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	Quarter	Site	Title	Description
05000250/FIN-2014007-01	2014Q1	Turkey Point	Failure to Properly Implement Time Critical Operator Action Program Procedure	The team identified a non-cited violation of Technical Specification 6.8.1, Procedures and Programs, for the licensees failure to implement procedure 0-ADM-232, Time Critical Action Program, to ensure time critical actions (TCAs) important to mitigate design basis events could be performed in the required time. The failure to implement this procedure was a performance deficiency. No documentation existed to demonstrate that the TCA to restore power to the battery chargers during a station blackout could be performed within the required time (30 minutes). The team also identified a TCA to locally isolate the auxiliary feedwater for a faulted steam generator that did not have a job performance measure to demonstrate the successful completion of the action. The licensee entered this issue into the corrective action program as action requests 01944453, 01945532, 01943321, 01943425, and 01943697. For TCAs where no validation documentation could be determined, the licensee completed tabletop exercises, simulator exercises, and field walkdowns to ensure that all of the TCAs to mitigate design basis events could be completed within the required action times. The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not implement 0-ADM-232 adequately to ensure that the TCAs listed in Attachment 1 of the procedure were properly validated; consequently, the licensee could not demonstrate that TCAs could be successfully executed in accordance with the design basis. The team determined the finding to be of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; and did not represent a loss of system and/or function. The team determined this finding was associated with the cross-cutting aspect of Procedure Adherence in the area of Human Performance because although the procedure was recently revised to include all necessary requirements to maintain the time critical action program, the licensee failed to follow procedure 0-ADM-232, which resulted in several TCAs not being properly validated.
05000259/FIN-2009008-03	2009Q4	Browns Ferry	Safety-Related Molded Case Circuit Breakers	The team indentified an Unresolved Item (URI) regarding safety-related molded-case circuit breakers (MCCBs) in safety-related applications. TVA had not implemented a test program to detect potential deterioration or to assure that all installed safety-related MCCBs would perform satisfactorily in service. TVA procedure 0-TI-395, Breaker Testing and Maintenance Program, required that critical molded case circuit breakers be subject to preventative maintenance (PM) activities. This included, in part, inspection for overheating, mechanical operation, enclosure inspection, overload trip testing, and instantaneous trip testing. The TVA program required that the testing and PM activities be performed every four to six years. UFSAR 8.6.4.1.1 stated, in part, that zero-resistance short circuits at the battery board or any point downstream can be cleared by the breakers operating within their ratings. To ensure this was satisfied by the installed equipment, degradation of breaker performance should have been detectable and acceptably controlled by periodic testing and preventive maintenance. Additionally, UFSAR 8.5.2.11 stated, in part, that the standby AC power system will meet or exceed the requirements of IEEE-308, Criteria for Class 1E Power Systems at Nuclear Generating Stations. This standard recommended that periodic tests be performed at scheduled intervals to detect deterioration of equipment and to demonstrate operability of components that are not exercised during normal operation. Four MCCBs on 250 VDC Battery Board 1 were selected by the team (breakers 607, 705, 712, 715) as part of the inspection sample. No records of PM activities were found on these four MCCBs. While addressing the extent of this condition, TVA identified a total of 622 safety-related MCCBs in both AC and DC applications for which no testing was being performed. As a result of the inspectors observations, TVA initiated PER 209095 and scheduled testing of the 622 safety-related MCCBs that were not being tested in accordance with the TVA PM program. As of December 28, 2009, TVA had successfully tested 30 of the 622 breakers. This represented a small sample of the total breaker population. The inspectors plan to review additional PM test results to determine if this performance deficiency is more than minor. This issue is unresolved pending further inspection to determine the extent of condition and impact of not implementing a test program to assure that all installed safety-related MCCBs would perform satisfactorily. (URI 05000259, 260, 296/2009008- 01, Safety-Related Molded Case Circuit Breakers
05000259/FIN-2010004-01	2010Q3	Browns Ferry	Uncontrolled Materials Adversely Impacted the Capability of the EDG Building Emergency Drainage System to Mitigate an Internal Flooding Event	The inspectors identified an unresolved item (URI) regarding a variety of materials left unattended, unanchored and improperly stored in the lower corridors of both of the Unit 1/2 and Unit 3 EDG buildings that could have adversely impacted the capability of the emergency drainage systems credited in the licensees internal flooding analysis for these buildings. During an internal flood protection walk-down of the Unit 1 and 2 EDG building, and the Unit 3 EDG building, the inspectors identified unattended and loose materials in the lower corridors that contain the EECW North and South supply header piping. The inspectors observed a 24-inch emergency drain in the Unit 1/2 EDG building lower corridor located in the Southwest corner. This drain emptied into the yard area just west of the Unit 1/2 EDG building and was shown on drawing 0-47E851-1, Rev. 29. In addition, the inspectors observed two 18-inch drains in the Unit 3 DG building lower corridor floor. These drains emptied into the yard area just east of the Unit 3 DG building and were shown on drawing 0-47E851-4, Rev. 13. The uncontrolled materials identified by the inspectors were of sufficient type and quantity to potentially obstruct the lower corridor emergency drains that were designed to mitigate the consequences of an internal flood due to the rupture of an EECW header. Furthermore, none of the doors that provide access to the four EDG rooms from the lower corridor were designed to be watertight. The inspectors reviewed the licensees Probabilistic Risk Analysis (PRA) for Internal Flooding as described in calculation number NDN-000-999-2007-0031, Rev 0. This analysis stated Flooding in the diesel-generator buildings could result from failure of the EECW headers that pass through the buildings. The diesel-generator buildings, however, are provided with a 24-inch emergency drain in the wall that empties into a culvert in the yard. It further states The common (i.e.- Unit 1/2 ) diesel generator building corridor has sump pumps and 24-inch drains. These are adequate to mitigate floods and major floods in the common corridors, so only one DG can be impacted by a flood. In the summary of qualitative screening results, the licensee concluded that a flood in the DG building lower corridor would result in no submergence due to large drains and no impacted SSCs. The inspectors also reviewed the licensees detailed design criteria document for the diesel generators, BFN-50-7082, Rev. 15. Section 3.7.4 stated that the EDG units shall be located in separate rooms to ensure that flooding, resulting from a postulated failure in the pressure boundary of any water systems, would not prevent the standby DG system from performing its safe shutdown function. The licensee initiated PER 256390 to evaluate the impact of the loose materials on the function of the EDG building lower corridor emergency drains. Additionally, the licensee removed the materials from the EDG buildings and added a daily requirement for the auxiliary unit operators (AUOs) to verify no unattended loose material in the EDG buildings lower corridors. This issue is unresolved pending further inspection to determine more specifically the adverse impact of the improperly stored materials in the EDG building lower corridors upon the drainage system, the availability and capability of other internal flooding mitigation features (e.g., sump level alarms), and the current licensing basis for moderate energy line breaks (MELB) in the EDG buildings. The URI for this issue is identified as 05000259, 260, and 296/2010004-01, Uncontrolled Materials Adversely Impacted the Capability of the EDG Building Emergency Drainage System to Mitigate an Internal Flooding Event.
