ML25129A092
| ML25129A092 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/13/2025 |
| From: | Robert Williams NRC/RGN-II/DORS |
| To: | Snider S Duke Energy Carolinas |
| References | |
| IR 2025001 | |
| Download: ML25129A092 (1) | |
Text
EAF-RII-2025-0058 Steven Snider Site Vice President Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752
SUBJECT:
OCONEE NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000269/2025001 AND 05000270/2025001 AND 05000287/2025001
Dear Steven Snider:
On March 31, 2025, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Oconee Nuclear Station. On May 1, 2025, the NRC inspectors discussed the results of this inspection with you 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 Oconee Nuclear Station.
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 May 12, 2025
S. Snider 2
Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Robert E. Williams, Jr., Chief Projects Branch 1 Division of Operating Reactor Safety Docket Nos. 05000269 and 05000270 and 05000287 License Nos. DPR-38 and DPR-47 and DPR-55
Enclosure:
As stated cc w/ encl: Distribution via LISTSERV Signed by Williams, Robert on 05/12/25
SUNSI Review x
Non-Sensitive
Sensitive x
Publicly Available
Non-Publicly Available OFFICE RII: DORS RII: DORS RII: ACES OE: EB RII: DORS NAME N. Smalley J. Seat M. Kowal J. Ortiz Rivera J. Peralta/D. Bradley R. Williams DATE 05/09/2025 05/09/2025 05/09/2025 05/12/2025 05/12/2025
Enclosure U.S. NUCLEAR REGULATORY COMMISSION Inspection Report Docket Numbers:
05000269, 05000270 and 05000287 License Numbers:
DPR-38, DPR-47 and DPR-55 Report Numbers:
05000269/2025001, 05000270/2025001 and 05000287/2025001 Enterprise Identifier:
I-2025-001-0028 Licensee:
Duke Energy Carolinas, LLC Facility:
Oconee Nuclear Station Location:
Seneca, South Carolina Inspection Dates:
January 01, 2025 to March 31, 2025 Inspectors:
D. Dang, Resident Inspector M. Meeks, Senior Operations Engineer E. Robinson, Resident Inspector C. Safouri, Senior Resident Inspector K. Schaaf, Operations Engineer N. Smalley, Senior Resident Inspector D. Willis, Team Leader Approved By:
Robert E. Williams, Jr., Chief Projects Branch 1 Division of Operating Reactor Safety
2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Oconee Nuclear Station, 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 Implement Post Maintenance Testing Procedures Appropriate to the Circumstances for the Standby Shutdown Facility Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000269/2025001-01 Open/Closed EAF-RII-2025-0058 None (NPP) 71111.12 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, was identified when the licensee failed to implement post maintenance testing (PMT) procedures appropriate to the circumstances, which resulted in inoperability of the standby shutdown facility (SSF) for Unit 1. Following replacement of a pressurizer (PZR) heater control switch in July 2022, PMT procedures failed to identify a degraded condition which prevented PZR heaters from operating when required. This resulted in a violation of technical specification (TS) 3.10.1, Standby Shutdown Facility (SSF), and TS 3.0.4, Limiting Condition for Operation (LCO) Applicability.
Additional Tracking Items Type Issue Number Title Report Section Status LER 05000269/2024-001-00 LER 2024-001-00 for Oconee Nuclear Station, Unit 1, Standby Shutdown Facility (SSF) Pressurizer Level Switch Configuration Caused by Legacy Procedure Deficiency Resulted in Condition Prohibited by Technical Specifications 71153 Closed LER 05000270/2024-001-00 LER 2024-001-00 for Oconee Nuclear Station, Unit 2, Common Cause Inoperability of Both Trains of Control Room Ventilation System Outside Air Booster Fans due to Supply Breaker Wiring Deficiency Resulted in a Condition that Could Have Prevented Fulfillment.
71153 Closed
3 PLANT STATUS Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection period.
Unit 2 operated at or near 100 percent RTP for the entire inspection period.
Unit 3 operated at or near 100 percent RTP for the entire inspection period.
INSPECTION SCOPES Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY 71111.01 - Adverse Weather Protection 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 winter weather on January 10, 2025.
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) 2A reactor building spray (RBS) train while 2B RBS out of service (OOS) for maintenance on February 11, 2025 (2)
Keowee Hydro Unit (KHU) #2 underground emergency power path with KHU #1 overhead OOS on February 18, 2025 (3) 3A low pressure injection (LPI) train while 3B LPI was out of service for maintenance on March 5, 2025 71111.05 - Fire Protection Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
4 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)
Fire zone 108: Unit 1 east penetration room on January 14, 2025 (2)
Fire zone 92: Unit 2 equipment room on February 10, 2025 (3)
Fire zone 101: Unit 3 cable room on February 10, 2025 (4)
Fire zone 90: Unit 2 auxiliary building 300 level hallway on February 25, 2025 (5)
Fire zone 34: Unit 1 4160V switchgear on March 4, 2025 Fire Brigade Drill Performance Sample (IP Section 03.02) (2 Samples)
(1)
The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on March 9, 2025.
(2)
The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on March 21, 2025.
71111.11B - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Requalification Program (IP Section 03.04) 71111.11B - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Requalification Program (IP Section 03.04)
An inspection was performed to assess the effectiveness of the facility licensee in implementing requalification requirements identified in 10 CFR Part 55, Operators Licenses. Each of the following inspection activities was conducted in accordance with IP 71111.11, Licensed Operator Requalification Program and Licensed Operator Performance.
Biennial Requalification Written Examinations The inspectors evaluated the quality of the licensed operator biennial requalification written examination administered on March 2025.
Annual Requalification Operating Tests The inspectors evaluated the adequacy of the facility licensees annual requalification operating test.
Administration of an Annual Requalification Operating Test The inspectors evaluated the effectiveness of the facility licensee in administering requalification operating tests required by 10 CFR 55.59(a)(2) and that the facility licensee is effectively evaluating their licensed operators for mastery of training objectives.
