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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000269/FIN-2018003-02 | 2018Q3 | Severity level IV | NRC identified | Failure to Make a 60 Day Notification of an Actuation of an Emergency AC Power Source | An NRC identified SL IV violation of 10 CFR 50.73(a)(2)(iv) was identified for the licensees failure to make a required 60-day notification. Keowee Hydro Unit 2 automatically started on May 7, 2018, following an electrical lockout of the Oconee Nuclear Station Unit 3 startup transformer. | |
| 05000269/FIN-2018003-01 | 2018Q3 | Severity level IV | NRC identified | Failure to Maintain the Effectiveness of the Emergency Plan | The inspectors identified a Severity Level IV (SL IV) NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(2), for the licensees failure to maintain the effectiveness of the Oconee Nuclear Station (ONS) Emergency Plan (E-Plan). Specifically, from December 2014 until January 2018, the licensee failed to perform an adequate 10 CFR 50.54(q) evaluation for their E-Plan when a corresponding change was made to their Protective Action Recommendation (PAR) Emergency Plan Implementing Procedure (EPIP). | |
| 05000287/FIN-2018003-03 | 2018Q3 | Severity level IV | Self-revealing | Main Steam Relief Valve As-Found Lift Pressure Prohibited by Technical Specifications | A self-revealed SL IV NCV of TS 3.7.1, Main Steam Relief Valves (MSRVs), was identified when a routine lift pressure test revealed that two of sixteen main steam relief valves were higher than allowed by TS SR 3.7.1.1 for a duration that was longer than the conditions TS required action completion time. | |
| 05000269/FIN-2018002-02 | 2018Q2 | Severity level IV | NRC identified | Failure to Coordinate a No-later-than Arrival Time for the Shipment of a Category 2 Quantity of Radioactive Material | The inspectors identified aSeverity Level IV NCV of 10 CFR 37.75(b) when the licensee failed to coordinate a no-later-than arrival time for a Category 2 shipment of radioactive material. Specifically, the licensee failed to recognize that a package of primary resin contained a Category 2 quantity of Cobalt-60 prior to shipment, and therefore failed to arrange a no-later-than arrival time with the receiving licensee. | |
| 05000287/FIN-2018002-01 | 2018Q2 | Green | P.5 | NRC identified | Failure to Perform ISI General Visual Examinations of Containment Moisture Barrier Associated with Containment Liner Leak Chase Test Connection Piping | The inspectors identified a Green NCV of 10 CFR Part 50.55a, Codes and Standards, involving the licensees failure to properly apply Subsection IWE, of ASME Section XI, for conducting general visual examinations of the leak chase test connection piping at the concrete floor interface which provides a moisture barrier to the containment liner seam welds. |
| 05000269/FIN-2018013-02 | 2018Q1 | TBD | NRC identified | Failure to Submit for License Review andObtain a License Amendment for a Modification | The licensee procedure Nuclear System Directive (NSD): 209 10 CFR 50.59 Process, committed to using Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation.The guidance in Nuclear Energy Institute (NEI) 96-07 Section 4.3.2, specified that if a change in likelihood of occurrence of a malfunction increases by more than a factor of two would need NRC approval, because certain changes that satisfy the factor of two limit exceed the minimal increase standard for accident/transient frequency under criterion 10 CFR 50.59(c)(2)(i). The guidance in NEI 96-07 Section 4.3.8, specified that the use of new or different methods of evaluation that are not approved by NRC for the intended application, such as the methods identified in the memo to File, ME Patrick (PJ North), dated 1/12/92, Single Failure Timing Licensing Basis, no file number given. (Note: Memo was actually written 1/12/93), would need NRC approval, because it was considered a departure from a method of evaluation described in the UFSAR. Based on this guidance, the team determined that the modifications associated with engineering changes (ECs), EC91880, Keowee Emergency Start Cable, revision 24 and EC91875, Keowee AC Power Supply Tie-Ins, revision 15, and EC91874, 13.8 KV Feed To PSW System from 100 KV APS, revision 7 would require NRC approval in accordance with 10 CFR 50.59(c)(2)Corrective Actions: TBDCorrective Action References: TBD | |
| 05000269/FIN-2018013-01 | 2018Q1 | TBD | NRC identified | Failure to Translate Design and Licensing Basis Requirements and Verify Adequate Design | The licensee did not correctly translate site design and licensing bases into the site specifications and procedures for the design and installation of plant modifications that included the re-configuration of electrical cables in electrical cable trench #3 between the Keowee Hydro Station (KHS) and transformer CT-4 at Oconee Nuclear Station (ONS) and the Protected Service Water (PSW) ductbank between CT-4 and the PSW building. The specific requirements of IEEE 279-1968 and single failure sections of IEEE 279-1971 were not fully implemented. Contrary to this requirement, the licensee placed Class 1E 125Vdc system cables adjacent to various medium voltage-high energy alternating current (ac) power distribution cables for the offsite and onsite power systems and introduced credible single failure conditions with the potential for exposure of the onsite redundant Class 1E dc power distribution and control systems (dc systems) to possible damaging peak voltage from the offsite and onsite AC power systems. Corrective Actions: The licensee reported this as an unanalyzed condition to the NRC in accordance with 10 CFR 50.73(a)(2)(ii) (B) in Licensee Event Report 269/2014-01 entered this issue into their corrective action program. The licensee also performed immediate and prompt determinations of operability in which they concluded a reasonable expectation of operability existed on the basis that the consideration of the specific hazards was not required by the site licensing basis. A number of plant modifications were implemented to address the concerns.Additional inspections of these corrective actions will be conducted as appropriate. For the limited areas where the concerns could not be addressed, on February 28, 2018, (ML180051B257) the NRC granted relief from the applicable Code and concluded that the proposed alternatives provided an acceptable level of quality and safety for the cable configurations and locations.Corrective Action Reference: PIP O-14-03190, PIP O-14-05125, PIP O-14-03915, and PIP O-14-02965 | |