05000259/FIN-2010004-02	2010Q3	Browns Ferry	Failure to Adequately Assess Online Risk Associated With Maintenance Activities on Risk Significant SSCs	The inspectors identified a non-cited violation of 10 CFR Part 50.65 (a)(4), for inadequate risk assessments of on-line risk associated with ongoing maintenance activities. Specifically, on July 21 and then again on September 16, 2010, the inspectors found that the licensee failed to perform a probabilistic risk analysis (PRA) evaluation of the multiple risk significant equipment that had been taken out of service for planned on-line maintenance. The licensee entered this issue into the corrective action program as problem evaluation reports (PERs) 241885 and 254000. In both instances the licensee subsequently performed the required PRA evaluations which determined the on-line risk to be Green. This finding affected the Mitigating Systems cornerstone and was determined to be greater than minor according to Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, because minor violations of 10 CFR 50.65(a)(4) have occurred repeatedly on five occasions and if continued to be left uncorrected would have the potential to lead to a more significant safety concern. The significance of this finding was evaluated using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Based on Appendix K, the inspectors determined that this finding was of very low safety significance (Green) because the licensees PRA evaluation concluded the actual risk deficit was less than 1E-6 for the incremental core damage probability deficit (ICDPD) and less than 1E-7 for the incremental large early release probability deficit (ILERPD). The cause of this finding was directly related to the cross cutting aspect of Procedural Compliance in the Work Practices component of the Human Performance area, because the licensee failed to follow the instructions in 0-TI-367 which required a PRA evaluation to be performed in accordance with SPP-9.1 (H.4(b)).
05000259/FIN-2010004-03	2010Q3	Browns Ferry	Failure to Adequately Test Molded Case Circuit Breakers	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for failure to establish a preventive maintenance (PM) test program for safety-related molded case circuit breakers (MCCBs) to demonstrate these breakers would perform satisfactorily upon demand. Since initial startup of all three units, the inspectors found that the licensee had not included 612 critical MCCBs, many of them safety-related, in their PM program which resulted in the MCCBs receiving no planned maintenance or testing. The licensee entered this issue into the corrective action program as problem evaluation report (PER) 209095. The licensees corrective actions included: identifying all critical MCCBs that required preventive maintenance, developing test procedures for these MCCBs, performing testing for all affected MCCBs, and conducting an extent-of-condition review of all safety-related components potentially excluded from the PM program. This finding was determined to be of greater than minor significance because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events, such as fire, that challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of a PM program for safety-related MCCBs resulted in no periodic planned maintenance or testing being performed since original installation, which in most cases was over thirty years. Based on operating experience, this could result in a breaker being slow to trip or sticking in the on position after an over-current condition. In accordance with IMC 0609, Significance Determination Process (SDP), Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to require a Phase 3 analysis since the finding represented an increase in the likelihood of a fire caused by an electrical fault at the MCCB compartment with the breaker not opening. A regional Senior Reactor Analyst conducted a Phase 3 SDP analysis, which concluded that the finding was of very low safety significance (Green). The cause of this finding was directly related to the cross cutting aspect of Appropriate Corrective Actions in the Corrective Action Program component of the Problem Identification and Resolution area, because the licensee did not adequately implement corrective actions to resolve the deficiencies previously identified by PER 131875 regarding certain Westinghouse MCCBs that were not in the PM program (P.1(d)).
