IR 05000348/2023004
| ML24029A267 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/09/2024 |
| From: | Mark Franke Division Reactor Projects II |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| EA 24-004 | |
| Download: ML24029A267 (17) | |
Text
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT-INTEGRATED INSPECTION REPORT 05000348/2023004 AND 05000364/2023004 AND EXERCISE OF ENFORCEMENT DISCRETION
Dear Jamie Coleman:
On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Joseph M. Farley Nuclear Plant. On February 6, 2024, the NRC inspectors discussed the results of this inspection with Mr. Edwin Dean III, Site Vice President and other members of your staff. The results of this inspection are documented in the enclosed report.
One Severity Level IV violation without an associated finding is documented in this report. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
No NRC-identified or self-revealing findings were identified during this inspection.
If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Joseph M. Farley Nuclear Plant.
For the Severity Level IV violation mentioned above, this inspection report documents an exercise of enforcement discretion in accordance with Section 3.10, Reactor Violations with No Performance Deficiencies, of the NRC Enforcement Policy. The issue aligned with a Severity Level III violation example in Section 6.1.c.2 of the Enforcement Policy and was considered for escalated enforcement action. The NRC Enforcement Policy can be found at the NRCs website at https://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html.
The inspectors concluded that there was no performance deficiency associated with the 2A pressurizer safety valve lifting outside of the allowed technical specification pressure tolerance band because the component failure was not avoidable by reasonable licensee internal procedures or management controls.
February 9, 2024 Based on the facts detailed in the enclosed report, and consultation with the Office of Enforcement and the Regional Administrator, I have been authorized to exercise enforcement discretion in accordance with Section 3.10 of the Enforcement Policy to categorize this violation as a Severity Level IV violation. The NRC concluded that the violation resulted in no, or relatively minimal potential safety impact because (1) the valve was susceptible to gross leakage for only a relatively short period of time before the unit was shut down to replace the valve, (2) other pressurizer pressure reducing components were available, and (3) an informational risk analysis was performed which indicated the violation was of very low risk significance.
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Mark E. Franke, Director Division of Reactor Projects Docket Nos. 05000348 and 05000364 License Nos. NPF2 and NPF8
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000348 and 05000364 License Numbers:
NPF-2 and NPF-8 Report Numbers:
05000348/2023004 and 05000364/2023004 Enterprise Identifier:
I2023-004-0015 Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Joseph M. Farley Nuclear Plant Location:
Columbia, AL Inspection Dates:
October 01, 2023 to December 31, 2023 Inspectors:
J. Bell, Senior Health Physicist B. Bowker, Reactor Inspector B. Caballero, Senior Operations Engineer M. Magyar, Reactor Inspector P. Meier, Senior Resident Inspector J. Rivera, Health Physicist S. Temple, Senior Project Engineer Approved By:
Mark E. Franke, Director Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Joseph M. Farley Nuclear Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Unit 2 A pressurizer code safety valve (PSV) Lift Pressure Outside of Technical Specification Limits Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000364/2023004-01 Open/Closed EA24-004 Not Applicable 71153 A self-revealed Severity Level (SL) IV non-cited violation (NCV) of Technical Specification (TS) Limiting condition for operations (LCO) 3.4.10, Pressurizer Safety Valves, was identified when the 2A PSV lifted at 2599 psig during asfound testing which is greater than the 2510 psig TS limit. Based on the Licensee Event Report 2023-001-00 submitted by the licensee regarding the issue and inspector evaluation, it was determined the 2A PSV setting was outside the TS limits at some period between July 8, 2022, to June 14, 2023, while the unit was in modes 1, 2, 3, and 4 with all reactor coolant system cold leg temperatures greater than 275°F (degrees Fahrenheit).
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000348/2023-003-00 LER 2023-003-00 for Joseph M. Farley Nuclear Plant, Unit 1, Emergency Diesel Generator (EDG) lube oil pump outlet coupling leak 71153 Closed LER 05000364/2023-001-00 LER 2023-001-00 for Joseph M. Farley Nuclear Plant, Unit 2, Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits 71153 Closed
PLANT STATUS
Unit 1 began the report period at approximately 100 percent rated thermal power (RTP) and remained at or near 100 percent RTP through the end of the report period.
