IR 05000400/2024003
| ML24311A145 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/07/2024 |
| From: | Matthew Fannon NRC/RGN-II/DORS |
| To: | Haaf T Duke Energy Progress |
| References | |
| EA-24-118 IR 2024003 | |
| Download: ML24311A145 (1) | |
Text
SUBJECT:
SHEARON HARRIS NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000400/2024003
Dear Thomas Haaf:
On September 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Shearon Harris Nuclear Plant. On October 23, 2024, 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.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Shearon Harris Nuclear Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Shearon Harris Nuclear Plant.
November 7, 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, Matthew S. Fannon, Chief Projects Branch 2 Division of Operating Reactor Safety Docket No. 05000400 License No. NPF-63
Enclosure:
As stated
Inspection Report
Docket Number:
05000400
License Number:
Report Number:
Enterprise Identifier:
I-2024-003-0022
Licensee:
Duke Energy Progress, LLC
Facility:
Shearon Harris Nuclear Plant
Location:
New Hill, North Carolina
Inspection Dates:
July 01, 2024, to September 30, 2024
Inspectors:
P. Boguszewski, Senior Resident Inspector
K. Henry, Regional Governmental Liaison Officer
M. Kay, Resident Inspector
P. Torres, Project Manager
Approved By:
Matthew S. Fannon, Chief
Projects Branch 2
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Shearon Harris Nuclear Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Verify Acceptability of 'C' Cold Leg Accumulator Temporary Modification Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000400/2024003-01 Open/Closed
[H.13] -
Consistent Process 71111.18 The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to provide design control measures for ensuring the adequacy of a temporary modification to the 'C' cold leg accumulator (CLA) relief valve. Specifically, the licensee failed to identify during the design process, that the relief valve discharge flange is rated for substantially less pressure than the CLA. This could have resulted in the pressurization of the relief valve discharge flange to a pressure greater than its rated capacity, leading to failure of the relief valve discharge flange and rapid depressurization of the 'C' CLA.
Failure to Justify Deviation from Manufacturers Specifications Results in Electrical Faults and a Reactor Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000400/2024003-02 Open/Closed EA-24-118
[H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated NCV of technical specification (TS) limiting condition for operation (LCO) 3.0.4, was identified when the licensee failed to follow procedure AD-EG-ALL-1132, Preparation and Control of Engineering Changes. Specifically, the licensee deviated from the manufacturers specifications by installing medium voltage direct current (DC) cables in an alternating current (AC) electrical system. The improper use of these cables revealed itself as electrical faults on the A startup transformer (SUT) and B unit auxiliary transformer (UAT). In review of past-operability of the transformers, this resulted in the licensee entering Mode 4 without meeting operability requirements of TS 3.8.1.1.a,
"A.C. Sources Operating," as required by TS 3.0.4.
Additional Tracking Items
None.
PLANT STATUS
Unit 1 operated at or near rated thermal power for the entire inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - 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)
'B' emergency service water during 'A' emergency service water system maintenance on July 9, 2024 (2)
'A' motor driven auxiliary feedwater during turbine driven auxiliary feedwater maintenance on September 25, 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) Fuel handling building north on August 14, 2024
- (2) Reactor auxiliary building 236' elevation on August 23, 2024
- (3) Fuel handling building spent fuel pool cooling area on August 23, 2024
- (4) Reactor auxiliary building 190' elevation on August 19, 2024
- (5) Reactor auxiliary building 216' elevation on August 19, 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 unannounced fire drill on September 10, 2024.
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 Maintenance Surveillance Test (MST)- I0151, "Steam Generator C Narrow Range Level Loop (L-0496) Operational Test," on August 20, 2024, and during nuclear instrument calibrations on August 21, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated simulator training on Abnormal Operating Procedure (AOP) 51, "Loss of Turbine Generator," and classroom training on AOP 004, "Remote Shutdown," on August 27, 2024.
