05000387/LER-2017-010-02, Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switches
| ML21182A041 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  (NPF-022) |
| Issue date: | 07/01/2021 |
| From: | Cimorelli K Susquehanna, Talen Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-7951 LER 2017-010-02 | |
| Download: ML21182A041 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ix)(A) |
| 3872017010R02 - NRC Website | |
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Kevin Cimorelli Site Vice President Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3795 Fax 570.542.1504 Kevin.Cimorelli@TalenEnergy.com SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-388/2017-010-02 UNIT 2 LICENSE NO. NPF-22 PLA-7951 TALEN~
ENERGY 10 CFR 50.73 Docket No. 50-388 Attached is the second supplement to Licensee Event Report (LER) 50-388/2017-010 that reported an event involving drifting of Reactor Pressure Steam Dome - Low pressure switches (Microswitch 2). The condition was reported as a condition prohibited by Technical Specifications in accordance with 10 CFR 50.73(a)(2)(i)(B), a condition that could have prevented fulfillment of a safety function in accordance with 10 CFR 50.73(a)(2)(v)(D), a common cause inoperability of independent trains or channels in accordance with 10 CFR 50.73(a)(2)(vii), and a single cause that could have prevented fulfillment of safety functions of trains or channels in different systems in accordance with 10 CFR 50.73(a)(2)(ix)(A). The reason for this supplement is a revised causal analysis that resulted in a modification to the cause and associated corrective actions.
There were no actual consequences to the health and safety of the public as a result of this event.
Attachment: LER 50-388/2017-010-02 Copy:
NRC Region I Mr. C. Highley, NRC Senior Resident Inspector Ms. A Klett, NRC Project Manager Mr. M. Shields, PA DEP/BRP July 1, 2021
Abstract
On December 5, 2017, the Unit 2 C and Unit 2 D Reactor Steam Dome Pressure - Low permissive pressure switches, Microswitch 2 (SW2), were found outside of the Technical Specification (TS) 3.3.5.1 allowable value. Both drifted outside of the upper allowable value which is intended to ensure that the reactor dome pressure has fallen to a value below the Core Spray and Residual Heat Removal (RHR)/Low Pressure Coolant Injection (LPCI) maximum design pressures to preclude over-pressurization of the low pressure systems prior to low pressure injection initiation.
Based on the information available, the condition existed for longer than allowed by TS 3.3.5.1 and TS 3.5.1. As such, this is a condition prohibited by TS and is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B). This is also considered a common cause inoperability of independent trains or channels (10 CFR 50.73(a)(2)(vii)), a condition that could have prevented fulfillment of a safety function (10 CFR 50.73(a)(2)(v)(D)), and a single cause that could have prevented fulfillment of safety functions of trains or channels in different systems (10 CFR 50.73(a)(2)(ix)(A)).
The cause was determined to be instrument setpoint drift due to temperature/humidity changes and mechanical hysteresis as a result of the original design inputs not accounting for temperature and humidity effects on the switch. Interim compensatory actions include performing calibration using the quarterly calibration procedures every 45 days on Unit 1 and every 30 days on Unit 2. Final corrective action will revise the allowable value for TS 3.3.5.1, Functions 1.c, 1.d, 2.c, and 2.d.
There were no actual consequences to the health and safety of the public as a result of this event.
CONDITIONS PRIOR TO EVENT
Unit 1 - Mode 1, approximately 100 percent Rated Thermal Power Unit 2 - Mode 1, approximately 100 percent Rated Thermal Power There were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.
