IR 05000348/2024002
| ML24212A115 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/12/2024 |
| From: | Mark Franke Division Reactor Projects II |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| EA-24-072 IR 2024002 | |
| Download: ML24212A115 (22) | |
Text
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000348/2024002 AND 05000364/2024002 AND EXERCISE OF ENFORCEMENT DISCRETION
Dear Jamie Coleman:
On June 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Joseph M. Farley Nuclear Plant. On July 23, 2024, the NRC inspectors discussed the results of this inspection with John Andrews, Plant Manager, and other members of your staff. The results of this inspection are documented in the enclosed report.
No NRC-identified or self-revealing findings were identified during this inspection.
One Severity Level (SL) IV violation associated with exercise of enforcement discretion in accordance with Section 3.10, Reactor Violations with No Performance Deficiencies, of the Enforcement Policy is documented in this inspection report. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
The violation involved the failure of the C main steam safety valve on the unit 2 B main steam line to lift within the pressure band allowed by the unit 2 technical specifications (TS). The inspectors concluded there was no performance deficiency associated with the violation. The violation was considered for escalated enforcement action because its circumstances aligned with an SL III violation example in Section 6.1.c.2 of the Enforcement Policy. The NRC Enforcement Policy can be found at the NRCs website at https://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html.
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 an SL IV violation. The NRC concluded that the violation resulted in no, or relatively minimal potential safety impact because (1) other pressure reducing components were available and (2)
an informational risk analysis performed by NRC staff indicated the violation was of very low risk significance.
August 12, 2024 If you contest the violation or the 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.
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. NPF-2 and NPF-8
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000348 and 05000364 License Numbers:
NPF-2 and NPF-8 Report Numbers:
05000348/2024002 and 05000364/2024002 Enterprise Identifier:
I-2024-002-0018 Licensee:
Southern Nuclear Operating Company. Inc.
Facility:
Joseph M. Farley Nuclear Plant Location:
Columbia, AL Inspection Dates:
April 1, 2024 to June 30, 2024 Inspectors:
A. Alen Arias, Senior Project Engineer J. Bell, Senior Health Physicist B. Bowker, Reactor Inspector J. Diaz-Velez, Senior Health Physicist B. Kellner, Senior Health Physicist P. Meier, Senior Resident Inspector C. Scott, Senior Project Engineer S. Temple, Senior Resident Inspector 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
Main steam safety valve lift pressure outside of TS limits due to setpoint drift Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000364/2024002-01 Open/Closed EA-24-072 Not Applicable 71153 A self-revealed Severity Level (SL) IV non-cited violation (NCV) of Technical Specifications (TS) Limiting Condition of Operation 3.7.1, Main Steam Safety Valves, was identified when a routine lift pressure test revealed that the lift setpoint for the C main steam safety valve, associated with the B main steam line, was higher than allowed by TS. Based on the Licensee Event Report 2023-002-00 submitted by the licensee regarding the issue and inspector evaluation, it was determined that MSSV Q2N11V0011C was inoperable for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> between April 30, 2019, to October 3, 2023, while the unit was in modes 1, 2, and 3.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000364/
2024-002-00 LER 2024-002-00 for Joseph M. Farley Nuclear Plant, Unit 2 for Manual Reactor Trip due to Loss of Power to the 2A 125 Volt DC Distribution Panel 71153 Closed LER 05000364/
2024-002-01 LER 2024-002-01 for Joseph M. Farley Nuclear Plant, Unit 2, Manual Reactor Trip due to Loss of Power to the 2A 125 Volt DC Distribution Panel 71153 Closed LER 05000364/
2023-002-00 LER 2023-002-00 for Joseph M. Farley Nuclear Plant, Unit 2, Main Steam Safety Valve Lift Pressure Outside of Technical Specifications Limits due to Setpoint Drift 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power. On April 7, 2024, the unit was shut down for scheduled refueling outage 1R32. On May 9, 2024, the unit was restarted and reached 100 percent rated thermal power on May 18, 2024. The unit remained at or near rated thermal power for the remainder of the inspection period.
