05000366/LER-2017-004

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LER-2017-004, 1 OF 3
Edwin I Hatch Nuclear Plant Unit 2
Event date: 6-30-2017
Report date: 08-24-2017
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
LER closed by
IR 05000366/2017004 (6 February 2018)
3662017004R00 - NRC Website

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as "(EIIS Code RV)"

Event Description

On June 30 2017, with Unit 2 at 100 percent rated thermal power (RTP), "as-found" testing of the 3-stage main steam safety relief valves (SRVs) (EDS Code RV) showed that two of the eleven main steam SRVs that were tested had experienced a drift in pressure lift setpoint during the previous operating cycle such that the allowable technical specification (TS) surveillance requirement (SR) 3 4.3.1 Omit of 1150 +1- 34.5 psig had been exceeded. Below is a table illustrating the Unit 2 SRVs that failed as found testing results after being removed from service during the Spring 2017 refueling outage.

MPL Drift 2B21-F013C - 39 psig 2B21-F013E - 49 psig Event Cause Analysis The SRV pilots were disassembled and inspected while investigating the reason for the drift. It was found that the abutment gap closed prematurely during testing using a linear variable differential transformer (LVDT) to measure pilot stroke distance The pre-mature abutment gap closure is most likely due to loose manufacturing tolerances leading to SRV setpoint drift.

Safety Assessment This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) because a condition occurred that is prohibited by TS 3.4.3. Specifically, an example of multiple test failures is given in NUREG-1022, Revision 3, "Event Reporting Guidelines 10 CFR 50.72 and 50.73", which describes the sequential testing of safety valves This example notes that "Sometimes multiple valves are found to lift with set points outside of technical specification limits." NUREG-1022 further states in the example that "discrepancies found in TS surveillance tests should be assumed to occur at the time of the test unless there is firm evidence, based on a review of relevant information (e.g., the equipment history and the cause of failure), to indicate that the discrepancy occurred earlier. However, the existence of similar discrepancies in multiple valves Is an indication that the discrepancies may well have arisen over a period of time and the failure mode should be evaluated to make this determination." Based on this guidance, the determination was made that this "as found" condition is reportable under the reporting requirements of 10 CFR 50.73(a)(2)(i)(B).

There are eleven SRVs located on the four main steam lines within the drywell in between the reactor pressure vessel (RPV) (EIIS Code RPV) and the inboard main steam isolation valves (MSIVs) (EIIS Code ISV). These SRVs are required to be operable during Modes 1, 2, and 3 to limit the peak pressure in the nuclear system such that it will not exceed the applicable ASME Boiler and Pressure Vessel Code Limits for the reactor coolant pressure boundary. The SRVs are tested in accordance with TS Surveillance Requirement 3.4.3.1 in which the valves are tested as directed by the In-Service Testing Program to verify lift set points are within their specified limits to confirm they would perform their required safety function of overpressure protection.

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CONTINUATION SHEET

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RUINER

004 The SRVs must accommodate the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code Limit of 1375 psig peak vessel pressure, has been defined by an event involving the dosure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV position switches with the reactor ultimately shutting down as the result of a high neutron flux trip (a scenario designated as MSIVF).

The two SRVs which failed to meet their Tech Spec required actuation pressure setpoint lifted early_ None of the eleven SRVs tested this cycle had as-found test results out of range high Therefore, since the two identified SRVs lifted earlier than expected, the ASME Code Limit of 1375 psig peak vessel pressure would be maintained under normal and accident conditions. The opening of one or more SRVs at lower pressures would result in a less severe transient with reduced peak vessel pressure Also the slightly lower actuating pressure does not pose a significant LOCA initiator threat because the reactor steam dome would not experience >1100 psig during normal operation; therefore, these valves would not have inadvertently opened Based on the observed setpoint drift slightly low, the overpressure protection system would have continued to perform its required safety function if called upon in its 'as found" condition Therefore, this event had no adverse impact on nudear safety and was of very low safety significance

Corrective Actions

The vendor specifications was revised to tighten as-left tolerances of abutment and pre-load gap, increase the minimum set for abutment pressure at the high end of specification, and tighten diametrical and face run-out tolerances for bellows assembly on pre-load spacer mounting end.

Previous Similar Events

specifications to tighten as-left tolerances of abutment and pre-load gap, increase the minimum set for abutment pressure at the high end of specification, and tighten diametrical and face run-out tolerances for bellows assembly on pre-load spacer mounting end.

2-stage SRVs with 3-stage SRVs which typically do not exhibit set point drift. The setpoint drift was out of spec high while the event discussed in LER 1-2016-004 have failed to meet acceptance criteria by drifting out of spec low.

2-stage SRVs with 3-stage SRVs which typically do not exhibit set point drift. The setpoint drift was out of spec high while the event discussed in LER 1-2016-004 have failed to meet acceptance criteria by drifting out of spec low.

2-stage SRVs with 2-stage SRVs whose pilot discs had undergone a platinum surface treatment which was considered at that time to be the long term fix for this corrosion bonding issue.