EVENT DESCRIPTION
Susquehanna Steam Electric Station (SSES) had been utilizing International Telephone and Telegraph (ITT)-Barton 288A pressure switches in the Reactor Steam Dome Pressure - Low channels that provide the injection permissive for the Core Spray system [EIIS System Identifier: BM] (Technical Specification (TS) 3.3.5.1, Function 1d) and the Residual Heat Removal (RHR)/Low Pressure Coolant Injection system (LPCI)
[EIIS System Identifier: BO] (TS 3.3.5.1, Function 2d). Each pressure switch has two microswitches, low (SW1) and high (SW2). All eight obsolete ITT-Barton 288A pressure switches were replaced with General Electric (GE) recommended Cameron-Barton 288A pressure switches between September 6, 2017 and November 15, 2017. The Cameron Barton 288A switches were bench tested prior to installation and calibration checked at the time of installation. Subsequent calibration checks were performed at intervals more frequent than the quarterly TS required calibrations. During these subsequent calibration checks, SW2 of the Unit 2 C (PIS-B21-2N021C) and Unit 2 D (PIS-B21-2N021D) pressure switches [EIIS System/Component Identifier: JE/PS], were found outside of the TS allowable value. A timeline of events is as follows:
September 20, 2017 - the obsolete ITT-Barton 288A pressure switch for PIS-B21-2N021C was replaced with a GE recommended Cameron-Barton 288A pressure switch.
October 9, 2017 - the obsolete ITT-Barton 288A pressure switch for PIS-B21-2N021D was replaced with a GE recommended Cameron-Barton 288A pressure switch.
October 20, 2017 - PIS-B21-2N021C and PIS-B21-2N021D were calibration checked and were found within the TS allowable value.
December 5, 2017 - PIS-B21-2N021C and PIS-B21-2N021D were found outside of the TS allowable value during the calibration check. Both drifted outside of the upper allowable value which is intended to ensure that the reactor dome pressure has fallen to a value below the Core Spray and RHR/LPCI maximum design pressures to preclude over-pressurization of the low pressure systems prior to low pressure injection initiation. PIS-B21-2N021C exceeded the TS allowable value by 1.3 psi. PIS-B21-2N021D exceeded the TS allowable value by 2.8 psi. Both switches were adjusted to within the TS allowable value.
After submitting LER 50-388/2017-010-00, two additional instances of drift outside of the TS allowable value were identified. Specifically, the Unit 1 B (PIS-B21-1N021B) switch was found outside the TS allowable value on June 5, 2018 and the Unit 2 C (PIS-B21-2N021C) switch was again found outside of the TS allowable value on June 6, 2018. These conditions are addressed in LER 50-387(388)/2018-005.
Based on the information available, the condition existed for longer than allowed by TS 3.3.5.1 and TS 3.5.1. As such, this is a condition prohibited by TS and is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B). In addition, since two Unit 2 channels were outside of the TS allowable values and impacted both Core Spray and LPCI functions, this is also considered a common cause inoperability of independent trains or channels (10 CFR 50.73(a)(2)(vii)), a condition that could have prevented fulfillment of a safety function (10 CFR 50.73(a)(2)(v)(D)), and a single cause that could have prevented fulfillment of safety functions of trains or channels in different systems (10 CFR 50.73(a)(2)(ix)(A)).
CAUSE OF EVENT
The direct cause was determined to be instrument setpoint drift due to temperature/humidity changes and mechanical hysteresis. The apparent cause involved the original design inputs not accounting for temperature and humidity effects on the switch.
ANALYSIS/SAFETY SIGNIFICANCE
An engineering evaluation concluded that all components in the RHR/LPCI and Core Spray would have been able to withstand a pressure of 457.8 psig, which was the maximum recorded pressure for the instrument drift. Therefore, Core Spray and RHR/LPCI would have been able to perform their safety function. This event will not be counted as a safety system functional failure for the NRC performance indicator based on the engineering analysis that shows there was no loss of ability to fulfill the safety function.
CORRECTIVE ACTIONS
Interim compensatory actions include performing calibration using the quarterly calibration procedures every 45 days on Unit 1 and every 30 days on Unit 2.
Final corrective action will revise the allowable value for TS 3.3.5.1, Functions 1.c, 1.d, 2.c, and 2.d.
COMPONENT FAILURE INFORMATION
The switches that drifted are Cameron-Barton 288A pressure indicating switches manufactured by Cameron.
PREVIOUS SIMILAR EVENTS
LER 50-388(387)/2015-001-01, Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Steam Dome-Low Switches, dated February 10, 2016.
LER 50-387(388)/2018-005-01, Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switches, dated December 16, 2019.