05000237/LER-2013-005
| Dresden Nuclear Power Station, Unit 2 | |
| Event date: | 11-14-2013 |
|---|---|
| Report date: | 01-13-2014 |
| 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 |
| 2372013005R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
Dresden Nuclear Power Station (DNPS) Units 2 is a General Electric Company Boiling Water Reactor with a licensed maximum power level of 2957 megawatts thermal. The Energy Industry Identification System codes used in the text are identified as [XX].
A. Plant Conditions Prior to Event:
Unit: 02 Event Date: 11-14-2013 Event Time: 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> CDT Reactor Mode: 5 Mode Name: Refueling Power Level: 000 percent
B. Description of Event:
On November 14, 2013, with Unit 2 shutdown for refueling outage D2R23, both the 2-0220-58B Feed Water [SJ] Inboard Check Valve and the 2-0220-62B Feed Water Outboard Check Valve failed Local Leak Rate Testing (LLRT) acceptance criteria. Specifically, the "as-found" leak rate for both valves was above the administrative acceptance criteria of 45.0 standard cubic feet per hour (scfh), and based upon the leakage rate observed, leakage was also determined to exceed the limits for primary containment [NH] leakage as specified in Technical Specification (TS) 5.5.12.c.
During testing, the in-series Feed Water "B" loop Containment Isolation valve volumes could not be pressurized with full flow service air through the test tap configuration. The injected air immediately passed through the check valve and out the test vent tap, indicating that the valve disc was not fully seated. Therefore, the check valve leakage rates could not be quantified, and based upon the diameter of the valves, it was determined that primary containment leakage rate limits would be exceeded through these valves. Repair and testing of the valves was successfully completed using improved maintenance practices identified in the stations root cause report.
These valves are considered primary containment isolation valves, and as such, are required to ensure that an adequate primary containment boundary is maintained. Therefore, this event is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(A), "any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded" and 10 CFR 50.73(a)(2)(v)(C) "any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.
C. Cause of Event:
A root cause analysis (RCA) was performed for D2R22 in 2012 that evaluated failures experienced during the previous refueling outage. Inspections of the failed valves have shown that the failures have been due to a failed disc-to-seat-ring seal. Specifically, the tilting disc has a tendency to hang in the open position prior to full closure when closed slowly and side-loaded. This failure mode is due to friction at the seat or hinge pin which is caused by wear and corrosion product buildup on the hinge pins, disc and seat ring ears, and seat assembly. The RCA performed determined that the best practices for installation and testing of the valves in order to maximize reliability had not been implemented.
D. Safety Analysis:
The Feed Water Primary Containment Isolation Valves (PCIV) provide containment isolation in the event of fission product release and reactor coolant pressure boundary isolation in the event of a feed water line break. The "B" Feed Water loop contains additional Safety-Related isolation valves upstream of the PCIVs to prevent High Pressure Coolant Injection (HPCI) backflow into the Feed Water system. These isolation valves were leak tested during the outage, and were determined to have been fully seated in the closed position and provided the ability to isolate containment.
It was conservatively determined that the unit was susceptible to a Feed Water line breech outside of containment with the Feed Water PCIVs failed. The Dresden FSAR describes any feedwater breech in the X-area (reactor building) as being bounded by the effects of a main steam line breech. The low probability of pipe rupture and location of the potential Feed Water line breech inside the X-area minimize the risk significance of the two primary containment isolation valve failures. The impact to the Core Damage Frequency and Large Early Release Frequency has been determined to be non risk significant.
E. Corrective Actions:
As a result of the root cause Dresden developed and implemented valve maintenance and testing best practices. Additionally, the station is evaluating long term actions to enhance the design of the Feed Water check valves and evaluating installing larger diameter test taps to provide a greater ability to test the valves at simulated accident conditions.
F. Previous Occurrences:
A search was performed to show previous occurrences for this event, with the following results:
Date Description Dresden Unit 3 — D3R16 LLRT failures of two feedwater violation.
Dresden Unit 3 — D3R21 LLRT failures of two feedwater IR 1135779 11/04/2010 check valves. Equipment Apparent Cause Analysis (EACE) performed.
Dresden Unit 2- D2R22 LLRT failures of all feedwater IR 1288720 11/10/2011 check valves; Root Cause Analysis (RCA) of the failures was requested. IR 1336479 initiated and documents the performance of the RCA.
The search revealed that there was one prior incident of feedwater check valve LLRT failures since the implementation of Option B of 10CFR 50 Appendix J, that specifically resulted in Appendix J requirements not being met. Additionally, there were two similar recent events where multiple feedwater check valves failed but did not result in not meeting Appendix J criteria; an EACE and a subsequent RCA were performed to determine the failure mechanism, cause, and perform corrective actions to improve reliability.
G. Component Failure Data:
Manufacturer Model Part Type Vendor Crane (Chapman) Model L973 N/A Tilting Disc Check Valve Crane Nuclear, Inc