05000259/FIN-2010004-04	2010Q3	Browns Ferry	Failure to Perform Functional Evaluations for Gas Identified During Venting	An NRC-identified Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to perform functional evaluations in accordance with procedure NEDP-22, Functional Evaluations, when gas was identified in the High Pressure Coolant Injection (HPCI) System during the Technical Specification required surveillance. The licensee has subsequently performed functional evaluations of the occurrences and entered the issue into their corrective action program as problem evaluation report (PER) 223067. This finding was considered more than minor because it adversely affected the Mitigating Systems Cornerstone objective of ensuring the availability and reliability of safety systems, and is related to the attribute of Procedure Quality (i.e.- Maintenance and Testing Procedures). Specifically, the failure to perform a functional evaluation or provide adequate justification for not performing one upon identification of gas during venting of the system could affect the operability, availability, and reliability of the HPCI system or could result in missing an opportunity to identify the source of voiding to preclude future inoperability. This deficiency also paralleled Inspection Manual Chapter 0612, Appendix E, Example 4.a, as the licensee routinely did not perform the required functional evaluations. The team assessed this finding using Inspection Manual Chapter 0609, Significance Determination Process, and determined that the finding was of very low safety significance (Green) because subsequent functional evaluations showed that the gas voids did not impact the operability of the HPCI system. The cause of this finding was directly related to the cross cutting aspect of Evaluation of Identified Problems in the Corrective Action Program component of the Problem Identification and Resolution area, in that the licensee failed to thoroughly evaluate gas voids such that the resolution addressed causes and extent of conditions, as necessary, and included the failure to thoroughly evaluate for operability and reportability conditions adverse to quality. (P.1(c))
05000259/FIN-2010004-05	2010Q3	Browns Ferry	Licensee-Identified Violation	TS LCO 3.1.7, Standby Liquid Control (SLC) System, in part required that two SLC subsystems be operable in Modes 1, 2, and 3 with an allowed outage time of 7 days for one inoperable SLC subsystem, or place the unit in Mode 3 within 12 hours and Mode 4 within 36 hours. However, during a routine TS required quarterly surveillance test, the licensee discovered that 3B SLC Pump would not start due to the improper engagement of the 480 VAC breaker racking sleeve. This resulted in 3B SLC subsystem being inoperable from April 7 to April 20, 2010, without the licensee taking the required TS 3.1.7 actions. The TS violation was entered into the licensees CAP as PER 225949. Even though the finding represented an actual loss of safety function of a single train of SLC for greater than its TS allowed outage time, the finding was determined to be of very low safety significance (Green) because the risk significance from the Browns Ferry SDP Phase 2 pre-solved table was green.
05000259/FIN-2010004-06	2010Q3	Browns Ferry	Licensee-Identified Violation	TS LCO 3.3.6.1, Primary Containment Isolation Instrumentation, in part, required that the Reactor Core Isolation Cooling (RCIC) Steam Line Flow-High Isolation Function for both the A and B Channels be operable in Modes 1, 2, and 3 with an allowed outage time of 24 hours for one inoperable channel; or isolate the affected penetration flow path within 1 hour; or place the unit in Mode 3 within 12 hours and Mode 4 within 36 hours. However, the licensee discovered that rubber boots had been inadvertently left installed on the channel B contacts for RCIC Steam Line Flow-High Isolation from April 9 to April 26, 2010. This rendered one of the two TS required channels of the RCIC Steam Line Flow-High Isolation Function as inoperable for a period much greater than the TS allowed 24 hours, without the licensee taking the required TS 3.3.6.1 actions. This TS violation was entered into the licensees CAP as PER 226666. Even though the finding represented an actual loss of function of a single channel of RCIC Steam Line Flow-High Isolation for greater than its TS allowed outage time, the finding was determined to be of very low safety significance (Green) because the redundant A Channel isolation function remained operable and would have isolated the Unit 2 RCIC system on a RCIC Steam Flow-High signal as needed.
05000259/FIN-2010004-07	2010Q3	Browns Ferry	Licensee-Identified Violation	TS LCO 3.6.1.3, Primary Containment Isolation Valves, required that each Primary Containment Isolation Valve (PCIV) be operable in Modes 1, 2, and 3, and when the associated instrumentation was required to be operable according to TS LCO 3.3.6.1, Primary Containment Isolation Instrumentation. For one or more inoperable excess flow check valves (EFCVs), TS 3.6.1.3 required the affected flow path to be isolated within 12 hours, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. However, during TS required surveillance testing during the U3C14 RFO, the licensee discovered that five of 15 EFCVs failed to isolate. Based on the existence of multiple failures the licensee concluded that one or more EFCV was inoperable during fuel Cycle 14 operation. This TS violation was entered into the licensees CAP as PER 222850. The finding was determined to be of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of primary containment and did not contribute to the increased potential of an reactor coolant system instrument line break.
05000259/FIN-2011002-01	2011Q1	Browns Ferry	Loss of Reactor Water Level during Unit 2 Reactor Reassembly due to a Mispositioned Valve	A self-revealing non-cited violation of Technical Specifications (TS) 5.4.1.a was identified for the licensees failure to adequately implement operations instruction 2-OI- 74, Residual Heat Removal System, to ensure the reactor cavity draindown flow path was isolated prior to suppression pool draindown. On March 25, 2011, Operations personnel inadvertently left a Residual Heat Removal (RHR) system drain valve in the open position which led to an uncontrolled draindown of the reactor pressure vessel (RPV) coolant to the suppression pool. Operators immediately identified the RPV level decrease and restored the valve lineup and water level. The licensees immediate corrective actions re-emphasized adherence to log keeping and turnover requirements; instituted shift manager challenges on activities that impact key safety functions including assessments of procedures, plant configuration, turnover information, and prejob briefs of personnel roles and responsibilities; and, for those same activities, instituted peer checks, marked up drawings, and supervisory review of completed field copies of procedures. This issue was entered into the licensees corrective action program as problem evaluation report (PER) 344533. This finding was considered more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. Specifically, a mispositioned RHR drain valve resulted in a loss of control of the RPV water level. This finding was determined to be of very low safety significance (Green) according to Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations, because the inadvertent loss in excess of 2 feet (approximately 40 inches) of reactor coolant inventory represented a loss of inventory control. Using IMC 0609, Appendix G, Attachment 3, Phase 2 Significance Determination Process Template for BWR During Shutdown, a Senior Reactor Analyst performed an analysis and determined the loss of inventory event was of very low risk significance (Green) due in part to automatic functions being available to isolate and mitigate the leak had it continued and remained undetected/uncorrected by the operators. The cause of this finding was directly related to the cross-cutting aspect of Work Activity Coordination in the Work Control component of the Human Performance area, because inadequate documentation and communication of plant system configuration by the control room operators resulted in a mispositioned valve and loss of RPV water level (H.3.(b)).
