05000391/LER-2020-002, Re Two Pressurizer Safety Valves Outside of Technical Specification Limits Due to Set Point Drift
| ML20352A005 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 12/17/2020 |
| From: | Anthony Williams Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| WBL-20-063 LER 2020-002-00 | |
| Download: ML20352A005 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3912020002R00 - NRC Website | |
text
uID Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 WBL-20-063 December 17, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391 10 CFR 50.73
Subject:
Licensee Event Report 391/2020-002-00, Two Pressurizer Safety Valves Outside of Technical Specification Limits Due to Set Point Drift This submittal provides Licensee Event Report (LER) 391/2020-002-00. This LER provides details concerning two Pressurizer Safety Valves outside of Technical Specification limits due to set point drift. This condition is being reported as an event or condition prohibited by Technical Specifications in accordance with 10 CFR 50.73(a)(2)(i)(B).
There are no regulatory commitments contained in this letter. Please direct any questions concerning this matter to Tony Brown, WBN Licensing Manager, at (423) 365-7720.
<----'1inthony L. Williams IV Site Vice President Watts Bar Nuclear Plant Enclosure: Licensee Event Report 391/2020-002-00, Two Pressurizer Safety Valves Outside of Technical Specification Limits Due to Set Point Drift cc: See Page 2
U.S. Nuclear Regulatory Commission WBL-20-063 Page 2 December 17, 2020 cc (Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant
Abstract
On November 2, 2020, while at 0 percent power level and defueled, it was discovered that two Unit 2 pressurizer safety valves (PSVs), which had been removed during the October/November 2020 refueling outage (U2R3) and shipped off-site for testing, failed their as-found lift pressure test. The PSVs lifted below the Technical Specification (TS) 3.4.10 allowable lift setting value. Set point drift was the cause of the PSVs failures. This set point drift did not result in the PSVs inability to perform their safety function.
The PSVs were replaced during the October/November 2020 refueling outage.
This condition is being reported as a condition prohibited by Technical Specifications in accordance with 10 CFR 50.73(a)(2)(i)(B).
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I.
Plant Operating Conditions Before the Event
Watts Bar Nuclear Plant (WBN) Unit 2 was defueled at 0 percent rated thermal power (RTP).
II.
Description of Event
A. Event Summary
During the Unit 2 October/November 2020 refueling outage (U2R3 ), while at 0 percent power level and with the Reactor Coolant System (RCS) [EIIS:AB] defueled, two pressurizer safety valves (PSV) [EIIS:RV] were removed as part of the routine In-Service Testing (IST) program and sent to an off-site testing facility for testing. On November 2, 2020 the site was notified that the as-found lift pressures of PSVs 2-RFV-68-563 and 2-RFV-68-565 were outside of the Technical Specification (TS) 3.4.10 allowable lift pressure settings of >/= 2410 psig and </= 2560 psig. The tested valves were not within the acceptance band of +/- 3 percent (2411-2559 psig).
This event is being reported to the Nuclear Regulatory Commission (NRC) under 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications.
B. Status of structures, components, or systems that were inoperable at the start of the event and that contributed to the event No inoperable structures, systems, or components contributed to this condition.
C. Dates and approximate times of occurrences
Date Time (EST)
Event 11/2/2020 1649 Condition Report (CR) 1649435 generated to document the as-found testing failure of 2-RFV-68-563 during U2R3. The set pressure is 2485 psig. The as-found lift pressure was 2380 psig.
11/2/2020 1704 CR 1649440 generated to document the as found testing failure of 2-RFV-68-565 U2R3. The set pressure is 2485 psig. The as-found lift pressure was 2398 psig.
D. Manufacturer and model number of each component that failed during the event
PSV 2-RFV-68-563 and PSV 2-RFV-68-565 are made by Crosby Valve & Gage Co with a part number of HB-BP-86. These are 6 inch valves.
E. Other systems or secondary functions affected
No other systems or secondary functions were affected.
Page 3 of 5 (08-2020)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
- 3. LER NUMBER Watts Bar Nuclear Plant, Unit 2 05000-391 YEAR SEQUENTIAL NUMBER REV NO.
2020
- - 002
- - 00
F. Method of discovery of each component or system failure or procedural error
During routine In-Service Testing, the vendor notified the site that the as-found lift pressures of two PSVs were below TS allowable lift pressure settings.
G. Failure mode, mechanism, and effect of each failed component Set point drift of the PSVs was determined to be the most likely cause of these failures.