| 05000269/FIN-2017004-01 | 2017Q4 | Green | H.5 | Self-revealing | Failure to Identify Sensitive Equipment During Modification Results in Loss of Safety Function | A self-revealing Green non-cited violation (NCV) of Oconee Nuclear Station Technical Specification (TS), Section 5.4, Procedures, was identified for the licensees failure to identify sensitive equipment in a work area that warranted implementation of compensatory measures as required by station procedure AD-EG-ALL-1180, Engineering Change (EC) Walkdowns. During the design and planning phase of a station modification, the licensee failed to identify sensitive components located in the subject work area and subsequently failed to implement adequate protective measures as defined in station procedures to prevent plant impacts during modification installation. The licensee entered this issue into their corrective action program (CAP) as nuclear condition report (NCR)02131608 and implemented corrective actions to identify other positionable components required for emergency power source operability that would require the use of protective measures, as defined by AD-OP-ALL-0204, Plant Status Control, in order to prevent inadvertent operation. The licensee created a formal Engineering department communication which included lessons learned from the event and familiarization with the EC walkdown checklist. The signs on the governor actuator cabinets were also revised to emphasize the sensitive nature of the equipment. The licensees failure to properly identify sensitive equipment and implement compensatory measures to prevent plant impacts as required by station procedure AD-EG-ALL-1180 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the human performance 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 performance deficiency resulted in the loss of the emergency AC power path function for 11 hours and 31 minutes. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. Inspection Manual Chapter 0609, Appendix A required a detailed risk evaluation because the finding represented a loss of system and/or function. A regional senior reactor analyst (SRA) performed the detailed risk evaluation using SAPHIRE Version 8.1.6 and a modified Version 8.50 of the SPAR Model for Oconee. The SRA developed two change sets to model the total exposure time for the finding. The first simulated a common cause failure of both Keowee units with an exposure time of 7 hours. The second simulated the failure of both Keowee units while the standby buses were energized by the Lee Station for 5 hours. The result was less than 1E-6 for each Oconee unit, which would be a finding of very low significance (Green). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the area of human performance, in that the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5) |
| 05000287/FIN-2017004-02 | 2017Q4 | Green | H.5 | Self-revealing | Failure to Properly Risk Screen Work Within Two Feet of a Single Point Vulnerability Component | A self-revealing Green NCV of Oconee Nuclear Station TS, Section 5.4, Procedures, was identified for the licensees failure to identify and properly risk screen work within 2 feet of a single point vulnerability (SPV) component in accordance with procedure AD-OP-ALL-0201, Protected Equipment. Specifically, the transmission and Oconee organizations failed to recognize that planned maintenance on a breaker in the 525 kilovolt (kV) switchyard was within 2 feet of an SPV component and, as a result, appropriate planning and oversight were not in place to prevent a plant trip during maintenance activities. The licensee entered this issue into their CAP as NCR 02138958. Corrective actions included revisions to station and transmission procedures to ensure inclusion of appropriate SPV program information, addition of the SY special emphasis code to all switchyard type work which require coordination of transmission resources, and the addition of the T1 trip/transient risk special emphasis code to all breaker failure relays in the 230 kV and 525 kV switchyard cabinets containing SPV components.The licensees failure to identify and properly risk screen the planned maintenance on PCB-57 as work within 2 feet of an SPV component in accordance with AD-OP-ALL-0201 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, human errors led to a Unit 3 main generator lockout, which resulted in a reactor trip. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. The inspectors determined the finding was of very low safety significance (Green) because the finding did not represent a transient initiator that caused both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (i.e. loss of condenser, loss of feedwater). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the human performance area, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5) |
| 05000269/FIN-2017001-02 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.3.8, PAM Instrumentation, requires CHRRMs, RIA-57 and RIA-58, to be operable in Modes 1, 2, or 3. Contrary to the above, from 1998 to October 2016, the licensee failed to maintain operability of the CHRRMs for all three units when they failed to provide reasonable assurance that the CHRRMs would provide accurate measurement of containment radiation levels during a HELB event in the east penetration room of the affected unit(s). The CHRRMs are utilized in the Oconee site emergency plan and implementing procedures to support assessment of the severity of an accident. The performance deficiency was determined to be more than minor because it was associated with the facilities and equipment attribute of the emergency preparedness cornerstone and adversely affected the cornerstone objective to ensure the licensees capability to implement adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued September 22, 2015, and determined the finding was of very low safety significance (Green) because no planning standard function failure occurred due to the availability of other parameters that could be used to validate the indications from the CHRRMs. The licensee has entered this issue into their corrective action program as NCRs 02069527 and 02077587. | |
| 05000269/FIN-2017001-03 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | Oconee Nuclear Station Technical Specification 3.0.4 requires that when a limiting condition of operation is not met, entry into a mode or other specified condition in the applicability shall not be made except when the associated actions to be entered permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time. Oconee Nuclear Station Technical Specification 3.3.7, Engineered Safeguards Protective System (ESPS) Automatic Actuation Output Logic Channels, requires eight ESPS automatic actuation output logic channels to be operable in Modes 1 and 2 and Modes 3 and 4 when associated ES equipment is required to the operable. Contrary to the above, Oconee Nuclear Station Unit 1 entered Mode 4 on November 24, 2016 with ES protective system voters 1 and 2 in an abnormal configuration (bypassed) for the plant mode of operation. Operations shift personnel discovered this abnormal configuration on November 25, 2016 and restored voters 1 and 2 to an operate condition which met Technical Specification 3.3.7. This failure to maintain ESPS channels in the correct mode of operation for the required mode of applicability was a performance deficiency and was determined to be more than minor. The issue is more than minor because it was associated with the configuration control attribute of the mitigating system 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 (i.e., core damage). Specifically, the issue challenged the configuration control attribute of ensuring operating equipment was available to respond to initiating events. The inspectors used IMC 0609, Att. 4, Initial Characterization of Findings, issued October 07, 2016, and IMC 0609, Appendix A, Significance Determination Process for Findings at Power, issued June 19, 2012, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time. The licensee has entered this issue into their corrective action program as NCR 02081523. | |
