05000321/LER-2016-004

LER-2016-004, Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

Edwin I. Hatch Nuclear Plant Unit 1
Event date:	3-30-2016
Report date:	5-26-2016
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3212016004R00 - NRC Website

LER 16-004-00 for Edwin I. Hatch Nuclear Plant Unit 1 RE: Safety Relief Valves as Found Settings Resulted in Not Meeting Tech Spec Surveillance Criteria

ML16147A344
Person / Time
Site:	Hatch
Issue date:	05/26/2016
From:	Pierce C R Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-0772 LER 16-004-00
Download: ML16147A344 (7)

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.

a DOCKET NUMBER 05000-321 a LER NUMBER

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EllS Code XX).

DESCRIPTION OF EVENT

On March 30 2016, with Unit 1 at 100 percent rated thermal power (RTP), "as-found" testing of the 3-stage main steam safety relief valves (SRVs) (EIIS 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 limit of 1150 +/- 34.5 (+1- 3%) psig had been exceeded. Below is a table illustrating the Unit 1 SRVs that failed as found testing results after being removed from service during the Spring 2016 refueling outage.

MPL

Lift Pressure Percent Drift 1621-F013D 1B21-F013E 1113 psig 1106 psig -3.2% -3.8%

CAUSE OF EVENT

The SRV pilots were disassembled and inspected while investigating the reason for the drift. SNC has determined that the linear variable differential transformer (LVDT) used to measure displacement confirmed the abutment gap closed pre- maturely. The pre-mature abutment gap closure is most likely due to loose manufacturing tolerances leading to SRV setpoint drift.

REPORTABILITY ANALYSIS AND 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).

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 intemet 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 informaton collection.

NO

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. 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 closure 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 (3.2% low and 3.8% low).

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.

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 nuclear safety and was of very low safety significance.

CORRECTIVE ACTIONS

The vendor specifications will be 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.

Edwin I. Hatch Nuclear Plant Unit 1 05000-321 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 intemet 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.

ADDITIONAL INFORMATION

Other Systems Affected: None Failed Components Information:

Master Parts List Number: 1621-F013D, E Manufacturer: Target Rock Model Number: 0867F Type: Relief Valve Manufacturer Code: T020 EIIS System Code: SB Reportable to EPIX: Yes Root Cause Code: B EIIS Component Code: RV Commitment Information: This report does not create any licensing commitments.

PREVIOUS SIMILAR EVENTS:

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.

Edwin I. Hatch Nuclear Plant Unit 1 05000-321