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{{Adams | number = ML24304A883 | issue date = 11/06/2024 | title = Integrated Inspection Report 05000327/2024003 and 05000328/2024003 | author name = Mckown L | author affiliation = NRC/RGN-II/DORS | addressee name = Erb D | addressee affiliation = Tennessee Valley Authority | docket = 05000327, 05000328 | license number = DPR-077, DPR-079 | contact person = | document report number = IR 2024003 | document type = Inspection Report, Letter | page count = 1 }} {{IR-Nav| site = 05000327 | year = 2024 | report number = 003 }} =Text= {{#Wiki_filter:==SUBJECT:== SEQUOYAH UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000327/2024003 AND 05000328/2024003 ==Dear Delson Erb:== On September 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah Units 1 and 2. On October 15, 2024, the NRC inspectors discussed the results of this inspection with Tom Marshall, 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. This finding involved a violation of NRC requirements. 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 Sequoyah Units 1 and 2. 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 Sequoyah Units 1 and 2. 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. November 6, 2024 Sincerely, Louis J. McKown, II, Chief Projects Branch 5 Division of Operating Reactor Safety Docket Nos. 05000327 and 05000328 License Nos. DPR-77 and DPR-79 ===Enclosure:=== As stated ==Inspection Report== Docket Numbers: 05000327 and 05000328 License Numbers: DPR-77 and DPR-79 Report Numbers: 05000327/2024003 and 05000328/2024003 Enterprise Identifier: I-2024-003-0028 Licensee: Tennessee Valley Authority Facility: Sequoyah Units 1 and 2 Location: Soddy-Daisy, TN Inspection Dates: July 01, 2024 to September 30, 2024 Inspectors: L. Bryson, Project Engineer P. Cooper, Senior Reactor Inspector W. Deschaine, Senior Project Engineer V. Furr, Reliability and Risk Analyst P. Meier, Senior Resident Inspector A. Nielsen, Senior Health Physicist A. Price, Resident Inspector D. Restrepo, Health Physicist J. Rivera, Health Physicist Approved By: Louis J. McKown, II, Chief Projects Branch 5 Division of Operating Reactor Safety =SUMMARY= The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Sequoyah Units 1 and 2, 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=== RCS leak caused by backfill of a RHR system flow transmitter sensing line Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000327/2024003-01 Open/Closed [H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated non-cited violation of Technical Specification 5.4.1.a was identified when the licensee failed to implement written procedures covering activities recommended in Regulatory Guide 1.33, Revision, 2, Appendix A. Specifically, the licensee failed to establish and implement a procedure for backfill of residual heat removal (RHR) system flow transmitter sensing lines that was appropriate for plant conditions. This action resulted in lifting the A RHR discharge header relief valve and discharging reactor coolant system (RCS) water to the pressurizer relief tank (PRT). ===Additional Tracking Items=== None. =PLANT STATUS= Unit 1 began the inspection period at 100 percent rated thermal power (RTP). On August 23, 2024, the unit entered mode three following an automatic trip due to a turbine trip (Event Notification number 57285). Mode four was entered on August 26, 2024, to address the cause of the reactor trip. On August 31, 2024, the unit entered mode three and reactor startup (mode two) occurred. Mode one was entered on September 1, 2024, and the main generator was synchronized to the grid on the following day, September 2, 2024. Full RTP was achieved on September 3, 2024, until September 29, 2024, when the unit entered mode three again following an automatic trip due to a turbine trip (Event Notification number 57351). The unit entered mode four on September 30, 2024, and remained there through the remainder of the reporting period to address the cause of the reactor trip. Unit 2 began the inspection period at 100 percent RTP. On July 15, 2024, reactor power was reduced to 45 percent RTP to address main generator issues. Full RTP was restored on July 18, 2024. On July 20, 2024, a controlled rapid load reduction to mode three was performed to address main generator issues. On that same day, Unit 2 entered mode four to perform main generator inspections and repairs. On July 27, 2024, the Unit 