| 05000269/FIN-2017001-01 | 2017Q1 | Green | H.1 | NRC identified | Failure to Comply with 10 CFR 55.49 | Green: A Green NRC-identified non-cited violation (NCV) of 10 CFR 55.49, Integrity of Examinations and Tests, was identified because the licensee engaged in an activity that compromised the integrity of examinations. Specifically, the licensee failed to ensure that current week simulator scenarios could not be predicted based on the previous weeks simulator scenarios during the annual operating exams required by 10 CFR 55.59, Requalification. While inspecting the annual operating examination schedules for the required simulator examinations for 2016 and 2017, the inspectors identified that one of the two scenarios that were administered during a single week of the annual exam cycle could be predicted for administration the following week. The licensee did not implement any immediate corrective actions because the exams were completed and there was no evidence of compromise. The licensee documented the issue in nuclear condition report (NCR) 2114313. This performance deficiency was more than minor because it was associated with the human 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, using predictable exam development and administration techniques adversely affected the integrity of the administration of the operating exams, which test licensed operator performance in order to ensure timely and correct mitigating actions during an event. Using the Licensed Operator Requalification Significance Determination Process, this finding was determined to be of very low safety significance (Green) because no known compromise of the examinations occurred. The inspectors determined the finding had a cross-cutting aspect of resources in the cross-cutting area of human performance because the licensee failed to ensure that adequate training procedures were available to meet industry standards and ensure that the potential for the compromise of regulatory examinations did not exist. (H.1) |
| 05000269/FIN-2017007-01 | 2017Q1 | Green | H.13 | NRC identified | Failure to Identify and Correct Broken Cable Trench Cover. | Green. The NRC identified a non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to assure that a condition adverse to quality associated with a damaged trench cover on the yellow trench was identified and corrected. Specifically, the seismic design function of the trench cover was not identified or recognized at the time of the licensees original identification of the issue and subsequent NCR generation, and, due to this error, appropriate corrective actions were not assigned or completed. In response to the issue, the licensee replaced the broken trench cover on the yellow trench with a temporary cover on March 22, 2017 , and planned work order 20147282 to replace it with a permanent cover to restore the design configuration . This performance deficiency was more than minor because it was associated with the Design Control Attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inadequate identification and correction of this condition adverse to quality adversely impacted the trench covers reliability and capability to perform its function during and following a seismic event. 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 (SSC), and the SSC maintained its operability or functionality. The team determined that the finding was indicative of current licensee performance, because the issue was first identified in January 2017. A cross -cutting aspect of Consistent Process (H.13.) in the Human Performance Area was assigned because individuals did not use a consistent, systematic approach to make decisions. |
| 05000269/FIN-2016004-01 | 2016Q4 | Green | NRC identified | 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 | Green | NRC identified | 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-2016405-01 | 2016Q3 | Green | NRC identified | Security | ||
| 05000269/FIN-2016003-01 | 2016Q3 | Green | H.6 | NRC identified | 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. |
| 05000287/FIN-2016003-02 | 2016Q3 | Green | Licensee-identified | 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). | |
| 05000287/FIN-2016002-04 | 2016Q2 | Severity level IV | NRC identified | Failure to Make a Non-Emergency Eight Hour Notification of a Loss of Safety Function | An NRC-identified Severity Level IV NCV of 10 CFR 50.72(b)(3)(v) was identified for the licensees failure to make a required non-emergency eight hour notification for a loss of the emergency AC power path function. On December 7, 2015 Oconee Nuclear Station Unit 3 experienced a loss of the emergency AC power path function for approximately 21 minutes. The licensee entered this issue into their corrective action program as NCR 01981762 and will evaluate their internal reportability procedures regarding the time of discovery. The failure to make an eight hour non-emergency report for a loss of the emergency AC power path function per 10 CFR 50.72(b)(3)(v) was a performance deficiency. This performance deficiency impacted the ability of the NRC to perform its regulatory oversight function and was dispositioned using traditional enforcement. This violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy, revised February 4, 2015. Using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72, the issue was determined to be a Severity Level IV violation. In accordance with IMC 0612, because this violation involved traditional enforcement and does not have an underlying technical violation that would be considered more than minor, a cross-cutting aspect was not assigned to this violation. | |
| 05000269/FIN-2016002-02 | 2016Q2 | Green | H.8 | NRC identified | Failure to Properly Control Transient Combustible Materials in the Oconee Main Control Rooms | An NRC-identified Green non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2, and 3 Renewed Facility Operating License Condition 3.D, Fire Protection, was identified for the licensees failure to adequately implement the requirements of the transient combustible material program. Specifically, the licensee failed to control the storage of transient combustible material in the Oconee main control rooms with the proper evaluation in accordance with procedure AD-EG-ALL-1520, Transient Combustible Control, Attachment 3, Allowed Combustible Materials in Level B and Level C Areas. The licensee removed the stored items from each of the main control rooms and entered this issue into their corrective program as nuclear condition reports (NCRs) 02012091, 02012290, and 02013990. The licensees failure to control the storage of transient combustible material in the Oconee main control rooms with the proper evaluation in accordance with procedure AD-EG-ALL-1520 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors 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 (i.e., core damage). Specifically, uncontrolled transient combustibles challenge the habitability requirements of the main control room in the event of a fire and the ability of licensed operators to respond to events using the systems designed to prevent undesirable consequences. The finding was screened in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings and IMC 0609 Appendix F, Fire Protection Significance Determination Process Task 1.3.1, and determined to be of very low safety significance (Green) because the finding did not prevent the reactor from reaching and maintaining a safe shutdown condition. The finding was determined to have a cross-cutting aspect of procedure adherence in the human performance cross-cutting area because the licensee failed to implement the requirements of station procedure AD-EG-ALL-1520, Transient Combustible Control. |