Unit 2 operated at or near 100 percent rated thermal power for the entire inspection 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 on-site 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
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal hot temperatures for the following systems during the month of May 2024:
- service water system
- component cooling water system
- containment coolers
- onsite AC power sources
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather due to forecasted severe thunderstorms and potential tornados on April 10, 2024, with potential to impact unit 1 refueling operations during the 2024 spring outage, 1R32 (procedure NMP-OS-017)
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Emergency diesel generator 1-2A alignment to unit 2 during the unit 2 'B' emergency diesel generator planned maintenance outage on April 1, 2024 (procedures FNP-0-SOP-38.0A and FNP-0-SOP-38.0C)
- (2) Unit 1 'B' train low head safety injection system following 1R32 on May 22, 2024 (FNP-1-SOP-7.0)
- (3) Unit 2 'A' train component cooling water system during maintenance and testing of the 'C' component cooling water pump on June 10, 2024 (FNP-2-SOP-23.0, and drawing D205002)
- (4) Unit 2 'B' train component cooling water system during maintenance and testing of the 'B' component cooling water pump on June 17, 2024 (FNP-2-SOP-23.0; D205002)
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 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 (FAs):
- (1) Unit 1 containment on April 26, 2024 (procedure FNP-1-FPP-3.0)
- (3) Unit 2 'B' train switchgear room (FA 2-021) on May 23, 2024 (FNP-2-FPP-1.0)
- (4) Unit 2 'A' train switchgear room (FA 2-041) on June 11, 2024 (FNP-2-FPP-1.0)
- (5) Unit 2 cable spreading room (FA 2-040) on June 11, 2024 (FNP-2-FPP-1.0)
- (6) Service water intake structure 'A' train switchgear room (FA-076) on June 11, 2024 (FNP-0-FPP-3.0)
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an announced fire drill (F-2024-007) in the protected area on June 12, 2024.
===71111.08P - Inservice Inspection Activities (PWR) The inspectors verified that the reactor coolant system boundary, reactor pressure 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 for unit 1 during refueling outage
==1R32 from April 8, 2024, to April 11, 2014.
==
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
- ALA1-4202-3-RB, pipe to elbow, ASME Class 1
- ALA1-4202-4-RB, valve to pipe, ASME Class 1
- ALA1-4207-3-RB, pipe to elbow, ASME Class 1
- ALA1-4207-4-RB, valve to pipe, ASME Class 1
- ALA1-4304-3-RB, pipe to elbow, ASME Class 1
- ALA1-4304-5-RB, pipe to elbow, ASME Class 1
- ALA1-4304-6-RB, valve to pipe, ASME Class 1
- ALA1-4305-3-RB, pipe to elbow, ASME Class 1
- ALA1-4305-4-RB, valve to pipe, ASME Class 1 Dye Penetrant Examination
- Welds 7F, 8F, 9F, and 10F, reactor coolant system pressurizer vent valve Q1B13V076A replacement, ASME Class 1 (work order [WO] SNC1183978)
Visual Examination
- ALA1-1100C-BMI-BMV, bare metal visual (BMV) of reactor pressure vessel bottom head mounted instrumentation, ASME Class 1
- VT-2, post-weld system operating pressure test for WO SNC1183978, ASME Class 1
- VT-2, post-weld system operating pressure test for WO SNC1183979, ASME Class 2 Welding Activities
- Gas Tungsten Arc Welding o
WO SNC1183978, reactor coolant system pressurizer vent valve Q1B13V076A, welds 7F, 8F, 9F, and 10F, ASME Class 1 o
WO SNC1183979, reactor coolant system pressurizer vent valve Q1B13V076B, weld 11F, ASME Class 2 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 - April 9, 2024
- Boric acid evaluations o
Condition Report (CR) 11049504, Corrosion Assessment Number:
CR 10999346, Corrosion Assessment Number: 2E21-2023-001 o
CR 11026039, Corrosion Assessment Number: 2E21-2024-001 o
CR 10924606, Corrosion Assessment Number: 1E11-2022-003
- Corrective actions performed for identified evidence of boric acid leaks o
