05000413/LER-2009-001
| Docket Number | |
| Event date: | 01-30-2009 |
|---|---|
| Report date: | 04-01-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 4132009001R00 - NRC Website | |
BACKGROUND
This event is being reported pursuant to 10'CFR 50.73(a)(2)(i)(B), Operation or condition prohibited by Technical Specifications (TS); and 10 CFR 50.73(a)(2)(v), Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to:
a) Shutdown the reactor and maintain it in a safe shutdown condition; b) Remove residual heat; c) Control the release of radioactive material; or d) Mitigate the consequences of an accident.
TS 3.7.7, KC, requires that two independent KC trains be operable in MODES 1, 2, 3, and 4. With one KC train inoperable, restore it to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
TS 3.6.6. (Containment Spray System (NS)) requires that two independent NS trains be operable in MODES 1, 2, 3, and 4. With one NS train inoperable, restore it to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />.
TS 3.5.2 (Emergency Core Cooling Systems (ECCS)) requires that two ECCS trains be operable in MODES 1, 2, and 3. With one or more ECCS trains inoperable, restore train(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 5 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The Centrifugal Charging Pumps (CCPs) are a subsystem of the ECCS.
TS 3.7.5 (Auxiliary Feedwater System (CA)) requires that three CA trains be operable in MODES 1, 2, and 3. With one CA train inoperable, restore it to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
TS LCO 3.0.3establishes requirements for actions when (1) an LCO is not met and the associated Actions are not met; (2) an associated ACTION is not provided, or (3) if directed by the associated ACTIONS. Entry into LCO 3.0.3 is not necessarily reportable under this criterion. However, it should be considered reportable under this criterion if the condition is not corrected within an hour, such that it is necessary to initiate actions to shutdown.
Catawba Nuclear Station (CNS) Unit 1 is a Westinghouse four-loop Pressurized Water Reactor (PWR) .[EIIS: .RCT]. .Unit 1 was operating in Mode 1 (Power Operation) at the time of this event.
KC [EIIS: CC] acts as a closed loop treated water system to dissipate waste heat from motor coolers and intersystem heat exchangers [EIIS:
HX] serving various plant systems supporting plant startup, normal, and shutdown activities. Specifically, KC dissipates waste heat from the NS [EIIS: BE], Chemical and Volume Control System (NV) [EIIS: CB] and the CA [EIIS: SJ] pump motor coolers. KC cools those heat exchangers in which a tube leak could allow radioactive fluid to enter the cooling water. Heat is then transferred to the RN [EIIS: BI] by the component cooling heat exchangers. This system serves as a boundary between the Reactor Coolant System (NC) [EIIS: AB] and RN; reducing the possibility of radioactive leakage to the environment.
RN provides assured cooling during normal and emergency conditions for all heat loads in the Auxiliary and Reactor Building except for the Ice Condenser (NF)System�BC]: RN supplies_assured make up to various systems, such as Containment Penetration Valve Injection, Spent Fuel Pool Cooling [EIIS: DA], CA and KC..
IRN-291 (KC Heat Exchanger 1A RN Control Valve [EIIS: FCV]) is a 12 inch, fail open, air operated control ball valve manufactured by Fisher Controls. Valve 1RN-291 is normally modulated to maintain a set KC, heat. exchanger outlet temperature. When Train 1A KC is inactive,. valve 1RN-291 can be aligned to the RN pump minimum flow controller. The valve will then modulate to allow the operating RN pump(s)to pass minimum flow.
NS provides containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values.
Reduction of containment pressure and the iodine removal capability of the spray reduce the release of fission product radioactivity from containment to the environment, in the event of a Design Basis Accident (DBA).
NS consists of two separate trains of equal capacity, each capable of meeting the system design basis spray coverage. Each train includes a
- NRC FORM 366A (1-2001) NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION (1-2001) FACILITY NAME (1) DOCKET (2) Catawba Nuclear Station, Unit 1 05000413 2009 - 001 - 00 4� OF 9 containment spray pump, one containment. spray heat exchanger, spray headers, nozzles, valves, and piping.