Requalification Examination Security The inspectors evaluated the ability of the facility licensee to safeguard examination
5 material, such that the examination is not compromised.
Remedial Training and Re-examinations The inspectors evaluated the effectiveness of remedial training conducted by the licensee, and reviewed the adequacy of re-examinations for licensed operators who did not pass a required requalification examination.
Operator License Conditions The inspectors evaluated the licensees program for ensuring that licensed operators meet the conditions of their licenses.
Control Room Simulator The inspectors evaluated the adequacy of the facility licensees control room simulator in modeling the actual plant, and for meeting the requirements contained in 10 CFR 55.46.
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 control rod movement testing, on February 24, 2025.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed and evaluated a simulator operator training exam in accordance with ASE-29 on March 4, 2025.
71111.12 - Maintenance Effectiveness Maintenance Effectiveness (IP Section 03.01) (3 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)
Nuclear condition report (NCR) 2534840, SSF PZR heater control circuit current switch incorrect configuration on November 8, 2024 (2)
NCR 02540577, review of grounds discovery and repair on unit 125V DC buses on January 14, 2025 and January 27, 2025 (3)
NCR 2546947, maintenance and restoration of train A chiller following trip during refrigerant evaluation testing on March 10, 2025 71111.13 - Maintenance Risk Assessments and Emergent Work Control Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)
6 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 green risk following loss of load center 1XL, on November 2, 2024 (2)
Unit 1 elevated green risk due to 1C LPI motor test work, on February 5, 2025 (3)
Unit 3 elevated green risk due to preventive maintenance on SSF auxiliary service water (ASW) emergency Unit 3 steam generator (SG) supply valves, 3CCW-268 and 287, on February 11, 2025 (4)
Unit 2 elevated green risk due to maintenance on LPI, on the week of February 24, 2025 (5)
Unit 3 elevated green risk due to maintenance on LPI, on March 19, 2025 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)
NCR 2533121, stop check valve 1HP-254 stuck open (2)
NCR 2541430, Unit 2 turbine driven emergency feedwater (TDEFW) pump test flow below required acceptance criteria on January 12, 2025 (3)
NCR 2541833, standby SSF battery did not meet minimum capacity during performance testing on January 26, 2024 (4)
NCR 2544934, 1C high pressure injection (HPI) pump motor cooler testing results (5)
NCR 2545420, KHU-1 and emergency power overhead path following transformer lockout restoration on February 26, 2025 (6)
NCR 2546711, nitrogen supply pressure for 1MS-93, TDEFW pump turbine steam admission valve, out of band on March 9, 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) (5 Samples)
(1)
PT/1/A/0251/001, "Low Pressure Service Water Pump Test," following motor replacement, on January 13, 2025 (2)
PT/1/A/0204/007, "1B Reactor Building Spray Pump Test," and train inspection following preventive maintenance, on January 15, 2025 (3)
PT/1/A/0202/011, "1C High Pressure Injection Pump Test," following preventive maintenance, on February 19, 2025 (4)
PT/0/A/0620/009, "Keowee Hydro Operation," following governor speed switch replacement, on February 21, 2025 (5)
PT/1/A/0600/012, "Unit 1 Turbine Driven Emergency Feedwater (TDEFW) Pump Test," following preventive maintenance, on February 27, 2025 Surveillance Testing (IP Section 03.01) (3 Samples)
7 (1)
PT/0/A/0620/016, "Keowee Hydro Emergency Start Test," on January 8, 2025 (2)
PT/0/A/0600/021, "Standby Shutdown Facility Diesel Generator Run," on January 14, 2025 (3)
PT/3/A/0204/007, "3B Reactor Building Spray Pump Test," on March 14, 2025 Inservice Testing (IST) (IP Section 03.01) (2 Samples)
(1)
PT/2/A/0600/13B, comprehensive test on 2B motor driven emergency feedwater pump, on January 27, 2025 (2)
PT/0/A/0400/005, "SSF Auxiliary Service Water Pump Test," on March 13, 2025 Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)
(1)
PT/1/A/0600/010, increased unidentified reactor coolant leakage on Unit 1, during the week of February 21, 2025 Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
(1)
FLEX testing, on the week of January 6, 2025 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) (3 Samples)
(1)
Unit 1 (January 1 through December 31, 2024)
(2)
Unit 2 (January 1 through December 31, 2024)
(3)
Unit 3 (January 1 through December 31, 2024)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (3 Samples)
(1)
Unit 1 (January 1 through December 31, 2024)
(2)
Unit 2 (January 1 through December 31, 2024)
(3)
Unit 3 (January 1 through December 31, 2024)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (3 Samples)
(1)
Unit 1 (January 1 through December 31, 2024)
(2)
Unit 2 (January 1 through December 31, 2024)
(3)
Unit 3 (January 1 through December 31, 2024)
MS07: High Pressure Injection Systems (IP Section 02.06) (3 Samples)
(1)
Unit 1 (January 1 through December 31, 2024)
(2)
Unit 2 (January 1 through December 31, 2024)
(3)
Unit 3 (January 1 through December 31, 2024)
8 71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03) (1 Partial)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1)
(Partial)
NRC inspectors assessed the Safety Conscious Work Environment (SCWE) within the Nuclear Supply Chain (NSC) department at Oconee Nuclear Station. The inspectors conducted interviews with all available staff in the department to assess the licensees environment for raising concerns, and to determine whether challenges existed to maintaining a SCWE.
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 05000269/2024-001-00, Standby Shutdown Facility (SSF) Pressurizer Level Switch Configuration Caused by Legacy Procedure Deficiency Resulted in Condition Prohibited by Technical Specifications (ADAMs Accession No. ML24354A337). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71111.12. This LER is Closed.
(2)
LER 05000270/2024-001-00, Common Cause Inoperability of Both Trains of Control Room Ventilation System Outside Air Booster Fans due to Supply Breaker Wiring Deficiency Resulted in a Condition that Could Have Prevented Fulfillment of a Safety Function (ADAMs Accession No. ML24354A312). The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified. The inspectors did not identify a violation of NRC requirements. This LER is Closed.