Unit 2 began the report period at approximately 93 percent RTP, coasting down in preparation for a planned refueling outage on October 8, 2023. On October 8, 2023, unit 2 entered mode 3 and began the outage (2R29). Following the refueling outage, on November 13, 2023, unit 2 reactor was taken critical and entered mode 1 on November 14, 2023. Later that same day on November 14, 2023, unit 2 was manually tripped due to rising steam generator levels (Event Report 56852). Unit 2 was restarted on November 15, 2023, and entered mode 1. Over a period of approximately five days, RTP was increased with various hold points for testing until November 20, 2022, when unit 2 achieved approximately 100 percent RTP. Unit 2 remained at approximately 100 percent RTP through the end of the report period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed onsite portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1 'B' EDG while the '1-2A' EDG was unavailable due to maintenance on October 23, 2023 (FNP0SOP38.0-1B)
- (2) Unit 1 'A' train component cooling water (CCW) while the 'B' CCW pump was inoperable due to unscheduled maintenance on December 16, 2023 (work order D175002)
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 2 containment on November 7, 2023 (FNP2FPP3.0)
- (2) Unit 2 'A' train electrical penetration room (FA 2-035; FZ 2333 / 2347) on December 18, 2023 (FNP2FPP1.0)
- (3) Unit 2 'B' train electrical penetration room (FA 2-034; FZ2334) on December 18, 2023 (FNP2FPP1.0)
- (4) Unit 2 auxiliary building 155 foot elevation of the radiation control area (FA 2-004; FZ 2405) on December 18, 2023 (FNP2FPP1.0)
===71111.08P - Inservice Inspection Activities (PWR) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from October 16,2023 to October 19, 2023.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic Examination
- Weld APR1-2100-4, Pressurizer (PRZ) bottom head to lower shell, Class 1
- Post-weld RT of FW4/FW5, Turbine Driven Auxiliary Feedwater (TDAFW)pump steam supply valve (Q2N120001A) replacement, Class 2 Magnetic Particle Examination
- APR1-4504-2RC-R142, pipe support, Class 1
- Bare metal visual of the Reactor Vessel Closure Head, N729-6 Welding Activities
- Gas Tungsten Arc Welding, o
FW4 and FW5, TDAFW pump steam supply valve (Q2N120001A)replacement, Class 2 PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection
Activities (IP Section 03.02) (1 Sample)
The inspectors verified that the license conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:
(1)
- Bare metal visual of the Reactor Vessel Closure Head, N729-6 PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:
(1)
- Boric Acid Walkdown - October 16, 2023
- Corrosion Assessment Number: 2E13-2023-001
- Corrosion Assessment Number: 2E21-2023-001
- Corrosion Assessment Number: G31-2023-001
- CR11015850
- CR11015852
- CR11015853
71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance
Requalification Examination Results (IP Section 03.03) (1 Sample)
The licensee completed the annual requalification operating test required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." The inspector performed an inoffice review of the overall pass/fail results of the individual operating test, and the crew simulator operating examinations in accordance with IP
===71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.
- (1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating test, which the facility licensee completed administering on August 4, 2023.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during the following activities:
- Unit 2 operators cross-connecting to unit 1 auxiliary steam and response to associated alarms on October 6, 2023 (FNP2SOP55.1)
- Unit 2 down power for refueling outage (2R29) on October 8, 2023
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01)===
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components remain capable of performing their intended function:
- (1) Unit 1 and unit 2 condensate pumps (CRs 10789478, 10932439, 11031178)
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 2 risk associated with the 'A' motor driven auxiliary feedwater pump and valve testing on October 5, 2023 (NMP-DP001)
- (2) Use of risk-informed completion times and planned yellow risk for units 1 and 2 associated with the '1-2L' load center planned maintenance outage on November 28, 2023 (NMP-DP001)
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Leading edge flowmeter issues identified on September 18, 2023 (CR11007690)
- (2) Unit 2 containment coolers extent of condition associated with the 2C containment cooler ground identified on September 28, 2023 (CR 11011003, 11017241)
- (3) Unit 1 'B' residual heat removal did not start during a surveillance test on October 2, 2023 (CR11011749)
- (4) Unit 2 'C' accumulator valve leak-by identified on October 21, 2023 (CR 11017400)
- (5) Unit 1 'B' component cooling water pump with elevated metal content in the outboard motor bearing oil on December 14, 2023 (CR 11029055, SNC1651307)
