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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000400/FIN-2011008-07 | 2011Q2 | Green | NRC identified | Failure to Test Safety-Related Molded Case Circuit Breakers | The team identified a Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to maintain the qualification bases for safety-related molded case circuit breakers (MCCBs). Immediate corrective actions included review of the MCCB testing and maintenance to validate current status. Permanent corrective actions are still being pursued by the licensee. The licensee entered this issue into the CAP as NCR 460900. The team determined that the failure to extend the qualified life of the installed Westinghouse MCCBs which were over 20 years old was a performance deficiency. The finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining qualified components in safety-related SSCs could lead to the inability to respond to design basis events. The finding was of very low safety significance because the finding was a design or qualification deficiency confirmed not to result in loss of operability or functionality. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. | |
| 05000400/FIN-2011008-08 | 2011Q2 | Green | NRC identified | Failure to Ensure Adequate Voltage for Safety Related Components | The team identified a Green, NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures to ensure safety related components had adequate voltage. The licensee entered this issue into the CAP as NCR 458640, NCR 458648 and NCR 460930, and initiated compensatory measures which included Standing Instruction 11-08 to explain that the alternate power supply to the safety related inverters could be subject to inadequate voltage. Permanent corrective actions are still being evaluated by the licensee. The licensees failure to perform an analysis to demonstrate that safety related components would have adequate voltage to operate during a design basis accident or transients was a performance deficiency. The finding was more than minor because it affected the Mitigating System Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to perform an analysis that demonstrated that the loads connected to Instrument Distribution Panels (IDPs) S-I, S-II, S-III and S-IV would have adequate voltage when the IDPSs are aligned to the output of their respective 7.5kVA safety related inverter or to their respective alternate sources. The finding was of very low safety significance because it was a design issue confirmed not to result in a loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. | |
| 05000400/FIN-2011008-09 | 2011Q2 | Green | P.2 | NRC identified | Inadequate Corrective Action Inadvertent Loss of Thermal Barrier HX Flow | The team indentified a Green finding for licensees failure to take adequate corrective action for the inadvertent closing of MOV 1CC-252 (Reactor Coolant Pump (RCP) Thermal Barrier Return Flow Isolation Valve) following the start of the standby Component Cooling Water (CCW) pump. As interim corrective action, the licensee revised operating procedures to reflect the issue and initiated compensatory measures which included Standing Instruction 11- 0012 to explain that during conditions where the standby CCW pump starts, a transient high flow can be expected that causes 1CC-252 to automatically close. Permanent corrective actions are still being evaluated by the licensee. The licensee entered this issue into the corrective action program (CAP) as NCR 460686. The licensees failure to take adequate corrective action for the inadvertent closing of MOV 1CC-252 following the start of the standby CCW pump was a performance deficiency. The finding was more than minor because it was affected the Equipment Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to evaluate the potential for the RCP thermal barrier to isolate following safety injection (SI) or de-energization of a safety bus upon the auto start of the standby CCW pump. The finding was considered to be of very low safety significance because assuming worst case degradation, the finding would not result in exceeding the Technical Specification (TS) limit for any reactor coolant system (RCS) leakage, result in the total loss of a safety function, did not contribute to both the likelihood of a reactor trip or the likelihood that mitigating equipment or functions would not be available, and did not increase the likelihood of a fire or internal/external flooding. Because the licensee failed to thoroughly evaluate problems such that the resolution(s) address causes and extent of conditions, as necessary, this finding is assigned a crosscutting aspect in the corrective action program of the Problem Identification and Resolution area. |
| 05000400/FIN-2011008-10 | 2011Q2 | NRC identified | Failure to Establish Adequate Preventative Maintenance Procedure for Safety-Related Tornado Dampers | The team identified a Green, NCV of TS 6.8.1 for the failure to implement an adequate preventative maintenance procedure to ensure reliable operation of the plants safety-related tornado dampers. Immediate corrective actions included procedure changes, testing of all dampers, and necessary corrective maintenance. In addition, the licensee submitted LER 2011-001 to address a discovered inoperable damper. Additional corrective actions are still being evaluated by the licensee. The licensee entered this issue into the CAP as NCRs 457949 and 458237. The licensees failure to implement an adequate preventative maintenance procedure to ensure reliable operation of the plants safety-related tornado dampers was a performance deficiency. This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of the safety-related ventilation system to respond to initiating events to prevent undesirable consequences and the cornerstone attribute of Protection against External Events, i.e. seismic, weather. Specifically, the failure of the dampers to function properly would impact the ability to maintain required ventilation during an external event. The inspectors assessed the finding using a Phase I SDP screening which determined a Phase III SDP evaluation was required due to the fact that the finding involved the loss or degradation of equipment specifically designed to mitigate a severe weather initiating event (e.g., tornado doors). A Phase III SDP evaluation was performed in accordance with NRC Inspection Manual Chapter 0609 Appendix A by a regional SRA using the NRC SPAR model. The analysis determined that the performance deficiency resulted in a core damage frequency (CDF) risk increase less than 1E-6/year. Therefore, the finding was characterized as having very low safety significance. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. | ||
| 05000400/FIN-2011009-01 | 2011Q3 | Green | P.2 | NRC identified | Failure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with the Quality Control Organizations Acceptance of Electrical Termination Errors. | An NRC identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to take adequate corrective action to preclude repetition of a significant condition adverse to quality associated with the oversight of the Quality Control (QC) organization at Harris. The licensee failed to take adequate corrective actions to address the cause of QCs acceptance of electrical termination errors which occurred during Refueling Outage (RFO) - 15 necessary to preclude related errors which occurred in RFO-16. The licensee entered this violation in their corrective action program as Nuclear Condition Report (NCR) 479478. The inspectors determined that failure of the licensee to take adequate corrective actions to address the cause of QCs acceptance of electrical termination errors which occurred in RFO-15 necessary to preclude related errors which occurred in RFO-16 was a performance deficiency (PD). The PD was determined to be more than minor because if left uncorrected, the PD has the potential to lead to a more significant safety concern. Specifically, failure to adequately correct the cause of QCs acceptance of electrical termination errors could result in unidentified wiring errors in safety related equipment associated with the Mitigating Systems cornerstone. In accordance with IMC 0609, Attachment 4, Table 4a, Phase 1 Initial Screening and Characterization of Findings , the finding was determined to be of very low safety significance (Green) because the finding is not a design deficiency, did not result in an actual loss of system or single train function, and was not potentially risk significant due to external events. The inspectors determined that this finding was directly related to the cross-cutting aspect of thoroughness of evaluation within the Corrective Action Program component of the Problem Identification and Resolution area because the licensee did not thoroughly evaluate the problems leading to QCs acceptance of electrical termination errors which occurred in RFO-15 and develop adequate corrective actions to address the cause, and as a result, corrective actions did not preclude repetition of similar QC errors in RFO-16. |
| 05000400/FIN-2011011-01 | 2011Q3 | Green | H.7 | NRC identified | Inadequate Procedure AOP-036.04 for Fire Area 1-A-BAL-C Post-Fire Safe Shutdown | The team identified a non-cited violation of Harris Nuclear Plant Technical Specification 6.8.1.a. for inadequate guidance in fire response abnormal operating procedure AOP-036.04, Fire Areas: 1-A-BAL-C, 1-A-BAL-D, 1-A-BAL-F, 1-G, FPYARD, Revision 17. Specifically, the procedure could not have been performed as written in that, AOP-036.04, Section 3.1, directed operators to implement a step in the procedure that did not exist. The licensee initiated Nuclear Condition Report 489092 to address this issue in the Corrective Action Program and subsequently revised the procedure. The team determined that inadequate fire response procedure guidance was a performance deficiency. This finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and it affected the cornerstone objective of protection against external events (i.e., fire). The team assessed this finding using IMC 0609, Appendix F, Fire Protection Significance Determination Process. The team assigned a low degradation rating to this finding because the abnormal operating procedure deficiency was compensated by available emergency operating procedure guidance, operator experience/familiarity, and training. It was likely that plant operators would have been able to assess plant parameters and would have taken the appropriate actions required to ensure post-fire safe and stable plant conditions. Therefore, this finding was of very low safety significance (Green). The cause of this finding was determined to have a cross cutting aspect in the Human Performance Area, Resources Component, because the licensees validation and verification process did not ensure that the procedure was adequate and accurate. |