2 reactor entered mode three and the reactor was restarted. On July 28, 2024, Unit 2 entered mode one and the main generator was synchronized to the grid on July 29, 2024. After achieving approximately 95 percent RTP on July 30, 2024, Unit 2 entered mode three that same day following an automatic trip due to a turbine trip (Event Notification number 57253). Unit 2 proceeded into mode six for a refueling outage (U2R26) while performing inspections and repairs to the main generator and remained in mode six through the remainder of the reporting 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 weather involving potential flash flooding and high winds on September 25 and 26, 2024. ==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 component cooling water system alignment with the 2B component cooling water pump aligned to the "B" train with the "C-S" component cooling water pump unavailable for planned maintenance on July 9, 2024 : (2) Unit 2 main steam isolation and safety valves on August 22, 2024 ===Complete Walkdown Sample (IP Section 03.02) (1 Sample)=== : (1) The inspectors evaluated system configurations during a complete walkdown of the Unit 1 turbine driven auxiliary feedwater system on September 12, 2024. ==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 2 containment annulus on August 20, 2024 : (2) Unit 2 lower containment on August 21, 2024 : (3) Auxiliary building compartment 714-01 (corridor 714 foot) on August 21, 2024 : (4) Unit 2 upper containment on September 12, 2024. : (5) Control Building elevation 706 on September 17, 2024 ===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 with a simulated fire in the Unit 1 auxiliary building 669 foot elevation on August 27, 2024 (OPDP-3-1). ===71111.08P - Inservice Inspection Activities (PWR) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation by reviewing the following activities in Unit 2 during the refueling outage from August 26 - 29, 2024. PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)=== {{IP sample|IP=IP 71111.08|count=1}} The inspectors verified that the following nondestructive examination and welding activities were performed appropriately: : (1) Ultrasonic Examination * Weld #1, flange to vessel weld, MRP-227 exam * 202 hot leg nozzle to shell weld, MRP - 227 exam * 22 hot leg nozzle to shell weld, MRP-227 exam Eddy Current Examination * Incore Flux Thimble Tubes PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection ===Activities (IP Section 03.02) (1 Sample)=== The inspectors verified that the licensee conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately: : (1) Bare Metal Visual of Penetrations 20, 27, and 35. 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) SQN-2-FCV-043-0023, RCS Hot Leg HDR Containment Isolation Valve, ASME Class 2 ==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 Unit 2 reactor start-up following a down power to address a turbine bearing issue on July 27, 2024. ===Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)=== : (1) The inspectors observed and evaluated a cycle 24-4 pilot crew simulator exam involving a steam generator tube rupture scenario on September 24, 2024 (Segment number OPL273E2441). ==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) Unit 1 control rod drive mechanism cooler issues associated with "A" control rod drive mechanism cooler leak and "B' control rod drive fan failure (CR1939435, CR1940005) : (2) Unit 1 and Unit 2 steam dump to condenser failures (CR 1955285, CR 1946817) ==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 1 "B" main feed pump high seal water differential pressure during a conservation operation alert during the week of July 1, 2024 (CR 1940617) : (2) Unit 2 "C-S" component cooling water pump maintenance outage during the week of July 8, 2024 : (3) Unit 2 turbine high vibrations and exciter ground troubleshooting and repairs during the week of July 15, 2024 (CR 1943971) ==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 1 control rod deviation (control bank A, rod H-10) alarm identified on June 24, 2024 (CR 1939857) : (2) Unit 1 leading edge flow meter issues associated with the plant computer system impacting the accuracy of indicated rated thermal power identified on July 2, 2024 (CR1941478) : (3) Unit 1 ice condenser 6A air handling unit blower not running identified on July 11, 2024 (CR 1943265) : (4) Unit 2 turbine driven auxiliary feedwater pump slowly spinning with