ML20117K695
| ML20117K695 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/05/1996 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20117K699 | List: |
| References | |
| 6730-96-2241, GL-95-07, GL-95-7, NUDOCS 9609120095 | |
| Download: ML20117K695 (6) | |
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September 5, 1996 Tel us-sn-m 6730-96-2241 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station (OCNGS)
)
Docket No. 50-219 Facility Operating License No. DPR-16 Response to Request for Additional Information - Generic Letter 95-07, " Pressure Locking and Thermal Binding of Safety-Related j
Power-Operated Gate Valves" i
NRC letter dated July 10,1996 (6730-96-3232) requested additional information regarding the OCNGS response to NRC Generic Letter 95-07 previously submitted in GPU Nuclear letter dated May 9,1996 (6730-96-2149).
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The attachment provides an itemized response to each of the NRC questions. If any additional information is required, please contact Mr. David J. Distel, GPU Nuclear
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Regulatory Affairs at (201) 316-7955.
Sincerely, l
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~-L e-Michael B. Roche Vice President and Director i
Oyster Creek Attachment DJD/plp c:
Administrator, Region I NRC Resident inspector g
NRC Project Manager gD i
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9609120095 960905 PDR ADOCK 05000219 P
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ATTACIIMENT RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI) ON GENERIC LETTER (GL) 95-07," PRESSURE LOCKING AND TIIERMAL BINDING OF SAFETY-RELATED POWER-OPERATED GATE VALVES."
OUESTION 1.
Regarding valves V-14-34, -35, Isolation Condenser Condensate Return, the licensee's submittal discusses an analysis which was performed to demonstrate that these valves have suflicient actuator capability to overcome the potential pressure locked condition. Please provide this analysis for our review.
In addition, the licensee's submittal states that high energy line break (HELB) analyses have shown that the reactor is isolated at 400 psi, which is greater than the calculated reactor pressure at which these valves are susceptible to pressure locking. Please briefly describe these HELB analyses, and discuss the operational configuration that isolates the reactor at 400 psi.
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RESPONSE
GPU Nuclear Calculation No. C-1302-211-5310-087, Rev. O documents the pressure lock analysis for Isolation Condenser System (ICS) valves V-14-34,35, and demonstrates that the potential pressure locked condition only occurs at reactor pressures less than 400 psi.
For V-14-34, potential pressure locking may occur at approximately 169 psi at motor undervoltage and at approximately 343 psi at motor undervoltage derated for accident temperatures (i.e., HELB in other ICS). For V-14-35, potential pressure locking may occur at approximately 193 psi at motor undervoltage and at approximately 330 psi at motor undervoltage derated for accident temperatures (i.e., HELB in other ICS).
Additionally, it is noted that the calculation conservatively assumes 0.20 stem friction (a basis has been developed to use 0.168) and both these valves will have their margin improved due to modifications planned for this outage (16R). This calculation is enclosed for your review.
GPU Nuclear Calculation Nos. C-1302-153-5450-083, Rev. 0; C-1302-411-5450-048, Rev. 0; C-1302-153-5450-070, Rev.1; and C-1302-153-5450-081, Rev. O document reactor vessel thermal-hydraulic analyses performed for various high energy line break (HELB) scenarios and establishes the reactor pressure at the time ofisolation of each postulated HELB. There is not an operational configuration that isolates the reactor at 400 psi; however, these transient analyses, as described below, have been performed to establish the reactor pressure condition at the time ofisolation.
