IR 05000348/2025001
| ML25126A165 | |
| Person / Time | |
|---|---|
| Site: | Farley, 07200042  |
| Issue date: | 05/13/2025 |
| From: | Alan Blamey NRC/RGN-II/DORS |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| IR 2025001 | |
| Download: ML25126A165 (19) | |
Text
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000348/2025001, 05000364/2025001, AND 07200042/2025001
Dear Jamie Coleman:
On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Joseph M. Farley Nuclear Plant. On April 22, 2025, the NRC inspectors discussed the results of this inspection with Edwin Dean III, Site Vice President and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations 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.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Joseph M. Farley Nuclear Plant.
May 12, 2025 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, Alan J. Blamey, Chief Reactor Projects Branch 3 Division of Operating Reactor Safety Docket Nos. 05000348, 05000364, and 07200042 License Nos. NPF-2 and NPF-8
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000348, 05000364, and 07200042
License Numbers:
Report Numbers:
05000348/2025001, 05000364/2025001, and 07200042/2025001
Enterprise Identifier:
I-2025-001-0024 and I-2025-001-0083
Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Joseph M. Farley Nuclear Plant
Location:
Columbia, AL
Inspection Dates:
January 1, 2025 to March 31, 2025
Inspectors:
B. Bowker, Senior Reactor Inspector
P. Cooper, Senior Reactor Inspector
P. Foley, Project Engineer
T. Griffin, Project Engineer
J. Hickey, Senior Project Engineer
K. Kirchbaum, Senior Operations Engineer
E. Morris, Senior Resident Inspector
D. Orr, Senior Project Engineer
R. Qian, General Engineer
M. Rich, Emergency Preparedness Inspector
S. Sandal, Senior Reactor Analyst
J. Tapp, Senior Transportation & Storage Safety Inspector
S. Temple, Senior Resident Inspector
J. Walker, Senior Emergency Preparedness Inspector
Approved By:
Alan J. Blamey, Chief
Reactor Projects Branch 3
Division of Operating Reactor Safety
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
Failure to Translate Controller Design Information into Unit 2 Turbine Driven Auxiliary Feedwater Pump Test Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000364/2025001-01 Open/Closed
[P.2] -
Evaluation 71152A A self-revealing Green finding and non-cited violation (NCV) of 10 CFR 50, Appendix B,
Criterion III, Design Control, was identified for failure to translate design requirements contained in system design procedures into test procedures. Specifically, the licensee failed to translate the turbine driven auxiliary feedwater pump (TDAFWP) controller design information found in FNP-2-SOP-22.0, Auxiliary Feedwater System, into test procedure FNP-2-STP-22.16, Turbine Driven Auxiliary Feedwater Pump Quarterly Inservice Test, which resulted in the controller failing to reset following a pump run and unplanned unavailability of the TDAFWP.
Unit 1 'B' Residual Heat Removal Pump Inoperable for Longer than Allowed by Technical Specifications Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000348/2025001-02 Open/Closed
[H.12] - Avoid Complacency 71153 A self-revealing Green finding and NCV of Technical Specification 5.4.1.a. Procedures, was identified for the licensees failure to follow licensee procedure NMP-GM 047, Plant Status and Configuration Control, Version 2.0, which resulted in an inoperable unit 1 B residual heat removal pump.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000348/2023-002-00 LER 2023-002-00 for Joseph M. Farley Nuclear Plant, Unit 1, Residual Heat Removal Pump Inoperable for Longer than Allowed by Technical Specifications 71153 Closed
PLANT STATUS
Unit 1 operated at or near 100% rated thermal power for the entire inspection period.
Unit 2 began the inspection period at 100% rated thermal power. On March 16, 2025, the unit was shutdown for a scheduled refueling outage and remained shutdown for the remainder of the 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
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 cold weather January 5-11, 2024 (NMP-OS-017, FNP-0-SOP-0.12).