INSPECTION RESULTS Failure to Implement Post Maintenance Testing Procedures Appropriate to the Circumstances for the Standby Shutdown Facility Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000269/2025001-01 Open/Closed EAF-RII-1025-0058 None (NPP) 71111.12 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, was identified when the licensee failed to implement PMT procedures appropriate to the circumstances, which resulted in inoperability of the SSF for Unit 1. Following replacement of a PZR heater control switch in July 2022, PMT procedures failed to identify a degraded condition which prevented PZR heaters from operating when required. This resulted in a violation of TS 3.10.1, Standby Shutdown Facility (SSF), and TS 3.0.4, Limiting Condition for Operation (LCO) Applicability.
9
==
Description:==
The SSF is designed as a standby, manually activated system to provide additional defense-in-depth protection, by serving as a backup to existing safety systems.
The SSF is provided as an alternate means to achieve and maintain the reactor in Mode 3 with reactor coolant system (RCS) temperature greater than or equal to 525F following certain fire, flooding, security, and station blackout (SBO) events. This is accomplished by re-establishing and maintaining reactor coolant pump (RCP) seal cooling, assuring natural circulation and core cooling by maintaining the RCS filled to a sufficient level in the pressurizer, while maintaining sufficient secondary side cooling water, and maintaining the reactor subcritical. The main components of the SSF are the SSF auxiliary service water (ASW) system, SSF portable pumping system, SSF reactor coolant (RC) makeup system, SSF power system, and SSF instrumentation.
The SSF ASW system is used to provide adequate cooling to maintain single phase RCS natural circulation flow in Mode 3 with an average RCS temperature = 525F, unless the initiating event causes the unit to be driven to a lower temperature. In order to maintain single phase flow, an adequate number of Bank 2 Group B and Group C PZR heaters must be operable. These heaters are needed to compensate for ambient heat loss from the PZR. As long as the temperature in the PZR is maintained, RCS pressure will also be maintained. This will preclude hot leg voiding and ensure adequate natural circulation cooling. Since the PZR heaters powered from the SSF during an SSF event do not have their own TS action statement, the SSF ASW system is declared inoperable when those PZR heaters are non-functional. The resulting inoperability of the SSF ASW system does not render other SSF systems inoperable.
On November 7, 2024, Unit 1 was in Mode 5 for a planned refueling outage when a Unit 1 PZR heater group did not turn on during the routine performance of a power transfer test.
During this test, control of Bank 2 Group B and Group C PZR heaters is transferred from the main control room to the SSF control room and functionality is verified. When Bank 2 Group B PZR heaters did not turn on, troubleshooting revealed an issue with the 1RC-IS-0072 current switch (level switch) in the heater control logic circuit. Upon inspection, the configurable jumpers on the level switch did not match the required configuration. This resulted in a condition in which a downstream relay from the level switch, PZR heater permissive relay (GD), would only close in, and therefore allow heaters to be energized, when PZR level was below the low level setpoint of 85 inches. This is the opposite logic of what was desired based on the functionality of the GD permissive relay.
On July 19, 2022, the licensee replaced the Unit 1 SSF PZR level switch with an approved acceptable substitute with the same fit and function but better power supply. However, during the replacement activity, maintenance technicians noted differences in the model numbers between the old and new level switches. Additionally, the work order instructions and procedures were then noted to include vague and confusing steps that required the technicians to interpret drawings with the help of technical support to configure the jumpers.
The PMT calibration procedure IP/0/A/0370/002 C, Standby Shutdown Facility RC System Pressurizer Level and Pressurizer Pressure, was also used to set the jumpers to match the contact status via the calibration data sheet in the procedure. Work was completed and signed off by a quality control (QC) representative. The PMT calibration procedure was performed, and acceptance criteria were met. The licensee later determined that the card was installed with the improper configuration at this time.
The licensees causal investigation determined that a legacy error existed in the calibration data sheet for the level switch in procedure IP/0/A/0370/002 C. This procedure was written
10 prior to 1999 and had not been used for this application before. No other units level switch had been replaced using this procedure. Acceptance criteria for the procedure utilized in IP/0/A/0370/002 C for the PMT listed a certain contact as close on PZR low level, which is the opposite logic that is needed functionally (open). The acceptance criteria were incorrect as written, but were met during the PMT, and therefore were documented as satisfactory.
Due to the plant configuration in July 2022 (Unit 1 was in Mode 1), the designated PMT did not identify the incorrectly configured level switch, and the system was returned to service.
The power transfer test conducted in November 2024, which is only performed during unit outages, energizes the downstream PZR permissive relay GD during the procedure and would have identified the incorrectly configured level switch. The function of controlling PZR heaters from the SSF is not tested by any other routine TS surveillance. This issue does not meet the criteria for an old design issue as described by the NRC Enforcement Policy.
Corrective Actions: The licensee corrected the jumper configuration for the affected level switch on Unit 1, completed an extent of condition review of Units 2 and 3, and revised the post maintenance test procedure to correct the acceptance criteria and improve procedure directions.
Corrective Action References: 253480, 2534824 Performance Assessment:
Performance Deficiency: The licensees failure to implement post maintenance testing using procedures appropriate to the circumstances following the replacement of the Unit 1 SSF PZR level switch was a performance deficiency (PD). Specifically, in July 2022, following replacement of the Unit 1 SSF PZR level switch, post maintenance testing did not identify that the level switch was configured such that the PZR heaters controlled by the SSF would not energize until PZR level was below 85 inches, instead of preventing operation at or below 85 inches.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control 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 PD resulted in the unavailability of normal pressurizer heater function and control from the SSF for over two years. The PD is also similar to example 5.b in IMC 0612 Appendix E, Examples of Minor Issues.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The condition was screened using IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power," and IMC 0609 Appendix F, Fire Protection Significance Determination Process. IMC 0609 Appendix A, Exhibit 2, question A2 and/or A3, can both be answered YES since the equipment was unable to perform its Probabilistic Risk Assessment (PRA) function for greater than the TS allowed outage time. Therefore, this issue screened to requiring performance of a detailed risk assessment. From IMC 0609 Appendix F, questions 1.4.7 A and B were answered NO, question 1.4.7 C was answered YES, and question 1.5.1 was answered NO since the condition was not modeled. Therefore, a Phase II evaluation was required.
