05000311/LER-2011-002
| Salem Generating Station - Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3112011002R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
Westinghouse — Pressurized Water Reactor (PWR/4) Reactor Protection System {JC/-} Reactor Coolant System {AB/-} * Energy Industry Identification System {EllS} codes and component function identifier codes appear as {SS/COO}
IDENTIFICATION OF OCCURRENCE
Event Date: April 10, 2011 Discovery Date: April 11, 2011
CONDITIONS PRIOR TO OCCURRENCE
Salem Unit 2 entered Mode 5 on April 10, 2011 at 02:21.
No structures, systems or components were inoperable at the time of the discovery that contributed to the event.
DESCRIPTION OF OCCURRENCE
Technical Specifications (TS) 3.4.10.3 "Overpressure Protection Systems" states "At least one of the following overpressure protection systems shall be OPERABLE:
a. Two Pressurizer Overpressure Protection System relief valves (POPS) with a lift setting of less than or equal to 375 psig, or b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 3.14 square inches.
Action b states:
"With one POPS inoperable in MODES 5 or 6 with the Reactor Vessel Head installed, restore the inoperable POPS to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or complete depressurization and venting of the RCS through at least a 3.14 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.
DESCRIPTION OF OCCURRENCE (cont'd) Action c states:
"With both POPSs inoperable, depressurize and vent the RCS through a 3.14 square inch vent(s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.
Salem Unit 2 entered Mode 5 on April 10, 2011 at 02:21.
On April 11, 2011, at 11:51 control room personnel entered TS 3.4.10.3 Act b to support testing of 2PR1 in accordance with procedure S2.0P-ST.PZR-0003 "Inservice Testing Pressurizer and Reactor Head Vent Valves." When the 2PR1 valve was demanded to open as the test key switch was turned to the test position, the channel 1 test light illuminated, but the valve 2PR1 did not respond as expected. 2PR1 was restored to its pretest position; however, 2PR1 (POPS Channel 1) remained inoperable and OFF.
At 12:18 TS 3.4.10.3.A Act c was entered to support testing on 2PR2. 2PR2 was tested satisfactorily and at 12:31 TS 3.4.10.3.A Act c was exited.
At 12:38 control room personnel attempted to open 2PR1 from normal control room bezel, but 2PR1 failed to open again. At this point it was determined that 2PR1 had been inoperable since the entry into Mode 5 on April 10 at 02:21, and that Salem Unit 2 had operated in a condition prohibited by Technical Specifications (TS 3.4.10.3 Act b).
The cause investigation determined that 2PR1 may have been inoperable during a portion of the cycle when Salem 2 Unit operated in Modes 1 through 3. Therefore, this LER also applies to TS 3.4.5 "Relief Valves," which requires two power operated relief valves (PORVs) operable. Action b states that with one inoperable PORV due to other causes than seat leakage, within one hour restore the PORV to operable or close the associated block valve.
Accordingly, this report is being made in accordance with 10CFR50.73 (a)(2)(i)(B) any operation or condition which was prohibited by the plant's Technical Specifications...
CAUSE OF OCCURRENCE
Based on the troubleshooting performed and the as-found condition upon disassembly of 2PR1 foreign material is the most likely cause of the failure of the valve to open upon initial demand.
CAUSE OF OCCURRENCE (cont'd) The disassembly inspection of the valve plug and cage found a longitudinal score between the mating surfaces of the plug outside diameter (OD) and the cage inside diameter (ID). The as-found condition of the plug and cage is typical when a foreign body gets lodged between the plug and cage. The foreign body was not recovered. A causal evaluation is in progress.
PREVIOUS OCCURRENCES
A review of LERs at Salem Station dating back to 2007 did not identify any similar events.
SAFETY CONSEQUENCES AND IMPLICATIONS
There was no actual safety consequence associated with this event.
In accordance with the IST Program Basis Document the valve has an active safety function in the open position to ensure that:
1. For all normal operating transients up to and including the rated 50% of full power step load reduction with steam dump, the maximum pressurizer pressure is below the pressurizer high pressure reactor trip set point, 2. Provides (in conjunction with the Residual Heat Removal suction relief valve) overpressure mitigation during plant shutdown conditions, as required by 10CFR50, Appendix G, and 3. Is used to depressurize the RCS in the event of a steam generator tube rupture during a loss of off-site power.
The valve has also an active safety function in the closed position to limit Reactor Coolant System inventory loss after an overpressure protection system actuation. The safety function in the closed position is unaffected by the failure of 2PR1 to open.
Safety Implication of the failure to open from the point of view of TS 3.4.10.3.A:
The OPERABILITY of two POPSs channels (2PR1 Channel 1, 2PR2 Channel 2) or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 312°F.
SAFETY CONSEQUENCES AND IMPLICATIONS (cont'd) Either POPS channel (2PR1 or 2PR2 valve) has adequate relieving capability to protect the RCS from Overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50°F above the RCS cold leg temperatures, or (2) the start of an Intermediate Head Safety Injection pump and its injection into a water solid RCS, or the start of a High Head Safety Injection pump in conjunction with a running Positive Displacement pump and its injection into a water solid RCS.
Procedures are in place to ensure that these criteria are met. When cooling down and depressurizing in accordance with S2.0P-10.ZZ-0006(Q) "Hot Standby to Cold Shutdown" actions are taken to preclude the inadvertent start of the intermediate and high head safety injection pumps. These actions include removing from service (cleared and tagged) both intermediate head safety injection pumps' breakers, and one of the two centrifugal charging pump's breakers if both pumps are operable. These steps are performed prior to the reactor coolant system cooling down below 312°F.
Additionally, procedure S2.0P-SO.RC-0001(Q) "Reactor Coolant Pump Operation" contains a precaution and limitation to verify that all steam generator secondary temperatures are less than 50°F before starting a reactor coolant pump. Note that reactor coolant pumps do not get any automatic start signal.
Therefore, based on the procedure controls the remaining POPS channel (2PR2) would have been able to perform its design safety function.
Safety Implication of the failure to open from the point of view of TS 3.4.5:
The Spurious Operation of the Safety Injection System at Power is the only accident which the PORVs are credited to successfully mitigate its consequences. The spurious operation of the safety injection system at power results in the pressurizer going water solid and pressure rapidly increasing to the PORV setpoint. The analysis concludes that only one PORV operating (relieving water), prevents the more serious event of opening one or more pressurizer safety valves with the possibility of one or more of them failing to reseat properly. The analysis also assumes that the operable PORV may be isolated by its block valve due to excessive seat leakage and credits operator action in reopening the block valve.
2PR2 was always operable; therefore, the valve would have actuated without the operator action delay assumption in the analysis.
SAFETY CONSEQUENCES AND IMPLICATIONS (cont'd) A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, did not occur. This event did not prevent the ability of a system to fulfill its safety function to either shutdown the reactor, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.
CORRECTIVE ACTIONS
1. A new trim set was installed into the valve, and the valve was tested satisfactorily.
2. A causal evaluation is in progress; any additional corrective actions associated with this event will be tracked in the PSEG Corrective Action Program.
COMMITMENTS
No commitments are made in this LER.