Each NS pump has two motor coolers on the KC essential 'header. The coolers are provided to maintain the motor temperature within
- acceptable limits when the pump(s) are operating. The coolers are required to prevent pump loss due to motor failure by overheating.
NV, during normal operation, provides letdown, charging and seal water, chemical control, purification and makeup. The accident mitigation function of NV is to provide high pressure injection and recirculation of borated water to NC as part of the ECCS during a DBA. During an accident, NV is isolated except for the CCPs, safety injection and NC pump seal injection flow paths.
Each CCP has two motor coolers on the KC essential header. The coolers are provided to maintain the motor temperature within acceptable limits when the pump(s) are operating. The coolers are required to prevent pump loss due to motor failure by overheating.
CA provides a nuclear safety related source of emergency feedwater to the steam generators to maintain secondary side level, at times when the normal feedwater system is not available. In this function, CA is relied upon to remove primary coolant stored,energy and residual core energy, and to prevent overpressurization of NC and the resultant reactor coolant expansion that could result in fuel damage.
CA consists of one Turbine-Driven Auxiliary Feedwater pump per unit and two Motor-Driven Auxiliary Feedwater pumps per unit. Each motor driven CA pump has two motor coolers on the KC essential header. The coolers are provided to maintain the motor temperature within acceptable limits when the pump(s) are operating. The coolers are required to prevent pump loss due to motor failure by overheating.
EVENT DESCRIPTION
(certain event times are approximate) Date/Time Event Description .
1/28/2009/---- Train 1A KC is in service. Valve 1RN-291 is modulating for a low flow rate through the 1A KC Heat Exchanger due to low temperatures in RN.
1/28/2009/1241 The Rod End Bearing Assembly of valve 1RN-291 separates from the valve shaft. Valve 1RN-291 fails in a slightly throttled position. (Note:
This information is based on a review of past OAC data.) 1/28/2009/2248 Train 1B NV declared inoperable and logged into Technical Specifications Action Item Log(TSAIL) due to scheduled maintenance activities.
1/29/2009/0145 Train 1B NV cleared from TSAIL.
1/29/2009/0415 Train 1B CA declared inoperable and logged into TSAIL due to scheduled maintenance, activities.
1/29/2009/0415 Train 1B NS declared inoperable and logged into TSAIL due to scheduled maintenance activities.
1/29/2009/1627 Train 1B CA cleared from TSAIL.
1/29/2009/1909 Train 1B NS cleared from TSAIL.
1/30/2009/1417 While swapping trains of KC on Unit 1, 1B KC heat exchanger outlet mode switch was placed in the "Temp Mode position.
1/30/2009/1425 After swapping trains, 1A KC heat exchanger outlet mode switch was placed in the "Mini Flow" position. Operations personnel noticed that valve 1RN-291 indicated open on the OAC with no increase in flow through the 1A KC heat exchanger. With the valve fully open, a flow rate of greater than 7,000 gpm is expected.
1/30/2009/1500 � Train lA KC declared inoperable and logged into TSAIL.
1/31/2009/1258 � Train 1A KC cleared fom TSAIL following replacement of the Rod End Bearing Assembly on 1RN-291 and completion of valve retest.
CAUSAL FACTORS
The event is attributed to the failure of the Rod End Bearing Assembly of valve 1RN-291. The Rod End Bearing Assembly provides the mechanical connection between the actuator stem and the valve shaft lever. The failure resulted in separation of the actuator and the shaft. The actuator stem (which actuates the switch for remote indication) moved to the expected open position even though the valve shaft lever did not reposition.
The failure mode of the Rod End Bearing Assembly is believed to be embrittlement of the Rod End Bearing Assembly body material with concurrent impact loading resulting from excessive bearing insert .
clearances due to bearing liner wear.