| 05000287/FIN-2016002-01 | 2016Q2 | Green | NRC identified | Failure to Perform ISI General Visual Examinations of Containment Moisture Barrier | An NRC-identified Green NCV of 10 CFR Part 50.55a, Codes and Standards, was identified for the licensees failure to conduct 100 percent general visual examinations of the moisture barriers to the containment liner in accordance with Subsection IWE of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. Specifically, the licensee failed to conduct visual examinations of the sealant applied to interior expansion joint locations in containment. In response, the licensee repaired the identified moisture barriers and confirmed the operability of the containment liner with the satisfactory results of the containment integrated leak rate test. The licensee entered this issue into their corrective action program as NCR 02027086. The failure to conduct a general visual examination of 100 percent of the moisture barriers intended to prevent intrusion of moisture against inaccessible areas of the containment liner was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inspectors determined that this finding was of more than minor significance because the failure to conduct required visual examinations and identify the degraded moisture barriers, which could allow the intrusion of water, if left uncorrected, had the potential to lead to a more significant concern. The inspectors used IMC-0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 3 Barrier Integrity Screening Questions, and determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined no cross-cutting aspect was associated with this finding because the finding was not reflective of present licensee performance. | |
| 05000287/FIN-2016002-03 | 2016Q2 | Green | P.2 | Self-revealing | Degraded power cables result in inoperable startup transformer and loss of Unit 3 safety function | A self-revealing Green violation of Oconee Technical Specification 5.4, Procedures, was identified for the licensees failure to establish adequate procedures to detect degradation of the startup transformer power cables. Station procedure IP/0/A/2400/002, Substation Insulators, Lighting Arrestors, CCVT, Transformer Drop Down Line, Bus Inspection and Maintenance, lacked sufficient detail for maintenance personnel to properly inspect power cables for cracks and fraying. This allowed undetected degradation of the Oconee startup transformer power cables to develop causing the Unit 3 startup transformer to become inoperable. The licensee performed repair activities on the degraded power cables to remove areas where strands of the power cables were severed and re-established proper connections. Also, the licensee created work orders in their work management process to replace the drop down lines on the Unit 1 and Unit 3 startup transformers. The licensee entered this issue into their corrective program as NCR 01733811. The licensees failure to establish an adequate procedure to detect degradation of startup transformer power cables during periodic maintenance was a performance deficiency. 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the power cable failure caused inoperability of the Unit 3 startup transformer. The finding was screened in accordance with IMC 0609, Significance Determination Process, Attachment 4 and Appendix A and determined to require a detailed risk evaluation. A senior reactor analyst performed a detailed risk evaluation of this condition and determined delta CDF was 3E-7 (Green). The finding was determined to have a cross-cutting aspect of evaluation in the problem identification and resolution cross-cutting area because the licensees corrective actions resulting from a degraded power cable in 2002 failed to incorporate sufficient detail into their procedures necessary to detect frayed cables. |
| 05000269/FIN-2016007-02 | 2016Q1 | Green | H.13 | NRC identified | Postulated Fire Affecting High Pressure Injection Pump Did Not Receive a VFDR Evaluation | The NRC identified a Green NCV of 10 CFR 50.48(c) and National Fire Protection Association Standard (NFPA) 805, Section 2.4.2.4 for the licensees failure to perform an adequate engineering analysis to determine the effects of fire on the ability to achieve the nuclear safety performance criteria, and consequently, did not add an associated variation from deterministic requirements (VFDR) into the Fire probabilistic risk assessment (PRA). Specifically, the licensees Nuclear Safety Capability Assessment (NSCA) failed to identify cables in the turbine building (TB) that could prevent the operation of the High Pressure Injection (HPI) Pumps. This item was entered into the corrective action program (CAP) as action request (AR) 02011673, and the licensee implemented compensatory measures in the form of hourly fire watches. The performance deficiency (PD) was more than minor because it was associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e. fire), 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 licensees failure to analyze the effects of fire damage on the HPI cables in the TB could result in fire damage adversely affecting the ability to achieve and maintain safe and stable conditions. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event (Task 1.4.5-B). A cross cutting aspect in the area of Human Performance, Consistent Process because the licensee did not use a consistent, systematic approach to make decisions, and did not incorporate appropriate risk insights (H.13). |
| 05000269/FIN-2016007-01 | 2016Q1 | Green | NRC identified | Pressure Boundary of Motor Operated Valves Could be Breached Due to Fire- Induced Hot Short | An unresolved item was identified regarding the licensees evaluation of certain motor operated valves (MOVs) in the NSCA. Specifically, based on the conclusions in the licensees NSCA, as well as subsequent evaluations, MOVs that are subject to a hot short that bypasses the torque or limit switch could result in damage to the valve that causes an unmitigated loss of reactor coolant system (RCS) inventory due to leakage through the damaged valves pressure boundary or the valves associated sealing components. Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, stated that fire damage could cause an electrical hot short that bypasses thermal overload protection for MOVs, and that this hot short could result in damage to the valve. As a part of the licensees transition to NFPA 805, the licensee identified a number of MOVs that could be susceptible to IN 92-18 type damage. These valves were classified as non-compliant components or variances from deterministic requirements (VFDRs). The subsequent evaluation of these valves by the licensees Fire PRA group determined that these VFDRs met the acceptance criteria of the Fire Risk Evaluation, as documented in OSC-9314, as being acceptable "as-is" and that no further action was required. These VFDRs and their FPRA dispositions were communicated to the NRC in the April 2010 Oconee NFPA 805 license amendment request (LAR). Subsequent to NRCs issuance of the SER, Oconee Valve Engineering determined that, due to the size of the installed motor/gearbox, 10 MOVs could potentially suffer IN 92-18 damage to the extent that the integrity of the valve body or bonnet could be compromised. Loss of valve integrity of the valve pressure boundary was not an assumption used in the FPRA evaluation. The licensee documented this condition in AR 01906086. After further evaluation, the licensee documented in AR 01999309 that 9 of the original 10 valves identified could potentially suffer IN 92-18 damage to the extent that the integrity of the valve body or bonnet could be compromised. For the 9 affected valves, the licensee has undertaken additional evaluations to determine whether some portion of the valve would fail before the valves pressure boundary is compromised, or that any possible leakage that may result can be bounded by the credited RCS make-up sourcein this case, the reactor coolant make-up pump. Inspectors determined that no immediate safety concern existed with this item because the licensee had implemented compensatory measures in accordance with the sites approved FPP. This item is unresolved pending inspector receipt and review of the licensees evaluation. | |
| 05000269/FIN-2015004-03 | 2015Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criteria III, Design Controls, requires in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures and instructions. Contrary to the above, since May 11, 1992, the licensee failed to ensure that applicable regulatory requirements and design basis were correctly translated into specifications, drawings, procedures and instructions for the SSF. Specifically, the licensees initial analytical assumptions were inadequate to demonstrate that the SSF could meet design requirements under all required operating conditions. Additionally, on multiple occasions the licensee failed to properly evaluate emergent issues and design changes to ensure the SSF continued to meet design requirements under all required operating conditions. The performance deficiency was more than minor because it was associated with the equipment performance and protection against external factors attributes 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. An NRC Region II senior reactor analyst evaluated both internal events and external events (e.g. fire, turbine building flooding, tornado) and determined the risk significance was very low (Green). The dominant contributors to the low risk result were: 1) the limited exposure time per year that an individual Oconee unit would spend in the vulnerable time-window immediately following shutdown, and 2) the low frequency of the external events that would demand the SSF. The licensee entered this condition into their CAP as NCR 01905088 and NCR 01905183. | |
| 05000270/FIN-2015004-02 | 2015Q4 | Green | H.8 | Self-revealing | Failure to Accomplish Activities Affecting Quality in Accordance With Station Instructions and Procedures Which Resulted in a Valid AFIS Actuation | A Green self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure to accomplish activities affecting quality in accordance with instructions and procedures established by the licensee. Specifically, the failure of station personnel to correctly close the Weidmuiller links on the feedwater control valves, in accordance with procedure PT/2/A/0152/020, AFIS Circuitry Test, Enclosure 13.2, AFIS Circuitry Verification and Valves Stroked on Refueling Frequency During FDW System Shutdown, Steps 1.22 and 1.23, caused feedwater flow oscillations. The feedwater flow oscillations resulted in a valid automatic feedwater isolation signal (AFIS) initialization. The licensee entered this issue into their corrective action program (CAP) as nuclear condition report (NCR) 01939072. The licensee verified all AFIS links on all units were closed and modified station procedures to include additional detail on ensuring that the links are fully closed. The licensees failure to follow procedure PT/2/A/0152/020, AFIS Circuitry Test, during the last AFIS circuitry testing on November 17, 2013 was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance and human performance attributes 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 (i.e., core damage). Specifically, the failure of station personnel to correctly close the Weidmuiller links on the feedwater control valves caused feedwater flow oscillations resulting in a valid AFIS initialization. Using NRC IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 Mitigating System Screening Questions Part B, dated July 1, 2012, the inspectors determined the finding to be of very low safety significance (Green) since the finding did not result in the loss of equipment specifically designed to mitigate a loss of feedwater flow. Specifically, the AFIS initiation was a valid actuation and as such, there was no loss of safety function. The finding had a cross-cutting aspect of procedure adherence in the area of human performance, because the licensee did not adequately follow processes, procedures, and work instructions. |
| 05000269/FIN-2015004-01 | 2015Q4 | Green | H.7 | NRC identified | Failure to Adequately Maintain Controlled Procedures in Emergency Response Facilities | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.47(b)(16), for the licensees failure to maintain the effectiveness of its emergency plan by ensuring procedures for use by the emergency response organization are maintained and up-to-date. Specifically, responsibilities for emergency plan implementing procedure distribution were not adequately maintained in multiple emergency response facilities because the procedures were not of the correct revision and may have been used had an emergency been declared. After the NRC inspectors informed the licensee of the discrepancy, the licensee entered the issue into their CAP as action request (AR) 01959550. The licensees immediate corrective actions were to perform an extent of condition review of all site EP procedures, including the corporate office and the other legacy Duke sites, and replace the procedures with the correct revision. The licensees failure to adequately maintain controlled procedures in the emergency response facilities was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because the performance deficiency was associated with the procedure quality attribute of the emergency preparedness (EP) cornerstone and adversely affected the associated cornerstone objective. The finding was evaluated using the EP significance determination process and was identified as having very low safety significance because it was a failure to comply with NRC requirements and was not a loss of the planning standard function. The finding was associated with a cross-cutting aspect in the documentation component of the human performance area because the licensee failed to maintain complete, accurate, and up-to-date documentation. |
| 05000269/FIN-2016008-01 | 2015Q4 | NRC identified | 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. | ||
| 05000269/FIN-2015002-01 | 2015Q2 | Green | H.12 | NRC identified | 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-2015002-03 | 2015Q2 | Green | H.4 | NRC identified | 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-2015002-02 | 2015Q2 | Green | H.12 | NRC identified | 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-2014005-02 | 2014Q4 | Green | P.3 | Self-revealing | 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-2014005-01 | 2014Q4 | Severity level IV | NRC identified | 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. | |