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)1CS-278, emergency boration isolation valve, failed to open from main control board (Nuclear Condition Report [NCR] 02512228) on August 8, 2024
- (2) Functional failures of residual heat removal pump breakers (NCR 02376252 and
===02301522) on September 20, 2024
- (3) Emergency DC lighting (battery powered) testing failures (NCR 02441330 and 02438075) on July 1, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01)===
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) Elevated risk due to planned 'A' emergency diesel generator (EDG) maintenance on July 9, 2024
- (2) Elevated risk due to planned 'B' EDG maintenance on July 25, 2024
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (2 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) NCR 02525809, 'C' CLA relief valve 1SI-227 potential over-pressurization at pancake flange, on September 10, 2024
- (2) NCR 02526890, fuel handling building wall penetration seal P3300 thickness discrepancies on September 6, 2024
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Engineering Change (EC) 424523, temporary relief valve for 'C' CLA, on September 11, 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) Operations Surveillance Test (OST)-1085, 1A-SA Diesel Generator Operability Test, on July 12, 2024
- (2) OST-1073, 1B-SB Emergency Diesel Generator Operability Test, on August 1, 2024
- (3) OST-1191, Steam Generator Pilot Operated Relief Valve and Block Valve Operability Test, on September 16, 2024
- (4) OST-1008, 1A-SA Residual Heat Removal Pump Operability Quarterly, on September 18th, 2024
- (5) OST-1831, Turbine Driven Auxiliary Feedwater (TDAFW) Pump Auto Start Response Time Test, following TDAFW governor replacement on September 27, 2024
- (6) OST-1411, TDAFW Pump Operability Test, following TDAFW governor replacement, and trip & throttle valve maintenance on September 30, 2024
Surveillance Testing (IP Section 03.01) (1 Sample)
- (1) Operations Periodic Test (OPT)-1531, Dedicated Shutdown Diesel Operability Test, on August 2, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) OST-1007, Chemical and Volume Control System/Safety Injection System Operability Train A Quarterly, on September 4, 2024
Diverse and Flexible Coping Strategies (FLEX) Testing (IP Section 03.02) (1 Sample)
71114.06 - Drill Evaluation
Required Emergency Preparedness Drill (1 Sample)
- (1) The inspectors observed and evaluated the conduct of an emergency preparedness drill involving a loss of all safety injection and a loss of coolant accident on August 6, 2024.
Additional Drill and/or Training Evolution (1 Sample)
The inspectors evaluated:
- (1) An emergency preparedness field monitoring team training scenario involving a loss of coolant accident and failure of the inner and outer emergency airlock door
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS06: Emergency AC Power Systems (IP Section 02.05)===
- (1) July 1, 2023, through June 30, 2024
MS07: High Pressure Injection Systems (IP Section 02.06) (1 Sample)
- (1) July 1, 2023, through June 30, 2024
MS09: Residual Heat Removal Systems (IP Section 02.08) (1 Sample)
- (1) July 1, 2023, through June 30, 2024
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1)
'A' containment spray discharge valve torque switch failure (NCR 02514152) on September 11, 2024
- (2) Gas void found in containment spray piping (NCR 02457523) on September 18, 2024
- (3) Plant vent stack radiation monitor repeat performance issues (NCR 02524459) on September 6,
INSPECTION RESULTS
Failure to Verify Acceptability of 'C' Cold Leg Accumulator Temporary Modification Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000400/2024003-01 Open/Closed
[H.13] -
Consistent Process 71111.18 The inspectors identified a Green finding and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to provide design control measures for ensuring the adequacy of a temporary modification to the 'C' CLA relief valve. Specifically, the licensee failed to identify during the design process, that the relief valve discharge flange is rated for substantially less pressure than the CLA. This could have resulted in the pressurization of the relief valve discharge flange to a pressure greater than its rated capacity, leading to failure of the relief valve discharge flange and rapid depressurization of the 'C' CLA.
