IR 05000348/2024001
| ML24108A093 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/24/2024 |
| From: | Alan Blamey NRC/RGN-II/DRP/RPB2 |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| IR 2024001 | |
| Download: ML24108A093 (15) | |
Text
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT-INTEGRATED INSPECTION REPORT 05000348/2024001 AND 05000364/2024001
Dear Jamie Coleman:
On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Joseph M. Farley Nuclear Plant. On April 22, 2024, 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.
One finding of very low safety significance (Green) is documented in this 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.
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.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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.April 24, 2024 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 2 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: 05000348/2024001 and 05000364/2024001
Enterprise Identifier: I2024001-0022
Licensee: Southern Nuclear Operating Company, Inc.
Facility: Joseph M. Farley Nuclear Plant
Location: Columbia, AL
Inspection Dates: January 01, 2024 to March 31, 2024
Inspectors: A. Alen Arias, Senior Project Engineer J. Bell, Senior Health Physicist J. Copeland, Reactor Inspector M. Donithan, Senior Operations Engineer K. Kirchbaum, Senior Operations Engineer P. Meier, Senior Resident Inspector
Approved By: Alan J. Blamey, Chief Reactor Projects Branch 2 Division of Reactor Projects
Enclosure 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
Inadequate post-modification testing associated with feedwater control system upgrades leads to unanticipated manual reactor trip Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.5] - Work 71153 FIN 05000364/202400101 Management Open A self-revealed Green Finding was identified for the licensees failure to establish and implement adequate post-modification testing associated with a unit 2 digital feedwater control system upgrade. Specifically, the licensee's post-modification testing failed to identify errors with the steam and feedwater flow signal in the design of the new system which led to an unanticipated feedwater system transient and a reactor trip.
Unit 2 Pressurizer Safety Valve Lift pressure Outside of Technical Specifications Limits Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71153 NCV 05000364/202400102 Open/Closed 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 a routine lift pressure test revealed that the unit 2 C pressurizer safety valve (PSV) asfound set pressure was higher than allowed by TS. This condition was determined to exist for a duration that exceeded the associated TS required action completion time.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000364/2024001-00 LER 2024001-00 for Joseph 71153 Closed M. Farley Nuclear Plant, Unit 2, Manual Reactor Trip due to Rising Steam Generator Levels
LER 05000364/2023003-00 LER 2023003-00 for Joseph 71153 Closed M. Farley Nuclear Plant, Unit 2, Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits
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 100 percent RTP and remained at or near 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
71111.01 - 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 thunderstorms and high winds on January 8, 2024 for forecasted conditions January 9, 2024 (NMP-OS017)
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1) Unit 1 component cooling water (CCW) system during the 'A' CCW heat exchanger inspection and 'B' CCW pump motor replacement during the week of January 29, 2024 (Dwg D175002)
(2) Unit 2 'A' train containment spray system during planned maintenance on the 'B' train containment spray system on February 27, 2024 (Dwg D205038; FNP2SOP9.0 A)
(3) Unit 1 'A' train penetration room filtration system during planned maintenance on the
'B' train primary room filtration system on February 28, 2024 (Dwg D175022; FNP1SOP60.0)
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1) The inspectors evaluated system configurations during a complete walkdown of the unit two motor driven auxiliary feedwater (AFW) system during the week of January 29, 2024 (Dwg D175007)
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 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 1 'B' train switchgear room (FA 1021) on March 7, 2024 (FNP1FPP1.0)
(2) Unit 1 'A' train vital DC room (FA 1018) on March 7, 2024 (FNP1FPP1.0)
(3) Unit 1 AFW control panel room (FA 1006) on March 15, 2024 (FNP1FPP1.0)
(4) Unit 1 'V' motor control center room (FA 1034) on March 20, 2024 (FNP1FPP1.0)
(5) 1C emergency diesel generator room (FZ DGB1C) on March 20, 2024 (FNP0FPP2.0)
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
(1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on February 9, 2023
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
(1) The inspectors evaluated internal flooding mitigation protections in the unit 2 piping penetration room 121 foot level on February 15, 2024 (calculation BM991932-001; U00FRIE-IF006)
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 for the following:
- Control room operations and annunciator response during a severe thunderstorm on January 9, 2024
- Unit 1 power increase following turbine valve testing on January 11, 2024
- General control room operations and annunciator response on January 30, 2024
- Unit 1 operators performing a unit 1 emergency diesel generator 24-hour surveillance run and transfer of fuel oil to the day tank on February 5, 2024
- Control room operator response to a unit 1 'A' inverter fault alarm response and restoration of instrument air after maintenance on the unit 1 '4A'
feedwater heater dump valve on February 8, 2024
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)
(1) The inspectors observed and evaluated a licensed operator continuing training graded simulator scenario involving a large break loss of cooling accident on February 1, 2024 (Segment 241 'As Left').