730%241. doc
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I The ICS removes fission product decay heat when the reactor vessel is isolated from the Main Condenser. The ICS Condensate Return Valves V-14-34 and 35 are normally closed and are designed to open on an isolation condenser actuation signal from either reactor vessel high pressure or low-low water level. These valves can also be opened -
from the Control Room to manually initiate the isolation condenser as directed by the Emergency Operating Procedures (EOPs) as an alternate method of reactor pressure vessel (RPV) pressure control. This is the condition being evaluated for potential pressure locking of V-14-34 and 35. The Appendix K LOCA analysis does not take credit for the ICS operation. Therefore, pressure locking is not a concern for any licensing basis LOCA analysis.
j The following HELB analyses were evaluated: ICS line break; Main Steam Line Break; i
and Reactor Water Cleanup System line break. In each of the HELB scenarios, except a cleanup line break with feedwater available, the reactor is isolated with~ pressure greater i
than 400 psi. Since potential pressure locking of V-14-34,35 could only occur at reactor l
pressures below 400 psi, the ICS would remain available for operator action or automatic initiation for these scenarios. The cleanup line break with feedwater available is funher described below.
The following discussion provides a brief description of the HELB analyses performed.
ISOLATION CONDENSER SYSTEM LINE BREAK i
i ICS line breaks were evaluated with both the ICS in standby mode and with the ICS in-l' service. The ICS line break with the ICS in-service is the bounding case since the system valve alignments are such that the reactor coolant escapes through both sides of the break resulting in the most rapid reactor depressurization of the two (2) cases. This case analyzed a break in the isolation condenser condensate return line following a loss-of-offsite power (LOOP) and ICS initiation due to high reactor pressure. The line break was I
assumed to occur after both isolation condensers were fully initiated which maximizes the break flow and reactor depressurization. The maximum isolation signal time delay and the maximum stroke time of DC-powered valves V-14-34 and 35 was assumed which is conservative. The reactor pressure at the time ofisolation is determined to be 660 psi which is well above the reactor pressure at which potential pressure locking could occur on ICS valves V-14-34 and 35 as determined in the enclosed Calculation No.
C-1302-211-5310-087.
Additionally, it is noted that once the valves open to initiate the ICS and begin reactor i
depressurization, the potential for pressure locking is alleviated.
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730%241. doc
MAIN STEAM LINE BREAK The Main Steam Line Break (MSLB) transient analysis considers a guilliotine break in one of two main steam lines, with simultaneous loss-of-offsite power which assumes that the reactor, turbine and recirculation pumps trip at time =0 and the main feedwater pumps coastdown in 10 seconds. This analysis also considers a 10 second delay in the tr2in steam isolation valve (MSIV) closure time and a 10 second MSIV stroke time. The results of this analysis show that at the time of reactor isolation, the reactor dome pressure is 590 psia which is well above the reactor pressure at which potential pressure locking could occur on ICS valves V-14-34 and 35.
REACTOR WATER CLEANUP SYSTEM LINE BREAK As stated above, the pressure at which potential pressure locking could occur for V-14-34 and 35 is approximately 169 psi and 193 psi at motor undervoltage, respectively. Post-accident temperatures for this scenario do not exceed 250 F, therefore additional derating for accident temperature is not applicable.
The Reactor Water Cleanup (RWCU) line break bounding transient analysis considers a guilliotine break in the RWCU line immediately downstream of the system isolation valve V-16-14. The isolation valve closes on low-low reactor vessel level signal and a valve stroke time of 60 seconds is accounted for. Main feedwater and recirculation pumps are assumed to trip at time =0. The results of this analysis show that at the time of reactor isolation, the reactor dome pressure is greater than 900 psia which is well above the reactor pressure at which potential pressure locking could occur on ICS valves V-14-34 and 35.
The RWCU line break transient analysis was also performed assuming main feedwater remains available. This transient is less bounding since continued feedwater addition mitigates core cooling concerns. Reactor pressure remains above 200 psia within the first 14 minutes of this scenario. Operator action to initiate ICS within this timeframe is reasonable if RCS depressurization is desired. Additionally, once the RWCU line break is isolated the continued addition of main feedwater in combination with reactor decay heat will maintain reactor pressure well above the pressure at which potential pressure locking of V-14-34 and 35 could occur. 730%241. doc
OUESTION 2.
Through review of operational experience feedback, the stafTis aware ofinstances where licensees have completed design or procedural modifications to preclude pressure locking or thermal binding which may have had an adverse impact on plant safety due to incomplete or incorrect evaluation of the potential effects of these modifications. Please describe evaluations and training for plant personnel that have been conducted for each design or procedural modification completed to address potential pressure locking or thermal binding concerns.