Description:
The reactor cooling system (RCS) CLAs are tanks of borated water pressurized with 585 to 665 pounds square inch gauge (psig) of nitrogen. One accumulator is attached to each of the cold legs of the RCS. The TS limits on accumulator water volume and nitrogen cover pressure ensure that the discharge from the accumulators is sufficient to limit core damage following a design-basis loss of coolant accident. Each accumulator has a relief valve to protect the accumulator from over-pressurization. During normal plant operations, each accumulator is isolated from the RCS by two seated check valves in series. During a loss of coolant accident, if the RCS pressure falls below the accumulator pressure, the check valves unseat, and borated water is immediately forced into the RCS by the expansion of the nitrogen volume. To ensure the accumulator pressure is maintained within the required TS limits, the licensee periodically adds nitrogen to the accumulator tanks.
In July 2024, the licensee noticed that the 'C' CLA required more frequent re-pressurization.
An investigation was performed based on this trend, and the licensee determined that the 'C' CLA relief valve was leaking/venting at a lower-than-expected pressure. To address the issue, on August 13, 2024, the licensee implemented an EC that involved adding a secondary relief valve at a new location. This temporary modification also included gagging the leaking relief valve and installation of a blind flange at the valve discharge flange to stop the leak.
The NRC inspectors reviewed EC 424523, Temporary Relief Valve for 1C-SA Accumulator (1SI-E011), Revision 0, and identified that a third-party review of the EC provided that if a blind flange is to be installed, pressure monitoring to the downstream end of the leaking relief valve shall be installed. The third-party review stated that the maximum allowable working pressure of the downstream end of the relief valve is only 245 psig at 100F. The inspectors could not find any information regarding pressure monitoring equipment installed at this location in the EC package and expressed this concern to the licensee. The licensees investigation determined that the discharge flange for the relief valve is an ANSI 150# flange, which is rated to 245 psig at 100F. Therefore, further leakage into the valve's discharge cavity has the potential to pressurize the discharge flange past the ANSI rating up to 665 psig (max accumulator pressure), which could result in a valve failure and has the potential to depressurize the 'C' CLA.
Based on this information, the licensee performed an operability evaluation of the 'C' CLA.
The licensees operability determination concluded that the C CLA was operable based on field reports that the installed modification had arrested the leakage. However, the licensee acknowledged that the potential for future pressurization of the discharge cavity of the relief valve remained unknown. The licensee determined that if a failure were to occur due to pressurizing the relief valve above its rated discharge pressure, it is most likely that a leak would develop at the discharge flange gasket prior to a destructive failure. However, due to the proprietary nature of the valve design, this did not rule out the possibility of a more impactful failure. To demonstrate that small amounts of leakage past the gag would leak out of the top of the valve and not pressurize the discharge flange, the licensee performed testing on a mockup of the installed modification in their maintenance shop. However, if leakage was to be significant, it is unknown if the discharge flange would become pressurized, potentially failing the relief valve.
The licensees engineering department also consulted the valve manufacturer, Emerson, to confirm the valve discharge flange pressure rating. Emerson confirmed the discharge flange is rated to a lower pressure than the valve inlet design. It was also confirmed that the weakest point on the valve pressure boundary is at the discharge flange. Based on Emersons recommendations, the licensee made the decision to revise the EC and remove the blind flange on September 11, 2024. The licensees attempt to remove the blind flange was unsuccessful due to the signs of potential pressure buildup at the valve gag, raising personnel safety concerns. In the interim, the licensee determined that the 'C' CLA remains operable and plans to perform additional evaluation to ensure the structural integrity of the system until the next refueling outage, when the licensee plans to remove the flange permanently.
Corrective Actions: The licensee entered the condition in their corrective action program and performed an immediate operability determination with engineering input. The licensees long-term solution is to remove the gag and the blind flange in next refueling outage, H1R26.
Corrective Action References: NCR 02525809
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure of design control measures that ensure adequacy of the design change to the 'C' CLA relief valve was a performance deficiency that was within the licensees ability to foresee and correct.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design 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 failure to ensure that design control measures provided verification of the adequacy of the design of a temporary modification to the 'C' CLA relief valve introduced the potential for a failure of the 'C' CLA vent valve, which could result in the rapid depressurization and inoperability of the 'C' CLA. The inspectors used IMC 0612, Appendix E, Examples of Minor Issues, dated November 1, 2023, to inform answers to the more than minor screening questions and found this condition consistent with more than minor Example 3.a.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the violation affected the design or qualification of a mitigating SSC, but the SSC maintained its operability.
