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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000269/FIN-2014005-01 | 2014Q4 | Oconee | Failure to Update FSAR for Mode 4 LOCA | An NRC identified Severity Level IV violation of 10 CFR 50.71(e), "Maintenance of Records, Making of Reports," was identified for the licensees failure to update the final safety analysis report (FSAR) after the licensee adopted the improved technical specifications (ITS). The licensee adoption of ITS introduced the possibility of a Mode 4 loss of cooling accident (LOCA), which was an accident of a different type than previously evaluated in the FSAR. The licensee initiated PIP O-15-00260 in order to determine future corrective actions. Continued non-compliance does not present an immediate safety concern because the inspectors assessed this as a very low safety significant issue. The licensees failure to update the FSAR as required by 10 CFR 50.71(e) was a performance deficiency. The performance deficiency impacted the ability of the NRC to perform its regulatory oversight function and was dispositioned using traditional enforcement. Specifically, a failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating license amendment requests, and in preparation for and conduct of inspection activities. This violation was determined to be a Severity Level IV violation per Section 6.1.d.3 of the NRC Enforcement Policy, revised July 9, 2013, because the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. The NRC Enforcement Policy also requires disposition of findings in the significance determination process, which determined the finding was not more than minor. Since this issue was dispositioned using traditional enforcement, there was no cross-cutting aspect associated with this violation. |
| 05000269/FIN-2014005-02 | 2014Q4 | Oconee | Keowee Hydro Unit 2 Inoperable for Longer Than Allowed TS Outage Time | A self-revealing Green NCV of Oconee Nuclear Station Technical Specification (TS) 3.8.1, AC Sources Operating, was identified for Keowee Hydro Unit 2 being inoperable for longer than allowed TS outage time. The licensee modified Keowee Hydro Unit 2 electrical protection circuitry with a faster response relay which was susceptible to an existing degraded system condition and ultimately caused Keowee Hydro Unit 2 to be inoperable. The licensee implemented engineering change (EC111358) which moved the 86E2X relay to another cabinet which was not susceptible to the vibration from the governor oil system. The licensee entered this issue in their corrective action program (CAP) as PIP-O-13-09152. The licensees failure to properly evaluate a modification to the electrical control circuit of the governor oil system, which resulted in Keowee Hydro Unit 2 being inoperable for longer than allowed TS outage time, was a performance deficiency. The issue is more than minor because it was associated with the equipment performance attribute of the mitigating system cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the modification of the governor oil system, including the addition of the 86E2X governor TXS catastrophic relay, resulted in Keowee Hydro Unit 2 being inoperable for longer than allowed TS outage time. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process (SDP), Attachment 4 and Attachment A and determined to require a detailed risk evaluation. A regional Senior Reactor Analyst performed a risk analysis of the performance deficiency which was found to be Green (CDF < 1E-6/year). The dominant accident sequence was a loss of offsite power where Keowee Unit 1 fails independently and unrelated to the performance deficiency and power is not successfully restored by Oconee operators. The influential factors in the Green result were the limited exposure time (19 days) and the ability to quickly restore power to the unit via the Lee Station gas turbines via the Fant Line. This finding was determined to have a cross-cutting aspect in the problem identification and resolution cross cutting area because the licensees organization failed to take effective corrective actions to address the issue in a timely manner commensurate with its safety significance. Specifically, the licensee failed to take effective corrective actions to address system interactions (i.e. high vibrations) which ultimately had an adverse effect upon modifications to the governor oil system of the Keowee Hydro Unit 2. |
| 05000269/FIN-2014011-01 | 2014Q2 | Oconee | Failure to Identify and Correct Weld Cracking in HPI Nozzle | A self-revealing potentially Greater than Green AV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified when the licensee failed to identify a crack in a weld located in the Unit 1 High Pressure Injection (HPI) system. In 2004, a procedure was developed for augmented in-service inspection program ultrasonic examinations which effectively removed reasonable assurance that HPI nozzle component cracking would be identified and corrected. NDE-995, Ultrasonic Examination of Small Diameter Piping Butt Welds and Base Material for Thermal Fatigue Damage, did not contain the necessary steps to achieve acceptable coverage for UT examinations when limitations were encountered. The inspectors determined that the failure to ensure that station procedure NDE-995 was adequate to identify and correct cracking in weld 1-RC-201-105 was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because it affected the Design Control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective in that an unidentified crack resulted in reactor coolant system pressure boundary leakage and a forced shutdown of Unit 1. The finding was determined to require a detailed risk analysis because the condition could have resulted in a leak which exceeded the reactor coolant system leak rate for a small-break loss of coolant accident. There was no immediate safety concern because the crack was repaired. The inspectors determined this finding has a cross-cutting aspect of H.7 in the Documentation component of the Human Performance area because the licensee did not create and maintain complete, accurate, and up-to-date documentation in procedure NDE-995 to ensure acceptable coverage for UT examinations. |
| 05000269/FIN-2015002-01 | 2015Q2 | Oconee | Inadequate Design Inputs for PSW Testing and Engineering Evaluations | The NRC identified a finding for the licensees failure to verify the adequacy of design inputs used in protected service water (PSW) testing and engineering evaluations to validate that the PSW system could perform its design function with respect to Milestone 4 of order EA-13-010, in accordance with the Duke Energy Carolinas Topical Report, Quality Assurance Program. The licensee entered this issue into their corrective action program as problem investigation program reports (PIPs) O-15-03630, O-15-03527, O-15-03529, O-15- 03631, O-15-03530, NCR 01930521, NCR 01929161, and PIP 0-15-4544. The performance deficiency was more than minor because it was associated with the design control attribute and adversely affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the errors identified in the hydraulic flow modeling software, Calculation OSC-9595, Protected Service Water (PSW) Hydraulic Model, Rev. 6, and supporting documentation required significant revision and reanalysis in order to determine that the PSW system was capable of meeting its design flow requirements for short term secondary heat removal capability. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of avoid complacency within the human performance area. Specifically, the licensee failed to utilize standard human error prevention tools to ensure critical reviews were performed for the PSW testing and engineering evaluations supporting the completion of Milestone 4 of order EA-13-010 dated July 1, 2013. |
| 05000269/FIN-2015002-02 | 2015Q2 | Oconee | Inadequate Acceptance Criteria for PSW Pump Surveillance Testing | The NRC identified a finding for the licensees failure to ensure that appropriate acceptance criteria was used during testing to verify PSW primary pump functionality in accordance with the Duke Energy Carolinas Topical Report, Quality Assurance Program. The licensee entered this issue into their corrective action program as PIP O-15-03190. The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, PSW pump surveillance PT/0/A/0500/001, Protected Service Water Primary and Booster Pump Test, Rev. 0, did not incorporate acceptance limits established by design documents, and as a result, the licensee could unknowingly consider the PSW primary pump functional beyond seven percent pump degradation. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its functionality. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of avoid complacency within the human performance area. Specifically, the licensee failed to utilize standard human error prevention tools to ensure critical reviews were performed for PSW pump testing. |