| 05000400/FIN-2011011-02 | 2011Q3 | Severity level Enforcement Discretion | NRC identified | Noncompliance for Providing Inadequate Procedural Guidance for Post-Fire Safe Shutdown | Shearon Harris License Condition 2.F, Fire Protection Program states, in part, that Carolina Power & Light Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (FSAR) for the facility as amended. Section 9.5.1, Fire Protection System, of the FSAR incorporates, by reference, Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1, Revision 3, dated May 7, 1986. o Section C.5.b (2) of Fire Protection Evaluation and Comparison to NUREG- 0800, BTP CMEB 9.5-1 requires one train of systems necessary to achieve and maintain hot standby conditions from either the control room or emergency control station(s) be free of fire damage by providing one of the means described in Section C.5.b (2) (i.e., use of spatial separation, passive fire barriers, and fire detection and an automatic fire suppression system). o Section C.5.b (3), of Fire Protection Evaluation and Comparison to NUREG- 0800, BTP CMEB 9.5-1 requires that alternative or dedicated shutdown capability be provided where the guidelines of Section C.5.b (1) and C.5.b (2) cannot be met. Section 9.5.1.5.4, Quality Assurance Program, of the FSAR states that the fire protection quality assurance program elements are included in Section 17.3 of the FSAR. Section 17.3.1.1, Methodology, of the FSAR states, in part, that the HNP quality assurance program prescribes measures for the control and accomplishment of activities for the operation of safety related and fire protection SSCs. Section 17.3.1.1 also commits to the requirements of 10 CFR 50, Appendix B. 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances. Contrary to the above, the licensee failed to meet the requirements of its documented fire protection program, in that: The licensee failed to protect redundant systems necessary to achieve and/or maintain hot shutdown conditions from the MCR or emergency control station(s) from fire damage by one of the means described in Section C.5.b(2) of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1. The licensee failed to ensure alternative shutdown capability was available for two fire areas where the guidelines for ensuring one redundant train for safe shutdown remain free of fire damage, detailed in Section C.5.b (1) and C.5.b (2) of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5- 1 could not be met. The licensee failed to provide adequate procedural guidance, in that the licensees fire safe shutdown procedure failed to incorporate instructions to alert operators concerning time constraints for restoring cooling to the RCP seals. Additionally, the licensees fire safe shutdown procedure included steps that were not appropriate to the circumstances in that a required procedural step may not have been feasible due to the presence of postulated smoke, under certain conditions. Because this issue relates to fire protection, and the associated noncompliances were resolved by compliance with 10 CFR 50.48(c), the NRC is exercising enforcement and reactor oversight process discretion for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, the licensee entered the noncompliances into their corrective action program and implemented appropriate compensatory measures. The noncompliances were not associated with a finding of high safety significance (Red), the noncompliances were not willful, and the licensee submitted a letter of intent stating its intention to transition to 10 CFR 50.48(c) by December 31, 2005. LER 05000400/2002-004-09, Unanalyzed Condition Due to Inadequate Separation of Associated Circuits; LER 05000400/2004-004-00, Unanalyzed Condition Due to Inadequate Separation of Associated Circuits; and URI 05000400/2005007-01, Fire Response Procedures May Not Be Adequate To Prevent RCP Seal Failure and Subsequent Seal Loss of Coolant Accident For a Fire in Certain Fire Areas, are closed. | |
| 05000400/FIN-2011402-01 | 2011Q2 | H.7 | Licensee-identified | Security | ||
| 05000400/FIN-2011402-02 | 2011Q2 | H.7 | Licensee-identified | Security | ||
| 05000400/FIN-2012002-01 | 2012Q1 | Green | P.3 | NRC identified | Failure to Correct the Reactor Auxiliary Building Emergency Exhaust System Dampers Failure to Close | The inspectors identified a Green Non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to correct a condition adverse to quality affecting the Reactor Auxiliary Building Emergency Exhaust System (RABEES). Specifically, the licensee failed to resolve the effects of the degraded condition of dampers that failed to close which resulted in RABEES being declared inoperable. The licensee experienced failures in RABEES dampers in 2009, and did not take adequate corrective actions to correct the condition adverse to quality. The licensee entered the violation into their Corrective Action Program (CAP) as Action Request (AR) #513163 and plans to change the springs in the actuators and to reevaluate the long term strategy for the dampers. The failure of the licensee to take adequate corrective actions to address the cause of the RABEES dampers failing to close was a performance deficiency (PD). The PD was more than minor because it affected the Barrier Performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, damper AVD52SB failed to close which led RABEES to be inoperable. The failure of the damper did not result in the loss of functionality of the RABEES system, however the licensee entered the Limiting Condition for Operation and declared RABEES inoperable. Using Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the issue was determined to have very low safety significance (Green), because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building. This finding has a crosscutting aspect in the area of Problem Identification and Resolution in the Corrective Action component because the licensee did not take appropriate corrective actions to address safety issues in a timely manner. |
| 05000400/FIN-2012002-02 | 2012Q1 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, states in part, measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Shearon Harris Nuclear Plant Unit One Technical Specifications (TS) 3.8.1, A.C. Sources, requires in part, that two separate and independent diesel generators shall be operable in modes 1, 2, 3 and 4. With one diesel generator inoperable the TS requires that the diesel generator be restored to operable status within 72 hours or be in at least hot standby within the next 6 hours and in cold shutdown within the following 30 hours. Contrary to the above, the licensee failed to promptly identify and correct the B EDG mechanical governor speed setting drift, a condition adverse to quality. The condition existed from April 2, 2011, to December 21, 2011. This resulted in the B EDG being inoperable for longer than the TS 3.8.1 Allowed Outage Time. The inspectors determined that an SDP Phase 2 analysis was required based on the finding representing an actual loss of safety function for a single train in excess of its TS allowed outage time. The Phase 2 analysis yielded a potential greater-than-green result and the issue was given to a Senior Reactor Analyst for a Phase 3 analysis to be performed. The Phase 3 analysis determined this issue to be of very low risk significance (i.e., Green) due to the low likelihood of events that could cause a sustained loss of offsite and onsite electrical power. Upon discovery of this violation the licensee put in place compensatory measures to ensure that the mechanical governor speed drift will not interfere with the electronic governor and plans to change the system for both the B and the A EDG. The licensee has entered the violation into their CAP as AR #505470. | |
| 05000400/FIN-2012003-01 | 2012Q2 | Green | Self-revealing | Inadequate Preventive Maintenance Results in Inoperability of the a Emergency Service Water System | A self-revealing Green NCV of Technical Specification (TS) 6.8.1, Procedures, was identified for the licensees failure to implement an adequate preventive maintenance procedure to identify a condition which led to the inoperability of the A Emergency Service Water (ESW) system. Specifically, the licensee failed to perform an adequate inspection of the grease in the lower gear box of the A ESW strainer motor, resulting in the strainer failing to function and the inoperability of the A ESW system. The licensee entered this issue into their CAP as AR #521946. As corrective action, the licensee revised PM-M0014 to include inspection of all similar gear boxes throughout the plant. The failure to implement an adequate preventive maintenance procedure to identify a condition which led to inoperability of the A ESW system was a performance deficiency. The performance deficiency was more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone, and it 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, it resulted in the unplanned inoperability of the A ESW train. Using IMC 0609, Significance Determination Process, Phase 1 screening worksheet of the SDP, this finding was determined to be of very low safety significance because it was not a design or qualification deficiency confirmed to result in a loss of operability or functionality, did not represent a loss of system safety function, did not result in a loss of safety system function for a single train for greater than TS allowed outage time, did not result in a loss of safety function of one or more non-TS trains of equipment designated as risk significant for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Due to the historic nature of the development of this preventive maintenance procedure and the fact that this procedure was not performed on either train of ESW within the past two years, this finding has no cross-cutting aspect. | |
| 05000400/FIN-2012003-02 | 2012Q2 | Green | H.12 | Self-revealing | Failure to Follow Fuel Handling Procedure | A self-revealing Green NCV of TS 6.8.1, Procedures, was identified for the licensees failure to follow procedure, FHP-014, Fuel and Insert Shuffle Sequence, during core offload resulting in inadvertently placing a spent fuel assembly in the wrong location in the spent fuel pool. Specifically, it resulted in spent fuel assembly HW40 being stored in a location for which it had not been analyzed for 22 days, until it was discovered on May 22, 2012. The licensee entered this issue into their CAP as AR #538457. As corrective action, the licensee verified that all other fuel assemblies moved during offload were located in their correct locations and performed a Human Performance Review Board. The failure to follow procedure FHP-014 during core offload resulting in inadvertently placing a spent fuel assembly in the wrong location in the spent fuel pool was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Barrier Integrity cornerstone, and it affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, it resulted in spent fuel assembly HW40 being stored in a location for which it had not been analyzed for 22 days. IMC 0609, Significance Determination Process, Phase 1 screening worksheet of the SDP, instructed the inspector to process this finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. Checklist 4 from IMC 0609, Appendix G, Attachment 1 was determined to be the most appropriate because the water level was greater than 23 feet and the time to boil was greater than two hours in the Spent Fuel Pool. Using Checklist 4, the inspector determined that the finding did not require a quantitative assessment because the licensee met the Technical Specifications for the spent fuel pool, specifically water level and boron concentration. Therefore, this finding was determined to be of very low safety significance (Green). The finding has a cross-cutting aspect of Human Error Prevention, as described in the Work Practices component of the Human Performance cross-cutting area because the designated human error prevention technique of concurrent verification failed to prevent this error. |