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 2 'A' and 'C' component cooling water heat exchangers while the 'B' component cooling water heat exchanger was out of service for inspection during the week of January 6, 2025 (D205002)
(2)unit 2 'A' and 'B' component cooling water pumps and 'A,' 'B,' and 'C' component cooling water heat exchangers while the 'C' component cooling water pump was out of service for a pump seal replacement during the week of February 17, 2025 (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:
(1)unit 2 component cooling water room (fire zone 2185) on January 7, 2025 (FNP-2-FPP-1.0)
(2)unit 2 turbine driven auxiliary pump room (fire zone 2193) on January 8, 2025 (FNP-2-FPP-1.0)
(3)unit 2 hot shutdown panel room (fire zone 2202) on January 10, 2025 (FNP-2-FPP-1.0)
(4)unit 1 component cooling water room (fire zone 185) on January 14, 2025 (FNP-1-FPP-1.0)
(5)unit 1 class 1E DC switchgear, 125 VDC battery, and battery charger rooms and adjoining corridors (fire zones 210 - 214 and 224 - 226) on February 27, 2025 (FNP-1-FPP-1.0)
(6)unit 2 class 1E DC switchgear, 125 VDC battery, and battery charger rooms and adjoining corridors (fire zones 2210 - 2214 and 2224 - 2226) on February 27, 2025 (FNP-2-FPP-1.0)
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during announced fire drill F-2025-002 in the auxiliary building on February 7, 2025.
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated internal flooding mitigation protections in the unit 1 auxiliary building 100' elevation auxiliary feedwater system rooms 189, 190, 191, 192, 193, 194, and 241 on January 28, 2025.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)unit 1 'B' component cooling water heat exchanger during the week of January 6, 2025 (work order (WO) SNC1539061)
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:
- unit 1 and unit 2 annunciator ground response on February 25, 2025
- unit 2 reactor shutdown for planned maintenance outage on March 16, 2025
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator continuing training simulator scenario As-Left Exam 2 on February 6. 2025.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components remain capable of performing their intended function:
(1)repetitive unit 1 4160V switchgear fast dead bus transfer failures previously under 50.65 a(1) in March 2025 (EVAL-F-R15-05789)
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (6 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)1-2K load center maintenance outage and associated use of risk-informed completion for both units on January 15, 2025 (NMP-OS-010)
(2)unit 2 elevated risk due to heavy machinery being used to add and level gravel in the low voltage switchyard just north of the 2A Startup transformer on February 11, 2025 (NMP-GM-021)
(3)unit 2 elevated risk due to pre-outage work pulling cables in the switch house of the high voltage switch yard on February 12, 2025 (NMP-GM-021)
(4)unit 1 elevated risk due to service water pump wet pit maintenance resulting in reduced service water pump availability on February 18, 2025 (NMP-OS-010)
(5)unit 2 elevated risk due to service water pump wet pit cage maintenance affecting emergency diesel generator availability on February 21, 2025 (NMP-OS-010-001)
(6)unit 2 risk during inservice testing of auxiliary feedwater valves and main steam safety valves on March 11 and 12, 2025 (NMP-OS-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 2 'B' containment spray pump breaker charging springs failed to discharge automatically on January 23, 2025 (CR11139963)
(2)unit 2 condensate storage tank Hi/Lo level alarm during cold weather on January 24, 2025 (CR11144415)
(3)unit 1 turbine driven auxiliary feedwater pump oil leak on January 31, 2025 (CR11145503)
(4)unit 2 residual heat removal heat exchanger flow control valve erratic readings on March 22, 2024 (CR 11163190)
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)monitoring instrumentation installation for intermittent voltage spikes on unit 2 N-42 power range nuclear instrumentation on January 28, 2025 (SNC2121065)
(2)installation of copper sleeves into the unit 2 'B' component cooling water heat exchanger tubes in January and February 2025 (SNC1794898)
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Partial)
(1)
(Partial) The inspectors evaluated refueling outage 2R30 activities from March 16, 2025 to March 31, 2025.