| 05000269/FIN-2015002-03 | 2015Q2 | Oconee | Failure To Translate The Design Basis Into Procedures Used To Test The HPI Motor Coolers | The NRC identified a finding for the licensees failure to translate the design requirements of the high pressure injection (HPI) pump motor coolers into the procedure used to verify adequate flow from PSW, in accordance with the Duke Energy Carolinas Topical Report, Quality Assurance Program. Specifically, the licensee failed to incorporate the fouling factor assumed in Calculation OSC-2042, HPI Pump Motor Upper Bearing Cooling Report, Rev. 8, into Procedure TT/1/A/05000/008, High Pressure Injection Motor Cooler Flow Test from PSW, Rev. 2. The licensee entered this issue into their corrective action program as PIPs O-15-03608 and O-15-04544. The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the low pressure service water (LPSW) and PSW flow test acceptance criteria could have been met without ensuring adequate heat transfer could be provided from the HPI motor coolers to PSW. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of teamwork within the human performance area. Specifically, the licensee failed to demonstrate a strong sense of collaboration and cooperation in connection with projects to ensure critical reviews were performed for the procedures used to test the HPI motor coolers. |
| 05000269/FIN-2016003-01 | 2016Q3 | Oconee | Failure to Translate Design Requirements to Prevent the Effects of Waterhammer | The NRC identified a finding for the licensees failure to translate the limiting flow rate design requirement into station procedures used to start and operate the alternate reactor building cooling (RBC) system, in accordance with the Duke Energy Carolinas Topical Report, Quality Assurance Plan (QAP). Specifically, the licensee failed to translate the limiting flow rate of 170 gallons per minute (gpm) into Procedure AP/0/A/1700/051, Alternate Reactor Building Cooling, Revision (Rev.) 2, to ensure prevention of waterhammer on the A reactor building cooling unit (RBCU) or connecting low pressure service water (LPSW) lines when starting the RBCU Hale pump. The licensee entered this issue into their corrective action program as Action Request (AR) 02049903 and revised Procedure AP/0/A/1700/051 to limit the RBCU Hale pump discharge flow to each affected unit to an initial fill rate of 120 gpm or less. The performance deficiency was determined to be more than minor because it adversely affected the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, opening the RBCU Hale pump discharge valve four turns, as specified in the procedure, would have resulted in filling the alternate RBC system at approximately 600-700 gpm and exceeding the design flow rate of 170 gpm established to prevent equipment and piping damage as a result of waterhammer. This provided a reasonable doubt that the alternate RBC system had the capability to reliably perform its intended safety function and, in turn, that the protected service water (PSW) system had the capability to meet its 30-day mission time during a turbine building fire that resulted in a loss of offsite power. The finding was determined to be of very low safety significance (Green) because the finding would not have resulted in a fire that caused secondary fires outside of the originating fire area due to circuit issues and did not affect the ability to reach and maintain a stable plant condition within the first 24-hours of a fire event. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of design margin, in the area of human performance. Specifically, the licensee failed to operate and maintain the alternate RBC system equipment within design margins when they did not translate design requirements from Engineering Change (EC) 110008 and Calculation OSC-8107 into station procedures. |
| 05000269/FIN-2016004-01 | 2016Q4 | Oconee | Failure to Perform Appropriate Evaluation of Motor Operated Valve Actuator Output Capability | Green. The NRC identified a non-cited violation (NCV) of Title 10 Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly determine the bounding degraded voltage to be assumed in the determination of motor operated valve (MOV) actuator output capability. Specifically, the licensee did not use appropriate transient voltages as input into the evaluation of the capability of the MOVs that are required to reposition in response to an accident signal. In response, the licensee entered the issue into their corrective action program as nuclear condition report (NCR) 2056895 and planned to formally revise their calculations to reflect the current plant configuration. This performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees programmatic failure to use bounding terminal voltage values in the evaluation of their automatically actuated, safety-related MOVs did not ensure they would be capable of mitigating accidents when powered from sources other than the 230kV switchyard, thus resulting in doubt on their capability to perform their intended safety function. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the most recent transient analysis that was performed for the sources other than the 230kV switchyard was performed in 2012. |
| 05000269/FIN-2016004-02 | 2016Q4 | Oconee | Inappropriate Voltage Band in Lee Combustion Turbine Unit Operating Procedure | Green. The NRC identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to identify appropriate procedural updates that were needed to ensure the Lee combustion turbine (LCT) procedures were appropriate for the circumstances and maintained current. Specifically, the licensee did not include appropriate operational limitations in procedures associated with the LCTs. In response, the licensee generated NCR 2058763, verified the LCT automatic voltage regulator setpoint was, and had been, 13.8kV, and generated a corrective action to revise the affected procedures limits to 13.78kV, a value bounded by station analyses. This performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees failure to limit the operating voltage band of the LCTs to an amount that was demonstrated as acceptable by analysis resulted in doubt on their capability to provide power to safety-related equipment during an accident. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the update to the procedures occurred in January and October 2007, after replacement of the LCTs. |
| 05000269/FIN-2016008-01 | 2015Q4 | Oconee | Potential lack of adequacy of the licensees maintenance program to detect substantial degradation of cables and their connections used on Oconee large oil filled stationary transformers | An URI was identified to determine if a performance deficiency exists regarding the adequacy of the licensees maintenance program to detect substantial degradation of cables and their connections used on the stations large oil filled stationary transformers. Description The inspectors developed an issue of concern related to the adequacy of the licensees maintenance program to detect substantial degradation of cables and their connections used on the stations large oil filled stationary transformers. The inspectors noted that all inspections required by the licensees surveillance and preventative maintenance programs used unaided visual inspections of bushings, surge arrestors, cable connections, T-connections, and cables on the stations large oil filled stationary transformers. The inspectors noted that the licensees metallurgical report associated with the failed power cable from the Unit 3 startup transformer identified degradation which likely occurred over a lengthy period of time. The inspectors determined that the following inspection activities should be pursued to determine if a performance deficiency exists: Review of the licensees completed cause determination Review of any additional testing and metallurgical reports Review of any license event report submitted by the licensee Review of requirements associated with emergency AC power paths and associated transformers This issue is identified as URI 05000269, 287/2016008-01, Potential lack of adequacy of the licensees maintenance program to detect substantial degradation of cables and their connections used on Oconee large oil filled stationary transformers. |
| 05000270/FIN-2014004-01 | 2014Q3 | Oconee | Review of FOD 50.59 Evaluation | An Unresolved Item (URI) was identified to review the licensees re-evaluation of the initial 50.59 evaluation for the Flood Outlet Device to determine if the performance deficiency is more than minor. In November 1998, the licensee identified that a HELB induced flood in the EPR could spread to other components in the Auxiliary Building (AB) and affect the ability of various safe shut down (S/D) equipment to perform its safety-related function as described in the Final Safety Analysis Report (FSAR). The licensee developed a modification package in April 2006, to install a Flood Outlet Device (FOD) which required a 50.59 evaluation. An initial 50.59 screening determined that the FOD modification did not require a detailed 50.59 evaluation. On August 21, 2006, the licensee conducted a review of the 50.59 screening and, as documented in PIP O-06-05726, ...were not able to conclusively determine if the correct conclusion had been made. A corrective action was identified in the corrective action document to perform an in-depth 50.59 screening and evaluation. The inspectors determined, through personnel interviews and review of documentation, that the licensee failed to perform this corrective action for a condition adverse to quality. The licensee is performing a revised 50.59 screening. The inspectors will evaluate the results of the screening to determine if a performance deficiency exists. This is identified as URI 05000270/2014004-01, Review of FOD 50.59 Evaluation. |
