05000352/LER-2016-003

LER-2016-003, Plant Shutdown Required by Technical Specification Due to a Pressure Boundary Leak

Limerick Generating Station, Unit 1
Event date:	03-20-2016
Report date:	05-18-2016
Reporting criterion:	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
Initial Reporting
ENS 51809	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown, 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
3522016003R00 - NRC Website

LER 16-003-00 for Limerick, Unit 1, Regarding Plant Shutdown Required by Technical Specification Due to a Pressure Boundary Leak

ML16139A428
Person / Time
Site:	Limerick
Issue date:	05/18/2016
From:	Libra R W Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LG-16-060 LER 16-003-00
Download: ML16139A428 (4)

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2. DOCKET

Unit Conditions Prior to the Event Unit 1 was in Operational Condition (OPCON) 1 (Power Operation) at approximately 14.5 percent power performing a planned Unit 1 shutdown to support a refueling outage. There were no structures, systems or components out of service that contributed to this event.

Description of the Event

On Sunday March 20, 2016, Limerick Unit 1 was operating at 14.5 percent power performing a planned Unit 1 shutdown to support a refueling outage (1R16). At 2154 hours0.0249 days <br />0.598 hours <br />0.00356 weeks <br />8.19597e-4 months <br />, the drywell leak inspection team identified a 0.5 gpm pressure boundary leak on the shutdown cooling (BO:EllS) testable check valve (ISV:EllS) equalizing line. The control room supervisor (CRS) entered Technical Specification (TS) 3.4.3.2 Reactor Coolant System - Operational Leakage Action "a" which requires being in at least Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Hot Shutdown TS Action was met at 0133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br /> and the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Cold Shutdown Action was met at 1401 hours0.0162 days <br />0.389 hours <br />0.00232 weeks <br />5.330805e-4 months <br />.

An investigation determined that Unit 1 Unidentified Leakage increased from 0.03 gpm to 0.06 gpm between January 16, 2016 and January 18, 2016. Leakage increased to 0.24 gpm on January 19, 2016 and continued to vary with reactor recirculation pump speed for the remainder of the cycle. The maximum leakage recorded was 1.2 gpm which stabilized at 1.0 gpm following power reduction to 95 percent. On March 20, 2016 Maintenance Technicians performed a planned drywell entry at 14.5 percent power and identified a leak on the 3/4 inch reactor side equalizing line of the 1A Shutdown Cooling (SDC) Return Line Inboard Testable Check Primary Containment Isolation Valve, HV-051-1F050A. The leak was located approximately 1/4 inch from the valve body weld and was determined to be pressure boundary leakage.

A 4-hour ENS (#51809) was completed at 2351 hours0.0272 days <br />0.653 hours <br />0.00389 weeks <br />8.945555e-4 months <br /> as required by 10CFR50.72(b)(2)(i) for the initiation of a plant shutdown required by TS. The ENS also reported an event that resulted in the condition of the plant principal safety barriers being seriously degraded per 10CFR50.72(b)(3)(ii). Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(i)(A) and 10CFR50.73(a)(2)(ii)(A).

Analysis of the Event

There was no actual safety consequence associated with this event. The potential safety consequences of this event were minimal. The Unidentified Leakage remained a small fraction of the 5 gpm TS 3.4.3.2 LCO. HPCI was unavailable less than 29 minutes due to testing and RCIC was unavailable for less than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> due to maintenance and testing during the three month period (January, February and March of 2016) of elevated drywell leakage. HPCI is designed to prevent the actuation of the automatic depressurization system (ADS) and ensure that the reactor core remains covered in the event of a small pipe break size of one-inch diameter or less.

The SDC return is a 12-inch diameter line equipped with an outboard motor operated valve (MOV) primary containment isolation valve (PCIV) and an inboard testable check valve PCIV.

The testable check valve has a bypass line that is equipped with an air operated valve (AOV) PCIV which is used to equalize the differential pressure across the testable check valve disc when the valve is opened during stroke testing. Testable check valves are also used in similar applications on low pressure coolant injection (LPCI) and core spray (CS) injection lines. The SDC valve is subject to vibration induced high cycle fatigue due to recirculation pump flow induced vibration. The LPCI and CS injection lines' socket welds are not subject to vibration induced high cycle fatigue.

The affected section of bypass line piping was replaced with a new socket weld with a 2x1 weld to improve pipe stability and minimize stresses at the toe as a result of the 2x1 weld configuration. A 2x1 weld was also applied at the similar valve body socket welds for the HV- 051-1F050A residual heat removal (RHR) side and HV-051-1F050B RHR and reactor side welds.

Cause of the Event

The Unit 1 `A' RHR Shutdown Cooling Return Check Valve equalizing line developed a crack at the toe of the weld due to high cyclic fatigue induced by vibration from the reactor recirculation system (Apparent Cause).

Corrective Action Completed The Unit 1 welds were reworked to EPRI 2x1 at select locations on the "A" and "B" RHR Shutdown Cooling Return check valve equalizing lines for HV-051-1F050A and 50B.

Corrective Action Planned The similar Unit 2 welds on equalizing lines for HV-051-2F050A and 50B will be examined and reinforced. The scope will be added into the next refueling outage (2R14) currently scheduled for April 2017.

Previous Similar Occurrences There were no previous similar occurrences of pressure boundary leakage in the past 5 years.

Component data:

System BO RHR/Low Pressure Coolant Injection System Component ISV Valve, Isolation Component number HV-051-1F050A Manufacturer A585 Weir Valves & Controls USA Inc.

Model number 50301-A Serial Number 2-50301-A