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) (4 Samples)
(1)unit 1 turbine driven auxiliary feedwater valve testing following feedwater pump to 'B' steam generator valve (HV3228B) maintenance on January 14, 2025 (WO SNC2662799)
(2)unit 1 pressurizer power operated relief valve 444B testing following hot shutdown panel local/remote control switch replacement on January 29, 2025 (WO SNC1001955)
(3)unit 2 'C' component cooling water pump testing following inboard bearing and inboard mechanical seal replacement on February 19, 2025 (WOs SNC1989535 and SNC2317920)
(4)unit 2 'B' service water pump testing following motor replacement from February 19, 2025 through March 2, 2025 (WOs SNC1080997 and SNC2343627)
Surveillance Testing (IP Section 03.01) (6 Samples)
(1)unit 1 B train service water pump surveillance on January 2, 2025 (FNP-1-STP-24.2)
(2)unit 1 "A" auxiliary feedwater pump surveillance on February 12, 2025 (FNP-1-STP-22.1)
(3)unit 2 'A' auxiliary feedwater pump surveillance on February 20, 2025 (FNP-2-STP-22.1)
(4)unit 1 'C' service water pump automatic starting circuity surveillance on February 26, 2025 (FNP-1-STP-24.10)
(5)unit 2 main steam safety valve testing on March 11 and 12, 2025 (FNP-2-STP-608.1)
(6)unit 2 containment integrated leak rate testing March 20 and 21, 2025 (FNP-2-STP-117.0)
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
(1)unit 1 'C' service water (swing) pump automatic start in-service test on February 10, 2025 (FNP-1-STP-2.4.10)
(2)unit 2 'A' and 'B' motor driven auxiliary feedwater pump and valve inservice testing on March 12, 2025 (FNP-2-STP-22.10, FNP-2-STP-22.12, FNP-2-STP-22.26)
71114.02 - Alert and Notification System Testing
Inspection Review (IP Section 02.01-02.04) (1 Sample)
- (1) The inspectors evaluated the maintenance and testing of the alert and notification system during the week of March 17, 2025.
71114.03 - Emergency Response Organization Staffing and Augmentation System
Inspection Review (IP Section 02.01-02.02) (1 Sample)
- (1) The inspectors evaluated the readiness of the Emergency Response Organization during the week of March 17, 2025.
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated submitted Emergency Action Level, Emergency Plan, and Emergency Plan Implementing Procedure changes during the week of March 17, 2025. This evaluation does not constitute NRC approval.
71114.05 - Maintenance of Emergency Preparedness
Inspection Review (IP Section 02.01 - 02.11) (1 Sample)
- (1) The inspectors evaluated the maintenance of the emergency preparedness program during the week of March 17,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===
(1)unit 1 (January 1, 2024, through December 31, 2024)
(2)unit 2 (January 1, 2024, through December 31, 2024)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
(1)unit 1 (January 1, 2024, through December 31, 2024)
(2)unit 2 (January 1, 2024, through December 31, 2024)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
(1)unit 1 (January 1, 2024, through December 31, 2024)
- (2) unit 2 (January 1, 2024, through December 31, 2024)
EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)
- (1) July 1 through December 31, 2024 EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13) (1 Sample)
- (1) July 1 through December 31, 2024 EP04: Emergency Response Facility and Equipment Readiness (ERFER) (IP Section 02.14)
This is a new NRC performance indicator, described in NEI 99-02, Revision 8 (ML24331A114). Licensees began collecting data for this performance indicator January 1, 2025. Therefore, at the time of inspection, there was no quarterly data compiled and submitted to the NRC.
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)training instructor qualification discrepancies (CR 11149288)
(2)unit 2 turbine driven auxiliary feedwater pump controller reset verification on January 6, 2025 (CR 10206376 and 11139838)
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 05000348/2023-002-00, Unit 1, Residual Heat Removal Pump Inoperable for Longer Than Allowed by Technical Specifications (ADAMS Ascension No.
ML23333A215). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is closed.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===60859 - Independent Spent Fuel Storage Installation (ISFSI) License Renewal Inspection 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 IMC 2690, Inspection Program for Storage of Spent Reactor Fuel and Reactor-Related Greater-than-Class C Waste at Independent Spent Fuel Storage Installations (ISFSI)and for 10 CFR Part 71 Transportation Packagings."