H. Operator actions
No operator actions were required.
I.
Automatically and manually initiated safety system responses
None required.
III.
Cause of the Event
A. Cause of each component or system failure or personnel error
Set point drift of the PSVs was determined to be the most likely cause of these failures.
There was no human performance cause related to this condition.
IV.
Analysis of the Event
There was no event. During the required IST and TS testing, these valves tested outside the limits specified in the TS. The valves were replaced with valves tested and set within +/- 1 percent of 2485 psig.
V.
Assessment of Safety Consequences
The PSVs provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), 2735 psig (2750 psia), which is 110 percent of the design pressure.
Overpressure protection is required in MODES 1, 2, 3, 4, and 5; however, in MODE 4 with any RCS cold leg temperature less than or equal to the Cold Overpressure Mitigation System (COMS) arming temperature specified in the Pressure and Temperature Limits Report, MODE 5, and MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, Cold Overpressure Mitigation System (COMS).
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Page 4 of 5 (08-2020)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
- 3. LER NUMBER Watts Bar Nuclear Plant, Unit 2 05000-391 YEAR SEQUENTIAL NUMBER REV NO.
2020
- - 002
- - 00 The combined capacity of two of the three safety valves is equal to or greater than the maximum surge rate resulting from complete loss of load, resulting from a turbine trip with a concurrent loss of main feedwater, without reactor trip or any other control. This objective is met without reactor trip or operator action by the opening of the steam safety valves when SG pressure reaches the steam-side safety setting. The two safety valves found with slightly low lift pressures does not adversely impact the specified safety function related to overpressure protection.
In addition, the Updated Final Safety Analysis Report (UFSAR) Section 15.2.12 describes an accidental depressurization of the reactor coolant system associated with a pressurizer safety valve opening, which may be a concern for a low setpoint impact. This scenario is still the same with two valves with a low set pressure because review of operating data showed that the valves remained closed during the cycle, when at nominal RCS pressure which is the starting point for the transient.
Additionally, the transient described in the UFSAR shows that no credit is given for reclosing of a safety valve once a reseating pressure is reached due to the continued reduction in pressure.
Based on as-found testing and inspection, there were no issues that would have resulted in the valves being incapable of reseating. Additionally, it is not industry practice to consider more than one valve remaining stuck open beyond reseat pressure. Therefore, an inadvertent depressurization would have still only occurred with inadvertent opening of a pressurizer safety valve with a resulting reactor trip, and either valve would have been capable of reseating; therefore, the condition remains bounded by the UFSAR analysis. The deviations between the as found and limiting values for pressurizer safety actuation were on the order of 1 percent; these differences are within the uncertainties of the WBN Probabilistic Risk Assessment (PRA) and therefore it can be said that no measurable risk resulted. A bounding calculation was performed increasing the likelihood that two of the pressurizer safety valves could lift and fail to reseat due to the lowered actuation pressure. The resulting increase in risk as measured by Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) was very small. Accordingly, there is minimal safety impact from this condition.
A. Availability of systems or components that could have performed the same function as the components and systems that failed during the event Not applicable.
B. For events that occurred when the reactor was shut down, availability of systems or components needed to shutdown the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident Not applicable.
C. For failure that rendered a train of a safety system inoperable, an estimate of the elapsed time from the discovery of the failure until the train was returned to service.
Not applicable.
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Page 5 of 5 (08-2020)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
- 3. LER NUMBER Watts Bar Nuclear Plant, Unit 2 05000-391 YEAR SEQUENTIAL NUMBER REV NO.
2020
- - 002
- - 00 VI.
Corrective Actions
This event was entered into the Tennessee Valley Authoritys (TVA) Corrective Action Program and is being tracked under Condition Reports (CRs) 1649435 and 1649440.
A. Immediate Corrective Actions
The PSVs were replaced during U2R3 with refurbished valves that were certified
+/- 1 percent of the set point. In addition, as required by the IST program, a sample expansion was performed and the third installed PSV (2-RFV-68-564) was sent for testing and replaced with a valve tested in accordance with TS.
B. Corrective Actions to Prevent Recurrence or to reduce probability of similar events occurring in the future Actions in CR 1649440 include documenting the vendors results of the disassembly and inspections of these PSVs and causes of failures. The corrective action plan will be revised, if necessary, based on the vendors results.
VII.
Previous Similar Events at the Same Site
No previous similar events were reported to NRC.
VIII.
Additional Information
There is no additional information.
IX.
Commitments
There are no new commitments.
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