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Unit 2 nuclear steam service system control and steam generator level control system digital upgrade performed during the unit 2 2023 fall refueling outage (2R29) in October 2023 (DCP SNC1083455)
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated unit 2 refueling outage 29 (2R29) activities from October 8, 2023, to November 16, 2023.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (3 Samples)
- (1) Unit 2 'B' inverter ten year component replacement performed during the unit 2 fall 2023 refueling outage (2R29) (WO SNC1124622)
- (2) Unit 2 'C' containment cooler following fan motor replacement on November 5, 2023 (CR 11011003, SNC1593127, SNC1582528)
- (3) Unit 2 steam generator level control system digital upgrade power ascension testing performed during the week of November 13, 2023 (FNP2SPETP005)
Surveillance Testing (IP Section 03.01) (1 Sample)
- (1) Unit 2 'B' train safety injection with loss of offsite power surveillance performed on October 17, 2023 (FNP2STP40.0B; CR 11015855)
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Unit 1 'C' charging pump quarterly inservice test on November 20, 2023 (FNP1STP4.3)
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) Containment Purge (penetration 12) limited leak rate testing on October 30, 2023 (FNP2STP627.2)
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
- (1) Licensee surveys of contaminated trash being removed from the unit 2 containment contaminated area (CA) during an outage
- (2) Workers doffing protective clothing at the unit 2 containment CA boundary during an outage
- (3) Workers exiting the plant radiologically controlled area during an outage
Radiological Hazards Control and Work Coverage (IP Section 03.04) (5 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Fuel movement during unit 2 refueling outage
- (2) Reactor head stud cleaning
- (3) Scaffold building activities in unit 2 containment and the auxiliary building
- (4) Unit 2 refueling cavity drain down activities
- (5) Unit 2 reactor coolant filter change out High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (4 Samples)
The inspectors evaluated licensee controls of the following high radiation areas and very high radiation areas (VHRAs):
- (1) Unit 2 volume control tank room
- (2) Unit 2 letdown regenerative heat exchanger cage
- (3) Unit 2 cavity drain line cage
- (4) Unit 1 reactor coolant filter cubicle Radiation Worker Performance and Radiation Protection Technician Proficiency (IP
Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &
Transportation
Radioactive Material Storage (IP Section 03.01)
The inspectors evaluated the licensees performance in controlling, labeling and securing the following radioactive materials:
- (1) Radioactive material in the Old Steam Generator Storage Facility.
- (2) Radioactive material in the Low Level Radwaste Building.
- (3) Sealed radioactive sources installed in instrument calibrators.
Radioactive Waste System Walkdown (IP Section 03.02) (1 Sample)
The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:
- (1) Solidification and Dewatering Facility (SSDF)
Waste Characterization and Classification (IP Section 03.03) (3 Samples)
The inspectors evaluated the following characterization and classification of radioactive waste:
(1)10 CFR Part 61 analysis for dry active waste, DAW082522V (2)10 CFR Part 61 analysis for liquid waste resin, PO689557-12V (3)10 CFR Part 61 analysis for primary resin, 120L21006V
Shipping Records (IP Section 03.05) (3 Samples)
The inspectors evaluated the following non-excepted radioactive material shipments through a record review:
- (1) Shipment no. 22-16, dry active waste, LSA-II
- (2) Shipment no. 23-07, liquid waste resin, LSA-II
- (3) Shipment no. 21-15, mixed bead resin, Type B
OTHER ACTIVITIES-BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS06: Emergency AC Power Systems (IP Section 02.05)===
- (1) Unit 1 (October 1, 2022 - September 30,2023)
- (2) Unit 2 (October 1, 2022 - September 30,2023)
MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)
- (1) Unit 1 (October 1, 2022 - September 30,2023)
- (2) Unit 2 (October 1, 2022 - September 30,2023)
MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)
- (1) Unit 1 (October 1, 2022 - September 30,2023)
- (2) Unit 2 (October 1, 2022 - September 30,2023)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) October 21, 2022 - October 19, 2023 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) October 1, 2022 - September 30, 2023
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Unit 2 digital rod position indication system advanced display system (DADS)indication failure on September 9, 2023 (CRs 11005823 and 11009204)
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000364/2023-001-00, Unit 2 pressurizer code safety valve lift pressure outside of Technical Specifications limits (ADAMS Accession No. ML23271A132) The circumstances surrounding this LER are documented in this inspection report, in the inspection results Section (NCV 05000364/2023004-01). This LER is Closed.
