and the FCS.
The RHRS suction lines are automatically isolated when RCS pressure exceeds 700 psig which is approximately 200 psig greater than the peak pressure resulting from the postulated heat input transient (Figure 3).
The general arrangement of the RHRS isolation valves and RHRSRVs is attached (Figure 1) and is shown in more detail in the Virgil C. Summer Nuclear Station Final Safety Analysis Report (FSAR), Figure 5.5-4.
Additionally, a comprehensive description of the RHRS can be found in the FSAR, Section 5.5.7.
The RHRSRVs are independent of any electric power supply or control circuitry and as such require no IEEE-279, Criteria 1971 qualification.
The design of the circuitry associated with the motor operated RHRS isolation valves meets the requirements of IEEE-279, Criteria 1971.
An alarm in the control room will be provided to inform the operator of any overpressurization transient exceeding the 450 psig setpoint of the RHRSRVs.
The RHRS piping and valves are classified as ASME Boiler and Pressure Vessel Code Section III, Code Class 2, Seismic Category I.
System components important to safety are classified in accordance with the August 1970 Draft of the American National Standard Institute (ANSI) N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."
The RHRS piping and valves are ANS Safety Class 2a.
Further discussion of compliance with Regulatory Guide 1.26 can be found in the FSAR, Appendix 3A.
2.0 OPERATING BASIS EARTHQUAKE The OMS is designed to function during an Operating Basis Earthquake (OBE) in addition to experiencing no degradation as a result of the OBE.
The RHRS piping and valves are designed and classified as Seismic Category I.
Compliance with Regulatory Guide 1.29 is discussed in the Virgil C.
Summer Nuclear Station FDAR, Appendix 3A.
3.0 PRESSURE TRANSIENT ANALYSIS The OMS design utilizing the RHRSRVs is capable of mitigating the consequences of cold overpressurization transients, thus maintaining RCS pressures within 10CFR50, Appendix G, limits.
The OMS design is based on the theoretical " worst case" mass input and heat input event.
The mass input transient analysis was performed assuming the most severe event involves a single centrifugal charging pump.
Specifically, a loss of instrument air is postulated, whereby the flow control valve on the charging line fails open and simultaneously the flow control valve on the RCS letdown line fails closed.
Assuming zero RCS pressure, a maximum delivery of 650 gpm for the centrifugal charging pump is expected. The charging flow rate at an RCS pressure of 450 psig (RHRSRVs setpoint) is approximately 595 gpm.
For the severe event, one (1) RHRSRV will allow RCS pressure to peak at approximately 478 psig (Figure 2).
This pressure is within 10CFR50, Appendix G, limits.
The heat input transient mechanism considered for analysis involves a Reactor Coolant Pump (RCP) startup in one (1) loop with a water solid condition and temperature asymmetry in the RCS.
The two most influential factors in the heat input transient are the magnitude of the secondary-to-primary temperature differential, and the initial temperature.
With the start of a RCP, a rapid acceleration in pressure will result.
This marked increase in pressure is due primarily to fluid expansion accompanying the RCS temperature increase.
For the most severe event, a 50 F secondary-to-primary temperature dif ferential and 300 F RCS temperature is assumed.
One (1) RHRSRV is capable of mitigating the consequences of the overpressurization transient resulting from the aforementioned pressure increase. The analysis results provided (Figure 3) demonstrate that the RCS pressure is within 10CFR50, Appendix G, limits.
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The OMS is designed to prevent RCS pressures in excess of 10CFR50, Appendix G, limits in the event of a single failure in addition to the occurrence initiating the transient. Power supplies for the motor operated RHRS isolation valves are designed to ensure that no single f ailure of the electrical system or loss of offsite power would isolate both of the RHRSRVs.
4.0 ADMINISTRATIVE PROCEDURES Although the system described previously mitigates the conse-quences of cold overpressurization transients, administrative procedures are recommended for minimizing the potential for any transient to challenge the OMS.
Of primary importance is the basic method of operation of the plant.
Normal plant operating procedures will maximize the use of a pressurizer cushion (steam / nitrogen bubble) during periods of low pressure, low temperature operation.
This cushion will dampen the plant's response to potential transient generating inputs, providing easier pressure control with slower response rates.
An adequate cushion reduces the severity of some potential transients such as RCP-induced heat input and slows the rate of pressure rise for others.
For modes of operation where water solid plant operation is possible, procedures further highlight precautions that minimize the potential for develop-ing overpressurization transients.
For example, Technical Specification 3.1.2.4 provides a clear indication that a maximum of one (1) centrifugal charging pump shall be operable whenever the temperature of one or more of the RCS cold legs is less than or equal to 300 F.
These procedures provide reasonable assurance that most potential transients can be terminated by operator action before the OMS is challenged.
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Q VIRGIL C.
SUMMER NUCLEAR STATION MASS INPUT TRANSIENT ANALYSIS RESULTS REACTOR RHRS RELIEF RHRS. RELIEF
- DESIGN REACTOR APPENDIX G COOLANT-VALVE VALVE MASS COOLANT REACTOR COOLANT SYSTEM SETPOINT BACK INPUT SYSTEM SYSTEM PRESSURE TEMPERATURE PRESSURE PRESSURE (GPM)
PEAK PRESSURE LIMIT
(*F)
(PSIG)
(PSIG)
(PSIG)
(PSIG) 350 450 100 595 475.7 NONE 300 450 100 595 476.2 NONE 250 450 100 595 476.6 1812 200-450 100 595 476.9 1088 150 450 100 595 477.2 700 l
100 450 100 595 477.4 575
- One Charging-Pump,, Delivery at 450 PSIG Reactor Coolant System Pressure Figure 2
e VIRGIL C.
SUMMER NUCLEAR STATION HEAT INPUT TRANSIENT ANALYSIS RESULTS
- REACTOR APPENDIX G STEAM REACTOR SECONDARY RHRS RHRS EXPANSION COOLANT REACTOR
' GENERATOR COOLANT TO PRIMARY RELIEF RELIEF EXCESS SYSTEM COOLANT SECONDARY SYSTEM TEMPERATURE VALVE VALVE VOLUME PEAK SYSTEM TEMPERATURE TEMPERATURE DIFFERENTIAL SETPOINT BACKPRESSURE RELIEF PRESSURE PRESSURE
(*F)
(*F)
(*F)
PRESSURE (PSIG)
(GPM)
(PSIG)
LIMIT (PSIG)
(PSIG)-
350 300 50 450 100 967 501.5 NONE
-300 250 50 450 100 768 488 1810 250 200 50 450 100 572 475 1088 200 150 50 450 100 347 462.3 700 150 100 50 450 100 256
- 458.8 575
- Relief Valve Flow Equal to Expansion Excess Volume Based on Steam Generator Tube Surface Area of 48,000 Ft.2 Figure 3
4 RESIDUAL HEAT REMOVAL RELIEF VALVES MANUFACTURER'S DATA Manufacturer Crosby Valve and Gage Valve Size & Type 3" Inlet x 4" Outlet Type B Valve Style JB-35-TD-WR Pressure Setpoint 450 PSIG Design Temperature 400 'F Pressure Class 600 PSIG x 150 PSIG Orifice Size / Area Size L/2.8 53 in.2 Design Accumulation 10%
Design Features Spring Loaded Bellows Type Tag Numbers XVR-8708A XVR-8708B Figure 4
'