RESPONSE
Isolation Condenser System Valves V-14-34 and V-14-35 have been replaced from Anchor Darling Flex Wedge gate valves to Anchor Darling Parallel Disc gate valves to preclude thermal binding. This modification was implemented during the OCNGS 13R outage (1991).
Plant cooldown procedures have also been modified to direct the operators to cycle the DC-powered ICS condensate return valves (V-14-34 and 35) during plant cooldown prior to initiating the isolation condensers in order to prevent pressure locking. Potential pressure locking could occur if during a normal cooldown or a plant transient the isolation condensers are selected for reactor pressure control and the reactor has already depressurized below 400 psig. This procedure revision instructs the operators to close the adjacent redundant AC-powered condensate return valves (V-14-36 and 37) between 500 and 550 psig reactor pressure, cycle V-14-34 and 35, then reopen V-14-36 and 37. This will reduce the pressure in the bonnet, preventing pressure locking and ensure isolation condenser availability. Operator requalification training has addressed this additional step during plant cooldown. Operator training also addresses the system conditions and parameters leading to valve pressure locking and thermal binding. The valve in each isolation condenser is cycled individually. Procedere requirements ensure that the redundant isolation condenser is operable and appropriate ICS Technical Specification requirements are satisfied. Therefore, this procedural activity will not adversely impact plant safety.
Core Spray System Valves V-20-15, V-20-21, V-20-40 and V-20-41 have been modified by drilling the discs such that the bonnet stays in communication with the reactor side of the valves. This modification was performed during the 15R outage (1994). OCNGS Valve Maintenance Procedure (2400-SMM-3917.03) has been revised to incorporate requirements to ensure correct disc orientation during valve reassembly. This modification does not adversely impact plant safety as described below: 73096241. doc
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During a LOCA event, with the RCS pressure at approximately 285 psig, the Core Spray i
h System will provide an alternate supply ofcooling water from the torus nto t e reactor pressure vessel via the open parallel injection valves and the core spray spargers. Per the j
Emergency Operating Procedures, the parallel injection valves will be cycled opened /
closed thereby performing a flow control function in order to maintain the reactor pressure vessel water level between 100" and 175" above top of active fuel. With the parallel injection valves closed, the core spray pumps will continue to operate on minimum recirculation flow. During a LOCA event and subsequent Core Spray System actuation, the core spray pump side of the valve disc does not have to be 100 % leaktight. Any Core l
Spray System water leakage (inward direction) past the closed parallel injection valves will flow into the reactor pressure vessel.
During normal plant operations, containment isolation is maintained by Core Spray System check valves (V-20-150, V-20-151, V-20-152,V-20-153) and the parallel injection valves.
l The parallel injection valves (i.e., core spray pump side of the valve disc) will prevent RCS j
water leakage (outward direction) into the Core Spray System.
During Core Spray System surveillance tests, procedural controls have been established to l
prevent Core Spray System water injection / leakage (inward direction) past the parallel injection valves and into the reactor pressure vessel. If the RCS pressure is less than or equal to 350 PSIG, Core Spray System Booster Pump discharge valves V-20-12 and V-l 20-18 are closed during surveillance test activities. These valves isolate the parallel injection valves from the Core Spray System surveillance test pressure. If the RCS i
pressure is greater than 350 PSIG, Core Spray System Booster Pump discharge valves V-20-12 and V-20-18 remain open during surveillance test activities. The parallelinjection valves will experience the Core Spray System surveillance test pressure which is acceptable. The RCS pressure is greater than the Core Spray System pressure, therefore, the net positive pressure effect is to maintain containment isolation by Core Spray System check valves (V-20-ISO, V-20-151, V-20-152,V-20-153) and the parallel injection valves.
The parallel injection valves (i.e., Core Spray Pump side of the valve disc) will prevent RCS water leakage (outward direction)into the Core Spray System. 730%241. doc
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