CRs 11066253 11066255, 11066256, 11066260, 11066265
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 main control room during the unit 1 shutdown for planned maintenance and refueling outage 1R32 on April 7, 2024.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 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 during a unit 2 'B' emergency diesel generator planned maintenance outage during the week of April 1, 2024
- (2) Unit 2 risk while the high voltage switchyard bus 1 was de-energized for an electrical maintenance outage April 17-20, 2024 (procedure NMP-GM-021)
- (3) Risk to units 1 and 2 with post-outage work ongoing in the low voltage switchyard while the '1-2A' emergency diesel generator was tagged out for planned maintenance on May 20, 2024 (NMP-GM-021)
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 1 'B' steam generator 'B' main steam isolation valve HV3370B exceeding acceptable stroke time identified on April 7, 2024 (CR 11065585)
- (2) Unit 1 'A' steam generator main steam line spring can support cracks identified on April 19, 2024 (CR 11069413)
- (3) Unit 1 main steam safety valves following the '1E' steam dump valve failure on May 9, 2024 (CR 11075471)
- (4) Unit 2 'B' emergency diesel generator with a jacket water leak discovered during surveillance testing on May 20, 2024 (CR 11080434)
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Permanent removal of the unit 1 pressurizer cubicle grating implemented during the unit 1 refueling outage 1R32 (design change package [DCP] SNC1356127)
- (2) Unit 1 nuclear steam service system control and steam generator level control system digital upgrade performed during the unit 1 refueling outage 1R32 in April 2024 (DCP SNC1083454)
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated refueling outage 1R32 activities from April 7, 2024, to May 9, 2024.
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) (6 Samples)
- (1) Unit 2 'B' emergency diesel generator jacket water flex hose replacement and jacket water pump seal replacement during the week of April 1, 2024 (WOs SNC965355 and SNC1129967)
- (2) Unit 1 source range detector (NI-32) repair during refueling outage 1R32 (CR 11070860)
- (3) Unit 1 power range nuclear instrument (NI-42) testing following replacement on May 1, 2024 (WO SNC1367753)
- (4) Unit 1 steam generator level control system digital upgrade power ascension testing performed during the week of May 6, 2024 (procedure FNP-1-SPETP-005)
- (5) Unit 2 'B' component cooling water pump testing following planned maintenance on June 17, 2024 (FNP-2-STP-23.2)
- (6) Emergency diesel generator '1C' jacket water coupling replacement on June 20, 2024 (FNP-0-SOP-38.0-1C)
Surveillance Testing (IP Section 03.01) (3 Samples)
- (1) Unit 1 'A' train loss of offsite power surveillance (FNP-1-STP-80.14; WO 480835) on April 8, 2024, and safety injection with loss of offsite power surveillance performed on April 12, 2024 (FNP-1-STP40.0A, WO SNC480835)
- (2) Unit 1 auxiliary building 'A' train battery modified performance testing on April 29, 2024 (FNP-1-STP-905.5; WO SNC715324)
- (3) Unit 1 'A' charging pump quarterly surveillance on June 17, 2024 (FNP-1-STP-4.1)
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Unit 2 'B' charging pump quarterly IST on April 15, 2024 (FNP-2-STP-4.2)
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) Containment purge exhaust isolation valve (penetration 13) on April 27, 2024 (FNP-1-STP-627.0)
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 and evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
- (1) Licensee surveys of potentially contaminated material leaving the radiologically controlled area.
- (2) Workers exiting the radiological controlled area during unit 1 refueling outage 1R32.
- (3) Licensee survey and labeling of bagged radioactive materials during unit 1 refueling outage 1R32.