| 05000280/FIN-2012005-01 | 2012Q4 | Surry | Submerged Cables Identified in SAFETY-RELATED Manhole | The inspectors identified a Green non-cited violation of Technical Specification 6.4.A.7, which requires appropriate corrective maintenance procedures which would have an effect on the safety of the reactor. Specifically, Dominion procedure 0-MCM- 1207-01, Pumping of Security and Electrical Cable Vaults, was inadequate to prevent or detect submerged cables in a safety-related manhole, which is a performance deficiency. The inspectors determined that Dominion procedure 0-MCM-1207-01, Pumping of Security and Electrical Cable Vaults was inadequate to accomplish its intended purpose, which constitutes a performance deficiency in accordance with Technical Specification 6.4.A.7, which requires appropriate corrective maintenance procedures which would have an effect on the safety of the reactor. The inspectors determined that the finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance 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, this condition could lead to cable degradation, increased likelihood of cable failure, and subsequent risk associated with the failure of safety-related equipment. The inspectors screened this finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and IMC 0609, Appendix A, SDP for Findings At-Power, dated June 19, 2012 and determined the finding was of very low safety significance, Green, since it was a deficiency determined not to have resulted in the loss of operability or functionality of a single train for greater than its TS allowed outage time. The finding had a cross-cutting aspect in problem identification and resolution, corrective action program, P.1(c), because the corrective actions taken to address previous NRC identified concerns in the same manhole did not thoroughly evaluate the problem such that resolutions addressed the causes. |
| 05000287/FIN-2016003-02 | 2016Q3 | Oconee | Licensee-Identified Violation | Technical Specification (TS) 5.4.1., Procedures, states, in part, written procedures shall be established, implemented, and maintained covering activities described in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Procedure MP/0/A/3009/017, Visual PM Inspection and Electrical Motor Tests is used by the licensee during maintenance of electric motors. Contrary to the above, on April 25, 2016, the licensee did not adequately implement maintenance procedure MP/0/A/3009/017. Specifically, the licensee incorrectly wired the 3C RBCU motor control center contactor leads during maintenance causing 3C RBCU fan to operate in the reverse direction. On June 16, 2016, during an engineer walkdown, the engineer noted anomalies in the RBCU inlet temperature readings. On June 28, 2016, while investigating the temperature readings the licensee discovered that the 3C RBCU fan was operating in the reverse direction and declared the 3C RBCU inoperable. The 3C RBCU was inoperable when the plant entered Mode 4 on May 14, 2016 until June 28, 2016 when the 3C RBCU was repaired (approximately 45 days). Technical Specification 3.6.5, Reactor Building Spray and Cooling Systems, requires all three trains of RBCU operable while in Modes 1, 2, 3, and 4. On May 14, 2016, Unit 3 was starting-up from the refueling outage and entered Modes 4 through 1 with one train of RBCU inoperable. This action of changing modes with the 3C RBCU inoperable is prohibited by TS 3.0.4. The licensee entered this condition into their corrective action program as NCR 02041501. The licensee also restored 3C RBCU operability, trained/counseled technicians, and incorporated a procedure change which will enhance configuration control for the lifted leads aspect in the maintenance procedure for this activity. This finding was assessed using IMC 0609, Phase 1 screening worksheet of Attachment 4, Appendix A, and Appendix H, and was determined to be of very low safety significance (Green). |
| 05000296/FIN-2017004-02 | 2017Q4 | Browns Ferry | Failure to Perform an IDO without delay for 3A EDG after Observing Indications of a Degraded Condition | The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to perform an immediate operability determination (IDO) for 3A Emergency Diesel Generator (EDG) upon discovering a degraded condition. Specifically, on December 19, 2017, the licensee failed to perform an IDO after identifying and confirming less than minimum cooling flow, thus leaving the EDG in an indeterminate state of operability.The performance deficiency is more than minor because it was associated with the equipment performance attribute and affected the associated cornerstone objective to ensure availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. As a corrective action, the licensee performed operations to restore flow within the acceptable range and performed an IDO. The violation was entered into the licensee's CAP as CR 1370601. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of H.13, Consistent Process, because the performance deficiency was caused by not following a consistent, systematic approach to making a decision concerning operability of the affected DG. |
| 05000298/FIN-2018011-01 | 2018Q2 | Cooper | Failure to Correct Extent of Condition of Surge Suppression Varistor Failures | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, occurred when the licensee failed to correct conditions adverse to quality associated with the corrective actions identified in Condition Report RCR 2002-1665 to verify that installed surge suppressor varistors were appropriately sized and that design information was correctly reflected in controlled drawings for the reactor protection system, diesel generator control circuits, and high pressure coolant injection control circuits. |
| 05000298/FIN-2018011-02 | 2018Q2 | Cooper | Failure to Ensure Adequate Design Control Measures are in Place Associated with RHR Service Water Booster Pump Room Cooling | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to incorporate malfunctions of the residual heat removal (RHR) service water booster pump (SWBP) room cooling temperature switch, which could cause environmental changes leading to functional degradation of system performance, into the design basis to verify the necessary protection system action be retained. |
| 05000298/FIN-2018011-03 | 2018Q2 | Cooper | Inadequate Design Basis Calculation for the EDG Rooms Temperature Distribution | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for the licensees failure to ensure design control measures provide for verifying or checking the adequacy of design of the emergency diesel generator room ventilation system by use of alternate or simplified calculation methods, or by a suitable testing program. Specifically, the licensee incorrectly extrapolated the results of the test program, which led to an incorrect room temperature profile. Additionally, the design calculation did not assume potential failures of the CO2 dampers. |
| 05000298/FIN-2018011-04 | 2018Q2 | Cooper | Incorrect Classification of Potential Safety-Related Components | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the inspectors identified three examples of the licensees failure to properly classify potential safety-related components in the emergency diesel generator ventilation system and RHR service water booster pump room cooling systems. |