trip and throttle valve shut due to leak-by identified on July 202024 (CR 1945067, CR 1945062) : (5) Unit 2 "A" motor driven auxiliary feedwater pump discharge check valve (2-VLV-003-0820) surveillance failure identified on August 12, 2024 (CR1950020) : (6) Unit 2 emergency core cooling containment sump suction valve to "A" residual heat removal pump (FCV-063-0072-A) failed to open during valve stroke identified on August 27, 2024 (CR 1953959) ==71111.20 - Refueling and Other Outage Activities== ===Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)=== : (1) The inspectors evaluated a Unit 2 forced outage for turbine vibrations on July 20, 2024 (CR1945095). : (2) The inspectors evaluated a Unit 1 forced outage to address main generator issues from August 23, 2024, through September 1, 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) "C" component cooling water pump maintenance outage during the week of July 8, 2024 (WO 123988231). : (2) Unit 2 transfer canal wafer valve testing following repairs on August 7, 2024 (WO ===123561121) : (3) Unit 1 "A" centrifugal charging pump auxiliary oil pump control switch calibration and testing on August 27, 2024 (WO 123762209) : (4) Unit 2 "A" charging pump motor replacement during the 2024 fall refueling outage, {{a|2R26}} ==2R26, in August 2024 (WO 117812692) (5)== Unit 2 loop two and loop four hot leg safety injection check valves (2-VLV-063-0559, 2-VLV-063-0558) replacements during the 2024 fall refueling outage, 2R26, in August 2024 (WO 122699707, WO 122699700) (6) "C" component cooling water pump testing following a motor replacement on September 21, 2024 (0-SI-SXP- 070-201.C; WO 117809394) Surveillance Testing (IP Section 03.01)=== {{IP sample|IP=IP 12356|count=2}} : (1) Unit 2 loss of offsite power with safety injection and emergency diesel generator test on August 9, 2024 (2-SI-OPS-082-026.B) : (2) Unit 1 pressurizer heater capacity test on September 11, 2024 (WO 124001066, 0-SI-OPS-068-297.0) ===Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)=== : (1) Unit 2 containment isolation valve local leak rate tests for the incore instrumentation room exhaust and purge isolation penetrations on August 20, 2024 (WO 123728783) ===Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)=== (1) "B" three mega-watt FLEX diesel generator full load two-hour surveillance performed on September 25, 2024 (WO124133451) ==71114.06 - Drill Evaluation== ===Additional Drill and/or Training Evolution (1 Sample)=== The inspectors evaluated: : (1) A licensed operator training scenario in the simulator associated with a steam generator tube rupture and emergency declaration that contributed to the drill and exercise performance indicator on September 24, ==RADIATION SAFETY== ==71124.06 - Radioactive Gaseous and Liquid Effluent Treatment== ===Walkdowns and Observations (IP Section 03.01) (5 Samples)=== The inspectors evaluated the following radioactive effluent systems during walkdowns: : (1) Auxiliary building exhaust gaseous monitoring system (0-RE-90-101) and associated system configuration and flow path. : (2) Liquid radwaste effluent monitoring system (0-RE-90-122) and associated system configuration and flow path. : (3) Auxiliary Building Gas Treatment System ventilation filtration system and associated system configuration and flow path. : (4) Auxiliary building exhaust routine weekly gaseous sampling. : (5) Cask Decontamination Collector Tank (CDCT) routine pre-release liquid sampling. ===Sampling and Analysis (IP Section 03.02) (3 Samples)=== Inspectors evaluated the following effluent samples, sampling processes and compensatory samples: : (1) Compensatory sampling for out-of-service (OOS) Essential Raw Cooling Water Liquid Effluent Monitors RM-90-133/140, work order (WO) 123277749. : (2) Compensatory sampling for OOS Liquid Radwaste Effluent Line Monitor RE-90-122, WO 123761595. : (3) Unit 1 upper containment non-routine sampling of noble gas and tritium in response to thermal power change. ===Dose Calculations (IP Section 03.03) (3 Samples)=== The inspectors evaluated the following dose calculations: : (1) CDCT liquid waste release permit 20241472.008.365.L : (2) Unit 1 containment vent gaseous waste release permit 20241308.027.419.G : (3) Unit 2 containment purge gaseous waste release permit 20241292.029.073.G ===Abnormal Discharges (IP Section 03.04) (1 Sample)=== The inspectors evaluated the following abnormal discharges: : (1) Unmonitored