05000259/FIN-2011002-02	2011Q1	Browns Ferry	Inadequate Corrective Actions To Address Unit 3 CR120A PCIS Relays That Exceeded Their Recommended Service Life	An NRC identified non-cited violation of 10 CFR 50 Appendix B, Criteria XVI, Corrective Action, was identified for the licensees failure to correct a condition adverse to quality related to Unit 3 primary containment isolation system (PCIS) logic relays exceeding their in-service life expectancy. Specifically, the licensee failed to replace numerous Unit 3 PCIS CR120A relays prior to exceeding their vendors recommended service lifetime. The licensee has entered this issue into their corrective action program as problem evaluation report (PER) 348160. This finding was determined to be more than minor because it was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the frequency of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, a relay failure could cause a reactor scram, engineered safeguards (ESF) actuation, and/or Group 1, 2, 3, or 6, primary containment isolation. The significance of the finding was evaluated using Phase 1 of the significance determination process in accordance with the Inspection Manual Chapter (IMC) 0609 Attachment 4, and was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions were not available. The cause of this finding was directly related to the cross cutting aspect of Appropriate Corrective Actions in the Corrective Action Program component of the Problem Identification and Resolution area, because the licensee failed to implement adequate corrective actions as part of PER 220336 to replace or extend the service life of the Unit 3 PCIS CR120A relays prior to exceeding their recommended service lifetime (P.1(d)).
05000259/FIN-2011002-03	2011Q1	Browns Ferry	Inadequate TS 5.5.2 Program for Primary Coolant Leaks Outside Containment	An NRC identified non-cited violation of Technical Specifications (TS) 5.5.2, Primary Coolant Sources Outside Containment was identified for the licensees failure to establish, implement, and maintain an adequate program for minimizing primary coolant leaks from systems (i.e., Core Spray, Residual Heat Removal, High Pressure Coolant Injection, and Reactor Core Isolation Cooling) outside containment, that could contain highly radioactive fluids during a serious transient or accident, to levels as low as practicable. The licensees corrective actions included identification, evaluation, and prioritization of all known primary coolant leaks outside containment; and development of a new program in accordance with 0-TI-578, Minimizing Primary Coolant Sources Outside Containment. This finding was entered into the licensees corrective action program as problem evaluation report (PER) 317464. This finding was determined to be more than minor because if left uncorrected it could have led to a more significant safety concern. Specifically, the licensees failure to effectively minimize and monitor primary coolant leakage outside containment could have resulted in increased main control room exposure and/or offsite dose during an accident due to excessive radioactive fission product releases into secondary containment. The finding was determined to be of very low safety significance (Green) according to IMC 0609, Appendix H, Containment Integrity Significance Determination Process, Section 6.0, Type B Findings, because the primary coolant leak rate into secondary containment was a small fraction of the leakage assumed in the design basis accident (DBA) safety analyses. The cause of this finding was directly related to the cross-cutting aspect Complete and Accurate Procedures in the Resources component of the Human Performance area because the licensees existing procedures were inadequate and incomplete for addressing the program requirements of TS 5.5.2 (H.2.(c)).
05000259/FIN-2011002-04	2011Q1	Browns Ferry	Failure to Identify Adverse Trend Resulted in Reactor Scram	A self-revealing finding (FIN) was identified for the licensees failure to adequately evaluate and take the required actions established by site standards to address an adverse system performance trend that had degraded below acceptable levels associated with the main generator exciter air coolers. Specifically, the licensee failed to identify that main generator exciter air cooler differential temperatures exceeded the licensee-defined limit of 10F, and did not initiate a PER as required by the licensees procedural guidance, Nuclear Engineering Department Procedure (NEDP) -20, Conduct of the Engineering Organization, Section 3.1, System Performance Monitoring. Subsequent licensee corrective actions included installing vents on the exciter air coolers to minimize air binding, establishing a process and frequency for venting the exciter air coolers, and increasing engineering supervisory oversight of the system monitoring process. The licensee captured this issue in the corrective action program as PER 301505. This finding is greater than minor because it is associated with the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability. Specifically, the finding resulted in a Unit 3 manual reactor scram due to elevated main turbine bearing vibrations caused by excessive main generator exciter air cooler differential temperatures. The significance of the finding was evaluated using Phase 1 of the significance determination process in accordance with the Inspection Manual Chapter (IMC) 0609 Attachment 4, and was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions were not available. The cause of this finding was directly related to the cross-cutting aspect of Corrective Action Program Implementation in the Corrective Action Program component of the Problem Identification and Resolution area, because the licensee failed to identify the adverse trend of excessive differential temperatures between the exciter air coolers in a timely manner and enter it into the corrective action program.