| 05000400/FIN-2012003-03 | 2012Q2 | Green | NRC identified | Failure to Use Appropriate Radioactive Sources to Calibrate Effluent Monitors | The inspectors identified two examples of a Green Non-Cited Violation (NCV) of TS 6.8.1, Procedures, for the licensees failure to implement an adequate Quality Assurance (QA) program for effluent monitoring. Specifically, the secondary calibration (transfer) sources used for effluent monitors 21WL-3541 (Waste Monitor Tanks Discharge) and RM21AV-3509-1SA (Plant Vent Stack Monitor) were not verified to be acceptable prior to use. The licensee has entered these issues into their CAP (AR 537505) and is currently evaluating corrective actions and extent of condition. The licensees failure to use appropriate secondary calibration sources to adequately calibrate REM-21WL-3541 and RM-21AV-3509-1SA was a performance deficiency. The finding was more than minor because it is associated with the Public Radiation Safety cornerstone attribute of plant equipment/process radiation monitoring and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The finding was assessed using Inspection Manual Chapter (IMC) 0609, Appendix D, Public Radiation Safety Significance Determination Process (SDP). The failure to use adequate secondary calibration sources does not represent a substantial failure to implement the radioactive effluents program since each batch release from a Waste Monitor Tank is sampled and analyzed prior to discharge and releases through the Plant Vent Stack are sampled and analyzed weekly. In addition, 10 CFR 20 and 10 CFR 50 dose limits to a member of the public were not exceeded. Therefore this finding was determined to be Green. No cross-cutting aspect was assigned for this finding because the performance deficiency does not represent current licensee performance | |
| 05000400/FIN-2012003-04 | 2012Q2 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. 10 CFR 20.1501 requires, in part, that each licensee make surveys that are reasonable under the circumstances to evaluate potential radiological hazards. Contrary to this, on June 6, 2012, during restart from refueling outage RFO17, two radiation workers and a HPT entered the reactor bio-shield wall (LHRA) to perform maintenance work near the A Reactor Coolant Pump and encountered radiological conditions that had not been evaluated. The work crew was briefed on radiological conditions in the travel path general areas of 5-20 mrem/hr corresponding to 0 percent reactor power (shutdown). However, immediately prior to entry, reactor power was raised above one percent and fluctuated between approximately 1.1 percent and 1.9 percent during their time inside the RCB. This caused an increase in the travel path general area dose rates inside the bio-shield to approximately 400 mrem/hr and resulted in multiple ED dose rate alarms. The increased dose rates were discovered by the accompanying HPT who briefed the workers in the field that ED alarms may be received upon exiting the bio-shield. This finding was of very low safety significance (Green) because there was no substantial potential for overexposure. This is due to the fact that the workers were accompanied by a HPT, the time spent inside the bio-shield was brief, and dose rates were not sufficiently high enough in the planned travel path which was adhered to by the work crew. The licensee entered the event into their CAP as AR 541773. | |
| 05000400/FIN-2012004-01 | 2012Q3 | Green | P.2 | NRC identified | Failure to Adequately Perform Containment Visual Inspection When Containment Integrity is Required | The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to adequately correct a previously identified issue associated with the performance of OST-1081, Containment Visual Inspection when Containment Integrity is Required. Specifically, on June 3, 2012 during an independent containment closeout inspection by the NRC resident inspectors, cables were identified as not having been analyzed for the impact on the operation of the containment sumps. The licensee did not identify or reconcile the unanalyzed cables in containment during the performance of OST-1081. The licensee removed a large portion of the cabling and then completed an operability evaluation, while in mode 3, on June 6, 2012 for the cables that remained. The evaluation concluded that the containment sump was fully operable, but with reduced margin because of the cables. The cables were further analyzed and recorded in Engineering Change 87249, with a similar conclusion. The issue was placed into the corrective action program (CAP) as action request (AR) #566201. The licensees failure to adequately identify and take prompt corrective actions to evaluate temporary cables in containment during OST-1081, which had not been previously analyzed was identified as a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it could potentially cause one or more Residual Heat Removal (RHR), Containment Spray (CT) pumps, and associated Emergency Core Cooling Systems (ECCS) trains to be inoperable in the event that the containment sump became clogged and lost the required Net Positive Suction Head (NPSH) to the pump, during certain accidents. Using IMC 0609, Significance Determination Process, this finding was determined to be of very low safety significance because it was not a design or qualification deficiency, did not represent an actual loss of function of at least a single train for greater than the Allowed Out-of-service Time (AOT) or two separate safety systems out-of-service for greater than the AOT, did not result in a loss of safety function of one or more non-Technical Specification (TS) trains of equipment designated as risk significant for greater than 24 hours, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors). The finding had a cross-cutting aspect of Evaluation of Identified Problems, as described in the Corrective Action component of the Problem Identification and Resolution cross-cutting area, because the licensee did not implement adequate corrective actions to prevent recurrence of unanalyzed material left in containment following the performance of OST-1081 |
| 05000400/FIN-2012004-02 | 2012Q3 | Green | H.7 | Self-revealing | B Startup Transformer Lockout Due to Loss of Oil Filled Cable Pressure | A self-revealing Green NCV of Technical Specification (TS) 6.8.1, Procedures, was identified for the licensees failure to develop an adequate procedure for maintenance on an oil filled cable. Specifically, the licensee failed to provide adequate instructions to prevent causing additional damage to the cable which resulted in the lockout of the B Startup Transformer (SUT) on June 25, 2012. This also resulted in unavailability of the preferred power source for the B safety related equipment for over two days. As corrective actions, the licensee repaired the cable, restored oil pressure and returned the B SUT to its normal standby configuration. Additionally, the licensee performed an investigation which concluded that the cable had been damaged at the site of a previous repair when it was handled during maintenance. The issue was placed into the CAP as AR #545920. The licensees failure to develop an adequate procedure to ensure proper handling of the cable and prevent inadvertently causing damage was a performance deficiency. The performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it resulted in the lockout of the B SUT and unavailability of the preferred power source for the B safety related equipment for over two days. Using IMC 0609, Significance Determination Process, this finding was determined to be of very low safety significance because it was not a design or qualification deficiency, did not represent an actual loss of function of at least a single train for greater than the TS AOT or two separate safety systems out-of-service for greater than the AOT, did not result in a loss of safety function of one or more non-TS trains of equipment designated as risk significant for greater than 24 hours, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors). The finding had a crosscutting aspect of complete, accurate and up-to-date procedures, as described in the Resources component of the Human Performance cross-cutting area, because the licensee did not develop adequate procedures to prevent further damage while performing maintenance on the SUT cables |
| 05000400/FIN-2012005-01 | 2012Q4 | NRC identified | Failure of the Primary Shield Supply Fan (S-2B-SB) to Remain Secure when Stopped | During monthly equipment swapping on October 26, 2012, the licensee attempted to secure the S-2B-SB from the main control board. The fan stopped when the switch was turned to the OFF position, but automatically restarted when the switch was released without a valid start signal. The inspectors identified that the failure of the S-2B-SB to remain stopped when secured from the main control board adversely affected compliance with TS Surveillance Requirements (SR) 4.8.1.1.2 F.4 which verifies operability of the B EDG, B Electrical bus and B sequencer. Additional inspection activities are needed to determine the extent of condition, relative to SR compliance, and if a performance deficiency exists. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000400/2012005- 01, Failure of the Primary Shield Supply Fan (S-2B-SB) to Remain Secure when Stopped. | ||
| 05000400/FIN-2012005-02 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified during the performance of OST-2044 (Radwaste Daily Operations Surveillance Test) on November 27, 2012, that the Waste Processing Building Stack 5 Accident Monitor (RM-WV-3546-1) was reading lower than expected and declared the monitor inoperable. An investigation revealed that at the conclusion of the maintenance activities performed the previous day, the monitor was returned to operable status while its database had incorrect settings. The inaccurate database was determined to have rendered the monitor inoperable. 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states in part that activities affecting quality shall be prescribed by documented instructions and procedures. Contrary to the above, on November 26, 2012, the licensees procedure OWP-RM-19 (Operations Work Procedure Radiation Monitor) failed to adequately prescribe the correct instructions to ensure Radioactive Gaseous Effluent Monitoring Instrumentation received the appropriate post maintenance test. The licensee entered this issue into their CAP as AR #574702 and the monitor database was restored and tested. Using IMC 0609, Significance Determination Process, this finding was determined to be of low safety significance because the finding did not represent an actual dose impact in excess of Appendix I to10 CFR Part 50 or 10 CFR 20.1301(e). | |