A regional senior reactor analyst (SRA) conducted a detailed risk assessment for this condition. The SRA identified that neither the NRCs Standardized Plant Analysis Risk
11 (SPAR) model nor Dukes Computer Aided Faulty Tree Analysis (CAFTA) model, appropriately modelled the standby shutdown facility powered pressurizer heaters (SSF PZR HTRs). The SSF PZR HTRs were required in order to overcome ambient losses from the pressurizer and maintain subcooling margin in the RCS, ensuring that single phase flow was maintained. When SSF PZR HTRs were unavailable, the SSF auxiliary service water system was considered inoperable per the plants technical specification basis. However, due to plant modifications such as installation of the protected service water system and a reconfiguration of the SSF RCS letdown line, the condition could no longer be accurately modelled using loss of SSF auxiliary service water as a surrogate and no other modeling tools were available.
Due to this fact the SRA used NRC Inspection Manual Chapter 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, to perform this risk assessment. A Planning Significance and Enforcement Review Panel (SERP) was conducted on February 28, 2025, to approve this approach. The dominant accident sequence was a large fire in the turbine building resulting in loss of onsite and emergency power (station blackout) and failure of the protected service water system. The full risk assessment can be found in Attachment A of this report. Plant risk was determined to be of very low safety significance (Green).
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Licensee procedure IP/0/A/0370/002 C, Standby Shutdown Facility RC System Pressurizer Level and Pressurizer Pressure, was used to replace and test the Unit 1 SSF PZR level switch, a safety-related component that supports operability of the SSF ASW system.
Oconee Technical Specification LCO 3.10.1 requires, in part, that the SSF Instrumentation and the SSF Auxiliary Service Water System shall be OPERABLE in Modes 1, 2, and 3. TS 3.10.1, Condition A, requires the SSF ASW system to be restored to OPERABLE within 7 days. If Condition A is not met for reasons other than maintenance, Condition G requires the plant must be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />.
Oconee Technical Specification 3.0.4 requires, in part, "When an LCO is not met, entry into a mode or other specified condition in the applicability shall only be made: a. When the associated actions to be entered permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time.
Contrary to the above, on July 19, 2022, the licensee failed to prescribe an activity affecting quality by documented instructions or procedures appropriate to the circumstances.
Specifically, the post maintenance testing procedure for replacing the Unit 1 SSF PZR level switch, IP/0/A/0370/002 C, did not contain appropriate identify that a level switch in the pressurizer heater control logic circuit was configured incorrectly during maintenanceance.
As a result, the SSF ASW system was rendered inoperable on Unit 1 from July 19, 2022, until November 8, 2024, while in Modes 1, 2, and 3. With the SSF ASW system in inoperable status, the licensee failed to perform the required actions specified in TS 3.10.1, Conditions A
12 and G, within the allowable completion times, and meet the mode entry requirements in TS 3.0.4 when Unit 1 entered Mode 3 on November 21, 2022, following a planned refueling outage.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Assessment 71152A SCWE Assessment of Oconee Nuclear Station Nuclear Supply Chain Department:
Based on interviews with Oconee Nuclear Supply Chain staff/managers and reviews of the latest safety culture survey results, the team did not identify any concerns with the safety-conscious work environment. The majority of employees interviewed appeared willing to raise nuclear safety concerns through multiple avenues. Most interviewees were aware of the licensee's employee concerns program and stated they would use the program, if necessary.
EXIT MEETINGS AND DEBRIEFS The inspectors verified that no proprietary information was retained or documented in this report.
On May 1, 2025, the inspectors presented the integrated inspection results to Steven Snider and other members of the licensee staff.
On March 20, 2025, the inspectors presented the operator requalification inspection results to Steven Snider and other members of the licensee staff.
13 DOCUMENTS REVIEWED Inspection Procedure Type Designation Description or Title Revision or Date 71111.01 Corrective Action Documents 2500888 71111.01 Miscellaneous Risk Profiles for Units 1, 2, and 3 for the week of January 6, 2025 71111.01 Procedures AD-OP-ONS-0120 Severe Weather Preparations 001 71111.04 Drawings OFD-102A-2.1 Flow Diagram of Low Pressure Injection System Borated Water Supply and LPI Pump Suction 63 71111.04 Drawings OFD-103A-2.1 Flow Diagram of Reactor Building Spray System (BS) 26 71111.04 Miscellaneous Clearance PRT-0-25-K1 OVH OOS-0048 71111.04 Miscellaneous OSS-0254.00 1034 (MECH) Design Basis for the Reactor Bldg Spray System 29 71111.04 Miscellaneous OSS-0254.00 2005 (ELECT) Keowee Emergency Power Design Basis Document 37 71111.04 Procedures AD-OP-ALL-0201 Protected Equipment 10 71111.04 Work Orders 20693858 71111.05 Calculations OSC-9314 NFPA 805 Transition Risk-Informed Performance-Based Fire Risk Evaluation 006 71111.05 Corrective Action Documents 02539605, 02538783 71111.05 Fire Plans CSD-ONS-PFP-1AB-0796 Pre-Fire Plan for U1 Auxiliary Building Elevation 796 001 71111.05 Fire Plans CSD-ONS-PFP-1AB-0809 Pre-Fire Plan For U1 Auxiliary Building Elevation 809 0