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Reactor head disassembly/reassembly during 1R32. Radiation work permit (RWP)24-1461 (including As-Low-As-Reasonably-Achievable [ALARA] Plan, RWP, and radiation protection hold points)
- (2) Reactor coolant pump maintenance during 1R32. RWP 24-1428 (including ALARA Plan, RWP, respirator use evaluation (Total Effective Dose Equivalent [TEDE]/ALARA evaluation), temporary shielding, and radiation protection hold points)
- (3) Unit 1 reactor water storage tank cleanout and inspection. RWP 24-1752 (including ALARA Plan, RWP, respirator use evaluation (TEDE/ALARA evaluation), temporary shielding, and radiation protection hold points)
- (4) Unit 1 seal table and in-core work. RWPs 24-1435 (Non-High Rad) and 24-1438 (High Rad)
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
- (1) Unit 1 reactor sump access door on 105-foot elevation in containment (VHRA/Grave Danger)
- (2) Unit 1 regenerative heat exchanger on 105-foot elevation in containment (locked high radiation area [LHRA])
- (3) Unit 2 volume control tank room 121-foot (LHRA)
- (4) HRA controls on the 121, 100, and 83-foot elevations for unit 1 crud burst
- (5) HRA controls for reactor cavity work on 155-foot elevation in containment 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.
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (2 Samples)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
- (1) Units 1 and 2 spent fuel pool ventilation system filtration units
- (2) Units 1 and 2 radiological waste area ventilation system filtration units
Temporary Ventilation Systems (IP Section 03.02) (2 Samples)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) High efficiency particulate air filter units HP-NPU-053 and HP-NPU-054, tested October 3, 2023
- (2) Vacuum units VAC-048 and VAC-049, tested October 3, 2023
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (3 Samples)
The inspectors evaluated the following internal dose assessments:
- (1) Whole body count of carpentry supervisor on June 15, 2023
- (2) Whole body count of radiation protection technician on June 15, 2023
- (3) Whole body count of worker unable to pass exit monitors on April 9, 2024
Special Dosimetric Situations (IP Section 03.04) (3 Samples)
The inspectors evaluated the following special dosimetric situations:
- (1) Declared pregnant worker on January 30, 2023
- (2) Declared pregnant worker on March 20, 2024
- (3) Extremity monitoring and shallow dose equivalent assigned on September 18, 2022
71124.05 - Radiation Monitoring Instrumentation
Walkdowns and Observations (IP Section 03.01) (10 Samples)
The inspectors evaluated the following radiation detection instrumentation during plant walkdowns:
- (1) Personnel contamination monitors, portal monitors, and small article monitors located at the radiologically controlled area exit.
- (2) Unit 2, 155-foot elevation, plant vent stack (low/mid/high range gas effluent monitor)
(R-29B)
- (3) Unit 2, 120-foot elevation, liquid radwaste processing radiation monitor (R-18)
- (4) Unit 2, 120-foot elevation, steam generator blowdown liquid radwaste monitor (R-23B)
- (5) Portable friskers located in various locations in the auxiliary building
- (6) Portable survey instruments at the radiation protection instruments calibration laboratory
- (8) Whole body counter at the dosimetry office in the support building.
- (9) Unit 1 main control room radiation monitoring system
- (10) Unit 2 main control room radiation monitoring system
Calibration and Testing Program (IP Section 03.02) (15 Samples)
The inspectors evaluated the calibration and testing of the following radiation detection instruments:
- (1) Mirion People Mover Whole Body Counter - Auxiliary Building, calibrated June 21, 2023
- (2) Canberra Argos 5AB, HP-IPC-036A, calibrated March 24, 2024
- (3) Canberra Argos 5AB, HP-IPC-037A, calibrated March 23, 2024
- (4) Canberra Argos 5AB, HP-IPC-040, calibrated July 24, 2023
- (5) Thermo Fisher Scientific - Small Article Monitor (SAM-12), HP-GSD-028A, calibrated August 17, 2023
- (6) Thermo Fisher Scientific - Small Article Monitor (SAM-12), HP-GSD-029, calibrated August 10, 2023
- (7) Canberra Gamma Exit Monitor (GEM-5), HP-GDS-025, calibrated August 2, 2023
- (8) Mirion Telepole II Dose Rate Meter, HP-GMT-289, calibrated March 21, 2024
- (9) Mirion Telepole II Dose Rate Meter, HP-GMT-299, calibrated March 19, 2024