| 05000400/FIN-2012007-01 | 2012Q2 | White | P.2 | NRC identified | Failure to Maintain an Adequate EOF to Support Emergency Response | The inspectors identified multiple examples of an Apparent Violation (AV) of 10 CFR 50.54(q) for the lack of facility oversight and control, coupled with component failures and removal of the Emergency Operations Facility (EOF) ventilation system from service (without adequate compensatory measures) which rendered the EOF non-functional on several occasions. Specifically, the licensee failed to ensure that adequate emergency response facilities and equipment were available as required by the Harris Nuclear Plant Emergency Plan, Section 3.1, revision 57, and 10 CFR 50.47(b)(8). The licensee restored the EOF ventilation system to a functional status on November 9, 2011, and entered this issue into their corrective action program (CAP) as Nuclear Condition Report (NCR) 504860. The lack of facility oversight and control, coupled with component failures and removal of the EOF ventilation system from service, which rendered the EOF non-functional on several occasions, was a performance deficiency. The finding was more than minor because it affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the Emergency Response Organization (ERO) Performance attribute was affected during the times when the EOF was not functional and it did not meet 10 CFR 50.47(b)(8) Planning Standard program elements. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, Appendix B Emergency Preparedness Significance Determination Process. Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure to comply; Loss of Risk Significant Planning Standard Function (RSPS), NO; RSPS Degraded Function, NO; Loss of Planning Standard Function, YES; results in a White finding. The NRC concluded that the significance of the finding is preliminarily low to moderate safety significance (White). The licensee restored the EOF ventilation system to a functional status on November 9, 2011, and entered this issue into their CAP as NCR 504860. This finding has a cross-cutting aspect in the Corrective Action Program component of the Problem Identification and Resolution area because the licensee did not identify the issues completely, accurately, and in a timely manner commensurate with their safety significance. Specifically, the licensee did not properly classify, prioritize, or evaluate for operability and reportability of the non-functional EOF. |
| 05000400/FIN-2012007-02 | 2012Q2 | Severity level III | NRC identified | Failure to Notify the NRC of the EOF Loss of Emergency Assessment Capability | The inspectors identified an AV of 10 CFR Part 50.72(b)(3)(xiii), for the failure to report the loss of emergency assessment capability in the EOF. Specifically, the EOF was unavailable to perform its intended function for periods greater than seven days on several occasions from August 2009 to November 2011. This issue was entered into the licensees CAP as NCR 492707. The failure to report the loss of emergency assessment capability in the EOF as required by 10 CFR Part 50.72(b)(3)(xiii) was a performance deficiency. The finding was more than minor because it impacted the regulatory process which depends on plant activities being properly reported. The inspectors evaluated this finding against NRC IMC 0609 Appendix B, Emergency Preparedness Significance Determination Process Section 7.3. The inspectors determined that traditional enforcement is applicable. The licensee failed to report an occurrence of a major loss of emergency assessment capability. Specifically, the licensee failed to maintain a fully functional EOF when portions of the ventilation system were removed from service without compensatory measures, and the licensee failed to report the occurrence as required. As discussed in the Enforcement Policy, the severity level of a violation involving the failure to make a required report to the NRC will be based upon the significance of and the circumstances surrounding the matter that should have been reported. In this case, and as discussed above, the NRC concluded that the failure to provide the required report is associated with a preliminarily White finding for the failure to maintain a fully functional EOF. In addition, the licensees failure to report the condition of the EOF from August 2009 to November 2011, as required by 10 CFR 50.72, impeded the NRCs regulatory process. Had the licensee reported the incident as required, NRC review and follow-up inspection likely would have occurred, which may have prompted the licensee to adopt compensatory measures and/or corrective actions, thereby precluding the incidents that followed after August 4, 2009. Based on the above, the NRC determined the severity level of this apparent violation is preliminarily Severity Level III in accordance with the NRC Enforcement Policy. | |
| 05000400/FIN-2012007-03 | 2012Q2 | Green | NRC identified | Failure to Properly Install the Electrical Power Feed Cables for the EOF | The inspectors identified a Green Non-Cited Violation (NCV) of 10 CFR 50.54(q) for the licensees failure to properly install the electrical power feed cables for the EOF in accordance with the national electrical code (NEC) as required by the Harris Emergency Plan, PLP-201, Revision 57, section 3.5.1.D. Specifically, the licensee failed to ensure that an adequate emergency response facility, EOF was available as required by the Harris Nuclear Plant Emergency Plan, Section 3.5, revision 57, and 10 CFR 50.47(b)(8). This issue was in the licensees CAP as NCR 381658. Upon completion of the corrective actions, the power feed cables and supports met the requirements of NEC Article 230.51 C. The licensees failure to properly install the electrical power feed cables for the EOF in accordance with the NEC as required by the Harris Emergency Plan, PLP-201, Revision 57, section 3.5.1.D was a performance deficiency. The finding was more than minor because it affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the Facilities and Equipment attribute was affected during the time when the EOF was degraded due to the power feed cables not being installed in accordance with the NEC, which resulted in not meeting the 10 CFR 50.47(b)(8) Planning Standard program elements. The finding was assessed for significance in accordance with NRC IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process. Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure to Comply; Loss of Risk Significant Planning Standard Function (RSPS), NO; RSPS Degraded Function, NO; Loss of Planning Standard Function, NO; results in a Green finding. The inspectors determined that this resulted in a low safety significance finding (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the performance deficiency occurred twelve years earlier when the power feed cables were initially installed and does not represent current licensee performance. | |
| 05000400/FIN-2012007-04 | 2012Q2 | NRC identified | Failure to Maintain an Adequate TSC to Support Emergency Response | The inspectors identified an AV of 10 CFR 50.54(q) for the licensees failure to provide a defensible technical basis for unfiltered air in-leakage, supported by sufficient experimental and empirical data for an input to a calculation used as the basis for TSC functionality. The compensatory measure established on February 16, 2012, was to issue a standing order (12-005) related to habitability and relocation of the TSC. The licensee has submitted an event notification (EN 47655), and entered this issue into their CAP as NCR 516120. The licensees failure to provide a defensible technical basis supported by sufficient experimental and empirical data for an input to the Alternate Source Term (AST) calculation, which was the basis for TSC functionality, was a performance deficiency. This failure resulted in the licensee being unable to meet the TSC habitability requirements as specified in the Harris Emergency Plan, PLP-201, Revision 57, section 3.3.1. The finding was more than minor because it affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the ERO performance attribute was affected during the times when the TSC was not functional, and it did not meet 10 CFR 50.47(b)(8) Planning Standard program elements. The finding was assessed for significance in accordance with NRC IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process. Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure to comply; Loss of Risk Significant Planning Standard Function (RSPS), NO; RSPS Degraded Function, NO; Loss of Planning Standard Function, YES; results in a White finding. The inspectors determined that this resulted in a preliminarily low to moderate safety significance finding (White). The inspectors did not identify a cross-cutting aspect associated with this finding because the performance deficiency occurred in 2001 and does not represent current licensee performance. | ||
| 05000400/FIN-2012008-01 | 2012Q2 | NRC identified | B AND C MSIVs FAIL TO CLOSE DURING SURVEILLANCE TESTING | The inspectors identified an URI associated with issues in the licensees MSIV maintenance and testing. These issues were potential contributing causes to the April 21, 2012, B and C MSIV failure to stroke close. Description: Several issues were identified regarding the licensees MSIV maintenance and testing. Some of the issues identified were: FnIn the last two refueling intervals, maintenance was making minor adjustments to the actuator hydraulic speed control system to decrease the time needed to shut the valves as a result of increasing stroke test closure time results. FnBeginning in 2001, work deficiency documents were initiated due to the MSIVs experiencing difficulty in opening during refueling outage cycling. There had not been any corrective maintenance conducted requiring valve internal disassembly and the licensee had not developed any periodic PMs to visually inspect the condition of valve internals. FnThe valve vendor manual recommended weekly valve partial exercising ten percent of its total stroke in order to assure that the actuator and valve was properly functioning. Prior to 2000, this partial exercising was being performed quarterly. In 2000, the licensee revised their IST program requirements to discontinue quarterly exercising in lieu of the 18-month cold shutdown TS stroke testing that was currently being conducted. FnPrior to the current MSIV failures; the MSIVs had never been tested as part of the licensees AOV program. Summary: The licensees root cause investigation was not completed at the conclusion of the special inspection; the determination as to whether these issues represented performance deficiencies was not completed. Pending completion of the licensees root cause evaluation (RCE) and subsequent NRC review to determine if a performance deficiency exists, disposition of these issues will be tracked via Unresolved Item (URI) 05000400/2012008-01, B and C MSIVs Fail to Close During Surveillance Testing. | ||
| 05000400/FIN-2012009-01 | 2012Q4 | Green | NRC identified | Technical Specification Inoperability of MSIVs Due to Failure to Conduct Diagnostic Testing | The inspectors identified a non-cited violation of Technical Specification (TS) 3.7.1.5, Main Steam Line Isolation Valves, due to one or more MSIVs being inoperable for a time greater than the allowed outage time and a plant shutdown was not completed in accordance with the action statement of TS 3.7.1.5. MSIV diagnostic testing in accordance with EGR-NGGC-0205, Air Operated Valve (AOV) Reliability Program, had not been conducted by the licensee. This contributed to the licensee not identifying long-term corrosion/oxidation of the valve piston rings that resulted in the B and C MSIV failure to initially close during stroke time testing on April 21, 2012. The licensee conducted repairs of all three MSIVs and restored them to an operable condition prior to entering Mode 4 following the completion of an ongoing refueling outage. The licensee entered this condition into their corrective action program (CAP) as Nuclear Condition Report (NCR) 531773. The failure to properly classify the MSIVs as risk significant and implement MSIV diagnostic testing in accordance with the AOV program procedure EGR-NGGC-0205 was a performance deficiency (PD). The PD is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objectives of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is also associated with the containment isolation barrier performance attribute of the Barrier Integrity cornerstone and 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 failure to conduct periodic diagnostic testing that would have identified long-term internal valve degradation due to unexpected corrosion/oxidation of the valve piston rings in all three MSIVs resulted in two MSIVs failing to initially close during TS stroke time testing on April 21, 2012, and excessive internal friction in all three MSIVs such that they may not have been capable of performing their safety-related closure function during certain design basis events. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At- Power, the inspectors determined there was an actual loss of safety function greater than the TS allowed outage time associated with the finding which required a more detailed risk evaluation. A detailed risk evaluation was performed by a regional senior reactor analyst. The result of the analysis of the risk of the PD was a delta core damage frequency (CDF) of <1E-6/year and a delta Large Early Release Fraction (LERF) of <1E- 7/year, a GREEN finding. No cross-cutting aspect was assigned to this finding because licensee decisions made in regard to classifying the MSIVs in the AOV program were made more than three years ago and therefore, not reflective of current plant performance. | |