gaseous release due to damaged and deteriorated ductwork in the auxiliary building exhaust; CR 1931728. ==71124.07 - Radiological Environmental Monitoring Program== ===Environmental Monitoring Equipment and Sampling (IP Section 03.01) (1 Sample)=== : (1) The inspectors evaluated environmental monitoring equipment and observed collection of environmental samples. ===Radiological Environmental Monitoring Program (IP Section 03.02) (1 Sample)=== : (1) The inspectors evaluated the implementation of the licensees radiological environmental monitoring program. ===GPI Implementation (IP Section 03.03) (1 Sample)=== : (1) The inspectors evaluated the licensees implementation of the Groundwater Protection Initiative (GPI) program to identify incomplete or discontinued program elements. ==OTHER ACTIVITIES - BASELINE== ===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below: MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)=== {{IP sample|IP=IP 71151|count=2}} : (1) Unit 1 (July 1, 2023 through June 30, 2024) : (2) Unit 2 (July 1, 2023 through June 30, 2024) ===MS06: Emergency AC Power Systems (IP Section 02.05) (2 Samples)=== : (1) Unit 1 (July 1, 2023 through June 30, 2024) : (2) Unit 2 (July 1, 2023 through June 30, 2024) ===MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)=== : (1) Unit 1 (July 1, 2023 through June 30, 2024) : (2) Unit 2 (July 1, 2023 through June 30, 2024) ===MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)=== : (1) Unit 1 (July 1, 2023 through June 30, 2024) : (2) Unit 2 (July 1, 2023 through June 30, 2024) ===MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)=== : (1) Unit 1 (July 1, 2023 through June 30, 2024) : (2) Unit 2 (July 1, 2023 through June 30, 2024) ===MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)=== : (1) Unit 1 (July 1, 2023 through June 30, 2024) : (2) Unit 2 (July 1, 2023 through June 30, 2024) PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample) : (1) March 1, 2023 through July 17, 2024 ===71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)=== {{IP sample|IP=IP 71152|count=1}} The inspectors reviewed the licensees implementation of its corrective action program related to the following issues: : (1) Unit 2 main steam stop valve surveillance requirements and plant conditions required identified on August 5, 2024 (CR 1948523) ===71153 - Follow Up of Events and Notices of Enforcement Discretion Event Follow up (IP Section 03.01)=== {{IP sample|IP=IP 71153|count=4}} : (1) The inspectors evaluated a Unit 2 rapid load reduction to 45 percent rated thermal power due to an exciter ground and high turbine vibrations and licensee's response on July 15, 2024 (CR 1943971). : (2) The inspectors evaluated a Unit 2 reactor trip due to an electrical trouble turbine trip and licensees response on July 30, 2024 (EN #57253; CR 1947208). : (3) The inspectors evaluated a Unit 1 reactor trip due to an electrical trouble turbine trip and licensees response on August 23, 2024 (EN #57285; CR 1953128). : (4) The inspectors evaluated a Unit 1 reactor trip from 100 percent RTP due to an electrical trouble turbine trip and licensees response on September 29, 2024 (EN #57351; CR 1962283). ==INSPECTION RESULTS== RCS leak caused by backfill of a RHR system flow transmitter sensing line Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000327/2024003-01 Open/Closed [H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated non-cited violation of Technical Specification 5.4.1.a was identified when the licensee failed to implement written procedures covering activities recommended in Regulatory Guide 1.33, Revision, 2, Appendix A. Specifically, the licensee failed to establish and implement a procedure for backfill of residual heat removal (RHR) system flow transmitter sensing lines that was appropriate for plant conditions. This action resulted in lifting the A RHR discharge header relief valve and discharging reactor coolant system (RCS) water to the pressurizer relief tank (PRT). =====Description:===== On April 24, 2024, Sequoyah Unit 1 was in Mode 5 on shutdown cooling with B train of RHR in operation and A train in standby. The RCS loops were filled, and two reactor coolant pumps (RCPs) were operating. Maintenance technicians were performing 1-PI-IFV-063-091.0, Backfilling sensing lines for SIS flow to the RCS cold leg Loop 2 and 3 flow transmitters 1-FT-63-91C, 1-FT-63-91B and 1-FT 63-91A, to ensure there were no voids in