71111.05 Fire Plans CSD-ONS-PFP-1TB-0796 Pre-Fire Plan for U1 Turbine Building Elevation 796 0
71111.05 Fire Plans CSD-ONS-PFP-2AB-0796 Pre-Fire Plan for U2 Auxiliary Building Elevation 796 002 71111.05 Fire Plans CSD-ONS-PFP-2TB-0775 Pre-Fire Plan for U2 Turbine Building Elevation 775 1
71111.05 Fire Plans CSD-ONS-PFP-3AB-0809 Pre-Fire Plan for U3 Auxiliary Building Elevation 809 0
71111.05 Miscellaneous O-0310-FZ-028 Turbine Building - Unit 1 Fire Protection Plan Fire Area and Fire Zone Boundaries Plan at Mezzanine EL 796+6 2
14 Inspection Procedure Type Designation Description or Title Revision or Date 71111.05 Miscellaneous O-0310-K-008 Auxiliary Building & Reactor Building - Unit 2 Fire Protection Plan & Fire Barrier, Flood & Pressure Boundaries Plan at EL 796+6
& EL 797+6 26 71111.05 Miscellaneous O-0310-K-012 Auxiliary Building - Unit 3 Fire Protection Plan & Fire Barrier, Flood & Pressure Boundaries Plan at EL 809+3 16 71111.05 Miscellaneous O-0310-L-002 Turbine Building - Unit 2 Fire Protection Plan and Fire Barriers, Flood, and Pressure Boundaries Plan at EL 775+0 12 71111.05 Miscellaneous O-0310-L-004 Turbine Building - Unit 1 Fire Protection Plan and Fire Barrier, Flood, and Pressure Boundaries Plan at Mezzanine EL 796+6 14 71111.05 Procedures AD-OP-ALL-0207 Fire Brigade Administrative Controls 007 71111.05 Procedures AP/0/A/1700/025 Standby Shutdown Facility Emergency Operating Procedure 070 71111.05 Procedures AP/0/A/1700/0403 Fire Brigade Response Procedure 012 71111.05 Procedures AP/2/A/1700/050 Challenging Plant Fire 006 71111.05 Work Orders 20280757 71111.11Q Miscellaneous ASE-29 Simulator Exercise Guide 0
71111.11Q Miscellaneous CSD-EP-ONS-0101-02 Oconee Nuclear Station Classification of Emergency 004 71111.11Q Procedures AD-OP-ALL-0103 Standards for Operations Continuous Performance Improvement 011 71111.11Q Procedures PT/2/A/0600/015 Control Rod Movement 032 71111.12 Corrective Action Documents 2534840, 2540577, 2534607, 2546947, 2546757, 2547209 71111.12 Drawings OEE-149-01 Pressurizer Heaters Arrangement & Legend 26 71111.12 Drawings OEE-149-12 Elementary Diagram SSF Press. HTR Group C Bank 2 8
71111.12 Drawings OEE-149-8 Elementary Diagram SSF Press. HTR Group B Bank 2 27 71111.12 Drawings OEE-163-16B Elementary Diagram Standby Shutdown Facility Control Transfer 4
71111.12 Drawings OEE-163-18 Elementary Diagram SSF 17
15 Inspection Procedure Type Designation Description or Title Revision or Date Transducer Power and Metering 71111.12 Drawings OM 201.0009.001 Unit 1 Pressurizer General Arrangement D17 71111.12 Miscellaneous OSC-3144 Pressurizer Heat Losses 21 71111.12 Miscellaneous OSS-0254.00 1033 (MECH) Design Basis Specification for Reactor Coolant System 058 71111.12 Procedures AD-EG-ALL-1103 Procurement Engineering Products 7
71111.12 Procedures AD-EG-ALL-1137 Engineering Change Product Selection 13 71111.12 Procedures AD-EG-ALL-1155 Post Modification Testing 008 71111.12 Procedures AD-EG-ALL-1311 Failure Investigation Process (FIP) 3 71111.12 Procedures AD-MN-ALL-1000 Conduct of Maintenance 25 71111.12 Procedures AP/0/A/1700/025 Standby Shutdown Facility Emergency Operating Procedure 070 71111.12 Procedures IP/0/A/0370/002 C
Standby Shutdown Facility RC System Pressurizer Level and Pressurizer Pressure 075 71111.12 Procedures IP/0/B/0200/037 C
Pressurizer Ambient Heat Loss Test 012 71111.12 Procedures OP/0/A/1107/008 Isolation of DC Systems Between Units 016 71111.12 Procedures PT/1/A/0600/024 SSF Comprehensive Control Transfer Verification 024 71111.12 Work Orders 20281743, 20705003, 20421049 71111.13 Corrective Action Documents 2533997, 2548087 71111.13 Drawings O-0703-D One Line Diagram Station Auxiliary Circuits 600V/208V/
L/C 1X5 & MCC 1XH, 1XK, 1XL & 1XT 066 71111.13 Drawings O-0703-E One Line Diagram Station Auxiliary Circuits 600V/208V L/C 1X6 & MCC 1XI, 1XN, 1XP & 1XQ 077 71111.13 Drawings OFD-102A-3.1 Flow Diagram of Low Pressure Injection System (Borated Water Supply and LPI Pump Suction) 65 71111.13 Drawings OFD-102A-3.2 Flow Diagram of Low Pressure Injection System (LPI Pump Discharge) 49
16 Inspection Procedure Type Designation Description or Title Revision or Date 71111.13 Miscellaneous Clearance PRT-3-25-3A LPIP OOS-0056 71111.13 Miscellaneous Risk Profile for Unit 1 for the week of February 5th 71111.13 Miscellaneous Defense-in-Depth Status Sheet for November 1, 2024, at 1600 71111.13 Miscellaneous Defense-in-Depth Status Sheet for November 2, 2024, at 0400 71111.13 Miscellaneous Risk Profile for Unit 3 for the week of March 17th, 2025 71111.13 Miscellaneous Clearance OPS-3-23-LPI-3A LPIP DRN-1215 71111.13 Miscellaneous CSD-WC-ONS-0240-00 ONS ERAT Guidance 002 71111.13 Miscellaneous OSC-6551 PRA Analysis of Maintenance Rule Availability Performance Criteria 002 71111.13 Miscellaneous OSS-0254.00 1006 (MECH) Design Basis Specification for the Spent Fuel Cooling System 034 71111.13 Miscellaneous Risk Profile for Unit 2 for the week of February 24th 71111.13 Miscellaneous Risk Profile for Unit 3 for the week of February 11th 71111.13 Procedures AD-NF-ALL-0501 Electronic Risk Assessment Tool (ERAT) 8 71111.13 Procedures AD-OP-ALL-0210 Operational Risk Management 004 71111.13 Procedures AD-PI-ALL-0106 Cause Investigation Checklists 10 71111.13 Procedures AD-WC-ALL-0240 On-Line Risk Management Process 6