- (2) LER 05000348/2003-003-00, Unit 1-Emergency Diesel Generator (1B) Lube Oil Pump Outlet Coupling, (ADAMS Accession No. ML23341A206). The inspection conclusions associated with LER are documented in NRC inspection report 05000348/2023091 (ML23263B166), issued on October 19, 2023. The circumstances surrounding the LER were further reviewed by an NRC supplemental inspection completed on December 14, 2023, and documented in inspection report 05000348/2023040 Section 95001 (ML24022A059), issued January 24, 2024. This LER is Closed.
INSPECTION RESULTS
Unit 2 A pressurizer code safety valve (PSV) Lift Pressure Outside of Technical Specification Limits Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000364/2023004-01 Open/Closed EA24-004 Not Applicable 71153 A self-revealed Severity Level (SL) IV non-cited violation (NCV) of Technical Specification (TS) Limiting condition for operations (LCO) 3.4.10, Pressurizer Safety Valves, was identified when the 2A PSV lifted at 2599 psig during asfound testing which is greater than the 2510 psig TS limit. Based on the LER submitted by the licensee regarding the issue and inspector evaluation, it was determined the 2A PSV setting was outside the TS limits at some period between July 8, 2022, to June 14, 2023, while the unit was in modes 1, 2, 3, and 4 with all reactor coolant system cold leg temperatures greater than 275°F (degrees Fahrenheit).
Description:
The Farley unit 2 A pressurizer code safety valve (2A PSV) was removed on June 14, 2023, due to leak-by and sent to a vendor for testing in accordance with TS 3.4.10, Pressurizer Safety Valves. On August 1, 2023, the licensee was informed that the 2A PSV set pressure results were high outside the TS asfound acceptance criteria of 2423 to 2510 psig. The 2A PSV lifted at approximately 2599 psig. The licensee determined that gross leakage past the valve seat due to internal steam cutting of the 2A PSV disc insert and nozzle caused the failure to lift within the acceptance criteria. On October 28, 2023, Farley submitted Licensee Event Report (LER) 2023-001-00 (ML23271A132) in accordance with 10 CFR 50.73.
On July 8, 2022, 2A PSV leakage was first identified based on high tailpiece temperature alarms and pressurizer relief tank indications. The licensee evaluated the risk associated with the leak in accordance with procedure NMP-OS003, revision 10.1, Operational Decision-Making Issue Evaluation Process (ODMI number FNP2-22-003; CR 1107795). They determined corrective actions to address the leak could be delayed until the next refueling outage in the fall of 2023 (2R29) or prepare for a forced outage should the leak exceed 0.5 gpm (24-hour average). A leak rate of 0.5 gpm was the point considered to be a concern for steam cutting. Leak rates varied from 0 to 0.1 gpm from the identified date of July 8, 2022, until May 29, 2023. The observed 2A PSV leak rates did not challenge TS 3.4.1, RCS Operational Leakage, limits for identified leakage of less than or equal to 10 gpm. There were no other specific requirements regarding acceptable PSV leakage. The main concern of operating with known leak-by on a PSV is the failure to perform its overpressure protection function due to the leak itself. Based on operating experience, the licensee determined there was no indication that the observed PSV leak-by would challenge the upper pressure limit. In fact, operating experience indicates PSVs lift pressures set with a loop seal will relieve at a lower pressure upon loss of the loop seal due to leakage (reference NRC Information Notice
[IN] 89-90). Additional information provided via IN 89-90, Supplement 1, indicates anything that causes an increase in the temperature of the valve, such as leak-by, may reduce the lift setpoint as well.
On May 29, 2023, the licensee observed a substantial 2A PSV leak increase to 0.3 gpm (24-hour average). This triggered a reevaluation of their decision making as previously documented from the original ODMI worksheet (ODMI number FNP2-22-003). The decision was made to perform a forced outage scheduled for June 13, 2023, to replace the 2A PSV.
However, on or around June 11, 2023, the 2A PSV leak rate was approaching 0.5 gpm which was an ODMI trigger to begin a controlled down power and enter a forced outage to replace the PSV. In consultation with Westinghouse, the licensee determined that the previous decision to shut down on June 13, 2023, even with a sustained leak rate at approximately 0.5 gpm for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, was sufficient time before severe steam cutting of the seating surfaces occurred. The leak rate was observed to reach approximately 0.5 gpm (24-hour average) on or around the evening of June 11, 2023. On June 13, 2023, at 20:15, the unit 2 reactor trip breakers were opened to commence the forced outage for the 2A PSV replacement. Unit 2 entered mode 4 early on June 14, 2023, at 03:24.