| 05000400/FIN-2012201-01 | 2012Q1 | Green | H.14 | NRC identified | Security | |
| 05000400/FIN-2012201-02 | 2012Q1 | Green | P.1 | NRC identified | Security | |
| 05000400/FIN-2012403-01 | 2012Q3 | Green | NRC identified | Security | ||
| 05000400/FIN-2013002-01 | 2013Q1 | Green | P.3 | NRC identified | Inadequate Correction Actions Involving the Incorrect Determination of Operability | An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action was identified for the licensees failure to take corrective actions related to incorrect operability determinations which resulted in violation of TS 3.8.1.1 (Electrical Power Sources) associated with the S-2B-SB failure to secure on October 26, 2012. The licensee entered the issue into their CAP as AR #569593. As corrective actions, on October 31, 2012, Operations opened the supply breaker (1B21-SB-4B) for the primary shield fan to remove any impact to the Emergency Diesel Generator (EDG) operability. Additionally, the licensee created AR #584473 to evaluate and correct issues associated with their operability determinations. The licensees failure to take timely, appropriate corrective actions for inadequate operability determinations was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to take timely, appropriate corrective actions could have resulted in a more safety significant violation of TS than the identified violation of TS 3.8.1.1 (Electrical Power Sources) associated with the S- 2B-SB failure to secure on October 26, 2012. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve a deficiency affecting the design or qualification of a mitigating system and did not represent a loss of system function. The cause of the finding was directly related to the cross-cutting aspect for appropriate corrective actions to address safety issues in a timely manner commensurate with their safety significance and complexity in the CAP component of the cross-cutting area of Problem Identification and Resolution, in that the licensee failed to take appropriate and timely corrective actions to address incorrect determinations of operability |
| 05000400/FIN-2013002-02 | 2013Q1 | Green | NRC identified | Failure to Implement Design Control Measures for the EDG Starting and Control Air System | The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, involving two examples. In one example, the licensee did not translate instrument uncertainties associated with the EDG low-pressure alarm and pressure indicator into operating and alarm response procedures. In the second example, the licensee failed to verify the design adequacy for blocking the EDG nonemergency generator trips during emergency operation. The licensee entered the first example into their CAP as ARs #586788, #586837, #588517, and #589308 and initiated a standing instruction to verify starting air pressure was maintained above 200 psig while evaluating appropriate corrective actions. The licensee entered the second example into their CAP as ARs #382359 and #412546, and implemented a facility change to correct the design deficiency. The failure to translate instrument uncertainties associated with the EDG low-pressure alarm and pressure indicator into operating and alarm response procedures, and failure to verify the design adequacy for blocking the EDG non-emergency generator trips were performance deficiencies. The performance deficiencies were more than minor because they were associated with the Design Control attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors assessed the finding using IMC 0609 Attachment 4, Initial Characterization of Findings; and IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, and determined the finding was of very low safety significance (Green) because the design deficiencies were confirmed not to result in loss of operability of the EDGs. The finding was reviewed for cross-cutting aspects and none were identified since the performance deficiencies were not indicative of current licensee performance. | |
| 05000400/FIN-2013002-03 | 2013Q1 | NRC identified | Solid State Protection System Digital Modification | The inspectors identified an unresolved item (URI) associated with the licensees implementation of a digital modification to the solid state protection system (SSPS) logic and control boards. This item remains unresolved pending NRC staff review of additional information to determine if the change could have been performed under a 10 CFR 50.59 evaluation and whether it should have been submitted to the NRC for prior approval. The SSPS logic and control boards provide the coincidence logic to produce actuation signals for operation of the reactor protection system (RPS) and the engineered safety features actuation systems (ESFAS). Engineering change 78484, Replace SSPS Boards with new Westinghouse Design Boards, Rev. 6, examined a digital modification to the existing SSPS logic and control boards. The original boards used fixed logic devices (transistor-transistor logic devices) whereas the replacement boards use reprogrammable logic devices (complex programmable logic devices (CPLDs)). The licensee performed a 10 CFR 50.59 Screening (AR 537776) using procedure REG-NGGC-0010, 10 CFR 50.59 and Selected Regulatory Reviews, Rev. 18. The procedure used the guidance in NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Rev. 1, as supplemented by NEI 01-01, Guidelines on Licensing Digital Upgrades, Rev. 1, to evaluate the design and implementation of digital modifications to instrumentation and control systems under 10 CFR 50.59. The licensees screening indicated (in summary) that the new design boards performed the same functions and were functionally tested; therefore, did not adversely affect the SSPS design bases functions previously evaluated in the UFSAR. The screening further determined the modification could be implemented without a more detailed 10 CFR 50.59 evaluation. The inspectors reviewed the screening using the licensees procedural guidance and determined the modification adversely affected the SSPS design functions described in the UFSAR because: (1) The response times of the new design boards were slower. Section 4.3.3 of NEI 01-01, Other Digital Issues in the Screening Process, indicates that performance changes from UFSAR described requirements (i.e. response time) should be screened in and require further evaluation under 10 CFR 50.59. (2) Human System Interface (HIS) features (i.e. dip switches, RS-232 communication ports, and indicating light-emitting diodes or LED) were added. Section 4.3.4 of NEI 01-01, Screening Human System Interface Changes, indicates that changes that create new potentional failure modes in the interaction of operators and maintenance personal with the system should be further evaluated for the potential increase in the likelihood of malfunctions. (3) The new boards were loaded with a data file (which NEI 01-01 defines as a type of base software) that configures the CPLD logic. Section 4.3.2 of NEI 01- 01 Software Considerations, indicates that digital modifications that involve the use of software applications should be conservatively treated as an adverse effect (requiring evaluation under 10 CFR 50.59) due to the potential introduction of new failure modes (software based failures, including Common Cause Failures (CCF)) not previously evaluated in the UFSAR, especially when modifications involve redundant high risk safety systems (i.e. RPS.ESFAS) In response to the inspectors questions, the licensee performed a 10 CFR 50.59 evaluation (AR #588797) and determined the change could be implemented without prior NRC review and approval. The licensee indicated that (1) the new boards still met the response time requirements for the SSPS as described in the UFSAR, (2) the HIS vulnerabilities were mitigated by configuration at the vendor facility, and (3) the CPLDs were not software-based and that the data files were simple logic files that were fully tested, verified, and validated to operated as expected. The licensee asserted that the development and quality assurance processes used, including design, verification & validation, and configuration control mitigated any potential increase in the likelihood of malfunctions due to software (or embedded data file) (10 CFR 50.59 criteria (c)(2)(ii)). The licensee also compared the hardware functional testing performed by the vendor with criteria in Branch Technical Position (BTP) 7-19, Guidance for Evaluation of Diversity and Defense-in-Depth in Digital Computer-Based I&C Systems, Rev. 6, section 1.9, to show that software CCFs required no further evaluation. Specifically, the licensee indicated that the functional testing for the boards was adequate for 100 percent testing for every possible combination of inputs and every possible sequence of device states were tested and all outputs were verified on the boards (and embedded software) to eliminate consideration of software based CCF. Based on this testing, the licensee concluded that the use of software did not create a possibility of malfunctions of the SSPS with a different result than previously evaluated in the UFSAR (10 CFR 50.59 criteria (c)(2)(vi)). After reviewing the 10 CFR 50.59 evaluation, the inspectors found that they did not have sufficient information to determine that NRC review and approval was not required prior to implementation of the modification. Specifically, the inspectors could not verify the licensees conclusions regarding the software reliability and the simplicity and testing of the new boards. Because the licensee claimed that the CPLDs were not softwarebased, the licensee did not address the software development processes described in NEI 01-01, section 5.3.3, Digital System Quality. Specifically, the inspectors noted that second and third party commercial vendors were involved in the manufacturing of the CPLDs and development of the base software data-file without a quality software development process as addressed in NEI 01-01. In addition, because of the licensees claim that the CPLDs were not software-based, the licensee excluded the possibility of software CCF as addressed in NEI 01-01, section 3.2.2, Software Common Cause Failure. The inspectors concluded that software CCF of the SSPS could introduce new failure modes not previously analyzed in the UFSAR. With respect to the simplicity and testing of the SSPS boards, the inspectors questioned the simplicity of the boards and the appropriateness of using testing to rule out consideration of CCFs. In addition, the testing performed by the licensee did not meet the guidance in BTP 7-19. The inspectors also concluded that the HSI features added to the SSPS boards provided additional risk of failures not associated with the original SSPS boards when used by operators and maintenance personnel. In order to determine if the change could have been performed under a 10 CFR 50.59 evaluation and whether it should have been submitted to the NRC for prior approval, this issue remains unresolved pending NRC staff review of additional information to be provided by the licensee to address the issues described above. This issue is being tracked as URI 05000400/2013002-03, Solid State Protection System Digital Modification. | ||