71111.13 Procedures AD-WC-ALL-0420 Shutdown Risk Management 8
71111.13 Procedures IP/0/A/2001/003 K Inspection and Maintenance of 600 Volt K-Line Breakers 039 71111.13 Procedures IP/0/A/2001/003 L Refurbishing 600 Volt K-Line Air Circuit Breaker 033 71111.13 Procedures IP/0/A/2001/015 PSW 13.8/4.16 kV Square D Type VR Vacuum Circuit Breaker Inspection and Maintenance 007 71111.13 Procedures OP/3/A/1102/008 On-Line Vale Lineup for MOV 035
17 Inspection Procedure Type Designation Description or Title Revision or Date Maintenance 71111.13 Procedures OP/3/A/1104/004 Low Pressure Injection System 172 71111.13 Work Orders 20699603, 20437227, 20706948, 20690548, 20687402, 20623800 71111.15 Calculations OSC-11505 HPI Pump Motor Upper Bearing Oil Cooler Performance Degradation Allowance 1
71111.15 Corrective Action Documents 02541430, 02541833, 02513617, 2527500, 02543252, 2533121, 2354722, 2545420, 2544934, 2424508, 2323274, 2546711, 2471779 71111.15 Drawings O-422-M-4 Instrument Details Steam to Emergency FDWP Trip Valve Control 14 71111.15 Drawings OFD-101A-1.4 Flow Diagram of High Pressure Injection System (Charging Section) 051 71111.15 Drawings OFD-101A-1.4 Flow Diagram of High Pressure Injection System (Charging Section) 052 71111.15 Drawings OFD-122A-1.4 Flow Diagram of Main Steam System Emergency Feedwater Pump Turbine Steam Supply and Exhaust 25 71111.15 Drawings OFD-127C-1.1 Flow Diagram of Nitrogen System (Nitrogen Supply to Close 1MS-93 During AFIS Actuation) 5 71111.15 Drawings ONTC-1-124B-0020-001 LPSW Flow to U1 HPI Pump Motor Coolers Test Acceptance Criteria 004 71111.15 Miscellaneous OM 251-0762.001 Outline Drawing For 6 CCI Drag Valve With Warming Disk, DMV-1265 1
71111.15 Miscellaneous OM 314.0586.001 Review of Pioneer HPI Pump Upper Motor Bearing Analysis and Certificate of Compliance 000 71111.15 Miscellaneous OSS-0254.00 1001 (MECH) High Pressure Injection and Purification &
Deborating Demineralizer Systems 066 71111.15 Miscellaneous OSS-0254.00 (MECH) Design Basis 064
18 Inspection Procedure Type Designation Description or Title Revision or Date 1004 Specification for Standby Shutdown Facility Reactor Coolant Makeup System 71111.15 Miscellaneous OSS-0254.00 1025 Design Basis Specification for the Instrument Air System 71111.15 Miscellaneous OSS-0254.00 4001 (MECH) Design Basis Spec for Reactor Building Containment Isolation 046 71111.15 Miscellaneous PTR001511 (4)
LCR-21 NUC Low Capacity, Lead-Acid Battery Laboratory Report 6438 71111.15 Procedures IP/0/A/3000/023 S SSF Battery DCSFS Performance Test 005 71111.15 Procedures IP/0/A/3000/023 S SSF Battery DCSFS Performance Test 005 71111.15 Procedures IP/1/A/0275/021 Unit 1 Emergency Feedwater System Nitrogen System Instrument Calibration 002 71111.15 Procedures MP/0/A/1200/132 A
Valve - Anchor Darling/Ladish - Flanged Bonnet - Swing Check -
Disassembly, Repair, and Assembly 033 71111.15 Procedures OP/0/A/1600/006 Operation of SSF KSF1/KSF2 Inverters And SSF CSF/CSFS Battery Chargers 033 71111.15 Procedures PT/1/A/0230/015 High Pressure Injection Motor Cooler Performance Test 062 71111.15 Procedures PT/1/A/0600/028 IMS-93 Nitrogen Supply Leakage Test 008 71111.15 Procedures PT/2/A/0600/012 Turbine Driven Emergency Feedwater Pump Test 099 71111.15 Work Orders 20690265, 20531943, 20267684, 20595784, 20376789, 20174297, 20540583, 20714044, 20284505 71111.24 Corrective Action Documents 02538150, 2540131, 2519656, 2500872, 2326549, 2513249, 2537631, 2390857, 2544935, 2201417, 2546947, 2546757, 2295170, 2539268 71111.24 Drawings OFD-102A-3.1 Flow Diagram of Low Pressure Injection System (Borated Water Supply and LPI Pump Suction) 065
19 Inspection Procedure Type Designation Description or Title Revision or Date 71111.24 Drawings OFD-102A-3.2 Flow Diagram of Low Pressure Injection System (LPI Pump Discharge) 049 71111.24 Drawings OFD-103A-3.1 Flow Diagram of Reactor Building Spray System 032 71111.24 Drawings OFD-133A-2.5 Flow Diagram of Condenser Circulating Water System SSF Aux Service 063 71111.24 Drawings ONTC-0-103A-0005-001 BS Pump Performance Test Acceptance for Pump Total Developed Head 4
71111.24 Miscellaneous CSD-EG-ONS-1619.1000 Diverse and Flexible Coping Strategies (FLEX) Program Document - Oconee Nuclear Station 005 71111.24 Miscellaneous OSS-0254.00 1005 (MECH) Design Basis Specification for the Standby Shutdown Facility Auxiliary Service Water System 047 71111.24 Miscellaneous OSS-0254.00 1034 Design Basis Specification for the Reactor Building Spray System 029 71111.24 Miscellaneous OSS-0254.00 2005 (ELECT) Keowee Emergency Power Design Basis 036 71111.24 Procedures IP/1/A/0400/049 KHU-1 Governor Speed Switch Instrument Calibration 20 71111.24 Procedures MP/0/A/1300/003 Pump - Ingersoll-Rand -
Low Pressure Service Water
- Rotating Assembly -
Removal and Replacement 039 71111.24 Procedures MP/0/A/1840/040 A
Pumps - Motors -
Miscellaneous Components -
Lubrication Post Maintenance Testing 004 71111.24 Procedures MP/0/A/3009/017 A