(2) The inspectors observed and evaluated the operators perform simulated time critical actions on March 21, 2024
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1) 2C emergency diesel generator (station blackout generator) coupling oil leak identified on February 8, 2024 (CR11048153)
(2) Unit 1 and unit 2 condensate storage tank to auxiliary feedwater pump isolation valves (Q1N23V501, Q1N23V502, Q2N23V501, Q2N23V502) due to identified corrosion (CR10977854)
Aging Management (IP Section 03.03) (1 Sample)
The inspectors evaluated the effectiveness of the aging management program for the following SSCs that did not meet their inspection or test acceptance criteria:
(1) Unit 1 low voltage switchyard fire protection header rupture identified on February 20, 2024 (CR11051746; CR10773474)
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 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 1 'C' service air compressor (instrument air) planned maintenance outage on January 3, 2024 (WO SNC1345889)
(2) Unit 2 risk during a 12A emergency diesel generator and unit 2 'A' motor driven auxiliary feedwater pump room cooler planned maintenance outages on January 22, 2024 (NMP-OS010)
(3) Fire main rupture repair in the low voltage switchyard resulting in isolation of fire water to the unit 1 turbine building on February 20, 2024 (CR11051746)
(4) Unit 2 risk due to 'B' train residual heat removal pump out-ofservice for planned maintenance on the pump's minimum flow line flow switch on March 7, 2024 (NMP-OS010, SNC1376529)
(5) Unit 2 risk during a 2C service air compressor and 1C emergency diesel generator planned maintenance outages on March 11, 2024 (WO SNC1297902, SNC1386514)
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1) Unit 2 pressurizer power operated relief valve (PORV) PCV445A isolated by its associated block valve (MOV8000A) due to leak-by identified on November 17, 2023 (CR 11025500; TE 1141914)
(2) Unit 1 'B' motor auxiliary feedwater pump main control board hand switch light indication not working identified on January 31, 2024 (CR11045269; Dwg D177186)
(3) Unit 1 'B' inverter frequency was found low during routine rounds on January 26, 2024 (CR 11043764; 11043820)
(4) Unit 1 containment sump to the 'A' residual heat removal pump suction valve (Q1E11V025A) packing leak and boric acid accumulation identified on February 20, 2024 (CR 11052099)
(5) 1C emergency diesel generator jacket water leak identified on March 18, 2024 (CR11059965)
(6) Unit 2 post accident containment vent flow transmitter out of calibration identified on February 11, 2024 (CR11049169)
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) (5 Samples)
(1) Unit 1 'C' steam generator steam supply valve (HV3235B) to the turbine driven auxiliary feedwater pump actuator air leak identified on January 3, 2024 (CR 11035858; WO SNC1680944)
(2) Unit 2 'B' containment spray pump room cooler service water outlet relief valve (Q2P16V0035B) replacement on January 10, 2024 (WO SNC1081607; CR11037910)
(3) 12A emergency diesel generator jacket water leak repair on January 22, 2024 (WO SNC1636999)
(4) Unit 1 'B' component cooling water pump motor during the week of January 29, 2024 (WO SNC1680872)
(5) Unit 1 'B' charging pump planned maintenance outage and balancing line replacement during the week of March 18, 2024 (WO SNC1162960)
Surveillance Testing (IP Section 03.01) (4 Samples)
(1) Unit 2 'B' charging pump surveillance on January 16, 2024 (FNP2STP4.2, WO SNC1506099)
(2) Unit 2 flux map on February 7, 2024 (FNP2STP121.0)
(3) The unit 1 Technical Specifications reactor coolant system specific activity surveillance on February 13, 2024 (WO SNC1531212)
(4) 'B' train control room emergency filtration system charcoal sampling on February 26, 2024 (FNP0STP950.1, FNP0STP590.2, SNC1265509, SNC1252105)
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
(1) Unit 1 turbine auxiliary feedwater pump quarterly surveillance run on January 4, 2024 (FNP1STP22.16, WO# SNC1478325)
(2) Unit 1 residual heat removal pump quarterly surveillance run on January 8, 2024 (FNP1STP11.2, WO# SNC1493187)
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
(1) Licensed operator continuing training simulator scenario involving an emergency declaration associated with a large break loss of coolant accident on February 1, 2024.