the sensing lines for the A RHR discharge header flow transmitters prior to the flow transmitters being required for operability in Mode 3. The backfilling of sensing lines is a frequently performed activity. The technicians used portable back fill source with a positive displacement 8000 psig pump to inject water into the sense lines. The procedure directs technicians to verify specific volumes (2.5 gallons) injected. This backfilling activity injected water into a portion of the residual heat removal (RHR) system that was hydraulically solid and bounded by two closed motor operated valves (MOV's) and an in-line 8-inch check valve. Injecting fluid into the system in this configuration resulted in seating the check valve closed and raising pressure from 350 psig to the 1A RHR discharge relief valve (1-VLV-063-626)setpoint of 600 psig. 1-VLV-063-626 lifted as designed and discharged to the PRT. Header pressure lowered when the relief valve opened. The relief failed to reseat, and system leakage continued through the open relief valve when the previously closed check valve opened due pressure in the 'A' RHR discharge header decreasing below RCS system pressure. RCS leak rate stabilized at 85 gpm. At 02:14, Operations personnel entered AOP-R.05, RCS Leak, due to decreasing pressurizer level. At 03:53, operators attempted to isolate 1-VLV-063-626 by isolating A RHR header. This resulted in lower leak rate into the PRT but did not stop the leak. At 04:12, operators stopped the running RCPs and commenced RCS depressurization. At 04:35, with RCS pressure 212 psig, 1-VLV-063-626 reseated and the leak stopped. Approximately 10,000 gallons of RCS water had been discharged into the PRT through the RHR discharge header relief valve. RCS volume is approximately 90,000 gallons and was maintained approximately constant with the makeup to the RCS during the transient. During the transient, insurges into the pressurizer resulted a maximum pressurizer cooldown rate of 223 degrees/hour and a maximum pressurizer heat up rate of 111 degrees/hour. These rates exceeded the pressurizer cooldown limit of 200 degrees/hour and heat-up limit of 100 degrees/hour identified in Technical Requirement Manual (TRM) Section 8.4.2 Pressurizer Temperature Limits. Reactor vessel heat up and cooldown limits were not challenged. Corrective Actions: The licensee stopped all backfill operations until extent of condition for the use of positive displacement pumps to inject into hydraulically isolated systems was determined. Pressurizer pressure was maintained below 500 psig in accordance with TRM 8.4.2 Contingency Measure B.2 until an engineering evaluation to determine the effects of the out of limit condition on the fracture toughness properties of the pressurizer was performed in accordance with TRM 8.4.2 Contingency Measure A.2. The lifted relief valve (1-VLV-063-626)was replaced. Following replacement of 1-VLV-063-626, normal Mode 5 RCS and RHR conditions were re-established. Corrective Action References: CR 1926614, CR 1926802, CR 1926579 =====Performance Assessment:===== Performance Deficiency: The inspectors determined that the failure to establish and implement a maintenance procedure for backfill of flow transmitter sensing lines in the RHR system that was appropriate to the circumstances was reasonably within the licensees ability to foresee and correct and should have been prevented. Specifically, Maintenance Instruction 1-PI-IFV-063-091.0, Backfilling sensing lines for SIS flow to the RCS cold leg Loop 2 and 3 flow transmitters 1-FT-63-91C, 1-FT-63-91B and 1-FT 63-91A, did not provide appropriate guidance on backfilling A RHR flow transmitter sensing lines with the RHR system in single B train operation and A train in standby. In this configuration, the procedure resulted in technicians injecting water into a hydraulicly isolated section of the A RHR discharge header. This action directly resulted in the opening of the A RHR discharge header relief valve and initiated a leak of the RCS to the PRT. Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality 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, maintenance instructions to backfill sensing lines for the A RHR pump discharge header flow transmitter while the header was in standby which was not appropriate because the system was hydraulically isolated and solid. This resulted in exceeding the lift setpoint of the A RHR discharge header relief valve and initiated draining of the RCS to the PRT. Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Exhibit 2 of 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, for Loss of Inventory (LOI) initiators. The inspectors concluded the finding screened to a Phase 2 evaluation because it represented a Loss of Inventory event that was not self-limiting. A regional senior reactor analyst conducted a Phase 2 evaluation in accordance with IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown. The finding was determined to represent a late outage time window LOI precursor that occurred in Plant Operational State (POS) 1. Because RCS temperature was below 200degF, Attachment 2 directs the use of the Loss of Level Control event tree and Worksheet 1 which were used to assess the Phase 2 significance of the finding. The screening concluded that the finding had the potential to result in conditional core damage probability sequences that were greater than E-06 and that further analysis would need to be performed as a Detailed Risk Evaluation (DRE). A headquarters senior reliability and risk analyst performed a DRE. A LOI event tree and supporting fault trees were created using SAPHIRE 8.2.9 and Sequoyah SPAR model 8.82 with the likelihood of the LOI initiator set to 1.0. NRC Shutdown Calculator, Version 4 was used to estimate the time to loss of RHR, time to RCS boil, and time to top of active fuel using the observed RHR leak rate. The event sequence yielded an estimated change in Core Damage Frequency (delta-CDF) of 8.6E-07/year which represented a finding of very low safety significance (Green). The dominant sequences included the LOI initiator followed by the failure of operator actions to adjust charging to maintain level, to isolate the leak, and to recover RHR. The risk of the finding was mitigated by the availability of mitigating equipment, the length of time prior to loss of cooling function, and the availability of operator cues during the event sequence. 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, backfilling of sensing lines is a frequently performed evolution and the senior reactor operator (SRO)authorizing the work did not challenge the scope of work and incorrectly thought the evolution to be non-intrusive to the RHR system. The technician assumed the SRO understood the back fill evolution and that the RHR system configuration would support the evolution. =====Enforcement:===== Violation: Technical Specification 5.4.1.a, requires, in part, that procedures shall be established, implemented, and maintained covering the applicable procedures recommended by Regulatory Guide 1.33, Appendix A, Revision 2. Section 9.a of Appendix A of Regulatory Guide 1.33, Revision 2, states, in part, that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, on April 24, 2024, the licensee failed to adequately establish and implement maintenance instructions related to backfilling flow transmitter sensing lines in the safety related RHR system. Specifically, technicians performing Maintenance Instruction 1-PI-IFV-063-091.0, Backfilling sensing lines for SIS flow to the RCS cold leg Loop 2 and 3 flow transmitters 1-FT-63-91C, 1-FT-63-91B and 1-FT 63-91A, over pressurized the A RHR system discharge header. This action directly resulted in the opening of the A RHR discharge header relief valve and draining approximately 10,000 gallons of RCS water to the PRT. 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 October 15, 2024, the inspectors presented the integrated inspection results to Tom Marshall, Site Vice President, and other members of the licensee staff. * On July 18, 2024, the inspectors presented the Radiation Protection inspection results to T. Marshall and other members of the licensee staff. * On August 29, 2024, the inspectors presented the Inservice Inspection inspection results to A. Jenkins, Director of Operations and other members of the licensee staff. =DOCUMENTS REVIEWED= Inspection Procedure Type Designation Description or Title Revision or Date 71111.08P Corrective Action Documents Resulting from Inspection CR 1953683 CR 1953696 CR 1954060 CR 1954061 CR 1954062 CR 1954065 CR 1954066 CR 1931728 Corrective Action Documents CR 1944627 71124.06 Radiation Surveys Release permit 231164.023.002.G Auxiliary building exhaust gas permit post-release data 09/20/2023 CR 1831595 71124.07 Corrective Action Documents CR 1944600 }}

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