| 05000321/FIN-2012008-01 | 2012Q3 | Hatch | Failure to Report a Degraded Primary Safety Barrier per 10 CFR 50.72(b)(3)(ii)(A) | The team identified a non-cited violation of 10 CFR 50.72(b)(3)(ii)(A), for the licensees failure to provide an 8-hour event notification to the NRC for the plant being in a condition that caused a principal safety barrier to be seriously degraded. The licensee generated condition report 489079 to document the failure to provide the required 8-hour notification. The team determined that the failure to report a seriously degraded principal safety barrier as required by 10 CFR 50.72(b)(3)(ii)(A) was a performance deficiency. Using the guidance of Inspection Manual Chapter 0612, Appendix B, Issue Screening, the team determined the performance deficiency involved a violation that could have impacted the regulatory process, therefore, it was dispositioned using the traditional enforcement process. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, a failure to make a report required by 10 CFR 50.72 is a Severity Level IV violation. Cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000321/FIN-2012008-02 | 2012Q3 | Hatch | Failure to Adequately Account for Potential Pump Discharge Check Valve Back-leakage | The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify or check the adequacy of design of the plant service water system including the pump discharge check valves allowable backleakage. As a result, the licensee entered the issue into their corrective action program as condition report 481741, performed an immediate determination of operability, and placed administrative control over the river level at which the pumps are declared inoperable to a level higher than the one specified in the plants technical specifications until more detailed analyses could be performed. The limit was reduced back to the original technical specification level following the results of the analysis. The failure to verify the adequacy of the plant service water system design through calculational methods or through a suitable test program as required by 10 CFR 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was more than minor because it affected the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the reliability, availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not implement a suitable test program to verify design inputs and ensure the capability of the system. The inspectors used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, for mitigating systems and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined the finding to be of very low safety significance (Green) because the finding was a design control deficiency issue that did not result in a loss of operability or functionality of the PSW system. The performance deficiency was indicative of current licensee performance since the system hydraulic model was verified in 2011, and was directly related to the complete documentation and labeling cross-cutting aspect of the resources component in the area of human performance because the licensee did not have accurate design documentation for the potential pump discharge check valve leakage that could cause reverse rotation of the pumps. |
| 05000321/FIN-2012008-03 | 2012Q3 | Hatch | Failure to Ensure Adequacy of Intake Structure Ventilation Design | The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that the licensee failed to verify or check the adequacy of the design of the intake structure ventilation support function for the plant service water and residual heat removal service water systems. Following the teams discovery, the licensee performed a bounding analysis and verified that the safety related components in the intake structure would not fail under the worst case high temperature conditions. The licensee entered the issue into their corrective action program as condition report 477809 to address the issue. The failure to verify the adequacy of intake structure ventilation design through calculational methods or through a suitable test program as required by 10 CFR 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was more than minor because it affected the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the reliability, availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not have adequate measures in place to ensure negative effects due to heat loading did not affect the reliability, availability, and capability of intake structure equipment. The inspectors used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, for mitigating systems and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined the finding to be of very low safety significance (Green) because the finding was a design control deficiency issue that did not result in a loss of operability or functionality of the plant service water and residual heat removal service water systems. During the inspection, it was determined that there was adequate margin to preclude component failures when conservative heat loading and single failure criteria were assumed. No cross-cutting aspect was assigned to this finding because the failure to provide an adequate calculation or test is not indicative of current licensee performance due to the age of the heat load analysis. |
| 05000321/FIN-2012008-04 | 2012Q3 | Hatch | Failure to Incorporate Appropriate Test Acceptance Criteria to Assure Satisfactory Steady State EDG Performance | The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the licensees failure to incorporate adequate acceptance limits in surveillance test procedures used to verify acceptable steady state output voltage of the emergency diesel generators. The licensee performed an immediate determination of operability to verify that the emergency diesel generators would reach and maintain a steady state voltage greater than the minimum 3,860 volts determined by the calculation and issued interim administrative limits for acceptable output voltage until technical specifications can be revised. The licensee entered this issue into their corrective action program as condition report 482310 to address the issue. The licensees failure to include the correct minimum steady state output voltage as surveillance test acceptance criteria for the emergency diesel generators was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding challenged the assurance that the acceptance criteria used during surveillance testing would ensure the emergency diesel generators could perform their intended safety function and remain operable. In accordance with IMC 0609.04, Initial Characterization of Findings, the team used the mitigating systems column, which resulted in screening the finding through Inspection Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power. The finding was determined to be of very low safety significance (Green) because it was not a design deficiency resulting in the loss of functionality or operability, did not represent an actual loss of system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. A cross-cutting aspect was not identified because this issue has existed since the implementation of Improved Technical Specifications on March 3, 1995, and is not indicative of current licensee performance. |
| 05000321/FIN-2012008-05 | 2012Q3 | Hatch | Failure to Provide Appropriate Acceptance Criteria for EDG Air-Start System Check Valves | The team identified a finding for the licensees failure to follow Regulatory Guide (RG) 1.155, Station Blackout, guidance for testing and test control for the emergency diesel generator (EDG) air start system check valves. The testing deficiency was entered into the licensees corrective action program as condition reports 490288 and 490210. The failure to implement the guidance in RG 1.155, to which the licensee was committed in the stations Final Safety Analysis Report, was a performance deficiency. The performance deficiency was more than minor because it affected the procedure quality attribute of the Mitigating Systems Cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the capability of the EDGs to start following a station blackout coping period was not ensured by the licensees test acceptance criteria for the air start check valves. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, for mitigating systems and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined a detailed risk evaluation was required, because the finding represented an actual loss of function of a non-Technical Specification train of equipment designated as high safety significant in accordance with the licensees maintenance rule program for greater than twenty-four hours. A regional senior reactor analyst performed an analysis to determine the risk associated with the finding. An actual loss of EDG function following a station blackout would require all of the Unit 1 EDGs to fail to start, because if any Unit 1 EDG ran and was connected to either emergency bus, even for a relatively short time, an air compressor would partially or fully recharge the 1A EDGs air start tank. The calculation showed that the portion of plant risk that came from common cause fail to start of the Unit 1 EDGs, and of the sites EDGs was less than the threshold for greater than green for conditional core damage frequency or large early release frequency in the SDP. Therefore, the finding is Green. There was no cross-cutting aspect associated with this finding because the performance deficiency is not indicative of current licensee performance due to the age of the established test acceptance criteria for the check valve leakage. |