Independent Spent Fuel Storage Installation (ISFSI) License Renewal Inspection===
The inspectors reviewed a sample of regulatory commitments, aging management Programs, and time limited aging analyses associated with the licensees implementation of the HI-STORM 100 Renewed Certificate of Compliance 1014, Amendment 2. The inspection took place from January 27 to 30, 2025, prior to the period of extended operation which begins on July 14, 2025. The inspectors reviewed implementing documents and conducted interviews with licensee staff to verify that the licensee completed the necessary actions to:
- (a) comply with the conditions stipulated in the renewed Certificate of Compliance and technical specifications; and
- (b) implement the aging management programs and time limited aging analyses as described in the NRC safety evaluation report and FSAR supplement.
The inspectors reviewed NMP-FOS-002, Dry Cask Storage Aging Management Program, which provides programmatic governance and implementation of the licensees HI-STORM Aging Management Program as required by Section 5.8 of Appendix A and Section 5.4 of Appendix C in the renewed HI-STORM Certificate of Compliance. Additionally, while onsite the inspectors directly observed activities associated with the MPC Aging Management Program and the HI-STORM Overpack Aging Management Program.
For those license renewal action items that were not completed at the time of this inspection, the team verified that there was reasonable assurance that such action items were on track for completion prior to the period of extended operation or in accordance with an established implementation schedule consistent with the conditions of the certificate of compliance, technical specifications, the NRC safety evaluation report, and the FSAR.
INSPECTION RESULTS
Failure to Translate Controller Design Information into Unit 2 Turbine Driven Auxiliary Feedwater Pump Test Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000364/2025001-01 Open/Closed
[P.2] -
Evaluation 71152A A self-revealing Green finding and non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for failure to translate design requirements contained in system design procedures into test procedures. Specifically, the licensee failed to translate the turbine driven auxiliary feedwater pump (TDAFWP) controller design information found in FNP-2-SOP-22.0, Auxiliary Feedwater System into test procedure FNP-2-STP-22.16, Turbine Driven Auxiliary Feedwater Pump Quarterly Inservice Test, which resulted in the controller failing to reset following a pump run and unplanned unavailability of the TDAFWP.
Description:
On January 6, 2025, the licensee performed FNP-2-STP-22.16, Turbine Driven Auxiliary Feedwater Pump Quarterly Inservice Test. The operators successfully started, ran, and stopped the TDAFWP using the C steam line. The operators then attempted to start the TDAFWP using the B steam line, and the pump failed to satisfy the test acceptance criteria.
The pump reached a speed of approximately 400 rpm, well below the acceptance criteria speed of greater than 3900 rpm.
Prior to the second start of the TDAFWP, the licensee determined that the pump was rotating at approximately 300 rpm due to leak by on valve Q2N12V006B (C steam line to TDAFWP isolation valve). This rotation allowed the TDAFWP controller display to read Ready to Start but did not reset the controller.
Licensee procedure FNP-2-SOP-22.0, Auxiliary Feedwater System, precaution and limitation 15 states in part, IF TDAFWP continues to roll after being shutdown and does NOT lower to less than 200 rpm (a dead band exists between 100-200 RPM for controller reset),
TDAFWP controller will not reset and complete the required shutdown sequence. As a result TDAFWP controller will not be in a configuration to receive the next start signal and TDAFWP will not start when demanded. The design information for the TDAFWP controller found in system procedure FNP-2-SOP-22.0 was not translated into the test procedure used on January 6, 2025. As a result, operators failed to verify that the controller reset after the pump run using the C steam line.
Corrective Actions: On January 7, 2025, following inspection and investigation of the TDAFWP and associated valves, the licensee reset the controller and successfully completed FNP-2-STP-22.16. On January 9, 2025, the licensee revised FNP-2-STP-22.16 and additional test procedures to include additional steps for speed verification of the TDAFWP and verification that the controller resets following each pump run.
Corrective Action References: Condition Reports 11139838 and 11139916
Performance Assessment:
Performance Deficiency: The licensees failure to translate the TDAFWP controller design information found in FNP-2-SOP-22.0 into procedure FNP-2-STP-22.16 was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the turbine driven auxiliary feedwater pump became unavailable to respond to initiating events due to the governor valve not resetting following testing.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the finding as having very low safety significance (Green) because, for Exhibit 2 (Mitigating Systems), all mitigating system screening questions were answered no.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, when FNP-2-SOP-22.0 was updated to include additional precautions and limitations related to resetting the TDAFWP controller, the extent of condition should have included translating this precaution into associated test procedures.