- (10) Mirion Alpha Beta Particulate Monitor (ABPM 203M), HP-LAS-104A, calibrated February 8, 2024
- (11) Mirion Alpha Beta Particulate Monitor (ABPM 203M), HP-LAS-107A, calibrated February 13, 2024
- (12) Ludlum 9-3 Ion Chamber Survey Meter, HP-IOC-202, calibrated August 7, 2023
- (13) Ludlum Model 30 Neutron Meter, HP-NDR-006, serial number 25014116 (detector Ludlum 42-49, serial number 369649), calibrated July 27, 2023
- (14) Ludlum 177-61 Benchtop Counter, HP-GMT-18C, calibrated January 24, 2020
- (15) Ludlum 177-61 Benchtop Counter, HP-GMT-011C, calibrated December 11, 2023 Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (3 Samples)
The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:
- (1) Unit 2, plant vent stack low/mid/high range gas effluent monitor (R-29B), calibrated December 23, 2023, and October 6, 2023
- (2) Units 1 and 2, liquid radwaste processing monitors (R-18), calibrated June 6, 2023, and May 13, 2023, respectively
- (3) U1, plant vent stack (particulate, iodine, and noble gases) (R-29C), calibrated November 10, 2021 and October 9,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS09: Residual Heat Removal Systems (IP Section 02.08)===
- (1) Unit 1 (April 1, 2023 - March 31, 2024)
- (2) Unit 2 (April 1, 2023 - March 31, 2024)
MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)
- (1) Unit 1 (April 1, 2023 - March 31, 2024)
- (2) Unit 2 (April 1, 2023 - March 31, 2024)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) October 1, 2023 - March 31, 2024
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 1 pressurizer relief tank slow pressure rise identified on May 5, 2024 (CR11076949)
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program for potential adverse trends in motor driven fire pump maintenance that might be indicative of a more significant safety issue.
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensees event reporting determinations to ensure it complied with reporting requirements.
- (1) Licensee event report (LER) 05000364/2024-002-00, Manual Reactor Trip due to Loss of Power to the 2A 125 Volt DC Distribution Panel, dated April 15, 2024.
Agencywide Documents Access and Management System (ADAMS) Accession No. ML24106A162. The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct, and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is Closed.
- (2) LER 05000364/2024-002-01, Manual Reactor Trip due to Loss of Power to the 2A 125 Volt DC Distribution Panel, dated June 19, 2024. ADAMS Accession No. ML24171A018. The inspectors reviewed the updated LER submittal. The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensee's ability to foresee and correct, and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified. This LER is Closed.
- (3) LER 05000364/2023-002-00, Main Steam Safety Valve Lift Pressure Outside of Technical Specification Limits due to Setpoint Drift, dated November 30, 2023.
ADAMS Accession No. ML23334A228. The inspection conclusions associated with this LER are documented in this report under Inspection Results section (NCV 05000364/2024002-01). This LER is Closed.
INSPECTION RESULTS
Main steam safety valve lift pressure outside of TS limits due to setpoint drift Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000364/2024002-01 Open/Closed EA-24-072 Not Applicable 71153 A self-revealed Severity Level (SL) IV non-cited violation (NCV) of Technical Specifications (TS) Limiting Condition of Operation 3.7.1, Main Steam Safety Valves, was identified when a routine lift pressure test revealed that the lift setpoint for the C main steam safety valve (MSSV) Q2N11V0011C, associated with the B main steam line, was higher than allowed by TS. Based on the Licensee Event Report 2023-002-00 submitted by the licensee regarding the issue and inspector evaluation, it was determined that MSSV Q2N11V0011C was inoperable for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> between April 30, 2019, to October 3, 2023, while the unit was in modes 1, 2, and 3.
Description:
The MSSVs provide overpressure protection for the secondary side main steam lines by limiting secondary system pressure to 110% (1195 psig) of the steam generators design pressure. The MSSVs also provide overpressure protection for the reactor coolant system (RCS) boundary by providing a heat sink for the removal of energy from the RCS if the preferred heat sink (i.e., condenser) is not available. The MSSVs are designed to limit RCS to 110% (2735 psig) of its design pressure.