| 05000400/FIN-2013002-04 | 2013Q1 | NRC identified | Number 1 Reactor Coolant Pump Seal Leakoff Line OVER-PRESSURIZATION | The inspectors identified a URI associated with licensees capability to meet their station blackout (SBO) mitigation strategy. This item remains unresolved pending the inspectors review of the additional information to determine compliance with 10 CFR 50.63, Loss of All Alternating Power. The reactor coolant pumps (RCP) No.1 seal leakoff line was designed to recover leakoff volume, at low pressure and temperature, and return it to the chemical and volume control system (CVCS). The leakoff lines (one per pump) join into a common header before exiting containment to the CVCS. In 1992, Westinghouse Technical Bulletin, NSD-TB-91-07-R1, Over-Pressurization of RCP No.1 Seal Leakoff Line, informed specific licensees (including Harris) of the potential over-pressurization of the No.1 seal leakoff line during high seal leakoff flow conditions as a result of abnormal performance of the No.1 RCP seal. Specifically, the leakoff line pipe segment downstream of the air operated valve (which fails open on loss of instrument air) and upstream of the flow element restriction orifice was designed to 150 psig and could overpressurize and fail under high flow conditions. While Harris had implemented recommendations contained in the bulletin, the licensee did not upgrade the piping for higher pressures nor evaluate the line capability to handle expected seal leakage flow rates associated with loss of seal cooling (LOSC) events documented in Westinghouse Owner Group Report WCAP-10541 Reactor Coolant Pump Seal Performance Following a Loss of All AC Power, Rev 2. The technical bulletin specifically stated that the validity of the information in WCAP-10541 was dependent upon the assumption that the integrity of the leakoff line was maintained. The inspectors reviewed WCAP-10541 and noted that the leakoff line could experience a pressure transient between 800-2000 psig during a LOSC event before seal leakage flow rates stabilize at approximately 21gpm and 800psig. The report also indicated that the backpressure provided by the leakoff line (upstream of the orifice) is what limits seal leagage to 21gpm and reduction of this backpressure would result in higher seal leakage flow rates. In 2003, Information Notice (IN) 2003-19, Unanalyzed Condition of Reactor Coolant Pump Seal Leakoff Line During Postulated Fire Scenarios or Station Blackout, informed licensees of specific LOSC events (SBO and fires coincident with loss of all AC events) that could over-pressurize and fail the leakoff line. The IN reemphasized the pressures that will be experienced by the leakoff line (800-2000 psig) and that the failure of the line would result in RCP leak rates in excess of the 21gpm determined by Westinghouse and the 25 gpm assumed in SBO coping analyses. The inspectors reviewed the licensees evaluation of IN 2003-19 (AR #1069790-09) and determined the licensees actions did not adequately address the potential for over-pressurizing the seal leakoff line. The licensee entered the issue into the CAP as AR #589248 and indicated that the alternate seal injection (ASI) system, installed in December of 2010 to meet the sites new fire protection program requirements (NFPA-805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants), automatically starts on a LOSC event (SBO or fire related) and will maintain seal cooling. By maintaining seal cooling, the RCP seal leakage flow rates are expected to remain at nominal operating values (2-5 gpm) and prevent seal leakage flow rates that would challenge the integrity of the No.1 seal leakoff line. The inspectors questioned the appropriateness of crediting the ASI system for SBO events and the systems capability to prevent overpressurization of the leakoff line. Specifically: The inspectors noted that the ASI system was not credited for meeting the current licensing bases for SBO. The ASI has a delayed start and the inspectors questioned whether seal cooling would be restored before seal leakage increases to the point of challenging the leakoff line. This issue remains unresolved pending the inspectors review of additional information to be provided by the licensee to address the issues described above and determine compliance with 10 CFR 50.63, Loss of All Alternating Power. This issue is being tracked as: URI 05000400/2013002-04, No. 1 Reactor Coolant Pump Seal Leakoff Line Over-Pressurization. | ||
| 05000400/FIN-2013002-05 | 2013Q1 | Green | P.5 | Self-revealing | Reactor Power Transient Due to Inadvertent Isolation of the 4B Feedwater Heater | A self-revealing Green finding (FIN) was identified for the licensees failure to establish and implement an adequate operating procedure (OP-136, Feedwater Heaters, Vents and Drains, Revision 41) to restore the 4B feedwater heater (FWH) alternate level control valve (1HD-323) to automatic operation. The licensee entered this issue into the Corrective Action Program (CAP) as Action Request (AR) #592336. The licensee took corrective action to reduce reactor power immediately and revise OP-136 to include a power reduction prior to restoring 1HD-323 to automatic operation. The licensees failure to establish and implement an adequate operating procedure (OP-136, Feedwater Heaters, Vents and Drains, Revision 41) to restore 1HD-323 to automatic operation was identified as a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, failure to establish and implement an adequate operating procedure resulted in a steam plant transient that caused an unplanned reactor power increase to 101.1 percent Rated Thermal Power (RTP). In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve the complete or partial loss of a support system that contributes to the likelihood of an initiating event and it did not affect mitigation equipment. The finding has a cross-cutting aspect of Implements and Institutionalizes Operating Experience, as described in the Operating Experience component of the Problem Identification and Resolution cross-cutting area because the licensee failed to institutionalize operating experience from the previous month. |
| 05000400/FIN-2013002-06 | 2013Q1 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being disposition as a Non-Cited Violation. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action requires, in part, that in the case of significant conditions adverse to quality corrective actions shall be taken to preclude repetition. Contrary to this requirement, corrective actions taken after the Containment Pre-Entry Purge Outside Containment Isolation Valve (1CP-1) failed EST-220, Type C Local Leak Rate Test on February 23, 2004 failed to preclude repetition (AR #119086). Specifically, the licensee failed to incorporate adequate guidance to re-torque the stud bolts on the seat clamping ring into procedure CMM0225. This resulted in 1CP-1 failing EST-220, Type C Local Leak Rate Test due to excessive leakage again on December 3, 2012. This violation was determined to be of very low safety significance (Green) because the finding did not represent an actual open pathway and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The licensee entered this issue into their CAP as AR #575878. As corrective actions, the licensee revised the seat replacement procedure, properly torqued the stud bolts and satisfactorily tested 1CP-1. | |
| 05000400/FIN-2013003-01 | 2013Q2 | Green | P.2 | Self-revealing | Power Transient due to a Main Feedwater Pump Oil Leak | A self-revealing Green finding (FIN) was identified for the licensees failure to adequately implement their procedure CAP-NGGC-0205, Condition Evaluation and Corrective Action Process, for two oil leaks from the B MFP which occurred on February 14, 2013 and February 17, 2013. Specifically, these failures resulted in a significant oil leak on the B MFP which required a rapid downpower to 55 percent RTP on March 29, 2013. The licensee entered this finding into their CAP as Action Request (AR) #598302. The licensee took corrective action to perform a design change to the breather to correct the plant issue. The licensees failure to adequately implement their procedure CAP-NGGC-0205 was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, this failure resulted in another oil leak on the B MFP which required a rapid downpower to 55 percent RTP on March 29, 2013. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not cause a reactor trip or the loss of mitigation equipment. The finding had a cross-cutting aspect of Evaluation of Identified Problems, as described in the Corrective Action component of the Problem Identification and Resolution cross-cutting area because the licensee failed to thoroughly evaluate the two oil leaks in February 2013 to ensure that the resolution addressed the cause, resulting in the transient on March 29, 2013. |
| 05000400/FIN-2013003-02 | 2013Q2 | NRC identified | Evaluate the Effects of Environmental Air Samplers Collecting Diluted Airborne Particulate Samples | The inspectors identified an Unresolved Item (URI) regarding environmental air samplers that may not be collecting particulate samples that are representative of actual airborne radionuclide concentrations. During an observation of environmental sample collection, the inspectors noted that three of the nine environmental air samplers (locations 63, 90, and 91) contained a housing fan that appeared to blow exhaust air directly onto the inlet of the particulate/iodine sampling head. The inspectors also noted that the interior walls of the air sampler housing contained a fibrous insulation media that appeared to act as a mechanical filter due to the particulate loading stains readily visible. Since the housing fan draws air from the interior of the enclosure, and blows the exhaust directly onto the air sampler intake, the potential exists for the air sampler to collect a non-representative (diluted) sample. This arrangement could result in a non-conservative measurement of airborne particulate radionuclide concentrations. Discussions with environmental monitoring staff indicated that the three air samplers in question have been in service for approximately five years. The inspectors concluded that more information is required to determine if there is a performance deficiency. URI 05000400, 2013003-02 Evaluate the Effects of Environmental Air Samplers Collecting Diluted Airborne Particulate Samples. | ||