Visual Inspection and Electrical Motor Tests Using baker Equipment 004 71111.24 Procedures MP/0/A/3009/020 B
Motor - QA - Electric -
Removal, Replacement, and Post Maintenance Testing 044 71111.24 Procedures OP/0/A/1106/019 Keowee Hydro at Oconee 114 71111.24 Procedures OP/0/A/2000/013 KHU-1 Generator 029 71111.24 Procedures OP/0/B/1106/033 Primary System Leak Identification 023 71111.24 Procedures OP/3/A/1104/004 Low Pressure Injection System 172 71111.24 Procedures OP/3/A/1104/005 Reactor Building Spray System 044
20 Inspection Procedure Type Designation Description or Title Revision or Date 71111.24 Procedures PT/0/A/0400/005 SSF Auxiliary Service Water Test 070 71111.24 Procedures PT/0/A/0600/021 Standby Shutdown Facility Diesel - Generator Operation 018 71111.24 Procedures PT/0/A/0620/009 Keowee Hydro Operation 056 71111.24 Procedures PT/0/A/0620/016 Keowee Hydro Emergency Start Test 055 71111.24 Procedures PT/1/A/0202/011 High Pressure Injection Pump Test 108 71111.24 Procedures PT/1/A/0204/007 Reactor Building Spray Pump Test 106 71111.24 Procedures PT/1/A/0251/001 Low Pressure Service Water Pump Test 114 71111.24 Procedures PT/1/A/0600/012 Turbine Driven Emergency Feedwater Pump Test 109 71111.24 Procedures PT/2/A/0600/013 Motor Driven Emergency Feedwater Pump Test 077 71111.24 Procedures PT/3/A/0204/007 Reactor Building Spray Pump Test 099 71111.24 Work Orders 20700430, 20692464, 20702695, 20702296, 20616064, 20614788, 20631913, 20614191, 20678484, 20600155, 20695656, 20283832, 20622468, 20713656, 20570704, 20700763, 20702807 71151 Corrective Action Documents NCR 2534818 71151 Miscellaneous MSPI Margin and Derivation Reports for the High Pressure Injection System for Unit 1 for the 4th quarter 2024 71151 Miscellaneous MSPI Margin and Derivation Reports for the High Pressure Injection System for Unit 2 for the 1st quarter 2024 71151 Miscellaneous MSPI Margin and Derivation Reports for the High Pressure Injection System for Unit 3 for the 4th quarter 2024 71152A Corrective Action Nuclear Condition Report(s) 2491256, 2493255
21 Inspection Procedure Type Designation Description or Title Revision or Date Documents 71152A Miscellaneous Independent Assessment of the Work Environment of the Oconee Nuclear Station Warehouse Operations and Support Team November 2024 71153 Corrective Action Documents 2534840, 2434572
Attachment A A-1 IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria (ADAMS Accession No. ML24192A216)
Exhibit 1: Results of the Initial Evaluation 1.
Describe the influential assumptions used in the initial evaluation: The degraded condition created by the performance deficiency is the rendering of the Unit 1 pressurizer heaters powered from the common standby shutdown facility (SSF) to be inoperable and unable to do their safety function. Without pressurizer heater input, ambient losses from the pressurizer would cause the pressurizer to cool and subcooling margin for the reactor coolant system would be lost, allowing boiling to occur in the core and allowing the bubble to collapse in the pressurizer. Two phase flow will continue to cool the core until hot leg voiding occurs and no natural circulation flow through the core will take place until the RCS transitions into boiler-condenser mode (steam cooling), at which time an equilibrium temperature will be reached as long as adequate make up in maintained. Although the TS basis requires the SSF auxiliary service water system to be considered inoperable when pressurizer heaters are unavailable, the SSF auxiliary service water system is still able to perform its probabilistic risk assessment (PRA) function to provide secondary side cooling until natural circulation is lost in the primary.
Since this takes between 13-15 hours to occur, it is not appropriate to use the SSF ASW system as a surrogate for this condition. The core damage sequences of concern are station blackout events (SBO), internal fire events which result in a SBO, and internal and external flooding events. The same basic condition was evaluated as a Yellow significance old design issue in 2011 although the detailed risk assessment had three orders of magnitude of uncertainty (from Red to White). The dominate accident sequence was a large fire in the turbine building. The pressurizer (PZR) heater function is not modelled in the Standardized Plant Analysis Risk (SPAR) model or the licensees Computer Aided Faulty Tree Analysis (CATFA) model. Since 2011 several key plant modifications have been implemented:
(a) The installation of the protected service water (PSW) system. This system provides an alternate means of feeding steam generators, provides an alternate source of power for plant equipment such as high pressure injection pumps, and powers alternative banks of PZR heaters. The system and cables do not pass through the turbine building. However, the alternate PZR heater power was not modelled in the SPAR or CAFTA models.
(b) The SSF letdown line was modified to change the suction location from the letdown heat exchangers off the cold leg of the RCS, to off the hot leg, and replaced the isolation valve from an orifice and a gate valve to an actual throttle valve. This makes the operators action to throttle letdown flow to match SSF RCS makeup flow more likely to be successful and avoid lifting of a pressurizer safety valve.