| 05000321/FIN-2015007-03 | 2015Q4 | Hatch | Failure to Assure that Class 1E Components were Qualified for Design Temperatures | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, for the licensees failure to ensure that adequate environmental test requirements were satisfied before relying on safety-related components to perform their intended safety functions. As an immediate corrective action, the licensee performed an operability evaluation and determined the components were operable. In addition, the licensee indicated that they planned to determine adequate corrective actions to restore full qualification of these commercial grade components, and entered this issue into their Corrective Action Program as Condition Report 10138133. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective, in that the licensee failed to verify the environmental qualification of safety-related components to ensure their performance up to the expected temperature of 150 degrees F. The finding was determined to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000321/FIN-2015007-04 | 2015Q4 | Hatch | Failure to Verify Design Basis Timing Margins for Safety Related Motor Operated Valves | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate if transients in control power voltage could affect the design basis margins for the timing of safety-related motor operated valves (MOVs). The licensee planned to perform corrective actions to ensure that the safety analysis remains bounded, and entered this violation into their Corrective Action Program as Condition Report 10138053. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective, in that the failure to evaluate transients that effect the timing margins for NOVs affected the established reliability and capability of the valves. The finding was determined to be of very low safety significance (Green) because the deficiency did not result in actual loss of safety function. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance (Section 1R21.2.b.4). |
| 05000324/FIN-2011004-01 | 2011Q3 | Brunswick | Inadequate Configuration Control Resulted in Rainwater Intrusion into the Unit 2 Reactor Building | A self-revealing Green non-cited violation of TS 5.4.1, Procedures, was identified for failure to implement procedural requirements of the equipment configuration control program to ensure that temporary power cables routed through an open manhole and into the reactor building north RHR (NRHR) room did not adversely impact the flood mitigation function of the storm drain system. This finding resulted in rainwater intrusion into the unit 2 reactor building. Upon discovery of this condition, the licensee resealed the manhole. The condition was entered into the licensees CAP as AR #483473. The failure to implement the requirements of the equipment configuration control program to ensure that the temporary cable routing did not adversely impact external flood protection features was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors - Flood Hazards and adversely affected the cornerstone objective in that the temporary change impacted the storm drain system which was credited for external flood protection. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Screening Worksheet, the finding screened as very low safety significance (Green) because it: (1) was not a design or qualification deficiency that was confirmed not to affect equipment operability; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; (4) did not represent a loss of risk significant non-Technical Specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event per table 4b of the worksheet because the leakage did not degrade the RHR system. The cause of the finding was directly related to the appropriately planning work activities cross-cutting aspect in the Work Control component of the Human Performance area because the licensee failed to incorporate environmental conditions which may impact plant structures, systems, and components into the temporary change. |
| 05000324/FIN-2011004-02 | 2011Q3 | Brunswick | Inadequate Corrective Actions for Control Building Air Conditioning Failures | The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criteria XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality related to the Control Room Air Conditioning (AC) system. Specifically, the licensee failed to identify and correct repetitive failures of nonconforming low ambient temperature damper actuators for the 2D control building air cooled condenser unit. This resulted in multiple control building AC refrigerant circuit failures. Upon discovery of the issue, the licensee placed the control building AC system in a safe condition for summer operation and initiated actions to procure acceptable damper actuators prior to the onset of low seasonal temperatures. The condition was entered into the licensees CAP as AR #462873. The inspectors determined that the licensees failure to promptly identify and correct the failures of the 2D control room AC system low ambient temperature damper actuators was a performance deficiency. This finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the finding reduced the reliability of the control building AC system and its ability to maintain control building equipment within specified temperature limits. The significance of the finding was evaluated using Phase 1 of the significance determination process in accordance with the Inspection Manual Chapter 0609 Attachment 4. The finding was determined to be of very low safety significance (Green) because the finding was a design or qualification deficiency that was confirmed not to affect equipment operability. The cause of this finding was directly related to the cross cutting aspect of thorough evaluation of problems in the Corrective Action Program component of the Problem Identification and Resolution area, because the licensee failed to promptly evaluate the failures of the low ambient temperature damper actuators and eliminate the adverse condition. |
| 05000324/FIN-2011004-03 | 2011Q3 | Brunswick | Inadequate Maintenance Results in Containment Isolation Valve Failure | A self-revealing Green finding was identified for inadequate maintenance on the overload relay of the unit 2 reactor water cleanup (RWCU) system inlet isolation valve 2-G31-F001. As a result of the inadequate maintenance, the overload relay actuated during operation of the valve under normal conditions, and the valve failed to shut. This was revealed while operators were attempting to isolate the RWCU system on August 2, 2011. After the valve failed to fully shut on August 2, 2011, the licensee shut the valve in series with 2-G31-F001 (2-G31-F004), repaired the overload relay for the 2-G31-F001 valve by installing the correct fasteners, returned the 2-G31-F001 valve to service, and entered the issue into their corrective action program (AR #480063). The inadequate maintenance on the 2-G31-F001 valve overload relay was a performance deficiency. The finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of structure, system, and component (SSC) and Barrier Performance, and it affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the finding prevented a primary containment isolation valve from shutting. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for Containment Barriers. The finding was determined to be of very low safety significance (Green) because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. The cause of this finding has no cross-cutting aspect because the maintenance took place in 1992 and is not indicative of current licensee performance. |
| 05000324/FIN-2011004-04 | 2011Q3 | Brunswick | Licensee-Identified Violation | Technical Specification 5.4.1, Procedures, requires that written procedures shall be implemented covering applicable procedures recommended in Regulatory Guide 1.33, Appendix A, November 1972 (Safety Guide 33, November 1972). Regulatory Guide 1.33, section I (Safety Guide 33, November 1972) requires written procedures for performing maintenance. Contrary to the above, the licensee identified that maintenance procedure 0CM-VFC500, Instructions for Repair, Reassembly, and Adjustment of the RCIC Terry Turbine Governor Valve, did not contain adequate guidance for assembling the unit 2 RCIC turbine governor valve. As a result, inadequate maintenance was performed on the unit 2 RCIC governor valve in 2009 in that proper spacing of the valve stem packing spacers was not maintained. This inadequate maintenance on the RCIC governor valve led to failure of the valve during quarterly surveillance testing on April 15, 2011. This finding was evaluated by the Regional Senior Reactor Analyst performing a Phase 3 significance analysis. The finding was determined to have a risk lower than 1E-6, and is GREEN. The short exposure time, and the availability of the severe accident mitigation alternative (SAMA) diesels for battery charging contributed to the low impact of the finding. The results were dominated by loss of the DC bus that powers HPCI, combined with automatic depressurization system (ADS) failures that could lead to high pressure core melt. External Events and Large Early Release Probability were found not to be major contributors to the risk of the finding. As corrective actions, the licensee revised the maintenance procedure and repaired the valve. This issue is in the licensees CAP as NCR #468283. |