Cross-Cutting Aspect: H.13 - Consistent Process: Individuals use a consistent, systematic approach to make decisions. Risk insights are incorporated as appropriate. Specifically, during the engineering change process, the licensee was mainly focused on the structural and code requirements associated with adding the new, equivalent, relief valve, to the 'C' CLA. Due to the increased focus on the configuration of the new relief valve, the licensee did not apply the same level of scrutiny to the third-party review information for the as-left configuration of the original, leaking, relief valve. Applying a consistent, systematic approach, to the review of vendor supplied information could have led to more rigorous review of the as-left configuration of the original relief valve.
Enforcement:
Violation: 10 CFR 50, Appendix B, Criterion III, Design Control, states in part that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
Contrary to the above, on August 13, 2024, the licensees design control measures did not provide for verifying or checking the adequacy of the design of a temporary modification to the 'C' CLA relief valve, such as by the performance of design reviews. Specifically, the licensee failed to identify that the relief valve flange is rated for substantially less pressure than the CLA during the design change process. This could have resulted in the pressurization of the relief valve discharge flange to a pressure greater than its rated capacity, leading to failure of the relief valve discharge flange and rapid depressurization of the 'C' CLA.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Justify Deviation from Manufacturers Specifications Results in Electrical Faults and a Reactor Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000400/2024003-02 Open/Closed EA-24-118
[H.12] - Avoid Complacency 71153 A self-revealed Green finding and associated NCV of TS LCO 3.0.4, was identified when the licensee failed to follow procedure AD-EG-ALL-1132, Preparation and Control of Engineering Changes. Specifically, the licensee deviated from the manufacturers specifications by installing medium voltage DC cables in an AC electrical system. The improper use of these cables revealed itself as electrical faults on the A SUT and B UAT. In review of past-operability of the transformers, this resulted in the licensee entering Mode 4 without meeting operability requirements of TS 3.8.1.1.a, "A.C. Sources Operating," as required by TS 3.0.4.
Description:
The licensees 6.9 kilovolt (kV) electrical system supplies Class 1E loads via emergency buses 1A-SA and 1B-SB for train 'A' and 'B' of safety-related equipment, respectively. There are three power sources available to each emergency bus. These three sources are the SUTs, the UATs, and the EDG. When the plant is online, power to the emergency buses is supplied by the UATs, which are fed by the main generator. On a plant trip or during shutdown and startup conditions, offsite power is supplied to the emergency buses by the SUTs. If both the UATs and the SUTs were to become unavailable, the emergency buses would be powered by the EDGs.
During the licensees refueling outage from April 10, 2024, to May 12, 2024, the licensee replaced the A SUT and B UAT. The new transformers were equipped with on load tap changers (OLTCs). OLTCs automatically adjust the transformers tap settings to maintain the transformer output voltage within the required range to ensure proper operation of plant normal and emergency loads following a unit trip or loss-of-coolant accident. The OLTCs were connected to the 6.9 kV non-segregated bus bars via cables that run through potential transformers (PTs). The PTs stepped the voltage down before sending it to the OLTCs as the sensing voltage. The OLTCs were installed in manual mode with the automatic function disabled. The licensee planned to place the OLTCs in automatic in the future when they would be necessary due to potential decreases in grid voltage caused by the planned retirement of other nearby power plants.
Following the replacement of the A SUT, the TS offsite power circuit was declared operable on April 30, 2024, at 5:00 PM. The A SUT was then loaded to support plant startup activities and remained loaded until the main generator was back online. The electrical loads were transferred to the A UAT per plant procedures on May 13, 2024, at 12:35 AM. Shortly after being unloaded, on May 13, 2024, at 2:45 AM, the plant experienced an electrical fault at the A SUT. As a result of the fault, one offsite power circuit was declared inoperable, and repairs were commenced. The licensee entered their failure investigation process to determine the cause of the electrical fault and concluded that the fault was the result of the failure of one of the cables going from the 6.9 kV non-segregated bus bars to the PTs in the OLTC circuit. The licensee made the decision to leave the OLTC circuit disconnected when completing repairs.
