05000374/LER-2017-004

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LER-2017-004, Two Main Steam Safety Relief Valves Failed Inservice Lift Inspection Pressure Test
Lasalle County Station, Unit 2
Event date: 02-17-2017
Report date: 07-14-2017
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3742017004R01 - NRC Website
LER 17-004-01 for LaSalle, Unit 2, Regarding Two Main Steam Safety Relief Valves Failed Inservice Lift Inspection Pressure Test
ML17195A306
Person / Time
Site: LaSalle, Lasalle Exelon icon.png
Issue date: 07/14/2017
From: Vinyard H T
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA17-063
Download: ML17195A306 (4)


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 2. DOCKET NUMBER - 01 004

PLANT AND SYSTEM IDENTIFICATION

LaSalle County Station Unit 2 is a General Electric Boiling Water Reactor with 3546 Megawatts Thermal Rated Core Power.

The main steam safety relief valves (SRVs) are designed to prevent over-pressurization of the reactor pressure vessel (RPV) during transients and abnormal conditions, which protects against a failure of the reactor coolant pressure boundary (RCPB).

There are thirteen SRVs installed on the four main steam lines, which discharge near the bottom of the suppression pool to condense the steam through SRV tailpipes that exhaust beneath the suppression pool surface.

CONDITION PRIOR TO EVENT

2017

DESCRIPTION OF EVENT

During the February 2017 Unit 2 refueling outage L2R16, two main steam SRVs did not pass the lift pressure requirements of the Inservice Testing (1ST) Program and Technical Specification (TS) Surveillance Requirement 3.4.4.1. Both SRVs lifted outside their tolerance and below their expected lift pressures. On February 16, 2017, SRV 2821-F013C was removed and set- pressure tested. SRV 2B21-F013C was required to lift within plus or minus three percent of 1175 psi (i.e., 1175 psi plus or minus 35.2 psi), but actually lifted at 1131 psi. On February 17, 2017 SRV 2B21-F013L was removed and set-pressure tested.

SRV 2B21-F013L was required to lift within plus or minus three percent of 1195 psi (i.e., 1195 psi plus or minus 35.8 psi), but actually lifted at 1130 psi.

CAUSE OF EVENT

Disassembly and inspection of valves 2B21-F013C and 2B21-F013L were performed at NWS Technologies to determine the cause for the failures. The vendor reported that for both valves all the spring tolerances were within the acceptance limits. There were no other signs of degradation or any other issue that would affect the set-points. Second lift tests for both valves were satisfactory and were within the plus or minus three percent tolerance of the set pressures.

The cause for 2621-F013C and 2B21-F013L to fail their SRV set pressure tests was found to be indeterminate.

REPORTABILITY AND SAFETY ANALYSIS

This condition was discovered while Unit 2 was outside the mode of applicability for TS 3.4.4, Safety/Relief Valves (Modes 1, 2 and 3); however, multiple test failures are reportable under 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the plant's Technical Specifications.

The safety significance of this condition was minimal. The out-of-tolerance lift pressures were discovered while the plant was in Mode 5 during a refueling outage and the SRVs were not required to be operable. Both SRVs lifted prior to their expected lift pressures, which is conservative in regards to maintaining reactor pressure vessel overpressure limits.

CORRECTIVE ACTIONS

Both SRVs 2B21-F013C and 21321-F013L were replaced during the outage. Vendor lab testing was performed that did not identify a cause of the failures.

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 2. DOCKET NUMBER - 01

PREVIOUS OCCURRENCES

A review of past events for both units identified a previous occurrence in 2015 on Unit 2 for multiple SRV test failures as follows.

During the February 2015 Unit 2 refueling outage L2R15, two main steam safety relief valves (SRV) did not pass TS Surveillance Requirement 3.4.4.1 and Inservice Testing Program lift pressure requirements. Both SRVs lifted below their expected lift pressures. SRV 2621-F013S was required to lift within plus or minus three percent of 1150 psi (i.e., 1150 psi plus or minus 34.5 psi) and actually lifted at 1099 psi. SRV 2B21-F013M was required to lift within plus or minus three percent of 1195 psi (i.e., 1195 psi plus or minus 35.8 psi) and actually lifted at 1145 psi. A failure analysis was conducted by a vendor testing laboratory, but the cause for the valves lifting below their set-point was indeterminate.

COMPONENT FAILURE DATA

Manufacturer: Crosby Device: Main Steam Safety Relief Valves, ASME Section III, Class 1 Component ID: Style HB-65-BP, Size 6R10 2017 004