| 05000324/FIN-2011004-05 | 2011Q3 | Brunswick | Licensee-Identified Violation | Technical Specification (TS) 3.3.6.1, Primary Containment Isolation Instrumentation, requires that the RWCU high differential flow instrumentation be operable in modes 1, 2, or 3. If the instrumentation is not operable, then TS 3.3.6.1 requires that the RWCU penetration flow path be isolated within 1 hour. Contrary to the above, the licensee identified that the RWCU high differential flow instrumentation was not operable and the penetration flow path was not isolated when the unit entered mode 1 on April 16, 2011 until August 2, 2011, because the RWCU inlet flow sensing element was installed backwards, causing the flow sensing element to be inaccurate. The resulting inaccuracy caused the instrumentation to be unable to isolate within the required TS limit of less than or equal to 73 gallons per minute differential flow. The finding was determined to be of very low safety significance per Appendix A of Inspection Manual Chapter 0609, Significance Determination Process, because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. Upon discovery of the condition, the licensee isolated the affected penetration flow path and installed the flow sensing element correctly. The issue is in the licensees CAP as NCR #479248. |
| 05000324/FIN-2013003-09 | 2013Q2 | Brunswick | Residual Heat Removal A Heat Exchanger Bypass Valve 2-E11-F048A Stud Failure | The inspectors are opening an URI to review the licensees evaluation of the operability of A RHR heat exchanger bypass valve 2-E11-F048A and determine if the performance deficiency associated with this issue is more than minor. On March 29, 2012, during Unit 2 refueling outage B221R1, maintenance personnel were going to repack A RHR heat exchanger bypass valve 2-E11-F048A. A member of maintenance hit one of the four valve yoke to bonnet hold down 134 studs with his foot and the stud sheared off at the nut. A second yoke hold down stud sheared off at the nut when maintenance personnel tried to remove the nut. The licensees corrective actions included replacing the four studs. The licensee determined the failure mechanism of the two studs was low stress, high cycle fatigue caused by vibration of the valve during throttling operations. The inspectors determined that the performance deficiency associated with this issue was the failure of the licensee to evaluate the effects of vibration on valve 2-E11-F048A when the valve was used for throttling, which resulted in the two studs sheering. The inspectors are opening an URI to review the licensees evaluation of the operability of valve F048A and determine if the performance deficiency is more than minor. The licensee entered this issue in the CAP as NCR 598294. This issue is being tracked as a URI: URI 05000324/2013003-09, Residual Heat Removal A Heat Exchanger Bypass Valve 2-E11-F048A Stud Failure. |
| 05000324/FIN-2013007-01 | 2013Q4 | Brunswick | Inadequate Acceptance Criteria for the Class 1E Station Battery Service Capacity Test Procedure | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to incorporate adequate acceptance criteria in the Class 1E station battery service test procedures. This failure to incorporate adequate acceptance criteria was a performance deficiency. The licensee entered this issue into their corrective action program as nuclear condition reports 632998 and 630621. The licensee performed a prompt determination of operability to verify that the batteries would be capable of supplying the necessary voltage to safety-related direct current loads at the required time intervals specified in design bases calculations. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, using an acceptance criterion of 105 volts direct current during the service test could result in incorrectly declaring a Class 1E station battery operable when greater terminal voltages, as specified in design bases calculations, were necessary for safety-related equipment to operate during the first minute of a design basis accident. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component which maintained its operability or functionality. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000324/FIN-2013007-02 | 2013Q4 | Brunswick | Inadequate DC System Calculations Three Examples | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with three examples, for the licensees failure to properly incorporate the design and licensing bases for the 125 volt direct current system into design calculations. This failure to properly incorporate the design and licensing bases for the 125 volts direct current system into design calculations was a performance deficiency. The licensee entered these issues into their corrective action program as nuclear condition reports 632998, 630621, 633538, and 633889. The licensee conducted a combination of prompt determinations of operability and engineering evaluations which provided reasonable expectation of operability of the direct current system pending final resolution. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether direct current system components would have adequate voltage to operate during design basis accidents. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component which maintained its operability or functionality. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000324/FIN-2013007-03 | 2013Q4 | Brunswick | Failure to Verify Adequacy of the Service Water Intake Structure Ventilation System | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the service water intake structure ventilation design through calculational methods. This failure to verify the adequacy of the service water intake structure ventilation design was a performance deficiency. The licensee entered this issue into their corrective action program as nuclear condition report 627708. The licensee performed a prompt determination of operability and implemented a number of compensatory actions to ensure safety-related components in the intake structure would not fail under the worst case high temperature conditions. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether safety-related components in the service water intake structure would be operable under design temperatures. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component which maintained its operability or functionality. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000324/FIN-2013007-04 | 2013Q4 | Brunswick | Failure to Scope Safety-related Components in the Maintenance Rule Program | The team identified a non-cited violation of 10 CFR 50.65(b)(1), for the licensees failure to scope the safety-related service water intake structure exhaust fan dampers into the Maintenance Rule program. This failure to scope safety-related service water intake structure exhaust fan dampers was a performance deficiency. The licensee entered this issue into their corrective action program as nuclear condition reports 630922, 627708, 630553, and 630993. The licensee has subsequently implemented corrective actions to include the dampers within the scope of the Maintenance Rule program. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, numerous dampers were found in degraded conditions such that effective control of performance or condition through appropriate preventive maintenance under 10 CFR 50.65(a)(2) could not be demonstrated. The team determined the finding to be of very low safety significance (Green) because the finding did not result in an actual loss of function of at least a single service water system train for greater than its technical specifications allowed outage time. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000324/FIN-2013007-05 | 2013Q4 | Brunswick | Failure to Follow Plant Procedure Directing the Performance of Preventive Maintenance on Safety-related Dampers | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow plant procedures specifying requirements for preventive maintenance of safety-related dampers. This failure to follow plant procedures was a performance deficiency. The licensee entered this issue into their corrective action program as nuclear condition reports 631376, 628132, 633710, and 631711. The licensee performed an immediate determination of operability to verify the as-found condition of the dampers did not affect operability of equipment inside the diesel generator building and implemented corrective actions to complete the missed preventive maintenance on the dampers. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the programmatic failure to perform preventive maintenance on the nine dampers resulted in decreased availability and reliability of the dampers such that multiple dampers were found in degraded conditions. The team determined the finding to be of very low safety significance (Green) because the finding did not result in an actual loss of function of at least a single emergency diesel generator for greater than its technical specifications allowed outage time. The team determined that this finding was associated with the cross-cutting aspect of Supervisory Oversight in the Work Practices component of the Human Performance area because Brunswick supervisors did not enforce the scheduled preventive maintenance nor did they ensure a justification for not performing preventive maintenance on safety-related components. |