05000259/FIN-2011002-05	2011Q1	Browns Ferry	Repeated Failure to Control Transient Combustibles in Proximity of the Independent Spent Fuel Storage Facility	A Severity Level IV, cited violation (VIO) of 10 CFR 72.212, Conditions of general license issued under 72.210, was identified by the inspectors for the licensees repetitive failure to adequately control transient combustible materials stored in the proximity of loaded dry casks on the ISFSI pad in accordance with site procedures. On February 3, 2011, while performing a routine walkdown of the ISFSI enclosed area, the inspectors observed seven storage cradles, multiple storage pallets and storage devices or cribbing located on or near the dry cask storage pad. The cradles, pallets and cribbing, were all constructed of wood products. The nearest items, were wood cradles located approximately 10 to 15 feet from the closest HI-STORM cask loaded with spent fuel. The other wood storage devices were approximately 20 feet from the closest loaded cask and were located both on and off the ISFSI pad. No apparent work was in progress at the time of discovery. The inspectors contacted responsible licensee personnel who promptly removed all the transient combustible material from the ISFSI exclusion area and initiated PER 318694. The licensee also performed an evaluation of the transient combustible loading for this material. This was the third occurrence identified by the inspectors of transient combustibles located in close proximity to HI-Storm casks loaded with spent fuel. The first two occurrences were on May 25, 2010 (see NCV 07200052/2010002-001, Transient Combustibles Stored Near Independent Spent Fuel Storage Facility in Excess of Amount Allowed), and on August 17, 2010 (see NOV 07200052/2010003-001, Transient Combustibles Stored Near Independent Spent Fuel Storage Facility in Excess of Amount Allowed), in both instances diesel fuel contained in vehicles left parked in close proximity to loaded HI-Storm casks was greater than the maximum allowed. According to NPG-SPP-18.4.7, Control of Transient Combustibles, the requirements and controls for handling and use of transient combustibles in proximity of the BFN ISFSI/Dry Cask Storage Pad were contained within drawings 0-47E201-1 and 0-47E201-2. In particular, drawing 0-47E201-2, ISFSI Fire Hazards Analysis Compensatory Actions, Item 11 stated that wooden structures facing the ISFSI were limited to a front face maximum height of 15 feet and a maximum width of 24 feet for a surface area total of 360 square feet, at a distance of 30 feet from the edge of the closest HI-STORM. Furthermore, General Operating Instruction (GOI) 0-GOI-300-1/ATT-12, Outside Operator Round Log, required operators to perform an inspection daily to ensure the ISFSI Pad and exclusion area were clear of the following: Flammable material such as wood, rags and plastic sheeting. If the ISFSI pad and exclusion area were not clear of these materials, then report the results to the Unit 3 Supervisor for evaluation of acceptability in accordance with drawing 0-47E201-2. Per 0-GOI-300-1/ATT-12 the ISFSI Pad exclusion area is defined as within 150 feet of the edges of the ISFSI Pad in all directions. Based upon discussion with the licensee and a review of work performed in the area, the inspectors determined that the licensee had allowed the wood cradles and cribbing to be left near a loaded HI-STORM cask for approximately one week from on or about January 26 to February 3, 2011. The licensee was performing work in the area to upright and inspect Multi Purpose Containers for the upcoming campaign. However, plant operators had not notified the Unit 3 US of the stored wooden material, and no evaluation had been performed on the acceptability of the transient combustible material as required by 0-GOI-300-1/ATT-12. Subsequent calculations by the licensee determined that the radiative heat load of the wood items was only about five percent of the allowed transient combustible loading limit.
05000259/FIN-2011002-06	2011Q1	Browns Ferry	Licensee-Identified Violation	Unit 3 Technical Specification 5.4.1.a. required that written procedures recommended in RG 1.33, Revision 2, Appendix A, shall be established, implemented, and maintained. Procedures for performing maintenance on safety related equipment were specifically listed as recommended procedures by Section 9 of Regulatory Guide 1.33, Appendix A. Preventive maintenance procedure EPI-0- 000-BKR15, 4KV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance, was established in part to specifically inspect the rubber bumper located on the MJ (52STA) switch actuator for damage/indention that would reduce the amount of travel of the MJ switch when the breaker was actuated. Contrary to this, the licensee determined that preventive maintenance was not performed on 3ED 4KV Shutdown Board Normal Feeder Breaker 1342 within the required six year periodicity, and was deferred an additional two years without adequate technical justification. On February 10, 2011, while performing thermography on the 3ED 4KV Shutdown Board, the licensee identified several deenergized relays resulting from unmade contacts due to age-related wear on the MJ switch actuation arm rubber bumper which would have prevented an automatic start of 3D Residual Heat Removal (RHR), 3D Core Spray (CS), and D1 RHR Service Water (RHRSW) Pumps. This finding was entered into the licensees CAP as PER 324038 and PER 322640 for the associated cause analysis. The finding was determined to be of very low safety significance because it did not constitute a total loss of system or train safety function since the pumps would have automatically started with the diesel generator supplying the electric board and manually started regardless of the electrical source. Therefore, Train D RHR and CS safety functions would have been available during a design accident condition.
05000259/FIN-2011002-07	2011Q1	Browns Ferry	Licensee-Identified Violation	Unit 1 Technical Specification 3.3.1.1, Reactor Protection System Instrumentation, required two operable drywell (DW) pressure channels per trip system in Modes 1 and 2, or place the inoperable channel in trip within 12 hours, or be in Mode 3 in 12 hours. Contrary to this, as described in the LER 259/2010-002, the licensee determined that DW pressure instrument 1-PT-064-56C was inoperable from October 2 until October 25, 2008, and again from November 30, 2008 until the December 9, 2008, due to accumulated water in the sensing line from improper line slope, without placing the channel in trip or shutting down Unit 1. This violation is not greater than Green because there was no loss of function due to the availability of redundant instrumentation. This issue was entered in the licensees CAP as PERs 159710, 219150, 242068 and 279760.