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) (2 Samples)
(1) Unit 1 (January 1, 2023, through December 31, 2023)
(2) Unit 2 (January 1, 2023, through December 31, 2023)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
(1) Unit 1 (January 1, 2023, through December 31, 2023)
(2) Unit 2 (January 1, 2023, through December 31, 2023)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
(1) Unit 1 (January 1, 2023, through December 31, 2023)
(2) Unit 2 (January 1, 2023, through December 31, 2023)
71152A - Annual Follow-up Problem Identification and Resolution
Annual Follow-up of Selected Issues (Section 03.03) (3 Samples)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1) Potential missed preventive maintenance on emergency diesel generator flexible and rubber hoses identified on December 13, 2023 (CR11031451)
(2) A potential defect reported by another licensee under 10 CFR 21.21(d)(3)(i) (i.e. Part 21) associated with fuel oil piping on the emergency diesel generators (Event Number
[EN] 56905; CR11038240)
(3) 10 CFR 21 notification regarding medium voltage circuit breakers with potential failure to close identified on June 20, 2022 (ML22229A515; ML23349A115; CR10889101)
71153 - Follow Up of Events and Notices of Enforcement Discretion
Event Report (IP Section 03.02) (2 Samples)
The inspectors evaluated the following licensee event reports (LERs):
(1) LER 05000364/2024001-00, Manual Reactor Trip due to Rising Steam Generator Levels (ADAMS Accession No. ML24011A217): The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is Closed.
(2) LER 0500364/2023003-00, Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits (ADAMS Accession No. ML23341A209): The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is Closed.
INSPECTION RESULTS
Inadequate post-modification testing associated with feedwater control system upgrades leads to unanticipated manual reactor trip Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green [H.5] - Work 71153 FIN 05000364/202400101 Management Open A self-revealed Green Finding was identified for the licensees failure to establish and implement adequate post-modification testing associated with a unit 2 digital feedwater control system upgrade. Specifically, the licensee's post-modification testing failed to identify errors with the steam and feedwater flow signal in the design of the new system which led to an unanticipated feedwater system transient and a reactor trip.
Description: On November 14, 2023, Farley unit 2 reactor was manually tripped at approximately 10% rated thermal power (RTP) while starting up after a refueling outage (2R29). The reactor was manually tripped due to rapidly increasing steam generator water levels caused by inaccurate steam and feedwater flow signals.
During 2R29, Farley unit 2 underwent a digital feedwater flow control (Ovation) system upgrade. At the time of installation, this new system had associated steam and feedwater flow signal errors that were unintentionally introduced during the design process. These errors were not discovered during the post-maintenance testing / post-modification testing (PMT) process and therefore left uncorrected before reactor startup. However, the flow signal errors were identified during reactor startup at roughly three percent RTP. The licensee had an opportunity at this time to stop and correct the errors before proceeding, but instead
decided to wait until 12 percent RTP to correct the errors.
The Ovation system is designed to work in two modes: low-power and high-power. The transition between the two is designed to occur based on a pre-determined threshold of Ovation feedwater indication flow of 20 percent equivalent RTP. However, due to the flow signal error, it entered high-power mode at 10 percent RTP because the Ovation feedwater flow indicated 20 percent equivalent RTP. This was not anticipated by the licensee as reactor power was being raised to 12 percent RTP. Upon entering high-power mode, the system swapped over from controlling feedwater flow on the bypass feedwater regulating valves (BFRVs) to the main feedwater regulating valves (MFRVs). When the mode switch occurred early, all three steam generators experienced rapidly increasing water levels because the MFRVs lack the ability to adequately control feedwater flow at a lower reactor power. Upon recognizing this, control room operators manually tripped the reactor prior to reaching the automatic trip setpoint.
Corrective Actions: Following the reactor trip, the licensee corrected the Ovation feedwater flow signal errors.