| 05000270/FIN-2014004-01 | 2014Q3 | NRC identified | 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. | ||
| 05000269/FIN-2014007-01 | 2014Q2 | Green | NRC identified | Failure to Evaluate the Under Voltage Relays at the Worst Case Minimum Drop Out Bus Voltage | The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that at the worst-case voltage, protective devices and thermal overload relays for safety-related loads would not trip prior to and after the transfer to the emergency power source. This transfer occurs for a sustained degraded voltage below the under voltage relay voltage settings for the duration of the time delay setting or the manual actions credited. The licensee revised their voltage calculations to account for previously unanalyzed loads. The licensee entered this issue into its corrective action program as problem identification program (PIP) O-14-2280. The team determined that the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the team identified that the voltages evaluated in the licensees analysis were nonconservative and could result in lower unanalyzed voltages that could result in connected safety-related loads stalling, becoming damaged, their protective devices tripping, or loads such as battery chargers being below their minimum operating voltages. The team determined that the finding was of very low safety significance (Green) because it was a design deficiency that did not result in a loss of off-site power operability. The team determined that no cross cutting aspect was applicable because this finding was not indicative of current licensee performance. | |
| 05000269/FIN-2014007-02 | 2014Q2 | Green | P.3 | NRC identified | Failure to Correct Issues with DC System Voltage Calculations and 120Vac Motor Control Center (MCC) Control Circuit Calculations | The team identified a Green non-cited violation (NCV), with two examples, of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct conditions adverse to quality. Specifically, the licensee (1) failed to correct voltage calculations for safety-related 4160 volt circuit breaker 125 volt-direct current control circuits and (2) failed to correct voltage calculations for safety-related 120 volt alternating current motor control center control circuits. The above issues were previously identified as NCV 05000269,270,287/2011010-04 and NCV 05000269,270,287/2011010-03, respectively. The incomplete corrective actions were newly entered in the licensees corrective action program as problem identification program (PIP) reports O-14-2781 and O-14-2811 to track their completion. The team determined that the performance deficiency was more than minor because it affected the Equipment Performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because the inadequate corrective actions did not result in losses of operability or function for either example. The violation was assigned the cross-cutting aspect of Resolution in the area of Problem Identification and Resolution because the licensee did not take effective corrective actions to address issues in a timely manner. |
| 05000269/FIN-2014007-04 | 2014Q2 | NRC identified | Degraded Voltage Relay Scheme | The team identified that the licensees degraded voltage relays did not monitor the safety-related 4.16kV buses, but rather they monitored the switchyard 230kV Yellow bus. This resulted in a lack of degraded voltage protection whenever the 4.16kV safety-related buses were not being fed through the start-up transformers. During normal power operation, the 4.16kV safety-related buses were supplied from the unit auxiliary transformers. Additionally, for degraded voltage detected on the 230kV switchyard Yellow bus with no accident signal present, the degraded voltage relay alarm in the main control room would have only resulted in manual actions to resolve the degraded voltage condition or to disconnect from the degraded source. It was estimated that the manual actions could take as long as 12 minutes to resolve the degraded voltage condition. The use of degraded voltage relays only on the 230kV switchyard Yellow bus and the use of manual actions for a degraded voltage condition appeared to be contrary to the design criteria for degraded voltage protection stated in an NRC letter to the licensee dated June 3, 1977 and NRC Regulatory Issue Summary 2011-12. Lastly, the team identified that Oconee currently credits operation of the loss-of-voltage relays monitoring the 4.16kV main feeder buses to disconnect from offsite power on a loss of voltage condition and subsequent re-connection to Keowee Hydro to meet the UFSAR Chapter 15 plant accident analyses. However, the loss of voltage relay setpoints and associated time delays were not included in the plant TS. This appeared to be contrary to 10 CFR 50.36(c)(2)(ii)(C) Criterion 3. The team determined that consultation with the Office of Nuclear Reactor Regulation was warranted for the NRC to determine: (1) whether Oconees existing licensing and design bases are adequate and meet all NRC regulations and requirements with their current degraded voltage relays design and off-site/station electric power system design, (2) whether the automatic actions for the loss-of-voltage relays meet the intent of the degraded voltage relays, and (3) whether the current plant TS meet the requirements of 10 CFR 50.36(c)(2)(ii)(C) which state, in part, that a TS limiting condition for operation of a nuclear reactor must be established for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The licensee entered this issue into their corrective action program as PIP O-14-2034. This issue is being tracked as URI 05000269/2014007-04, 05000270/2014007-04, 05000287/2014007-04, Degraded Voltage Relay Scheme. | ||
| 05000269/FIN-2014007-05 | 2014Q2 | NRC identified | Potential Unanalyzed Condition Associated with Emergency Power System | During a review of Oconees engineered safeguards protection system (ESPS) emergency power start control for the KHUs, the team noted that the 125Vdc control cables for train A of the ESPS and cables for supervisory control of both KHUs were recently modified. The team also noted that these 125Vdc control cables were installed in the same underground concrete raceway systems as the 4160Vac auxiliary power cables, 13.8kVac power cables for both emergency power and protected service water (PSW), and were in close proximity to these power cables. The team was concerned that a short circuit (which the licensee considered outside their design basis) in the 13.8kVac cables could induce voltage and currents in the dc control system which could potentially impact the functionality of the emergency power system which is required to mitigate certain design basis events. A similar issue exists in Manhole 6 of the PSW underground raceway where the new power supply to the PSW (adjacent to the 125Vdc control emergency power system) could short circuit or fault to ground. The licensee had not performed an analysis to determine the effects of such failures on the ability of the emergency power system to perform its safety function, thus the team questioned whether the plant was in an unanalyzed condition. Although the licensee did not agree that these failures were part of their licensing basis, they reported this as an unanalyzed condition to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B) in Licensee Event Report 269/2014-01. In response to the teams concerns, the licensee entered this issue into their corrective action program, and performed immediate and prompt determinations of operability in which they concluded a reasonable expectation of operability exists. The team has requested assistance from subject matter experts in the Office of Nuclear Reactor Regulation via a Task Interface Agreement1 to review the emergency power system licensing basis to determine the acceptability of the licensees design. If the design is found to be noncompliant with the licensing basis, the licensee will be required to implement corrective actions to restore compliance. This issue is being tracked as URI 05000269/2014007-05, 05000270/2014007-05, 05000287/2014007-05, Potential Unanalyzed Condition Associated with Emergency Power System. | ||