(c) The licensee adopted National Fire Protection Association (NFPA) 805 and developed a Regulatory Guide 1200 Fire PRA. For this significance determination process (SDP), the licensee made a modification to model the PSW PZR heaters. The dominant human error probabilities (HEPs) were 1)
Attachment A A-2 Failure to Restore Pressurizer Heaters via Protected Service Water Power (1NPZRPSWDHE) and 2) Operators Failing to Throttle SSF Letdown to prevent lifting Pressurizer Safety Valves (1NPZRPSVDHE). PSW PZR heaters are not considered to be available for fires in the east penetration room and fires which result in a main control room evacuation (as the heaters are controlled from the main control room.)
(d) The licensee performed a thermal hydraulic analysis for this condition and identified the following. Without SSF pressurizer heaters it would take approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> to lose subcooling margin (SCM). At this point the pressurize bubble would collapse and the plant would go into solid plant operations. Once SCM was lost, two phase flow would begin in the RCS and after approximately 45 minutes, natural circulation flow would be lost. At this point, the RCS would rapidly begin to heat up and cause inflows into the pressurizer causing rapid pressure increases making it challenging to maintain pressure and prevent lifting a pressurizer relief valve. After about 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> without natural circulation, the RCS level would lower until the transition to boiler-condenser mode occurred and temperature would stabilize via steam cooling.
During a loss of coolant accident (LOCA), this transition is relatively short; however, during an SBO scenario this transition is longer and the possibility of lifting safety relief valves and losing RCS inventory is more likely during this time frame. Based on this, core damage can be prevented by restoring PSW supplied pressurizer heaters or by troubleshooting and repairing the SSF supplied pressurizer heaters during the first 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Operators would have an indication that SSF powered pressurizer heaters failed to energize during the first hour of the event. (The heaters energized light would not come on when the action was directed by procedure to use heaters to maintain RCS Pressure.)
2.
Sensitivities: The SRA compared the SPAR model results and cutsets from the detailed risk assessment performed in 2011 and the SPAR model internal events results and cutsets using the current SPAR model revision 8.82. The current model includes the PSW water supply to the steam generators and the backup power supply to the high pressure injection pumps. The 2011 SPAR model was dominated by a loss of instrument air event. This event has been significantly refined in the current model. Internal events are about two orders of magnitude lower in the current model. Qualitatively this can be applied to previous fire SDP results as well.
The licensees CAFTA model, using the model changes discussed above, used baseline values of HEPs 1NPZRPSWDHE and 1NPZRPSVDHE set at 1E-2. The model was most sensitive to 1NPZRPSVDHE. Given the uncertainty related to the success of this term, once natural circulation is lost under solid plant conditions, values of 1E-1 to 1E-3 would be reasonable, and result in risk results of 2E-6 to 2E-8.
3.
Identify any information gaps in defining the influential assumptions used in the initial evaluation: The HEP assumptions are extremely difficult to quantify. The plant conditions are not covered in normal training and actions would be contrary to the same actions taken before subcooling was lost. However, there would be substantial time to brief operators and restore power to the heaters.
Initial Evaluation Result: Bounding result is 2E-6 using best available data and surrogates.
A-3 Exhibit 2: Considerations for Evaluation of Decision Attributes Table 1:
Qualitative Decision-Making Attributes for NRC Management Review Decision Attribute Basis for Input to Decision - Provide qualitative and/or quantitative information for management review and decision making.
Defense-in-Depth Defense in depth has been enhanced with the addition of the PSW system. Although a weather related or grid related loss of offsite power (LOOP) would also assume PSW to be unavailable, PSW significantly mitigates the impact of any other LOOP events and major fire in the turbine building since the PSW cables and piping do not run through the turbine building.
Safety Margin Safety margin is higher than it was in 2011.
During the 2011 Detailed Risk Evaluation, it was concluded that the Thermal Hydraulic Code of Record was not effective in modeling two-phase flow, so core damage was assumed to occur early in the event. The current Thermal Hydraulic Code used for this evaluation shows operators have a substantial amount of time for recovery and/or repair efforts before natural circulation flow is lost between 13-15 hours.
Extent of Condition The Unit 2 and Unit 3 SSF PZR heaters and PSW powered PZR heaters for all three units were not affected by this performance deficiency; however, since it was a maintenance related event, common mode still must be considered.
Degree of Degradation The heater block assembly was installed incorrectly, not allowing the SSF powered PZR heater to energize until the low-level set point was reached (85), vice deenergizing them at that point. The error was troubleshot and repaired in approximately five hours after discovery, demonstrating that the condition was potentially recoverable via repairs.
Exposure Time The condition was present since 2022. For SDP purposes, the maximum 1-year exposure time was applied.
A-4 Recovery Actions
- 1) Recover PZR heater function using the PSW system. The PSW heaters use an independent circuit and power different heater banks. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> available. Note: Procedures exist for placing both PSW and SSF system in service in parallel during an event.
When both systems are available, operators will secure one of the systems. However, the SSF PZR heater availability is not considered in the procedure.
- 2) Repair SSF heaters by replacing the mis-wired heater block or rewiring. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> available.
Additional Qualitative Considerations The SSF letdown line modifications make throttling significantly easier decreasing the HEP for the first 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, but from 13-17 hours it would be a more significant challenge.
The modification also allows for some bleeding of steam from the RCS, which would likely delay the loss of natural circulation for a short additional period of time.
The SRA assumed an HEP of 1E-3 for hours 0-13, and HEP of 1E-1 for hours 13-17, and an HEP of 1E-2 for hours 17-24. This would give a time-weighted HEP of 2E-2 for the 24-hour PRA mission time.
==
Conclusion:==
Given the plant modifications and improved level of thermal hydraulic modelling which defines the substantial amount of time available before SSF throttling becomes significantly more difficult, the SRA recommends using the adjusted value of 1NPZRPSVDHE at 2E-2. This would also conservatively account for the repair option.
Using the licensees baseline data and adjusting for the HEP, the Delta CDP would be less than 1E-6, which corresponds to a finding of Very Low Safety Significance (GREEN).
Result of management review (COLOR):
GREEN