05000352/LER-2012-007
| Limerick Generating Station, Unit 1 | |
| Event date: | 08-20-2012 |
|---|---|
| Report date: | 10-19-2012 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| Initial Reporting | |
| ENS 48334 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded, 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material |
| 3522012007R00 - NRC Website | |
Unit Conditions Prior to the Event Units 1 and 2 were in Operational Condition (OPCON) 1 (Power Operation) at approximately 100% power. There were no structures, systems or components out of service that contributed to this event.
Description of the Event
On Monday, August 20, 2012, an evaluation of loss of coolant accident (LOCA) transient voltage was being performed. During the evaluation, it was identified that there was a potential for some primary containment (EIIS:NH) isolation valves (PCIVs)(EIIS:ISV), with closing stroke times of approximately 5 seconds, to fail to fully close during a design basis accident (DBA) LOCA with offsite power available.
At 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, six normally closed drywell purge PCIVs in the CAC system were declared inoperable. Technical Specification 3.6.3 Primary Containment Isolation Valves, action "a" was entered and met.
Also, since the Unit 1 and Unit 2 drywell purge exhaust PCIVs were declared inoperable, TS 3.6.5.2.1 Reactor Enclosure Secondary Containment Automatic Isolation Valves and TS 3.6.5.2.2 Refueling Area Secondary Containment Automatic Isolation Valves actions were entered and met.
It was later identified that LOCA break sizes that are less than the DBA LOCA could affect additional PCIVs with closing stroke times that exceed 5 seconds. This resulted in additional PCIVs being declared PCIVs and 15 Unit 2 PCIVs were affected.
During a LOCA, most of the affected PCIVs are designed to automatically close on a high drywell pressure signal. The reactor water clean-up (RWCU) PCIVs do not close on high drywell pressure but do close on a reactor water level 2 signal. The emergency core cooling system (ECCS) LOCA initiation at reactor level 1 initiates a load shed of 480 VAC loads which results in interruption of PCIV automatic closure for approximately 3.5 seconds; when power is restored the remaining open PCIVs should resume closure.
The PCIV closing circuit design includes a dead zone as the PCIV approaches full closure. This condition exists in the valve stroke after the valve indicates closed but prior to full valve closure and the opening of the closing torque switch. If the valve is in this dead zone it will not resume closure when 480 VAC power is restored.
In addition the valve indicating lights indicate the valve is fully closed (i.e., red light extinguished) when the valve is in the dead zone.
It was identified that there was a potential during a LOCA for the reactor level 1 LOCA 480 VAC load shed to occur as PCIVs enter the dead zone. This would result in a failure of the PCIV to resume closing when power is restored. Due to differing valve stroke times, it is likely that only one or no penetrations would be impacted, depending on the event time line.
A design change was implemented that installed a jumper on the LS-8 contact on the 24 affected valves and eliminated the PCIV vulnerability. Three affected Unit 1 CAC valves were restored to operable status by reducing the stroke time which results in full valve closure prior to the LOCA load shedding. Three Unit 2 CAC valves remain inoperable and will be restored to operable when their stroke times are reduced.
An ENS (#48334) notification to the NRC was completed on Monday, September 21, 2012, at 2129 ET. The event was reported as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material per 10CFR50.72(b)(3)(v)(C). The event was also reported as a degraded primary containment principle safety barrier that significantly degrades plant safety per 10CFR50.72(b)(3)(ii)(A).
This event involved a condition that could have prevented fulfillment of the primary containment safety function. Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(v)(C).
This event involved a condition that resulted in the condition of the nuclear power plant, including its principle safety barrier of primary containment, being seriously degraded. Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(ii)(A).
This event involved a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(ii)(B).
This event involved a common-cause inoperability of independent trains. Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(vii).
This event involved conditions prohibited by TS. Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(i)(B).
The actions for the following TS were not met from initial operation since the condition was not previously identified:
TS 3.6.3 Primary Containment Isolation Valves TS 3.6.1 Primary Containment Integrity TS 3.6.1.2 Primary Containment Leakage TS 3.6.5.1.1 Reactor Enclosure Secondary Containment Integrity TS 3.6.5.1.2 Refueling Area Secondary Containment Integrity TS 3.6.5.2.1 Reactor Enclosure Secondary Containment Automatic Isolation Valves TS 3.6.5.2.2 Refueling Area Secondary Containment Automatic Isolation Valves
Analysis of the Event
There was no actual safety consequence associated with this event.
The potential safety consequences of this event were more than minimal since the plant operated with multiple inoperable PCIVs on normally open penetrations since initial operation. Overall primary containment leakage would have been adversely affected by a failure of both PCIVs in an open penetration to fully close with an isolation signal present.
The LS-8 limit switch contact feature is required on the PCIVs without locking gear sets. The open LS-8 contact prevents multiple close demands on the valve when the torque switch relaxes. The CAC containment purge PCIVs do not have locking gear sets. A modification is being evaluated for the CAC purge valves to install the locking gear sets, jumper the LS-8 contact and restore the original stroke times.
PCIVs in the following systems were affected on each unit as follows:
1.The RWCU suction line is a normally open 6-inch penetration.
Both the inboard and outboard PCIVs were affected by this condition.
2.Two loops of drywell chilled water (DWCW) are normally in service using four 8-inch penetrations. The four inboard isolation PCIVs were affected. The outboard PCIVs were not affected.
3.Three normally closed CAC 24-inch penetrations used for primary containment purge were affected. Each of the three lines use air operated PCIVs in series with the affected PCIVs. The air operated PCIVs were not affected.
Two additional normally closed CAC 2-inch penetrations used for primary containment venting to the reactor enclosure equipment compartment exhaust were affected. Each of the two lines use air operated PCIVs in series with the affected PCIVs. The air operated PCIVs were not affected.
Two additional normally closed CAC 1-inch penetrations used for primary containment nitrogen makeup were affected. The lines use two inboard solenoid valve PCIVs in series with the one affected PCIV. The inboard solenoid valve PCIVs were not affected.
4.The instrument gas compressor suction line is a normally open one-inch penetration. The inboard PCIV was affected. The outboard PCIV is an air operated PCIV which was not affected.
5.The suppression pool clean-up pump suction line is a normally closed 6-inch penetration. Both the inboard and outboard PCIVs were affected.
Cause of the Event
The original design for certain PCIVs contained a design flaw that could potentially prevent the equipment from performing the intended safety function.
Corrective Action Completed Following the identification of the design deficiency a design change was implemented on 24 affected PCIVs. The design change installed a jumper which removed the LS-8 contact feature. This eliminated the dead zone on these PCIVs. In addition, three Unit 1 affected CAC drywell purge PCIVs' stroke times were reduced to remove the vulnerability.
Corrective Action Planned Three inoperable Unit 2 affected CAC drywell purge PCIVs' stroke times will be reduced to remove the vulnerability.
Previous Similar Occurrences There were no previous similar conditions in the prior 5 years of PCIVs being declared inoperable due to electrical design deficiencies.