Enforcement:
Violation: 10 CFR 50 Appendix B, Criterion III, Design Control, requires in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.
Licensee system procedure FNP-2-SOP-22.0, Auxiliary Feedwater System, precaution and limitation 15 states in part, IF TDAFWP continues to roll after being shutdown and does NOT lower to less than 200 rpm (a dead band exists between 100-200 RPM for controller reset),
TDAFWP controller will not reset and complete the required shutdown sequence. As a result, TDAFWP controller will not be in a configuration to receive the next start signal and TDAFWP will not start when demanded.
Contrary to the above, as of January 6, 2025, the licensee failed to assure that design requirements of the TDAFWP controller were translated into procedure FNP-2-STP-22.16, Turbine Driven Auxiliary Feedwater Pump Quarterly Inservice Test. Specifically, the reset criterion of the TDAFWP controller related to pump speed was not translated into the test procedure, allowing the controller to remain in a configuration where it would not receive the next start signal, resulting in unplanned unavailability of the unit 2 TDAFWP.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Unit 1 'B' Residual Heat Removal Pump Inoperable for Longer than allowed by Technical Specifications Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000348/2025001-02 Open/Closed
[H.12] - Avoid Complacency 71153 A self-revealing Green finding and NCV of Technical Specification (TS) 5.4.1.a. Procedures, was identified for the licensees failure to follow licensee procedure NMP-GM 047, Plant Status and Configuration Control, Version 2.0, which resulted in an inoperable unit 1 B residual heat removal (1B RHR) pump.
Description:
On October 2, 2023, the unit 1 B residual heat removal (RHR) pump failed to manually start from the control room during a surveillance test. This pump was last successfully started and stopped from the control room on July 5, 2023, for a routine surveillance. There is no record of any other pump or breaker manipulations between July 5, 2023, and October 2, 2023. The licensee found that the 4160 Volt (4KV) breaker charging spring was discharged with the breaker in the open position. Upon further investigation, the licensee found the racking release handle was in the trip free position which
- (1) prevents the breaker from closing and
- (2) prevents charging spring from being able to charge. It takes deliberate manual action consisting of four distinct steps to move the release handle from its normal position to the trip free position. Upon discovery of the issue, the licensee entered TS 3.5.2 Emergency Core Cooling Systems, Condition A, which requires restoration of the pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The licensee restored the release handle to the normal position and the spring was recharged and the breaker successfully closed. The licensee successfully completed the 1B RHR pump surveillance and declared the pump operable later the same day on October 2, 2023.
The licensee launched an independent investigation to determine how the racking release handle for the 1B RHR pump breaker was manipulated to the trip-free position. As previously noted, the independent investigation did not identify any work on the 1B RHR breaker, however, they did identify work done on the adjacent cubicle, the 1B emergency diesel generator (EDG) breaker, on August 28, 2023. Also, an NRC investigation identified that workers had entered the 1B RHR pump breaker cubicle during planned work on the 1B EDG breaker on August 28, 2023. There were no work instructions or approved written guidance allowing access or work on the 1B RHR pump breaker cubicle. This was contrary to licensee procedure NMP-GM-047, Plant Status and Configuration Control, Version 2.0, Step 4.1 which requires, in part that power plant components and equipment are manipulated only with approved written guidance (e.g. procedures, tagouts, work order instructions) and authorized to be implemented.
Corrective Actions: The licensee restored the release handle to the normal position and the spring was recharged allowing the breaker to successfully close resolving the issue on October 2, 2023. The licensee subsequently launched an Incident Response Team investigation to determine the root cause of the breaker racking release handle misposition.
Corrective Action References: The licensee captured the failure of the 1B RHR pump to start in Condition Report (CR) 1101749 and the restoration of the 1B RHR pump to operable status in work order SNC1586330. A root cause determination report was generated in CAR 540397.