05000259/FIN-2011002-08	2011Q1	Browns Ferry	Licensee-Identified Violation	Unit 1 Technical Specification 3.3.1.1, Reactor Protection System Instrumentation, required two operable drywell (DW) pressure channels per trip system in Modes 1 and 2, or place the inoperable channel in trip within 12 hours, or be in Mode 3 in 12 hours. Contrary to this, as described in the LER 259/2010-002, the licensee determined that DW pressure instrument 1-PIS-064-56B was inoperable from May 25, 2010 to October 6, 2010, due to accumulated water in the sensing line from improper line slope, without placing the channel in trip or shutting down Unit 1. This violation is not greater than Green because there was no loss of function due to availability of redundant instrumentation. This issue was entered in the licensees CAP as PERs 159710, 219150, 242068 and 279760.
05000259/FIN-2011002-09	2011Q1	Browns Ferry	Licensee-Identified Violation	10 CFR 26.205(d) states, in part, that the licensee shall control the work hours of personnel performing maintenance, as identified by 10CFR26.4(a)(4), to ensure maintenance personnel individually meet the work hour limits, minimum break periods, and minimum days off prescribed by 10CFR26.205(d)(1) thru (3). However, from April to November 2010, due a programmatic breakdown in tracking overtime hours, the licensee failed to control individual work hours, breaks, and minimum days off for numerous Maintenance Department personnel in a manner that complied with the aforementioned regulatory requirements and the licensees program SPP-1.5 (later changed to NPG-SPP-03.21), Fatigue Management and Work Hour Limits. These violations were captured in the licensees CAP as PERs 255792 and 322569. This finding was determined to be of very low safety significance (Green) because no significant human errors occurred that were attributable to fatigue while these individuals were in violation of the Nuclear Fatigue Rule.
05000259/FIN-2013007-01	2013Q2	Browns Ferry	Failure to Verify the Capability of HPCI to Achieve Required Flow and Pressure within 30 Seconds Under Accident Conditions	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure that post-maintenance and post-modification testing of the high pressure cooling injection (HPCI) pump adequately demonstrated that it could achieve design basis flow within 30 seconds from a cold, non-oil-primed, turbine quick start under design basis conditions. This was a performance deficiency. The test configuration was less limiting than the design basis accident configuration, and the licensee had not verified by calculation or testing that the acceptance criteria in the test was adequate to demonstrate the HPCI pump could perform its function under design basis conditions. The licensee performed an operability review and documented the results in the corrective action program as Problem Evaluation Report 690086. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the HPCI pumps. Specifically, using procedure 3-SR-3.5.1.7, the licensee failed to demonstrate that the HPCI pump could achieve the required flow and discharge pressure under accident conditions as required by the design basis. Additional analysis was required to verify system operability. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since the original design of the plant and was not indicative of current licensee performance.
05000259/FIN-2013007-02	2013Q2	Browns Ferry	Failure to Evaluate the Effects of the Failure of Non-Class 1E Load Center Transformer Cooling Fans on the Class 1E 4160-480V Load Center Transformers and 480V Shutdown Boards	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, involving the failure to evaluate the effects of a postulated failure of the load center transformer non-safety-related, non-Class 1E cooling fans, which includes the fan power wiring and fan control equipment, on the safety-related Class 1E shutdown board load center transformers and 480V shutdown boards. This was a performance deficiency. The licensee tested the fans and performed an operability evaluation as documented in Problem Evaluation Report 682254 to provide reasonable assurance that the safety-related transformers would not be damaged from postulated failures from the non-safety-related fans and be capable of operating when required for the design basis accident conditions. The performance deficiency was determined to be more than minor because the finding affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the load center transformers TS1A and TS1B and the 480V shutdown boards 1A and 1B respectively. Specifically, the licensee had not evaluated the effects of the failure of non-safety-related transformer cooling fans, on both the safetyrelated load center transformer and 480V shutdown board and resulted in a reasonable doubt of operability. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since November 2004; therefore, not indicative of current licensee performance.
05000259/FIN-2013007-03	2013Q2	Browns Ferry	Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to perform analyses demonstrating that the degraded voltage relay (DVR) set points specified in technical specifications (TS) would ensure adequate voltage to safety-related equipment. This was a performance deficiency. The licensee entered this issue into their corrective action program as PERs 676678 and 696876. As immediate corrective actions, the licensee performed a sensitivity study to verify that the voltage at the DVR set points specified in TS could provide adequate starting voltage to a sample of limiting safety-related equipment. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the 4160 volts alternating current buses. Specifically, the finding challenged the assurance that safety-related loads had adequate motor starting voltage during required degraded voltage scenarios. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since 1993 and was not indicative of current licensee performance.
05000259/FIN-2013007-04	2013Q2	Browns Ferry	Failure to Adequately Identify, Evaluate, and Correct the EECW Strainers Degraded/Nonconforming Condition	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and take corrective actions to address a non-conforming condition adverse to quality related to three faulted strainers in the safety related Emergency Equipment Cooling Water system. This was a performance deficiency. The licensee initiated Problem Evaluation Report 677627 to perform a new operability evaluation since the operability evaluation in Problem Evaluation Report 208636 was found to be inadequate. The licensee concluded that there were no current operability issues. The performance deficiency was determined to be more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the core spray system to respond to initiating events, in that, if left uncorrected could result in the plant not being able to sustain short-term heat removal under specific conditions. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. The team evaluated the finding for cross-cutting aspects and determined the finding was associated with the corrective action program component of the problem identification and resolution area, because the licensee did not perform a thorough evaluation of identified problems such that the resolutions addressed the underlying causes and extent of condition.