| 05000324/FIN-2013007-06 | 2013Q4 | Brunswick | Inadequate Evaluation of Vibration on 2A RHR Heat Exchanger | A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to verify or check the adequacy of design of the Unit 2 A residual heat removal heat exchanger bypass valve 2-E11-F048A. This failure to verify or check the adequacy of design of the bypass valve was a performance deficiency. The licensee entered this issue into their corrective action program as nuclear condition report 598294. The licensees corrective actions included replacing the four valve yoke to bonnet hold down studs and initiating long term corrective actions to perform a design change to reduce vibration on the valve. The performance deficiency was determined to be more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, continued fatigue of the studs could have resulted in a more degraded state than the actual as-found condition, which could have affected the ability of the valve to operate for its safety-related function. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component which maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000327/FIN-2016007-01 | 2016Q2 | Sequoyah | Failure to Implement the Design Change Process when Modifying the Safety-Related Fire Dampers | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to use the design change process to make modifications to the Emergency Diesel Generator EDG room inlet dampers as required by NPG-SPP-9.3, Plant Modifications and Engineering Change Control. The licensee entered the issue into the corrective action program and implemented compensatory measures, while implementing plans to modify each of the affected inlet and exhaust fire dampers. This performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee modified the dampers to include the wrong brackets, which could adversely affect the dampers ability to remain open to provide cooling during EDG operation and support EDG reliability and availability. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000327/FIN-2016007-02 | 2016Q2 | Sequoyah | Failure to Install Safety-Related Components that are Designed to Withstand the Effects of a Design Basis Tornado | The NRC identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to install emergency diesel generator components that could withstand the effects of a design basis tornado as required by Section 3.1.2 of the Update Final Safety Analysis Report (UFSAR). The licensee entered the issue into the corrective action program and implemented compensatory measures to protect the affected components. This performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the capability of the equipment to withstand the effects of a tornado was not ensured. The team determined the finding to be of very low safety significance (Green) because of the low frequency of tornados/high winds and the potential for recovery by the operators on site. This finding was not assigned a crosscutting aspect because the issue did not reflect present licensee performance. |
| 05000327/FIN-2016007-03 | 2016Q2 | Sequoyah | Inadequate Monitoring of the 480V Shutdown Transformers | The NRC identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have documented procedures in place to ensure effective monitoring of the 480V Shutdown Transformers as required by Section 5.3.2.(4) of IEEE 308-1971. The licensee entered the issue into the corrective action program and planned to put additional transformer testing/monitoring in place to detect degradation prior to equipment failure. This performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to perform adequate maintenance on the shutdown transformer, which could result in the inability to detect the deterioration of the shutdown transformer toward an unacceptable condition. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect present licensee performance. |
| 05000327/FIN-2016007-04 | 2016Q2 | Sequoyah | Failure to Energize Hydrogen Igniters during Extended Station Blackout | The NRC identified a finding (FIN) for the licensees failure to meet their docketed commitment to revise the back-up generators to include supplying one train of containment hydrogen igniters per unit in response to Generic Safety Issue 189, Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen Combustion During a Severe Accident. The licensee entered this issue into their corrective action program and completed immediate corrective actions to revise procedure FSI-5.01, Initial Assessment and Flex Equipment Deployment, Rev. 0, to ensure the hydrogen igniters would be energized during an extended station blackout (SBO) event. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to energize the hydrogen igniters during an extended SBO event could result in containment failure. The team determined the finding to be of very low safety significance (Green) because the risk was mitigated by the low frequency of SBO conditions and the high likelihood of operator recovery given the obvious diagnosis of the performance deficiency. The team determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of operating experience (OE), in the area of Problem Identification and Resolution, because the licensee failed to effectively collect, evaluate, and implement relevant internal OE before implementing their new station procedures to use the FLEX diesels as the power supply to the hydrogen igniters. |
| 05000335/FIN-2015007-01 | 2015Q1 | Saint Lucie | Failure to Submit a License Amendment Request for Unit 1 RPS | An NRC-identified severity level IV (SL IV) non-cited violation (NCV) of 10 CFR 50.59(c)(2)(ii) and an associated finding of very low safety significance (Green) was identified for the licensees failure to obtain a license amendment prior to implementing a change to the Unit 1 reactor protective system (RPS). The failure to obtain a license amendment for the change resulted in the implementation of a modification that did not conform with the licensees current licensing basis. The licensees failure to obtain NRC approval prior to implementing the change to the Unit 1 RPS was determined to impact the regulatory process because the change required NRC review and approval prior to implementation. The licensee entered this issue into their corrective action program as action requests (ARs) 2029652 and 2030820, planned to restore the RPS configuration into conformance, and performed a prompt operability determination which concluded that there was a reasonable expectation that the RPS channels remained operable and could perform their required design basis functions. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the modification did not ensure the reliability of the RPS to respond to a design basis event because the design requirements for physical separation of RPS channels A and C were not met and resulted in a condition where revision or rework would be required to resolve the physical separation concerns. The team determined the finding to be of very low safety significance (Green) because the finding did not affect a single RPS trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators. The traditional enforcement violation was evaluated using the NRC Enforcement Policy dated January 28, 2013, and revised February 4, 2015. The inspectors determined the violation was SL IV per Section 6.1.d.2 because the associated finding was evaluated by the SDP as having very low safety significance (i.e., Green). The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of change management, in the area of human performance, because the licensee did not use a systematic process for evaluating and implementing a change such that nuclear safety remained the overriding priority. |
| 05000335/FIN-2015007-02 | 2015Q1 | Saint Lucie | Failure to Establish Appropriate Procedural Limitations to Prevent Exceeding Non-LOCA Event Analysis Assumptions for Steam Generator Blowdown Flow Rate | An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to assure that design basis assumptions for steam generator blowdown (SGBD) flow rate were translated into procedural guidance. Specifically, procedures 1-NOP-23.02 and 1-AOP-09.03 for Unit 1, and 2-NOP-23.02 and 2-AOP-09.03 for Unit 2, allowed SGBD flow rates significantly in excess of the assumed values in non-loss of coolant accident (LOCA) event analyses. The licensee entered the issue into their corrective action program as action requests (ARs) 2030177, 2031217, and 2031218. The licensees immediate corrective actions included performing a functionality assessment of the SGBD systems for both units, which included; re-performing the event analyses, issuing an operations department night order to temporarily provide operators appropriate direction for limiting the SGBD system flow, and plans to update the analyses of record, plant procedures, and the UFSAR with new system limitations. The performance deficiency was determined to be more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of the secondary side heat removal systems to respond to design basis non-LOCA events because analysis assumptions were not translated into procedural limitations for the SGBD system. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The inspectors determined that the issue was indicative of present licensee performance because the analyses were performed in 2013. The finding was associated with the cross-cutting aspect of design margins, in the area of human performance, because the organization did not operate and maintain equipment within design margins. |