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to follow licensee procedure NMP-GM047, Plant Status and Configuration Control, Version 2.0 was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 1B RHR pump could not perform its PRA function for greater than the TS allowed outage time.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors screened the significance of the finding using Exhibit 2 of IMC 0609 Appendix A for the mitigating systems cornerstone and determined that a detailed risk evaluation was required because the failure of the 1B RHR pump breaker represented inoperability of one train of a multi-train system for greater than the TS allowed outage time. A detailed risk evaluation was performed by a regional Senior Reactor Analyst using Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.2.12 and NRC Farley Standardized Plant Analysis Risk model Version 8.85. Fire and internal flooding risk insights were obtained from the licensees PRA model. A conditional analysis was performed to evaluate the risk increase due to the failure of the 1B RHR pump breaker using a condition exposure period of 35 days. The dominant cutsets involved small and medium sized loss of reactor coolant accidents that were accompanied by failure of the A-train RHR pump. Fire initiators were the most significant contributor to the overall risk estimate. The change in risk was mitigated by the relatively short exposure period of the finding and the amount of time that would have been available to operators to identify and correct the misposition of the racking release lever mechanism prior to depletion of the volume of water in the refueling water storage tank. The analysis determined that the increase in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) was less than 1E-06/year for delta-CDF and less than 1E-07/year for delta-LERF, representing a finding of very low safety significance (GREEN).
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, the electrical maintenance technicians did not utilize proper error reduction tools while performing work activities in the adjacent breaker cubicle on August 28, 2023, which was the most likely cause of the wrong component being mispositioned.
Enforcement:
Violation: Technical Specification 5.4.1.a requires in part, that written procedures shall be established, implemented, and maintained for the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, 1.c. requires administrative procedures for equipment control.
Technical Specifications Limiting Condition for Operation (LCO) 3.0.1 requires, in part, that LCOs shall be met during modes of applicability. TS LCO 3.5.2, Emergency Core Cooling Systems, requires, in part, two emergency core cooling system trains shall be OPERABLE.
Licensee procedure NMP-GM-047, Plant Status and Configuration Control, version 2.0, requires, in part that power plant components and equipment are manipulated only with approved written guidance (e.g. procedures, tagouts, work order instructions) and authorized to be implemented.
Contrary to the above, between August 28, 2023, and October 2, 2023, the 1B RHR pump breaker racking release handle was manipulated without approved written guidance or authorization in accordance with procedure NMP-GM-047 resulting in one train of the emergency core cooling system being inoperable for greater than its TS allowed outage time.
Specifically, the breaker racking release handle was taken to the trip-free position which prevents the 1B RHR pump breaker from closing upon demand. There was no evidence that approved written guidance or authorization was provided to manipulate the position of the racking release handle of the 1B RHR breaker.
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 that no proprietary information was retained or documented in this report.
- On January 30, 2025, the inspectors presented the ISFSI License Renewal Inspection Debrief Meeting inspection results to Edwin Dean III, Site Vice President, and other members of the licensee staff.
- On March 20, 2025, the inspectors presented the Emergency Preparedness Exercise Inspection results at the Exit Meeting to Edwin Dean III and other members of the licensee staff.
- On April 22, 2025, the inspectors presented the integrated inspection results to Edwin Dean III and other members of the licensee staff.
THIRD PARTY REVIEWS Inspectors reviewed Institute of Nuclear Power Operations report that was issued March 2025.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Southern Nuclear Operating Company Standard Emergency
Plan
Procedures
Standard Emergency Plan Annex for Farley Nuclear Plant,
Units 1 and 2
Corrective Action
Documents
206376,
11139838,
11139844,
11139916
FNP-2-SOP-22.0
Auxiliary Feedwater System
89.0
Procedures
FNP-2-STP-22.16
Turbine Driven Auxiliary Feedwater Pump Quarterly
Inservice Test
74.1
SNC1545487
FNP-2-STP-22.16 (TDAFW 18 MOS STM SUPPLY EAC)
01/07/2025
Work Orders
SNC1826506
FNP-2-STP-22.16 - U2 TDAFW Pump Quarterly (72hr RAS)
01/07/2025