Corrective Actions:
1. Replaced the 2A PSV with a previously tested valve (work order SNC1348879).
2. Procedurally limit pressurizer safety valve leakage at 0.3 gpm to minimize chance
of steam cutting. Changes were made to the alarm response procedures (FNP1ARP1.8, FNP2ARP1.8) associated with a leaking PSV and the precaution and limitations in the reactor coolant system normal operating procedure (FNP1SOP1.1, FNP2SOP1.1).
3. Revised the startup procedures (FNP1UOP1.1, FNP2UOP1.1) to incorporate
industry practice and Electric Power Research Institute (EPRI) recommendations for reactor coolant system pressure soaking to thermally equalize the PSVs before reaching full pressure.
Corrective Action References: Technical Evaluation (TE) 1136890, TE 1133293, TE 1124475, TE 1063582
Performance Assessment:
The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Enforcement:
The NRC exercised enforcement discretion in enforcement action (EA)-24-004, in accordance with Section 3.10, Reactor Violations with No Performance Deficiencies, of the NRC Enforcement Policy for an issue that aligned with a Severity Level III violation example in Section 6.1.c.2 of the Enforcement Policy and was considered for escalated enforcement action. The inspectors concluded that there was no performance deficiency associated with the 2A pressurizer safety valve lifting above the allowed technical specification pressure tolerance band because the component failure was not avoidable by reasonable licensee internal procedures or management controls.
Severity: The inspectors categorized this violation as a Severity Level IV violation. The NRC concluded that the violation resulted in no, or relatively minimal potential safety impact because
- (1) the valve was susceptible to the gross leakage for only a relatively short period of time before the unit was shut down to replace the valve,
- (2) other pressurizer pressure reducing components were available, and
- (3) an informational risk analysis was performed which indicated the violation was of very low risk significance.
Violation: Farley unit 2 TS LCO 3.0.1 requires, in part, that LCOs shall be met during the modes of applicability.
TS LCO 3.4.10, Pressurizer Safety Valves, requires, in part, three operable PSVs with lift settings greater than or equal to 2423 psig and less than or equal to 2510 psig while unit 2 is modes 1, 2, 3, and 4 with all reactor coolant system cold leg temperatures greater than 275°F.
With one pressurizer safety valve inoperable, Action Statement, condition A. Required Action A.1, required restoration of the valve to operable status within 15 minutes. If the required action and associated completion time is not met, Action Statement, Condition B, required that the unit be in mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and mode 4 and less than 275°F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Contrary to the above, during as-found testing the 2A PSV lifted at 2599 psig which is greater than the 2510 psig TS limit. Based on the LER submitted by the licensee regarding the issue and inspector evaluation, it was determined the 2A PSV setting was outside the TS limits at some period between July 8, 2022, to June 14, 2023, while the Unit was in modes 1, 2, 3, and 4 with all reactor coolant system cold leg temperatures greater than 275°F. This period coincides to when the 2A PSV leak was first identified until the unit was placed in mode 4 and less than 275°F.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On February 6, 2024, the inspectors presented the integrated inspection results to Mr.
Edwin Dean III, Site Vice President and other members of the licensee staff.
- On October 19, 2023, the inspectors presented the Radiation Protection inspection results to Delson Erb and other members of the licensee staff.
- On October 19, 2023, the inspectors presented the In-Service Inspection (ISI) Exit Meeting inspection results to Delson Erb and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
ALARA Plans
MAP 23-2496
U2 RCS filter manual change out and storage
10/19/23
11016513
NRC observation of temporary scaffolds without survey tags
10/18/23
11016526
NRC observation of containers not marked or labeled
10/18/23
11016531
NRC observation of Radiation Area posting
10/18/23
11016532
NRC observation of Contaminated Area boundary
10/18/23
Corrective Action
Documents
Resulting from
Inspection
11016982
NRC observation during U2 RCS filter change out
10/19/23
Miscellaneous
Air sample
gamma
spectroscopy
results #282272
U2 RCS filter room
10/19/23
NMP-HP204
ALARA Planning and Job Review
Version 10.4
Procedures
NMP-HP300
Radiation and Contamination Surveys
Version 5.8
Work Orders
WO SNC1111412
WO SNC666037
WO SNC997661
WO SNC1114961
WO SNC963853
WO SNC1078293