| 05000335/FIN-2015007-03 | 2015Q1 | Saint Lucie | Adequacy of 10 CFR 50.59 Screening Performed for Unit 1 SGBD Maximum Flow Evaluation Test | An unresolved item (URI) was identified regarding the adequacy of a 10 CFR 50.59 screening that was completed for the performance of a test on the Unit 1 SGBD system. A violation of 10 CFR 50.59(d)(1) was identified for the licensees failure to perform a full written 10 CFR 50.59 evaluation which provided the basis that the test or experiment did not require a license amendment. Specifically, the test introduced operating conditions that were inconsistent with the analyses described in the stations UFSAR, and a full 10 CFR 50.59 evaluation was not performed. The URI is being opened to provide for additional inspection of the licensees past operability evaluation of the test conditions, and corresponding event re-analyses, to determine if the violation of 10 CFR 50.59 was more than minor. On November 11, 2011, the licensee performed a test using procedure 1-LOI-23.01, Steam Generator Blowdown Maximum Flow Evaluation Test, Rev. 1. During the test, SGBD flow was increased to 160 gpm on each steam generator. Prior to the performance of the test, a 10 CFR 50.59 screening was performed for the activity, which determined that the proposed activity did not involve a test or experiment not described in the UFSAR, where an SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC or is inconsistent with analyses or descriptions in the UFSAR. The inspectors determined that at the time the 10 CFR 50.59 screen was completed, Chapter 15 of the UFSAR identified that the assumed SGBD flow rate during the loss of normal feedwater event was 40 gpm per steam generator. Another event involving a loss of feedwater with no AFW flow, described in UFSAR Chapter 10, identified that the SGBD flow rate was assumed to be 35 gpm. The inspectors determined that the SGBD flow rate of 160 gpm allowed by 1-LOI-23.01 was inconsistent with the UFSAR analyses assumptions for the SGBD system. Following the inspectors identification of the discrepancy, the licensee planned to evaluate the test conditions to determine if analysis acceptance criteria could be met when the SGBD flow rate input was increased to values allowed during the test. Additional inspection of this re-analysis is needed to determine if the full 10 CFR 50.59 evaluation, had it been performed, would have concluded that a license amendment should have been pursued prior to implementing the activity. This issue will be identified as URI 05000335/2015007-03, Adequacy of 10 CFR 50.59 Screening Performed for Unit 1 SGBD Maximum Flow Evaluation Test. |
| 05000335/FIN-2016008-01 | 2016Q1 | Saint Lucie | Failure to Consider Elevated Temperature Effects on MOV Actuator Output Capability | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to consider the impact of elevated ambient temperatures on motor operated valve (MOV) actuator output. The licensee entered the issue into the corrective action program and also evaluated the elevated ambient temperature effects on several affected station MOVs and determined the MOVs remained operable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of several MOVs scoped into their MOV program because they did not consider reduced actuator output torque due to elevated temperatures. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was assigned a cross-cutting aspect of Evaluation in the Problem Identification and Resolution Area because the finding was indicative of current licensee performance, and the licensee did not thoroughly evaluate the issue identified in AR 2030822, such that the design issue of accounting for elevated temperature was resolved (P.2). |
| 05000335/FIN-2016008-02 | 2016Q1 | Saint Lucie | Failure to Update UFSAR to Reflect Station Blackout Coping Time Basis | The NRC identified a non-cited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, for the licensees failure to update the Updated Final Safety Analysis Report (UFSAR) to reflect the offsite power design characteristic group and emergency alternating current power configuration group for station blackout coping duration. The licensee entered the issue into the corrective action program in order to update the information. The failure to update the UFSAR was dispositioned using the traditional enforcement process because it had the potential to impact the regulatory process. The team determined the violation was more than minor because not accurately classifying the offsite power design characteristic group and emergency alternating current power design characteristic group could have a material impact on licensed activities. The team determined the violation to be a Severity Level IV violation because the lack of upto- date information has not resulted in any unacceptable change to the facility or procedures. This violation was not assigned a cross-cutting aspect because crosscutting aspects are not assigned to traditional enforcement violations. |
| 05000335/FIN-2016008-03 | 2016Q1 | Saint Lucie | Inadequate Testing of 125VDC MCCBs | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to perform testing for safety-related 125 volts direct current (VDC) molded case circuit breakers (MCCBs) to detect deterioration. The licensee entered the issue into the corrective action program and plans to make changes to the procedure to ensure deterioration of the safety-related 125VDC MCCBs is adequately detected. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, cycling the breakers multiple times before electro-mechanical testing could mask degradation of the circuit breakers and thus decrease the reliability of the breakers to perform their safety function when called upon. The team determined the finding to be of very low safety significance (Green), because it was not a deficiency affecting the design or qualification of a structure, system, or component which did not maintain its functionality; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) allowed outage time or two separate safety systems out-of-service for greater than its TS allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000335/FIN-2016008-04 | 2016Q1 | Saint Lucie | Failure to Identify Degraded Condition of Unit 1 Electrical Equipment Room Supply Fan Gravity Dampers | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify a condition adverse to quality, which prevented the Unit 1 electrical equipment room (EER) supply fan dampers from performing their safety-related function to close. The licensee entered the issue into their corrective action program and implemented compensatory measures to prevent reverse flow of air through the degraded dampers in the event of a failure of their supply fan. This compensatory measure will remain in place until the licensee is able to replace both gravity dampers. This performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inability of the gravity dampers to close upon failure of one of the supply fans would result in room temperatures above the design temperature of 104oF. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000335/FIN-2016008-05 | 2016Q1 | Saint Lucie | Failure to Verify the Adequacy of Design of Unit 1 Electrical Equipment Room Ventilation System | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the Unit 1 electrical equipment room (EER) ventilation system design when performing a design calculation. The licensee entered the issue into the corrective action program and plans to re-balance flow rates in the EERs or revise the equipment qualification temperatures for equipment located in the EERs. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the re-analysis of the ventilation system resulted in a reduction in temperature margin, which could impact the reliability and capability of emergency electrical equipment in the EERs. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000335/FIN-2016008-06 | 2016Q1 | Saint Lucie | Failure to Provide a Missile-Protected Intertie | The NRC identified a finding for the licensees failure to properly provide a completely missile-protected intertie from the Unit 1 diesel oil transfer pumps to the Unit 2 diesel oil storage tanks. The licensee entered the issue into the corrective action program. The performance deficiency was determined to be more than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone objective which of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, a postulated tornado missile could fail the unprotected section of piping, rendering the intertie unable to complete its intended function, thereby reducing the licensees capability to mitigate a design basis tornado event. The team determined the finding to be of very low safety significance (Green) because it did not involve the total loss of any safety function, nor was it identified by the licensee through probabilistic risk assessment, Individual Plant Evaluation of External Events (IPEEE), or similar analysis that would have contributed to external event initiated core damage accident sequences. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |