05000263/LER-2017-003

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LER-2017-003, Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits
Monticello Nuclear Generating Plant
Event date: 04-20-2017
Report date: 06-14-2017
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
LER closed by
IR 05000263/2017003 (6 November 2017)
2632017003R00 - NRC Website
LER 17-003-00 for Monticello Regarding Main Steam Isolation Valve Leakage Exceeds Technical Specification Limits
ML17166A137
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 06/14/2017
From: Gardner P A
Northern States Power Co, Xcel Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-17-046
Download: ML17166A137 (4)


comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by inlernet 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.

EVENT DESCRIPTION

On April 20, 2017, with the plant at 0% power in Mode 5 (Refueling), during refueling outage 1R28, Local Leak Rate Testing (Appendix J) of AO-2-86C, "13 Outboard Main Steam [SB] Isolation Valve [ISV]," had an unacceptable as-found leak rate. The measured leak rate was 187.8 standard cubic feet per hour (scfh) which exceeds the Monticello Nuclear Generating (MNGP) Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.3.12 limit of 100 scfh. AO-2-86C was declared inoperable and the valve was subsequently disassembled to make repairs. The valve's stem, discs, upper/lower wedges, disc retainer, and wedge pin were replaced and retested. The as-left leak rate after completion of the work was 2.64 scfh.

This component failure is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS 3.6.1.3, "Primary Containment Isolation Valves," since AO-2-86C likely had been inoperable for greater than the TS 3.6.1.3, Required Action A.1, Completion Time of

8 hours
9.259259e-5 days
0.00222 hours
1.322751e-5 weeks
3.044e-6 months

to isolate a main steam line, and the Completion Time for TS 3.6.1.3, Required Action F, to be in Mode 3 in

12 hours
1.388889e-4 days
0.00333 hours
1.984127e-5 weeks
4.566e-6 months

and Mode 4 in

36 hours
4.166667e-4 days
0.01 hours
5.952381e-5 weeks
1.3698e-5 months

when the completion time of A.1 is not met.

The basis for the reportable condition is the change in wear rate associated with AO-2-86C valve internals. In 2011, AO-2-86C (Anchor Darling model W9324183 18"-900 venturied double) was disassembled and showed unexpected accelerated wear and excessive damage. The stem, upper and lower wedges, disc retainers and discs were replaced. The last as-found leak rate for AO-2-86C was 64.1 scfh in the 2015 refueling outage (1R27). After the actuator was replaced in 1R27 the as-left leakage was 4.4 scfh. Based on these data points it is concluded that the leak rate increased during the cycle and the valve likely had exceeded the TS SR limits during the cycle preceding 1R28.

EVENT ANALYSIS

The event was determined to be reportable in accordance with 10 CFR 50.73 (a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications." Specifically, this component failure is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS 3.6.1.3, "Primary Containment Isolation Valves," since AO-2-86C likely had been inoperable for greater than the TS 3.6.1.3, Required Action A.1, Completion Time of

8 hours
9.259259e-5 days
0.00222 hours
1.322751e-5 weeks
3.044e-6 months

to isolate a main steam line, and the Completion Time for TS 3.6.1.3, Required Action F, to be in Mode 3 in

12 hours
1.388889e-4 days
0.00333 hours
1.984127e-5 weeks
4.566e-6 months

and Mode 4 in

36 hours
4.166667e-4 days
0.01 hours
5.952381e-5 weeks
1.3698e-5 months

when the completion time of A.1 is not met.

This event is not classified as a safety system functional failure as the inboard valve was fully operational.

003 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet 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.

Main Steam line "C" (AO-2-80C) was tested for both leak rate and closing time over the past cycle and each test was completed satisfactorily. Therefore, the primary containment isolation capability of the main steam lines remained operable which ensured the required isolation safety function was maintained.

CAUSE

The failure was attributed to oscillating of the disc and wear on the trunnion pin. The oscillation caused wear between the downstream disc trunnion and mating upper wedge hole. As the wear increased, the disc dropped, increasing the gap between the disc retainer and disc groove thereby allowing further rotation of the disc. Eventually, the corner at the end of the disc groove started to contact one of the ends of the retainer plate and wear into it. This resulted in interference between the downstream disc groove area and the bottom corner of the disc retainer. This interference prevented or restricted the ability of the upper part of the downstream disc to move axially towards its corresponding body seat thereby resulting in a gap or reduced seating force in portions of the seat. Based on this, the increased leakage of the valve is attributable to wear which led to reduced seating force or a gap (due to interference) in the valve disc as it contacts the valve body seat.

CORRECTIVE ACTION

The entirety of the internal disc pack was replaced. This includes the stem, discs, upper/lower wedges, disc retainer, and wedge pin. A modification was made to hard face the trunnion outer diameter, upper wedge hole inner diameter, and disc grooves with a Stellite 21 overlay to help reduce the amount of wear. Other improvements were made to the disc retainers as well.

PREVIOUS SIMILAR EVENTS

There were no previous similar licensee event reports in the past three years.

ADDITIONAL INFORMATION

The Institute of Electrical and Electronics Engineer codes for equipment are denoted by [XX].

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