| 05000400/FIN-2013004-01 | 2013Q3 | Green | H.5 | NRC identified | Failure to Compensate for a Blocked Open Fire Door | The inspectors identified a Green non-cited violation (NCV) of the Shearon Harris Nuclear Power Plant Operating License NPF-63 condition 2.F, Fire Protection Program, and 10 CFR 50.48(c), National Fire Protection Association (NFPA) Standard 805, for failing to implement required compensatory measures per licensee procedure FPP-013, Fire Protection. Specifically, the licensee failed to establish an hourly fire watch and stage backup fire suppression equipment for a blocked open fire door (FD- 241) between the A and B safety related switchgear rooms on September 9, 2013. The licensee took corrective action by restoring the fire door to functional. The licensee entered this into the corrective action program (CAP) as Action Request (AR) #627493. The failure to implement fire compensatory measures in accordance with licensee procedure FPP-013, Fire Protection during the two hour exposure period when the fire door between Switchgear Rooms A and B was propped open was determined to be a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this failure inadvertently bypassed a three hour fire barrier and created the potential for a fire to affect both safety related switchgear rooms. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609 attachment 4 which determined that an evaluation using NRC IMC 0609 Appendix F (Fire SDP) was required. The propped open door constituted a high degradation condition per NRC IMC 0609 appendix F Attachment 2 which required a detailed risk evaluation. A bounding phase 3 risk analysis was done by a regional SRA using a hand calculation and guidance from NRC IMC 0609 Appendix F. The major analysis assumptions included a duration factor of 2 hours, an ignition frequency of 2E- 2/year, a base case conditional core damage probability (CCDP) of 0.1 (assumed large single SWGR room fire would require alternate safe shutdown), a non-conforming case CCDP of 1.0 (assumed a dual SWGR fire scenario would result in core damage), and a probability of non-suppression (PNS) of 1E-3. The dominant sequence was a challenging SWGR room fire which remained unsuppressed long enough to develop into a damaging hot gas layer scenario which would fail SSD equipment in both SWGR rooms A and B due to the open fire door and result in core damage. The licensees fire PRA produced similar results. The short exposure period, ability to close the fire door, and the low PNS mitigated the risk. The analysis result was an increase in core damage of < 1E-6/year, a finding of very low safety significance (Green). The finding had a cross-cutting aspect of Work Planning, as described in the Work Control component of the Human Performance cross-cutting area because the licensee failed to identify the need for a compensatory action due to the blocked open fire door. |
| 05000400/FIN-2013005-01 | 2013Q4 | Green | P.6 | NRC identified | Failure to Maintain Environmental Qualification for Electric Equipment | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.49 for the failure to adequately implement the environmental qualification (EQ) program for electric equipment important to safety. Specifically, between September 2013 and November 2013, multiple EQ program deficiencies were identified including design documentation and the qualification of electric equipment installed in the plant. The licensee took corrective action to repair or schedule repair for all of the identified issues. The licensee entered these issues into the CAP as AR #663071. The inspectors determined that the failure to completely implement the EQ program as required by 10 CFR 50.49 was a performance deficiency. Specifically, between September 2013 and November 2013, multiple EQ program deficiencies were identified including design documentation and the qualification of electric equipment installed in the plant. This finding was more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern if the functions of other components in the EQ program are challenged. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 2- Mitigating Systems Screening Questions, the inspectors determined this finding to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of equipment. The finding had a cross-cutting aspect of Conducts Self- Assessments, as described in the Self and Independent Assessments component of the Problem Identification and Resolution cross-cutting area because the licensee failed to identify these issues during their recent self-assessments. |
| 05000400/FIN-2013005-02 | 2013Q4 | NRC identified | Potential Performance Deficiency Associated with Oxygen Concentration in the Waste Gas System | On November 11, 2013, the licensee was performing a degassing evolution to the waste gas system while the oxygen and hydrogen analyzers were inoperable and the hydrogen recombiner was not functional. The inspectors identified that the licensees procedures and compensatory measures for degassing without all equipment in service may have been inadequate to properly control the oxygen concentration in the waste gas system. The inspectors are awaiting the completion of the licensees root cause evaluation to determine the licensees compliance with applicable procedures and TS 3.11.2.5, which limits oxygen concentration in the waste gas system. Additional inspection activities are needed to determine the extent of condition, relative to procedural and TS compliance during operation of the waste gas system. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000400/2013005-02, Potential Performance Deficiency Associated with Oxygen Concentration in the Waste Gas System. | ||
| 05000400/FIN-2013005-03 | 2013Q4 | Green | H.6 | Self-revealing | Inadequate Transformer Preventive Maintenance Procedure | The inspectors determined that inadequate testing prescribed by procedure NGGPMB- XFM-02 and performed under procedure PM-E0015 was a performance deficiency. Specifically, licensee procedure ADM-NGGC-0107, Equipment Reliability Process Guideline, resulted in the determination that the 1E2 transformer was a critical component. Licensee procedure NGG-PMB-XFM-02, Equipment Reliability Template for Dry-Type Transformers, states that critical components are maintained to not allow any failure that would result in a trip, transient, or significant challenge to continued safe operation. However, implementing procedure PM-E0015, failed to contain steps to identify degradation in the 1E2 transformer windings prior to failure. This finding was more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operation. Specifically, a power transient resulted from the 1E2 failure. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 1- Initiating Events Screening Questions, the inspectors determined this finding to be of very low safety significance (Green) because the finding did not cause 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 (e.g. loss of condenser, loss of feedwater). The finding had a cross-cutting aspect of Long Term Safety, as described in the Resources component of the Human Performance cross-cutting area because the licensees evaluation of the transformer PM program in March 2012 removed additional testing which might have indicated that the transformer windings had experienced insulation degradation. |
| 05000400/FIN-2013009-01 | 2013Q3 | Green | H.2 | NRC identified | Failure to Submit a License Amendment Request for a Digital Modification to the Solid State Protection System | The inspectors identified a SL IV Green NCV of 10 CFR 50.59, Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment before implementing a change that created the possibility of a malfunction of a system, structure, or component important to safety with a different result than previously evaluated. The licensee did not follow guidance in Nuclear Energy Institute document NEI 01-01, Guidelines on Licensing Digital Upgrades, Rev. 1, (referenced in licensee Procedure EGR-NGGC-0157, Engineering of Plant Digital Systems and Components, Rev. 7), which resulted in the licensee implementing a change that created the possibility of common cause software malfunctions of the reactor protection system and engineered safety features actuation systems not previously evaluated in the Updated Final Safety Analysis Report. This failure to follow NEI guidance when implementing a change was a performance deficiency. The licensee entered this issue into their corrective action program, performed an evaluation that provided a reasonable expectation of operability, and initiated development of a license amendment request. 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 (i.e., core damage). Additionally, in accordance with the guidance in the NRC Enforcement Manual, the 10 CFR 50.59 violation was more than minor because there was reasonable likelihood that the change would require NRC approval prior to implementation. The inspectors evaluated the significance of the finding using IMC 0609, The Significance Determination Process, and determined the finding was of very low safety significance (Green). In accordance with the Enforcement Policy, the violation of 10 CFR 50.59 was determined to be a SL IV violation because it resulted in a condition evaluated as having very low safety significance (i.e., Green) by the SDP. The finding had a cross-cutting aspect in the Decision Making component of the Human Performance area because the most significant causal factor of the performance deficiency was that the licensee failed to oversee the work activities of vendors such that nuclear safety was supported. |
| 05000400/FIN-2013010-01 | 2013Q2 | Severity level IV | NRC identified | Failure to Report a Degraded Primary Safety Barrier Per 10 CFR 50.73(A)(2)(II)(A) | The inspectors identified a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.73(a)(2)(ii)(A) for the licensees failure to submit a 60-day Licensee Event Report (LER) for a condition in which one of the plants principal safety barriers was seriously degraded. The licensee generated Action Request 00606893 to document the failure to provide the required 60-day LER. The inspectors determined that the failure to report a seriously degraded principal safety barrier as required by 10 CFR 50.73(a)(2)(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 50.73 is a Severity Level IV violation. Cross-cutting aspects are not assigned to traditional enforcement violations. | |
| 05000400/FIN-2013010-02 | 2013Q2 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, which requires, in part, that measures shall be established to assure that conditions adverse to quality, such as deficiencies, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to this requirement, while performing ultrasonic examinations on the reactor pressure vessel head during the spring 2012 refueling outage as required by 10 CFR 50.55a(g)(6)(ii)(D), the licensee failed to identify an unacceptable indication in Nozzle 49 that overlapped the J-groove weld and exhibited characteristics of primary water stress corrosion cracking. This finding was determined to be of very low safety significance because subsequent visual and volumetric examinations performed did not detect any leakage and sizing of the indication determined that structural integrity of the vessel head was not compromised. Additionally, the licensee reanalyzed 100% of the spring 2012 inspection data and did not discover any further missed indications. The licensee entered this condition in their corrective action program as Action Request 00606317. | |
