05000277/LER-2014-003
| Peach Bottom Atomic Power Station Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(vii)(C), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| Initial Reporting | |
| ENS 50571 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 2772014003R00 - NRC Website | |
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Peach Bottom Atomic Power Station 14 003 00 Unit Conditions Prior to the Event Unit 2 was shut down and in Mode 5 (refueling) for the P2R20 Refueling Outage when this condition was discovered on 10/29/14. There were no structures, systems or components out of service that contributed to this event.
Description of the Event
On 10/29/14 at approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />, Engineering personnel determined that the N-9A primary containment (EIIS: BD) penetration pathway had leakage that exceeded the maximum allowable primary containment leakage rate (La) value required by Technical Specification (TS) 5.5.12, Primary Containment Leakage Rate Testing Program. This determination was based on as-found leakage through the seats of two redundant feed water (ENS: SJ) check valves (CHK-2-06-28A and CHK-2-06- 96A, EIIS: ISV) that are Primary Containment Isolation Valves (PCIVs). CHK-2-06-28A is the inboard PCIV (located within the primary containment) and CHK-2-06-96A is the outboard PCIV (located just outside the primary containment) for the 'A' feed water line to the Reactor Pressure Vessel (RPV).
The valves were tested, as scheduled, on 10/28/14 using a Type C local leak rate test as required by TS 5.5.12. During the leak test, the check valves were closed but exhibited high through-seat leakage that was determined to exceed the La value of 175,584 sccm. This condition also resulted in not meeting the final as-found leakage rate for the Integrated Leak Rate Test (ILRT) of containment that was performed during P2R20.
As a result of this condition, initial notification of this event was made to the NRC (EN# 50571) on 10/29/14 at 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br /> in accordance with 10CFR 50.72(b)(3)(ii)(A) as a condition that resulted in the degradation of one of the plant principal safety barriers (i.e., primary containment, EIIS: BD).
This LER is being submitted to satisfy the below reporting requirements:
- 10CFR 50.73(a)(2)(ii)(A) — Degradation of a principal safety barrier since the TS 5.5.12 limit for La was exceeded.
- 10CFR 50.73(a)(2)(v)(C) and (D) — An event or condition that could have prevented the fulfillment of the safety function of structures or systems needed to control the release of radioactive material and mitigate the consequences of an accident since primary containment was discovered to be inoperable per TS Limiting Condition for Operation (LCO) 3.6.1.1.
- 10CFR 50.73(a)(2)(i)(B) - A condition prohibited by TS LCO 3.6.1.1 since it is reasonable to assume that this condition existed during plant power operations.
- 10CFR 50.73(a)(2)(vii)(C) and (D) - An event where a single cause or condition caused two independent trains to become inoperable in a single system designed to control the release of radioactive material and mitigate the consequences of an accident since both the inboard and outboard PCIVs for penetration N-9A were degraded due to the same cause.
CHK-2-06-28A and CHK-2-06-96A were repaired during the P2R20 Refueling Outage. The final as-left leakage rate for the I LRT was 33.9% of La.
Analysis of the Event
There were no actual safety consequences associated with the event. There were no actual plant operating transients during the Cycle 20 operating period that required the isolation of the primary containment. Unit 2 operated continuously during its Cycle 20 period of operation.
The safety objective of the primary containment is to contain the released steam in the event of the design basis loss-of-coolant accident (LOCA) to limit the release of fission products to secondary containment during a design basis event. Pipes or ducts which penetrate the primary containment and which connect to the reactor primary system, or are open to the drywell free gas space, generally are provided with at least two isolation valves in series. Valves in this category are designed to close automatically.
For the N-9A containment penetration (`A' feed water line to the RPV), CHK-2-06-28A and CHK-2-06-96A are considered as PCIVs for that penetration. CHK-2-06-28A (inboard PCIV) and CHK-2-06-96A (outboard PCIV) do not provide a safety function in the open direction. During power plant operations, the check valves are normally open to allow flow of feed water to the RPV. However, CHK-2-06-28A and CHK-2-06-96A do provide a containment isolation function in the closed position to ensure containment of radioactive material during design basis events. Also, CHK-2-06-96A performs a safety function in the closed direction to ensure that High Pressure Coolant Injection (HPCI) flow is directed to the RPV during design basis events requiring HPCI. The HPCI piping feeds into the 'A' feed water piping to the RPV in between the CHK-2-06-96A and CHK-2-06-28A check valves. Based on a review by Engineering, the leakage back through CHK-2-06-96A did not significantly affect the HPCI safety function and therefore, HPCI would have been operable for design basis events.
For design basis events that involve HPCI operation, there would not have been a concern with primary containment leakage from penetration N-9A since HPCI would be operating and feeding coolant into the RPV. For design basis events that do not involve the operation of HPCI, excessive leakage through the primary containment PCIVs (CHK-2-06-28A and CHK-2-06-96A) could have occurred. However, additional feed water system equipment would be available to minimize leakage from the containment, thereby minimizing any dose consequences to main control room personnel or the public.
Both swing check valves are supplied by Weir Valves & Controls USA (Model No. 20857-H, 24" X 20" X 24", Mark NSP214W).
Cause of the Event
The cause of the leak has been determined to be due to operational wear on the swing check valve pivot shaft and associated bushings. This operational wear resulted in a misalignment between the check valve disc and seat. Previous to this event, these check valves have demonstrated a good operational history.
Further causal analysis is being performed in accordance with the station's corrective action program to determine any underlying causes to this condition.
Corrective Actions
CHK-2-06-28A and CHK-2-06-96A were repaired during the P2R20 refueling outage. An as-left Type C local rate test was successfully performed during P2R20, thereby proving appropriate leak-tightness of the check valves in the closed position. An ILRT was performed during P2R20. The as-left results determined during the ILRT were satisfactory to assure primary containment leakage limits were met.
Additional corrective actions are being assessed as part of a causal analysis being performed in accordance with the station's corrective action program.
Previous Similar Occurrences There were no previous LERs identified involving a failure of redundant check valves resulting in exceeding the La primary containment leakage limit. The last noted local leak rate failure of either valve was CHK-2-06-96A in 1996.