On October 3, 2023, while in mode 1, before the Farley Nuclear Plant Unit 2 October 2023 refueling outage 2R32, the as-found lift pressure for MSSV Q2N11V0011C (i.e., C safety valve on the B main steam line) did not meet the acceptance criteria of +/- 3 percent of the setpoint (1102 psig) required for operability per TS Table 3.7.1-2. The valve lifted high at 1140 psig, which is 5 psig outside of its acceptance range of 1069 to 1135 psig. The valve was installed and placed in service in unit 2 on April 26, 2016 (work order [WO] SNC82187).
Additionally, the valve was found within the +/- 3 percent acceptance criteria during the most recent lift test surveillance on April 2, 2019 (WO SNC786539).
On November 30, 2023, Farley submitted Licensee Event Report (LER) 05000364/2023-002-00, Main Steam Safety Valve Lift Pressure Outside of Technical Specification Limits due to Setpoint Drift, (ADAMS Accession No.ML23334A228) in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.73, Licensee event report system.
Corrective Actions: The MSSV setpoint was adjusted to within the TS as-left requirement of
+/-1 percent of the setpoint (1102 psig) on October 3, 2023. The same valve was then replaced with a refurbished and tested valve during the unit 2 refueling outage 2R32 (WO SNC1167341). The licensee conducted a cause evaluation (documented in corrective action report [CAR] 529355) that included evaluation of a failure analysis report (technical evaluation
[TE] 1141155) conducted by an offsite vendor. Other than normal setpoint drift, no extenuating circumstance or conditions were identified as the cause of the drift. The licensee conclusions were consistent with the results of the failure analysis. Per industry operating experience, absent any identifying cause, the small drift above setpoint observed on the valve are not outside the range of historical performance for this type of valve and is attributed to normal setpoint drift.
Corrective Action References: CR11012189 and CAR 529355
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-24-072, in accordance with Section 3.10 of the Enforcement Policy, because the violation was not associated with a licensee performance deficiency. Specifically, the violation was not attributable to equipment failures that were avoidable by reasonable licensee quality assurance measures or management controls. Therefore, inspectors concluded that there was no performance deficiency associated with MSSV Q2N11V0011C failure to lift within the TS acceptance criteria. The licensee adjusted the valve to within TS limits and completion times, subsequently replaced the valve during the upcoming unit 2 refueling outage (2R32)and conducted a cause evaluation that determined that the observed drift was within the range of historical performance for this type of valve. This enforcement discretion will not be considered in the assessment process of the NRCs Action Matrix.
Severity: The inspectors assessed the severity of the violation using Section 6.1 of the Enforcement Policy. The circumstances of the violation aligned with an SL III violation example in Section 6.1.c.2 of the Enforcement Policy and was considered for escalated enforcement action. However, the NRC concluded that the violation resulted in no, or relatively minimal potential safety impact because
- (1) other main steam and RCS pressure reducing components were available to support the systems function, and
- (2) an informational risk analysis performed by NRC staff indicated the violation was of very low risk significance. Therefore, the NRC staff and determined the significance is appropriately characterized at SL IV.
Violation: Farley Nuclear Plant Unit 2 TS LCO 3.7.1, Main Steam Safety Valves, requires five operable MSSVs per steam generator with a lift setting of +/- 3 percent of the value corresponding to each valve in accordance with TS Table 3.7.1-2, while the unit is in modes 1, 2, and 3. Table 3.7.1-2 specifies that the corresponding lift setpoint of MSSV Q2N11V0011C is +/- 3 percent of 1102 psig. With one MSSV inoperable, Required Action A.1, requires reactor power to be reduced to less than or equal to 87 percent rated thermal power within four hours. If the required action and associated completion time is not met, Required Action C.1, requires that the unit to be in mode 3 within six hours.