05000259/FIN-2013007-05	2013Q2	Browns Ferry	Security
05000261/FIN-2010011-01	2010Q4	Robinson	Failure to Establish Proper In-service Testing Acceptance Criteria to Prevent Reverse Rotation of the SDAFW Pump	The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to ensure the in-service testing(IST) of the discharge check valve of the Auxiliary Feedwater (AFW) steam driven pump applied an acceptance criterion that is in accordance with the limits established in design documents. The licensee revised the IST procedure during the inspection and is tracking further action in the corrective action program under NCR 419768. The failure to establish proper acceptance criteria for the Steam Driven (SD) AFW discharge check valve was a performance deficiency. This finding was more than minor because it affected the mitigating systems cornerstone attribute of procedure quality to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to incorporate the proper acceptance criteria could result in a failure of the test to identify a check valve degraded to a condition where its back leakage will cause reverse rotation of the SD AFW pump. This finding was of very low safety significance because it was not a test issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the 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 the failure to establish a proper acceptance criteria for the discharge check valve of the SDAFW pump was determined to not be indicative of current licensee performance.
05000261/FIN-2010011-02	2010Q4	Robinson	Failure To Ensure that the Full Range of Emergency Diesel Generator Frequency is Accounted for in the Safety Analyses	The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for the high range of Emergency Diesel Generator (EDG) frequency allowed by technical specifications (TS) in the safety analysis. While no immediate operability issues were identified, the licensee entered this issue into the corrective action program as NCR 420058. The failure to evaluate the effect of high EDG frequency was a performance deficiency. This finding was a more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. This finding also closely parallels IMC 0612, Appendix E, Example 3.j, Not Minor: If the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component, or if significant programmatic deficiencies were identified with the issue that could lead to worse errors if uncorrected. Specifically, failure to account for an allowable diesel frequency of 61.2 Hz (60 +2%) for all safety related pumps may result in operating at a higher flow rate and a higher developed suction head. This finding was of very low safety significance because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding the TS allowed outage time, and did not affect external event mitigation. The team also evaluated the finding for cross-cutting aspects and determined it to involve the area of Problem Identification and Resolution associated with Operating Experience for the licensees failure thoroughly evaluate NRC Information Notice 2008-02, which specifically identified high diesel frequency as a potential problem for AC motor-operated pumps (P.2(a)).
05000261/FIN-2010011-03	2010Q4	Robinson	Failure to Demonstrate the Capability of the Fuel Oil Storage Tank and the Service Water Pumps to Fulfill Their Safety Functions Under All Conditions	The team a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to have calculations supporting the design bases of safety related components, specifically for the EDG fuel oil storage tank with respect to tornado wind loadings, and the net positive suction head (NPSH) of the service water pumps. No immediate operability issues were identified and the licensee entered this issue into the corrective action program as NCR 422985 and NCR 423985. The failure to demonstrate the adequacy of the design for safety related components, specifically regarding the capability of the fuel oil storage tank to withstand tornado wind loading and the failure to demonstrate that the NPSH available to the service water pumps was greater than the required NPSH, was a performance deficiency. This finding was more than minor because it affected the mitigating systems cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance because the licensee performed a simplified evaluation indicating that this condition was not a design issue resulting in loss of function, it did not represent an actual loss of a system safety function, did not result in exceeding the 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 the failure to demonstrate the adequacy of the design was determined to not be indicative of current licensee performance.
05000261/FIN-2010011-04	2010Q4	Robinson	Failure to Correctly Translate EDG Starting Air System Design Requirements into TS	The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design basis of the EDG air start system into specifications. Specifically, the licensee did not properly translate the lowest air pressure for the EDG air start receiver that would provide a single EDG start into the TS (150 psig). The licensee reviewed the low pressure alarm history of the EDGs and did find any instance where they failed to declare the EDG inoperable based on the new operability setpoint. Further actions are being tracked in the corrective action program under NCR 423776. The licensees failure to correctly translate the design basis of the EDG air start system into the TS was determined to be a performance deficiency. The finding was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the EDG starting air receiver pressure could fall below 150 psig, but the TS would not direct the licensee to declare the EDG inoperable. The finding is of very low safety significance as it was determined not to have resulted in the loss of operability or functionality. The team determined that no cross cutting aspect was applicable to this performance deficiency because the failure was determined to not be indicative of current licensee performance.
05000261/FIN-2010011-05	2010Q4	Robinson	Inadequate Criteria to Prevent Spurious Actuation of Amptector Trip Devices	The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, in that the licensee failed to verify the adequacy of the design for Amptector trip devices installed on safety related 480V circuit breakers. The licensee reviewed the latest calibration records and contacted the vendor for further guidance and additional information. Further actions are being tracked in the corrective action program under NCR 423795. The team determined that the failure to establish an adequate minimum setting for Amptector trip devices was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Also, this finding closely parallels NRC IMC 0612, Appendix E, Example 3.j in that the condition resulted in reasonable doubt of the operability of the safety related 480V system pending re-analysis. Specifically, the licensee failed to evaluate margins needed to prevent spurious tripping during accident loading conditions. The team determined the finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability or functionality. The team also evaluated the finding for cross-cutting aspects and determined it to involve the area of Human Performance, because this condition is related to the component of resources which requires complete, accurate and up-to-date design documentation, specifically calculations, to assure nuclear safety (H.2(c)).