| 05000269/FIN-2014007-06 | 2014Q2 | Severity level IV | Licensee-identified | Licensee-Identified Violation | 10 CFR 50.71(e) requires, in part, that each person licensed to operate a nuclear power reactor, shall update periodically, the FSAR originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. This submittal shall include the effects of all changes made in the facility or procedures as described in the FSAR. Contrary to the above, since December 6, 2012, after updating the UFSAR to reflect the new licensing basis under NFPA-805, several items applicable to the Fire Protection System were incorrectly removed. Traditional enforcement was applicable because the violation could impact the regulatory process, and was evaluated using the NRCs Enforcement Policy. This violation was determined to be a Severity Level IV violation because the lack of up-to-date information did not result in an unacceptable change to the facility or procedures. This violation was documented in the licensees corrective action program as PIP O-13-09302. | |
| 05000269/FIN-2014007-03 | 2014Q2 | Severity level IV | NRC identified | Failure to Update the UFSAR with Current Battery Testing Standards | The team identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) for the licensees failure to include in the latest Updated Final Safety Analysis Report (UFSAR) changes made to the sites licensing bases with respect to station battery testing made during the Technical Specification conversion to Integrated Technical Specifications. Specifically, the UFSAR did not identify the standards by which the testing was conducted. The licensee entered this issue into its corrective action program as problem identification program report O-14-2338 and planned to include the omitted battery testing standards to the UFSAR during an upcoming update cycle. The team dispositioned the performance deficiency using the traditional enforcement process because failing to update the UFSAR had the potential to adversely impact the NRCs ability to perform its regulatory function. The performance deficiency was characterized as a Severity Level IV violation in accordance with the NRC Enforcement Policy, Section 6.1.d.3 as the lack of up-to-date information did not result in any unacceptable change to the facility or procedures. In accordance with IMC 0612, Power Reactor Inspection Reports, no cross-cutting aspects are assigned to traditional enforcement violations. | |
| 05000269/FIN-2014011-01 | 2014Q2 | White | H.7 H.7(n) | Self-revealing | 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-2014002-01 | 2014Q1 | H.7 H.7(n) | NRC identified | Inadequate Procedure to Ensure Adequate Piping Weld Inspections | A NRC-identified potentially Greater than Green Apparent Violation (AV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to ensure that procedure NDE-995, Ultrasonic Examination of Small Diameter Piping Butt Welds and Base Material for Thermal Fatigue Damage, was adequate to achieve acceptable coverage for the ultrasonic (UT) examination of weld 1-RC- 201-205. NDE-995 did not contain the necessary steps to achieve acceptable coverage for UT examinations when limitations were encountered. The licensee entered this finding into their corrective action program as PIP O-13-13168. The failure to ensure that station procedure NDE-995 was adequate to achieve acceptable coverage for the UT examination of weld 1-RC-201-205 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 undetected crack resulted in reactor coolant system pressure boundary leakage and a forced shutdown of Unit 1. The inspectors determined that detailed risk analysis was required. 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-2014403-01 | 2014Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000269/FIN-2013005-01 | 2013Q4 | Green | H.7 H.10 | NRC identified | Failure to properly maintain a fire barrier penetration seal | An NRC-identified non-cited violation (NCV) of 10 CFR 50.48(c) and National Fire Protection Association Standard 805 (NFPA 805), Section 3.11.4, was identified for the licensees failure to comply with the fire barrier penetration sealing and inspection requirements of the approved fire protection program (FPP). The annular space between the fire barrier opening and the 2 conduit was not properly sealed. The licensee entered the issue in their CAP as PIP O-13-09104, initiated a work order to repair the seal, and implemented an hourly fire watch as required by Oconee Selected Licensee Commitment (SLC) 16.9.5. The licensees failure to comply with the fire barrier penetration sealing and inspection requirements of the approved fire protection program was a performance deficiency. This performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and adversely affected the cornerstone in that the fire barrier could not be relied upon to fully perform its function. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of very low safety significance (Green) because safety significant equipment was located a sufficient distance from the degraded penetration and the reactors ability to reach and maintain a safe shutdown condition was not impacted. The cause of this finding was determined to have a cross-cutting aspect of H.2(c) in the Resources component of the Human Performance area because the licensee did not ensure that complete, accurate, and up-to-date design documentation and procedures were available because adequate guidance was not included in the maintenance inspection procedures to allow personnel to identify a degraded condition. |
| 05000269/FIN-2013005-02 | 2013Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50, App. B, Criterion XVI, required in part that conditions adverse to quality, such as non-conformances, are promptly identified and corrected. NSD-203, Operability/Functionality, required entry into the operability determination process (ODP) upon the discovery of circumstances that call into question the operability of any TS SSC including degraded/non-conforming conditions. NSD-203 also requires that actions to confirm if the SSC is degraded or non-conforming should be completed in a timeframe that is commensurate with its safety significance. Contrary to the above, a potential non-conforming condition was identified on December 30, 2012; however, the ODP was not entered until November 26, 2013, and corrective actions generated to correct the non-conforming condition. The finding was not greater than Green because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components and did not involve an actual reduction of hydrogen igniters in containment. This violation was entered into the CAP as PIP O-13-14547. | |
