05000461/LER-2017-004

LER-2017-004, Main Steam Isolation Valve Local Leak Rate Test Limit Exceeded During Refueling Outage

Clinton Power Station, Unit 1
Event date:	05-12-2017
Report date:	11-22-2017
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
4612017004R01 - NRC Website

LER 17-004-00 For Clinton Power Station, Unit 1 Re: Main Steam Isolation Valve Local Leak Rate Test Limit Exceeded During Refueling Outage

ML17194A824
Person / Time
Site:	Clinton
Issue date:	07/10/2017
From:	Stoner T R Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SRRS 5A.108, U-604358 LER 17-004-00
Download: ML17194A824 (5)

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the infor1nation collection.

3. LER NUMBER

2017 - 01 004

PLANT AND SYSTEM IDENTIFICATION

General Electric—Boiling Water Reactor, 3473 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX]

EVENT IDENTIFICATION

Main Steam Isolation Valve Local Leak Rate Test Limit Exceeded During Refueling Outage A. Plant Operating Conditions before the Event Unit: 1 Mode: 5 Event Date: May 12, 2017 Event Time: 0045 Mode Name: Refueling Reactor Power: 00 percent

B. DESCRIPTION OF EVENT

On May 12, 2017 0045 CDT during Refueling Outage C1R17, Clinton Power Station (CPS) performed a local leak rate test (LLRT) on its Main Steam Isolation Valves (MSIV) and discovered that the as-found leakage for the `D' main steam line (MSL) exceeded the Technical Specification (TS) 3.6.1.3, Primary Containment Isolation Valves, Surveillance Requirement (SR) 3.6.1.3.9 limits. During Modes 1, 2, and 3, TS SR 3.6.1.3.9 requires MSIV leakage for a single MSL to be less than or equal to 100 standard cubic feet per hour (scfh) (47,195 standard cubic centimeters per minute (sccm)) and requires the combined leakage rate for all four MSLs to be less than or equal to 200 scfh (94,390 sccm) when tested at 9 psig. The as-found leakage for the `D' MSL was 53,921.61 sccm for the `D' inboard MSIV (1621F022D) and 59,698.8 sccm for the `D' outboard MSIV (1621F028D). The as-found combined min-path leakage for all four MSLs was 102,463 sccm. It was also identified that MSIV 1621F028A exceeded the administrative allowable limit established by plant procedures.

An event investigation determined the as-found condition of MSIVs 1621F028A, 1621F022D, and 1621F028D did not reveal any damage, only normal wear indications were noted. The valve dimensions were within tolerance. The extent of the condition identified in this report was limited to the CPS MSIV LLRTs and the frequency with which maintenance is performed. Prior LLRT results were evaluated for component failure that would be indicated by an abrupt rise in leak rates, however, no failures were noted. Evidence of gradual increase in LLRT rates were apparent on all valves tested.

The internal maintenance for the MSIVs are based on as-found LLRT test and diagnostic test results. There are no internal valve preventative maintenance tests that are performed at a given frequency.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 01 004

C. CAUSE OF EVENT

The apparent cause the leakage rate of affected valves exceeding TS SR limits is expected wear. An event investigation determined the as found condition of MSIVs 1621F028A, 1621F022D, and 1621F028D did not reveal any damage only normal wear indications.

D. SAFETY ANALYSIS

There were no safety consequences associated with this condition. This event is reportable under the provisions of 10 CFR 50.73(a)(2)(ii)(A) for the condition of the nuclear power plant including its principal safety barriers being seriously degraded. The event described in this report is also a condition prohibited by TS under the provisions of 10 CFR 50.73(a)(2)(i)(B). The MSIV LLRT leakage values observed during the surveillance during C1R17 exceeded TS limits.

A plant shutdown was not required since the plant was in Mode 5 at the time of discovery.

Discovery of the reportable conditions was the result of a planned activity to perform leak rate testing on the MSIVs. Systems necessary to maintain the plant per TS requirements during the performance of refueling outage activities in Mode 5 remained available to perform their safety function.

This event report does not identify any safety system functional failures.

E. CORRECTIVE ACTIONS

The MSIVs were refurbished to rectify the condition that caused local leak rate test failures. The cause of the leakage values which exceeded TS requirements was expected wear. CPS repaired valves 1B21F028A, 1621F022D, and 1B21F028D so that as-left leakage values complied with TS SR 3.6.1.3.9. Repair activities included cleaning or replacing valve internals and repacking valves to correct the condition which caused the event.

F. PREVIOUS SIMILAR OCCURENCES

On February 3, 2010, after entering Mode 2 (Startup) following refueling outage C1R12, it was discovered that a primary containment LLRT performed on feedwater check valve, 1621-F032B, exceeded its acceptance criteria. Technical Specification SR 3.6.1.3.11 requires that the combined leakage rate for both primary containment feedwater penetrations comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 01 004 to be less than or equal to 2 gallons per minute (gpm) for the worst of the isolation valves.

The measured leakage rate for 1621-F032B was reported to be 2.5 gpm.

This leakage rate is greater than that assumed in the plant safety analysis. The cause of the 1B21-F032B check valve to fail its leak rate test was age-related degradation of the lubrication causing increased friction in the actuator. Corrective action for this event included establishing preventive maintenance activities to lubricate and overhaul the actuators.

Program and Initial Plant Operation Results in Unacceptable Main Steam Isolation Valve Leakage Rates On March 20, 1988, at 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br />, with the plant in Mode 4 (Cold Shutdown) and the reactor at approximately 150 degrees Fahrenheit and atmospheric pressure, local leak rate testing by test engineers identified that the primary containment leakage rates of the MSIVs on MSL D exceeded TS limits of 13,214 sccm (28 scfh) per line. The cause of the excessive leakage has been attributed to component wear based on service seen during the power ascension program and initial plant operation. The corrective action included reworking the two MSIVs by lapping the seats and machining the poppets. Subsequent leak rate testing of MSL D was satisfactorily performed.

G. COMPONENT FAILURE DATA

Main Steam Isolation Valves 1621F028A, 1621F022D, and 1621F028D are manufactured by Atwood & Morrill Company. The model number for the valves is 40012.