Contrary to the above, for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> within the period from April 30, 2019, to October 3, 2023, MSSV Q2N11V0011C was inoperable because it failed to meet the lift setpoint established in TS Table 3.7.1-2 while the unit was in modes 1, 2, and 3. With MSSV Q2N11V0011C inoperable, the licensee failed to take Required Action A.1 to reduce power within four hours and Required Action C.1 to place the unit in mode 3 within six hours.
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 July 23, 2024, the inspectors presented the integrated inspection results to John Andrews, Plant Manager, and other members of the licensee staff.
- On April 11, 2024, the inspectors presented the radiation safety inspection results to Edwin Dean, Site Vice President, and other members of the licensee staff.
- On April 11, 2024, the inspectors presented the inservice inspection results to Edwin Dean, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Condition Report
(CR)
CR 11079418
Procedures
NMP-GM-025
Seasonal Readiness Process
05/13/2024
Corrective Action
Documents
CR 11080434
Corrective Action
Documents
CR 11071519
Procedures
FNP-1-STP-228.6
NIS Power Range Channel N42 Calibration N1C55NE0042
Ver. 81.0
Procedures
FNP-1-STP-4.1
1A Charging Pump Quarterly Inservice Test
Ver. 81.0
Procedures
FNP-1-STP-627.0
Local Leak Rate Testing of Containment Penetrations
Ver. 63.0
Procedures
FNP-2-STP-23.2
2B Component Cooling Water Pump Quarterly Inservice Test
Ver. 39.0
Work Orders
SNC1048142
OHI -N1C55NI0042-Replace TRIAX connectors for U1 NI42
NIS/Excore Detectors
05/01/2024
Work Orders
SNC1237420
U1 7300-Q1H11NGPIC2505-Power Ascension Test-Process
I&C System 7300
06/10/2024
Work Orders
SNC1367753
OHI - N1C55NE0042A and N1C55NE0042B - Replace
Upper and Lower Detectors IAW FNP-1-IMP-228.5
05/01/2024
Work Orders
SNC1642064
Q2P17P001B - FNP-2-STP-23.2 l 2B CCW PMP Quarterly
06/18/2024
Work Orders
SNC1772453
AEA - Penetration 13 - CTMT Purge Exhaust excessive
LLRT leakage
ALARA Plans
Farley Nuclear Plant U2R29 Refueling Outage ALARA
Report (10/8/23-11/17/23)
03/12/2024
Calculations
Fleet Technical Position Paper for the Conversion of Smear
Beta-Gamma Activity From mRad/h to Disintegrations Per
Minute (dpm)
Corrective Action
Documents
CRs
11018918, 11019626, 11022219, 11030825, and 11032018
Various
Miscellaneous
High Radiation/Locked High Radiation Area Key Log
04/04/2024
Miscellaneous
Facility Alpha Characterization Study - Farley
01/25/2024
Miscellaneous
Unit 1 and 2 Spent Fuel Pool Inventory Log [Non-Fuel/Trash]
01/20/2022
Procedures
FNP-0-RCP-126
Special Radiological Controls for Reactor Coolant System
(RCS) Degas and Crud Burst
Ver. 1.0
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
NMP-AD-029
Preparation and Reporting of Regulatory Assessment
Performance Indicator Data and the Monthly Operating
Report
Ver. 3.2
Procedures
NMP-HP-204
ALARA Planning and Job Review
Ver. 10.5
Procedures
NMP-HP-205-001
Farley Site Specific Temporary Shielding Information
Ver. 1.0
Procedures
NMP-HP-302-001
Radiological Key Control
Ver. 6.0
Radiation
Surveys
Plant Farley Routine Radiological Survey Frequency List
04/06/2024
Radiation
Surveys
Radiological
Information
Survey (RadIS) #
2964
Spent Fuel Storage lnstallation (ISFSI)
05/25/2023
Radiation
Surveys
RadIS # 163822