The valve and actuator for 1RN-291 are rebuilt on a 4R frequency(approximately every six years). The maintenance procedure provides removal and reinstallation instructions for the Rod End Bearing assembly. The rebuild procedures require a general inspection of all parts for damage. A review of previous work orders revealed that no damage was identified for this part. Prior to failure, the next scheduled rebuild for this valve was November 3, 2009.
Valve 1RN-291 is IWV stroke tested on a quarterly basis and IWVR stroke tested once every two years.� The. stroke tests present opportunities to identify conditions that are potentially similar to those that led to valve failure. Quarterly stroke times for this valve have been in the acceptable range. This test was last completed on December 16, 2008.
CORRECTIVE ACTIONS
Immediate:
1. TS LCO 3.7.7 was immediately entered due to Train lA KC being declared inoperable.
2. Train 1B was placed in service to provide cooling for KC.
Subsequent:
1. The Rod End Bearing Assembly for valves 1(2) RN-291, 1(2)RN 351(KC Heat Exchanger RN Control Valves) and 2KC-057A (ND Heat Exchanger flow control) were replaced.
2. The failed Rod End Bearing Assembly was sent to the Duke Metallurgy Laboratory for testing and cause analysis.
Planned:
1.The Rod End Bearing Assembly for 1KC-057A, 1(2) KC-082B (ND Heat Exchanger flow control) will be replaced.
2.Periodic maintenance procedures will be revised to ensure replacement of the Rod End . Bearing Assemblies for the aforementioned valves occurs on an established frequency.
The planned corrective actions are being addressed within the Catawba Corrective Action Program. There are no NRC commitments contained in this LER.
SAFETY ANALYSIS
The specified safety function of KC is to provide an intermediate nuclear safety related heat sink for components of various systems essential to the mitigation of DBA which require ECCS operation. The function of KC is to transfer heat from these systems to RN which is the ultimate heat sink. More specifically, KC dissipates waste heat from the NV, CA, and NS pump motor. coolers. The failure of valve 1RN 291 caused a loss of heat sink capability for Train 1A KC. Train 1P1' CCP is equipped with a backup non-safety related cooling water supply from the drinking water system (YD). Upon loss of normal KC cooling to CCP A, the supply from YD could have been manually aligned to provide the cooling water for the CCP A motor coolers, Pump Bearing and Speed Reducer Oil Coolers. Train lA RN was fully capable of providing heat sink capabilities to other Train A systems, subsystems and components.
Additionally, at some point during this time period Train 1B of NV, CA and NS were taken out of service for routine maintenance activities.
The core damage significance of this event was quantitatively evaluated by considering the following:
■ Inoperability of valve 1RN-291 and Train lA KC Heat Exchanger ■ Actual plant configuration and maintenance activities on Train 1B NV and CA during the inoperability of valve 1RN-291; NS is not modeled in the PRA and therefore is not evaluated for impact on CDF (Core Damage Frequency) or Large Early Release Frequency (LERF) ■ Current PRA model accounting for any necessary updates since the last revision The resulting analysis concluded that the incremental conditional core damage probability of this event is less than 1E-6 and has an insignificant impact on core damage risk.
The large early release significance of'this event has also'been quantitatively evaluated using the same considerations as for the CDF impact. The event has an insignificant impact on the large early release risk.
As such, it is reasonable to conclude that this event was inconsequential from a plant risk perspective and there was no actual impact on the .health and safety of the public.
ADDITIONAL INFORMATION
Within the last three years, there were no LERs that involved both trains of a safety .related system ,being inoperable and the cause attributed to component failure. Therefore, this event was determined to be non-recurring.
Energy Industry Identification System (EIIS) codes are identified in the text as [EIIS: XX]. This event is reportable to the Equipment Performance and Information Exchange (EPIX) program.
This event is considered to be a Safety System Functional Failure.
There were no releases of radioactiVe materials, radiation exposures, or personnel injuries associated with this event.