The A SUT was repaired, and the TS required offsite power circuit was declared operable on May 22, 2024, at 10:38 PM.
Due to the belief that the cable failure could have been caused by a one-off manufacturing defect or damage during installation, the licensee did not immediately de-energize and modify the B UAT to remove the OLTC circuit on the B UAT. The licensee performed additional testing on cables of the same material and engaged with the cable vendor to better understand the cause of the PT cable failure while assessing the likelihood of the same cables on the B UAT failing in the same manner. However, no corrective actions were taken on the B UAT prior to May 30, 2024. On May 30, 2024, at 7:49 PM, a PT cable failed in the OLTC circuit of the B UAT resulting a main generator lockout, turbine trip, and reactor trip.
Further investigation and discussions with the manufacturer revealed that the PT cables used, were only rated for DC applications and did not have an approved AC rating. The cables were used in an AC circuit on both the A SUT and B UAT with an applied voltage of 4 kV phase to ground. The licensee confirmed through testing that at 4 kV applied voltage, the cables in question will fail prematurely. The licensees root cause evaluation concluded that that they did not adequately evaluate the use of the DC cables in an AC circuit prior to the approval and implementation of the transformer replacements.
The inspectors concluded that the licensee failed to follow procedure AD-EG-ALL-1132, Preparation and Control of Engineering Changes, Section 5.2.4 (Design Inputs), Step 33, which stated, [i]f the EC deviates from Manufacturers specifications, then consult the Original Equipment Manufacturer or successor to discuss the deviation, or to justify the deviation in the BASEDOCS attribute, including feedback from the manufacturer on the risks associated with the deviation. Responsible Engineering Manager approval of the deviation occurs as part of the overall EC Approval. Specifically, the licensee deviated from the manufacturers specifications by installing medium voltage DC cables in an AC electrical system without justification and approval in the engineering change documentation. As discussed above, the improper use of these cables revealed itself as electrical faults on the A SUT and B UAT. In review of past-operability, this resulted in the licensee transitioning from Mode 5 to Mode 4 without meeting operability requirements of TS 3.8.1.1.a, which requires two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, as required by TS 3.0.4.
The inspectors determined that the April 30, 2024 cable installation rendered the A SUT inoperable, as it could not have performed its design function with the incorrect cables installed. Thus, only one of two TS required offsite power circuits was available until the repair to the transformer was completed on May 22, 2024.
Corrective Actions: The licensee repaired the damaged electrical equipment, removed the PT cables, entered the condition into their corrective action program, and performed a root cause evaluation.
Corrective Action References: NCR 2517946
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to adhere to procedure AD-EG-ALL-1132, Preparation and Control of Engineering Change, when performing a design change on the A SUT and B UAT, was a performance deficiency that was within the licensees ability to foresee and correct. Specifically, the licensee deviated from the cable manufacturers specifications by using electrical cables rated for medium voltage DC in an AC circuit without providing the procedurally required justification and approval of the deviation. As a result, an electrical fault and inoperability of the A SUT on May 13, 2024, and an electrical fault on B UAT, resulted in a turbine trip and subsequent reactor trip on May 30, 2024.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance 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 condition affected the availability and reliability of the A SUT from April 30, 2024, to May 22, 2024, and an electrical fault on the 'B' AUT resulted in a turbine trip and subsequent reactor trip on May 30, 2024. The inspectors used IMC 0612, Appendix E, Examples of Minor Issues, dated November 1, 2023, to inform answers to the more than minor screening questions and found this condition consistent with more than minor Example 4.b.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined a detailed risk evaluation (DRE) was required because the degraded condition resulted in an actual partial loss of a support system (AC Power). A regional Senior Reactor Analyst (SRA) performed a DRE in accordance with IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, and the Risk Assessment Standardization Project (RASP) Manual, Volume 1.
Exposure Time Per the RASP Manual, the exposure time is T + Repair Time since the A SUT and B UAT were in continuous operation when the failures occurred. RASP Manual, Volume 1, Section 2.6, Exposure Time for Continuous Component Operation Failures, states:
- For failure of a component that is normally in continuous operation while at-power (e.g., normally operating service water pump), the exposure time should be the PRA mission time (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).