Spent Fuel Storage lnstallation (ISFSI)
07/14/2023
Radiation
Surveys
RadIS # 168039
Solidification and Dewatering Facility (SSDF)
03/07/2024
Radiation
Surveys
RadIS # 168474
Unit 1 155 ft. Containment [Initial Entry after shutdown]
04/07/2024
Radiation
Surveys
RadIS # 168475
Unit 1 129ft Containment [Initial Entry after shutdown]
04/07/2024
Radiation
Surveys
RadIS # 168476
Unit 1 105ft Containment O/S Biowall [Initial Entry post
shutdown]
04/07/2024
Radiation
Surveys
RadIS # 168477
Unit 1 105ft Containment Inside Biowall [Initial Entry post
shutdown]
04/07/2024
Radiation
Surveys
RadIS # 168488
Unit 1 A RCP Cubicle [Initial Entry after shutdown]
04/07/2024
Radiation
Surveys
RadIS # 168520
Unit 1 105ft Containment [Inside bio-wall to downgrade
posting after crud burst clean-up]
04/08/2024
Radiation
Surveys
RadIS # 168538
Unit 1 A/B 121ft piping penetration room [Down grade
posting after Crud Burst]
04/08/2024
Radiation
Surveys
RadIS # 168540
Unit 1 A/B 100ft piping penetration room [Down grade
posting after Crud Burst]
04/08/2024
Radiation
Surveys
RadIS # 168542
Unit 1 A/B Monday 83ft [Down grade posting after Crud
Burst]
04/08/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Radiation
Surveys
RadIS #168543
Unit 1 RHR Pump Rooms [Down grade posting after Crud
Burst]
04/08/2024
Self-Assessments
Pre-Fleet RP NOS audit Check-In Self-Assessment
05/25/2023
Self-Assessments Fleet-RP-2023
Audit of Radiation Protection (Farley, Hatch, and Vogtle 1&2)
06/06/2023
Corrective Action
Documents
CRs
10926363, 10957472, 11066290, and 11002557
Various
Corrective Action
Documents
CRs
11043344 and 11022219
Various
Corrective Action
Documents
Resulting from
Inspection
CR 11063976
NRC identified a procedure enhancement for a Whole-Body
Counting
04/01/2024
Corrective Action
Documents
CRs
10938833, 10957472, 10999269, and 11029306
Various
Corrective Action
Documents
Resulting from
Inspection
CRs
11063188, 11063200, 11063976, 1106332, 11064328, and
11063187
Various
71151
Miscellaneous
Electronic Dosimeter (ED) Dose and Dose Rate Alarm Log
(October 1, 2023 through April 10, 2024)
Various
71152S
Corrective Action
Documents
CRs
11085954 and 11090238
Various
Corrective Action
Documents
Corrective Action
Report (CAR)
29355
B Loop Main Steam Safety Valve (MSSV) Q2N11V0011C
failed high during setpoint verification testing
2/12/2024
Drawings
D-205033
Unit 2 - P&ID - Main Steam and Auxiliary Steam System
Ver. 44.0
Engineering
Evaluations
APR-TANL-TM-
AA-000002
Farley Unit 2 Evaluation of Main Steam Safety Valve
Elevated Pressure for a Justification of Past Operability
(JPO)
06/25/2024
Miscellaneous
NMS Traveler 23-
380
NWS Technologies Failure Analysis Report: Farley Dresser
MSSV, Model 3707RA-RT22 (SN: BP09820)
Rev. 0
Procedures
FNP-2-STP-608.1
Main Steam Safety Valve Operational Test by Vendor
Ver. 19.0
Work Orders
SNC1074173
IST - Q2N11V0011C. B Loop MS Safety - FNP-2-STP-608.1
10/16/2023
Work Orders
SNC1081243
IST - Q2N11V0011D. B Loop MS Safety - FNP-2-STP-608.1
10/15/2023
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Work Orders
SNC1120197
IST - Q2N11V0011E. B Loop MS Safety - FNP-2-STP-608.1
10/05/2023
Work Orders
SNC1167341
IST - Q2N11V0011C. B Loop MS Safety - Remove for
Inspection/Repair
10/20/2023
Work Orders
SNC786539
IST - Q2N11V0011C. B Loop MS Safety - FNP-2STP-608.0
04/02/2019