| 05000269/FIN-2013407-01 | 2013Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000269/FIN-2013501-01 | 2013Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.4.1(a), Procedures, required in part that written procedures be established, implemented, and maintained covering the applicable procedures in Regulatory Guide 1.33, Rev. 2, Appendix A, February 1978. Procedure OP/1/A/1102/008, Enclosure 4.35, On Line Valve Lineup for MOV Maintenance, Step 2.5, stated, in part, for the operator to cycle 1LP-22 (1B LPI BWST suction). Contrary to the above, on June 26, 2013, the licensee operator failed to follow written procedure when he closed 1LP-21 (1A LPI BWST suction) which isolated the operable LPI train from the BWST rendered Unit 1 LPI inoperable. The licensee restored the LPI A train to its proper alignment within thirteen minutes. The finding was determined not to be greater than Green because the loss of function of at least a single train did not exceed its TS allowed outage time. The licensee entered the issue into their CAP as PIP O-13-06879. | |
| 05000269/FIN-2013501-02 | 2013Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.4.1(a), Procedures, required in part that written procedures be established, implemented, and maintained covering the applicable procedures in Regulatory Guide 1.33, Rev. 2, Appendix A, February 1978. Procedure OP/0/A/1107/016, Enclosure 4.4, Removal and Restoration of 230KV Switchyard PCB, Step 2.2.4, stated, in part, Ensure locked closed PCB (27) Yellow (Red) Bus Side Disconnect. Contrary to the above, on October 22, 2012, the licensee failed to ensure PCB27 was locked closed. The licensee discovered and corrected this condition on April 24, 2013. The finding was determined to represent a loss of system and/or function which required a risk evaluation by a Senior Reactor Analyst (SRA). The SRA estimated the likelihood of faults that could lead to damage of the disconnect and multiplied these by the change in conditional core damage probability due to a loss of the transformer impacted. Dominant cutsets involved failure of one Keowee hydro unit in conjunction with LOOP sequences, operators failure to recover offsite power, or the Keowee faults within 4 hours, and failure of EFW. The risk impact was less than 1E-7 for the exposure period. In addition, the risk impact of seismic events was estimated not to be a major contributor to the change in risk. Because the risk impact was less than 1E-7, the finding was determined not to be greater than Green. Licensee personnel entered the issue into their corrective action program as PIP O-13-04503. | |
| 05000269/FIN-2013007-03 | 2013Q3 | Green | H.14 | NRC identified | Fire Protection Program Change did not Meet Oconee License Condition Requirements for NFPA 805 Chapter Three | An NRC-identified Apparent Violation (AV) and associated traditional enforcement violation of Oconee Nuclear Station Renewed Facility Operating License Condition 3.D for Units 1, 2, and 3 was identified for the licensees failure to implement and maintain in effect all provisions of the approved fire protection program (FPP) that comply with 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. The licensee made a change to the approved FPP involving control of combustible materials when the definition of transient fire loads was revised to exclude fire retardant scaffolding materials as transient fire loads, which would not require the licensee to track these items as combustible fire loads. The licensee also failed to submit the FPP change to the NRC for review and approval prior to implementation which impacted the ability of the NRC to perform its regulatory oversight function. The licensee entered this issue into the corrective action program as Problem Investigation Program O-13-08584. This finding did not represent an immediate safety concern because the licensee implemented compensatory measures in the form of combustible tracking impairments and fire watches in the high safety significant fire zones which contained the scaffolding. Failure to comply with Oconee Operating License Condition 3.D for a change to the approved FPP involving control of fire retardant scaffolding materials was a performance deficiency. This performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events (i.e. fire), and it adversely affected the cornerstone objective in that the change to the FPP had the potential to adversely affect the ability to achieve and maintain safe and stable plant conditions due to the increased transient fire load in the affected fire zones. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609 Appendix F, Fire Protection Significance Determination Process, review was required as the finding affected fire prevention and administrative controls. The performance deficiency applied to most fire zones within the plant because the licensee stopped tracking the use of the fire retardant scaffolding materials. The team determined that a systematic plant-wide assessment effort was beyond the intended scope of the fire protection SDP. Therefore additional analysis is required to assess the significance of this finding. The cause of this finding was determined to have a cross-cutting aspect of H.1(b) in the Decision- Making component of the Human Performance area because the licensee used nonconservative assumptions in the decision making associated with this FPP change. Additionally, the licensees failure to submit the FPP change to the NRC was a traditional enforcement violation. The severity level of the traditional enforcement violation will be assigned based on the significance determination of the associated finding. |
| 05000269/FIN-2013007-04 | 2013Q3 | Green | H.8 | NRC identified | Failure to Evaluate Unapproved Combustibles in Accordance With Procedures | An NRC-identified Green non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2, and 3 Renewed Facility Operating License Condition 3.D was identified for the licensees failure to follow procedures for the control of transient combustible materials. The team identified five examples where the licensee failed to follow procedure Nuclear System Directive (NSD) 313, Control of Transient Fire Loads, in that unapproved combustible materials were stored in fire areas/fire zones without proper evaluation and without appropriate compensatory actions being implemented. The licensee entered these issues into the corrective action program as Problem Investigation Program documents O-13-07896, O-13-07897, O-13-07989, O-13-08051, and O-13-08459; and initiated immediate corrective actions to remove the unapproved combustibles from the identified fire areas/fire zones. The licensees failure to follow procedure NSD 313 for storage of transient combustibles in fire areas/fire zones was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external events (i.e. fire), 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. The finding was determined to be of very low safety significance (Green) because it only affected the ability to reach and maintain cold shutdown conditions. The cause of this finding was determined to have a cross-cutting aspect of H.4(b) in the Work Practices component of the Human Performance area, because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures. |