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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 53646 | 5 October 2018 05:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | En Revision Imported Date 10/25/2018 | EN Revision Text: MAIN STEAM ISOLATION VALVES EXCEEDED PRIMARY CONTAINMENT LOCAL LEAK RATE ACCEPTANCE CRITERIA At 0520 (CDT), on October 05, 2018, it was discovered that a Primary Containment local leak rate test performed on Main Steam Isolation Valves (MSIV) exceeded its acceptance criteria. During Mode 1, 2, and 3, Surveillance Requirement 3.6.1.3.10 requires MSIV leakage for a single MSIV line to be less than or equal to 106 standard cubic feet per hour (scfh) when tested at 29 psig and Surveillance Requirement 3.6.1.3.12 requires the combined leakage rate for all MSIV leakage paths to be less than or equal to 212 scfh when tested at 29 psig. As-found for the 'C' MSIV line leakage results were unquantifiable and gave a (minimum) path value greeter than 160 scfh. This leakage rate lead to Surveillance Requirement 3.6.1.3.10 and 3.6.1.3.12 limits to be exceeded. This event is being reported as a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded per 10 CFR 50.72(b)(3)(ii)(A) since the Primary Containment Isolation Valves leakage limits for MSIVs were exceeded. The NRC Resident Inspector has been notified.
CNS (Cooper Nuclear Station) is retracting the 8-hour non-emergency notification made on October 5, 2018 at 0520 CDT (EN# 53646). Subsequent evaluation concluded that overall as-found 'C' MSIV leakage rate was not at a level that exceeded the surveillance requirement 3.6.1.3.10 and 3.6.1.3.12 limits and thus the Primary Containment Isolation Valve leakage rate limits for the MSIVs were not exceeded. The NRC Senior Resident Inspector has been notified. Notified the R4DO (Drake). | Main Steam Isolation Valve Primary containment | |
| ENS 42939 | 27 October 2006 08:37:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Containment Pathway Declared Inoperable Following Local Leak Rate Testing | This condition is being reported in accordance with the requirements of 10CFR50.72(b)(3)(ii) as an event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. During local leak rate testing (LLRT) of the 'B' reactor feed line to the reactor, both primary containment boundary check valves (RF-CV-13CV and RF-CV-14CV) failed LLRT. Efforts to quantify the leak rate were unsuccessful. The maximum leakage indicated on the test equipment used is 424 standard cubic feet per hour (SCFH). The allowable leakage (La) for Primary Containment is 317.41 SCFH. The affected primary containment pathway was declared inoperable. CNS is in MODE 5 for refueling with vessel level greater than 21 feet above the flange and the spent fuel pool gates removed. No fuel movements or operations with the potential to drain the reactor vessel were in progress. Primary containment is not required per CNS Technical Specifications in this mode of operation. The NRC Senior Resident Inspector has been notified.
This notification is being made to retract Event Notification #42939 which reported a Primary Containment isolation pathway being declared inoperable following Local Leak Rate Testing (LLRT); specifically, the 'B' reactor feed line containment boundary inboard and outboard check valves. CNS has determined that the LLRT test method established test conditions that were not representative of the conditions under which the check valves would be required to perform their safety function. Inspection, in-situ testing and evaluation demonstrated the reactor feed line check valves would have performed their isolation functions during credible accident conditions. Thus, the condition was not a serious degradation of a principal safety barrier. During the normal plant shutdown to refueling outage RE23 with gradually decreasing feed flow, the check valve discs had not completely settled into the valve seats. Due to the slow pressurization rate of the LLRT working medium (air), sufficient differential pressure across the valve could not be developed which resulted in failing the test. Inspection of the check valves in their as-found condition revealed that a slight contact between the disc and the top of seat left a crescent shaped gap which, at the maximum (the bottom), was no more than 1/2 inch. An in-situ test measured the force needed to settle the disc into the seat and found it equated to a flow related pressure drop of less than 1 psid. A review of accident sequences concluded that post-accident pressurization of the feed lines would have closed the check valves for all credible accident scenarios. Check valve disc to pivot arm spacing adjustments were made to ensure sufficient disc float to reseat and fully close the valve from any open position. Subsequently, LLRT results were satisfactory. The licensee notified the NRC Resident Inspector. R4DO (Johnson) notified. | Primary containment |