| 05000400/FIN-2014002-01 | 2014Q1 | Green | H.1 | NRC identified | Failure to Adequately Perform the New Fuel Oil Surveillance Requirement | The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 6.8.1.a, Procedures and Programs, for the licensees failure to have an adequate surveillance test to implement the requirements of SR 4.8.1.1.2.c, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, Appendix A, Section 8.b. Specifically, licensee procedure RST-209, Technical Specification Surveillance of New Diesel Fuel Oil (DFO), did not adequately ensure a representative sample of the DFO to confirm the required properties prior to addition to the B diesel fuel oil storage tank (DFOST). This created the potential for DFO of an unacceptable quality to be introduced to the B emergency diesel generator (EDG) on December 4, and 6, 2013. The licensee took corrective action by testing the fuel oil in the A and B DFOSTs and EDG day tanks to verify that the DFO met the required properties as outlined in TS. Additionally, the licensee planned to revise RST-209 and established interim actions to prevent adding new fuel oil prior to obtaining a representative sample. The inspectors determined that the failure to have an adequate surveillance test to implement the requirements of SR 4.8.1.1.2.c. on December 4, and 6, 2013 was a performance deficiency. Specifically, this created the potential for fuel oil of an unacceptable quality to be introduced to the B EDG. This finding was more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern in that it could have affected operability of the EDGs. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined this finding to be of very low safety significance (Green) because the finding is not a deficiency affecting the design or qualification and does not represent an actual loss of system and/or function. The finding had a cross-cutting aspect of Resources, as described in the Human Performance crosscutting area because the licensee failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, procedure RST-209 Revision 23 inappropriately permitted the use of data from a sample that was 20 months old to meet SR 4.8.1.1.2.c. |
| 05000400/FIN-2014002-02 | 2014Q1 | Green | P.3 | Self-revealing | Failure to Prevent Recurrence of a Signifiant Condition Adverse to Quality | A self-revealing Green finding was identified for the failure to implement an adequate corrective action to prevent recurrence (CAPR) for a Significant Condition Adverse to Quality (SCAQ) as required by licensee procedure CAP-NGGC-0205, Condition Evaluation and Corrective Action Process, resulting in the failure of the 1D2 transformer on January 18, 2014. Specifically, after the 1E2 transformer failed on August 8, 2013, the licensee determined the event to be a SCAQ, but failed to implement an adequate CAPR to prevent the failure of the 1D2 transformer. The licensee entered this issue into the corrective action program (CAP) as Action Request (AR) #663324. As corrective action, the licensee is replacing the 1D2 transformer and other similar transformers and implemented additional testing to aid in the identification of degradation prior to transformer failure. The inspectors determined that the failure to implement an adequate CAPR for a SCAQ was a performance deficiency. This finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operation. Specifically, a manual reactor trip resulted from the 1D2 failure. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 1- Initiating Events Screening Questions, the inspectors determined this finding to be of very low safety significance (Green) because the finding did cause a reactor trip but did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of condenser, loss of feedwater). The finding had a cross-cutting aspect of Resolution, as described in the Problem Identification and Resolution cross-cutting area because the licensee did not implement effective corrective actions to address the issue in a timely manner commensurate with their safety significance. Specifically, the licensees CAPR for the August 8, 2013, event did not resolve the cause for transformer failures. |
| 05000400/FIN-2014002-03 | 2014Q1 | Green | H.8 | NRC identified | Failure to Comply With Technical Specification | The inspectors identified a Green NCV of TS 3.11.2.5, Explosive Gas Mixture, for the failure to implement the actions of the limiting condition for operation (LCO). Specifically, during shutdown plant operations in November 2013, the licensee identified oxygen concentrations in the gaseous radwaste treatment system (GRTS) of greater than two percent oxygen, with hydrogen concentration greater than four percent and did not enter nor take the actions of TS LCO 3.11.2.5. The licensee entered the issue into their CAP as AR #651188 and reduced the oxygen concentration to less than two percent on December 11, 2013. The licensees failure to enter and implement the actions of TS LCO 3.11.2.5, once oxygen concentrations exceeded two percent, with hydrogen concentrations greater than four percent within the GRTS was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, it would have the potential to lead to a more significant safety concern such as an explosive gas mixture. Specifically, on November 11, 2013, SR 4.11.2.5 was performed unsatisfactorily; Operations was unaware of the results and did not implement the actions of TS LCO 11.2.5. Using IMC 0609, SDP, Appendix A, Exhibit 2-External Event Mitigation Systems Screening Questions, the inspectors determined this finding to be of very low safety significance (Green) because it was a deficiency that did not result in a degradation or loss of system function. The finding had a cross-cutting aspect of Procedure Adherence, as described in the Human Performance cross-cutting area because the licensee failed to comply with RST-202, Hydrogen and Oxygen Surveillance of the GRTS, and notify Operations of the unsatisfactory test result. |
| 05000400/FIN-2014003-01 | 2014Q2 | Green | H.13 | NRC identified | Failure to Adequately Implement a Plant Modification | The inspectors identified a finding of very low safety significance (Green) when the licensee did not adequately implement the procedural requirements of ADM-NGGC-0106, Configuration Management Program Implementation, during the installation of a temporary modification to install temporary air compressors on May 31, 2014. The licensee entered the issue into their Corrective Action Program (CAP) as Action Request (AR) #690371 and revised procedure OP-151.01 several times to address the procedural issues. The inspectors determined that the failure to adequately implement ADM-NGGC-106 was a performance deficiency. This performance deficiency was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, ADM-NGGC-0106, Section 9.2.39A, was not adequately implemented which resulted in OP-151.01, Attachment 7 being inadequate to implement a temporary modification for the use of three temporary air compressors supplying plant air to equipment and components which can cause plant transients. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 1 - Initiating Events Screening Questions, the inspectors determined this finding to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause an initiating event and affected mitigation equipment. The finding had a cross-cutting aspect of Consistent Process, as described in the Human Performance cross-cutting area because the licensee failed to comply with ADM-NGGC-106 and correct the inadequate operating procedure (H.13). |
| 05000400/FIN-2014004-01 | 2014Q3 | NRC identified | Potential Impact of Sump Pumps out of Service | The inspectors identified an URI associated with an equipment clearance that inadvertently resulted in all sump pumps in the EDG and DFOST buildings being nonfunctional. This item is unresolved pending review and evaluation of the licensees evaluation to determine the impact of a potential internal flood and if a performance deficiency exists. On June 26, 2014, the licensee placed equipment under clearance to support installation associated with an Engineering Change (EC). This clearance removed all sump pumps in the EDG and DFOST buildings from service. Inspectors identified this issue and informed the licensee, who restored the sump pumps to service. Additional inspection activities are needed to determine the impact of a potential internal flood and if a performance deficiency exists. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000400/2014004-01, Potential Impact of Sump Pumps out of Service. | ||
| 05000400/FIN-2014004-02 | 2014Q3 | Green | NRC identified | Loss of Emergency Planning Sirens | The NRC identified a Green NCV associated with emergency preparedness planning standard 10 CFR 50.47(b)(5), which requires in part, that the means to provide alert and notification and clear instruction to the populace within the plume exposure pathway Emergency Planning Zone (EPZ) have been established. Specifically, on April 3, 2014, the licensee unintentionally initiated a complete loss of sirens while responding to a siren system alarm. The licensee entered this issue into the corrective action program (CAP) as Action Request (AR) #679984. As corrective action, the licensee replaced a failed circuit card and restored functionality of the siren system. The licensees failure to comply with WCP-NGGC-0300, Work Request Initiation, Screening, Prioritization and Classification, was a performance deficiency. Specifically, this failure combined with the circuit card failure caused a complete loss of siren functionality for approximately two hours. This finding was more than minor because if left uncorrected, loss of Alert Notification System function has the potential to lead to a more significant safety concern and is associated with the emergency preparedness cornerstone attribute of Facilities and Equipment (Availability of ANS). This ANS unavailability affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Using Manual Chapter 0609 Appendix B, Emergency Preparedness Significance Determination Process (Section 5.5) Failure to Comply with 10 CFR 50.47(b)(5), the inspectors determined this finding to be of very low safety significance (Green) because the loss of siren function was of short duration and did not reach the Degraded Risk Significant Planning Standard (RSPS) threshold. The finding had a crosscutting aspect of Procedure Adherence, as described in the Human Performance crosscutting area because the EPTs failed to comply with the procedural guidance of WCPNGGC- 0300 (H.8). | |
| 05000400/FIN-2014004-03 | 2014Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 6.8.1 requires the procedures recommended in RG 1.33 to be established, implemented, and maintained. Regulatory Guide 1.33 requires implementation of an RWP system. Specifically, RWP #1014, Task 4, Valve Maintenance RCB (No HRA Access), required HP to be notified prior to the start of work and for HP to be present and perform surveys when breaching a contaminated system. Contrary to these RWP requirements, on November 22, 2013, two workers entered the containment building and cut out two primary CS valves (CS-761 and 762) without having HP present to perform surveys when breaching a contaminated system. After they exited containment, HP discovered the valves on the ground with removable beta-gamma contamination levels up to 200,000 dpm/100 cm2. This finding was of very low safety significance (Green) because there was no substantial potential for overexposure. This was due to the fact that the external dose rates were low and the contamination levels were not high enough to constitute a substantial potential for overexposure. The inspectors noted that no personnel were contaminated as a result of this event. The licensee entered the event into their corrective action program as AR #648061. |