- The analysis of some conditions may involve fault tree modeling of a support system initiating event. In this case, mission times for the normally running components may be more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In this case, the failure of the A SUT occurred on May 13, 2024. The licensee promptly entered their risk-informed completion time (RICT) program as permitted by TS, and established risk management actions (RMAs) as required by the program. The A SUT was returned to service on May 22, 2024, after a repair time of 212 hours0.00245 days <br />0.0589 hours <br />3.505291e-4 weeks <br />8.0666e-5 months <br />. Therefore, the exposure time for the SUT was from May 12, 2024, until May 22, 2024, for a total of 236 hours0.00273 days <br />0.0656 hours <br />3.902116e-4 weeks <br />8.9798e-5 months <br />. After this time, the A SUT no longer had the non-conforming condition.
The B UAT was in service with the non-conforming condition from May 13, 2024 (two hours prior to the A SUT failure) until it failed on May 30, 2024.
Therefore, four time periods were considered in this analysis:
1) 24-hours prior to the A SUT failure until 2-hours prior to the A SUT failure 2) 2-hours prior to the A SUT failure until the A SUT failure 3) The 212-hour repair period when the licensee was in a RICT for the A SUT 4) The 8-day period from the A SUT being returned to service until the B UAT failure and plant trip The SRA conducted a plant walkdown the week of September 3, 2024, to validate FLEX strategies and investigate whether any recovery actions or repairs could reasonably be credited in the risk analysis. As a result of the walkdown, the SRA identified that the 480VAC 1D2 bus could be repowered via cross-connecting with the 1E2 bus for the original dominant accident sequence of a steam generator tube rupture (SGTR). The 1D2 bus powers the demineralized water transfer pumps and would lose power when the A SUT failed. There is a plant procedure to perform this cross-connection and it was an evolution performed during the recent outage to support electrical bus work. The SRA reviewed the licensees modular analysis program analysis data and plant logs to determine there would be adequate time to perform this action in a SGTR scenario.
Therefore, the SPAR model was adjusted to include this action.
The SRA used SAPHIRE Version 8.2.10 and the Harris Unit 1 SPAR model, Version 8.83, dated September 21, 2023, to perform the analysis for each of the four periods for internal events, internal fires, internal flooding, seismic, and tornado/high winds. The dominant accident sequence was a Period 3 Loss of the 6.9 KV AC emergency bus 1B-SB with a failure of the A EDG to run and a failure of the TDAFW pump to run. The delta Core Damage Frequency (CDF) was approximately 9 E-7, and the delta Large Early Release Frequency (LERF) was 9.67 E-8; therefore, the finding is characterized as very low safety significance (Green).
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, during the engineering change process, the licensee engineers perceived the PT cables as a minor part of the modification in comparison to the transformer design. Due to this perception, individuals did not plan for the inherent risk associated with a failure of the PT cables.
Enforcement:
Violation: TS LCO 3.0.4 requires, in part, that all LCOs be met when entering a Mode or other specified condition stated in the applicability section. TS LCO 3.8.1.1.a requires that two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system be operable in Modes 1 through 4.
Contrary to the above, the licensee failed to ensure all LCOs were met before entering a Mode of applicability. Specifically, the licensee transitioned from Mode 5 to Mode 4 on May 4, 2024, and again on May 10, 2024, without meeting operability requirements of TS 3.8.1.1.a, with only one of two offsite power circuit operable, as required by TS 3.0.4.
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 23, 2024, the inspectors presented the integrated inspection results to Thomas Haaf and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
Miscellaneous
CAR-SH-AS-60
Ebasco Specification Penetration Barrier Fire Stop
Assemblies and Fire Breaks
Revision 2
Work Orders
20632305
Installation of Cat 6 Cabling for the Operations Camera
Project
Revision 24
Corrective Action
Documents
Resulting from
Inspection
2525809
Corrective Action
Documents
2521547,
2521548,
2521564,
2522439,
2523301,
2524459