Corrective Action References: Work Order SNC1629442 Performance Assessment:
Performance Deficiency: The failure to follow the standards for PMTs for modification packages as required by licensee procedure NMP-MA014, Post Maintenance Testing/Post Modification Testing, was a performance deficiency. Specifically, section 2.6 of NMP-MA014 states, in part, that post-modification tests for modification packages are required to verify that the final configuration meets all the design and operational requirements following the modification.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to verify the modified feedwater control system met the intended design and operation requirements resulted in a feedwater transient causing an unplanned manual reactor trip.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding was determined to be of very low safety significance (Green) because when screened as a transient initiator under the initiating events cornerstone screening questions, it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser or feedwater).
Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. The inspectors determined the licensee failed to adequately implement the work activities associated with the feedwater control modification despite the multiple plant departments (i.e.
engineering, maintenance, operations) involved with coordination.
Enforcement: Inspectors did not identify a violation of regulatory requirements associated with this finding.
Unit 2 Pressurizer Safety Valve Lift pressure Outside of Technical Specifications Limits Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71153 Applicable NCV 05000364/202400102 Applicable Open/Closed 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 a routine lift pressure test revealed that the unit 2 C pressurizer safety valve (PSV) asfound set pressure was higher than allowed by TS. This condition was determined to exist for a duration that exceeded the associated TS required action completion time.
Description: During the Farley unit 2 fall 2023 refueling outage (2R29) on October 20, 2023, the C PSV (Q2B13V0031C) was removed from service for routine testing at an off-site facility. On October 24, 2023, the site was notified that the asfound set pressure was at 2522 psig, which was high and outside the TS maximum allowable lift pressure setting of 2510 psig. Licensee Event Report (LER) 05000365/2023003-00, Pressurizer Safety Valve Lift Pressure Outside of Technical Specification Limits, (ADAMS Accession Number ML23341A209) was submitted by the licensee for this event. The PSV had been installed and placed in service at Farley Nuclear Plant unit 2 during the 2019 spring (work order (WO)
SNC694181) refueling outage (2R26) and remained in service during three complete 18-month fuel cycles. Based on the LER and inspector evaluation, it was determined the C PSV setting was outside the TS limit at some period between April 29, 2019, to October 8, 2023, while the unit was in modes 1, 2, 3, and 4 with all reactor coolant system (RCS) cold leg temperatures greater than 275°F (degrees Fahrenheit).
Although the TS lift pressure was exceeded, the asfound setpoint remained bounded by the accident analysis, as described in the final safety analysis report (FSAR). Therefore, the valves safety function to prevent the RCS from exceeding the safety limit of 2733 psig was not affected. Upon further examination, it was determined that the valve failed due to setpoint drift.
Corrective Actions: The valve was replaced with a similar operable refurbished valve on October 18, 2023, during the 2R29 refueling outage prior to plant startup.
Corrective Action References: Condition report 11017969 and work order SNC1078293.
Performance Assessment: The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Specifically, random setpoint drift is a recognized valid phenomenon that can occur despite routine testing and maintenance.
Enforcement: Traditional Enforcement is being used to disposition this violation with no associated Reactor Oversight Process performance deficiency per section 3.10 of the Enforcement Manual.
Severity: The inspector assessed the severity of the violation using Section 6.1 of the Enforcement Policy and determined the significance is appropriately characterized as SL IV, due to the low potential safety consequences. Specifically, the valves asfound lift pressure
remained bounded (i.e., below) by the accident analysis in the FSAR and did not have an adverse impact on RCS over-pressure protection.
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, that three PSVs shall be operable with lift settings greater than or equal to 2423 psig and less than or equal to 2510 psig while unit 2 is in modes 1, 2, 3, and 4 with all RCS cold leg temperatures greater than 275°F. Action Statement, Condition A, required restoration of an inoperable PSV 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, the unit 2 C PSV lift setting was determined to be greater than the TS limit of 2510 psig for longer than allowed by TS during the last three operating cycles, between April 29, 2019, and October 8, 2023, while unit 2 was in modes 1, 2, 3, and 4 with all RCS cold leg temperatures greater than 275°F. This period coincides with when unit 2 achieved mode 4 and greater than 275°F following the C PSV installation during the 2R26 refueling outage until the unit was placed in mode 4 and less than 275°F for the 2R29 refueling outage.
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 April 22, 2024, the inspectors presented the integrated inspection results to Edwin Dean III, Site Vice President, and other members of the licensee staff.
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