05000261/FIN-2010011-06	2010Q4	Robinson	Failure to Translate Vendor Recommendations Into Procedures for 480V Circuit Breakers	The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, Instructions and Drawings, for failure to follow procedure EGRNGGC- 006, Vendor Manuals, which requires performance of reviews to determine technical accuracy and potential changes to procedures, processes or equipment; specifically for safety related 480V Breakers and reactor trip breakers. The licensee performed a gap analysis, reviewed the discrepancies, and concluded that they did not impede the ability of the breakers from performing their associated function. Further actions are being tracked in the corrective action program under NCRs 422184 and 422976. The team concluded that the failure to perform reviews to determine technical accuracy and potential changes to procedures for circuit breaker vendor manual changes was a performance deficiency. This finding is more than minor because it affects the mitigating systems cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events and is associated with the attribute of procedure quality, in that procedure inconsistencies were identified in procedures MST-012-1, Maintenance and Testing of A Reactor Trip Breaker, and PM-466, Westinghouse Type 50DH350E 1200 Amp 4160V Air Circuit Breaker Maintenance. This finding also closely parallels IMC 0612, Appendix E, Example 4.a., in that the procedure discrepancies indicate that the licensee routinely failed to perform reviews EGRNGGC- 006, Vendor Manuals, which requires performance of reviews to determine technical accuracy and potential changes to procedures, processes or equipment. The team determined the finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability or functionality. The team also evaluated the finding for cross-cutting aspects and determined it to involve the area of Problem Identification and Resolution associated with Operating Experience for the licensees failure to thoroughly evaluate vendor recommendations, as well as NRC Information Notice 2008-02 which identified an issue relating to improper maintenance of circuit breakers involving failure to follow vendor maintenance recommendations (P.2(b)).
05000261/FIN-2010011-07	2010Q4	Robinson	Failure to Implement Adequate Post Maintenance Test of Residual Heat Removal Valve Interlock Function	The team identified a finding having very low safety significance (Green) involving the failure to perform a post maintenance test to verify functionality of valve position permissive interlocks associated with the reactor coolant system (RCS) hot leg loop isolation valves. The licensee performed a visual inspection to verify that the associated contacts for the valve position permissive interlock function were in their expected open position, and is tracking further actions in the corrective action program under NCR 422032. The failure to perform a post maintenance test to verify functionality of the permissive interlock associated with the RCS hot leg loop isolation valves following replacement of relays which affected that function was a performance deficiency. The finding was more than minor because it adversely affected the RCS and barrier performance attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to verify functionality of the permissive interlocks for the RCS hot leg loop isolation valves following intrusive maintenance, challenged the assurance that the interlocks design function would be available to prevent opening of the RCS hot leg isolation valves with a flow path established to the RWST and; therefore, prevent a loss of RCS water inventory to the RWST. The finding was determined to be of very low safety significance because the finding would not have likely affected other mitigation systems resulting in a total loss of their safety function. Further, this finding did not constitute a violation of NRC requirements since the interlock function and associated components the licensee failed to test were not safety-related. The finding is assigned a crosscutting aspect in the resources component of the human performance area in that complete, accurate, and up-to-date work packages were not provided (H.2(c)).
05000261/FIN-2012005-01	2012Q4	Robinson	Adequacy of Preventative Maintenance for the Dedicated Shutdown Diesel Generator Cooling System	On October 2, 2012, during monthly testing of the DSDG in accordance with OST-910, Dedicated Shutdown Diesel Generator Monthly, the control room received a DSDG Trouble alarm. Shortly after the alarm was received, the DSDG tripped. The licensee determined that the DSDG automatically tripped due to an engine jacket water over temperature condition. After the trip, licensee personnel inspected the engine and discovered that the drive belts for the belt driven radiator fan had come off the pulleys which prevented proper heat removal from the engine cooling system. All three drive belts were found to have varying degrees of wear and degradation. The last visual inspection of the fan belts was performed on September 12, 2011 and the last satisfactory surveillance run was performed on August 28, 2012. The DSDG is required to supply back-up power during a 10 CFR 50.65 Station Blackout condition and Appendix R conditions. Following the discovery of the thrown belts, the licensee replaced all three belts and performed a root cause evaluation. The root cause team determined that the cause of the failure was the lack of a time based replacement of the fan belts. The belts were last replaced in 2003. The inspectors reviewed the licensees root cause and asked additional questions regarding the expected service life of the fan belts. Additional inspection is required to review the licensees response to the inspectors questions and determine if a performance deficiency exists.
05000261/FIN-2012005-02	2012Q4	Robinson	Failure to Effectively Implement Gas Intrusion Program	The inspectors identified a Finding for the licensees failure to perform the 18- month pre-refueling outage (RO) ultrasonic testing (UT) examinations on 47 potential gas accumulation locations required by plant operating manual PLP-085, Emergency Core Cooling Systems Gas Management Program (GL 2008-01). Compliance with PLP-085 ensures the capability of the safety injection (SI), residual heat removal (RHR), and containment spray (CS) systems to perform their safety-related functions, and effectively implements the licensees gas management program as committed to the NRC in response to Generic Letter 2008-01. The licensee entered the issue into the corrective action program (CAP) as nuclear condition report (NCR) 575063, and is evaluating corrective actions. The failure to perform pre-RO UT examinations on 47 potential gas accumulation locations, as required by PLP-085 was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, if the licensee continued to miss pre-RO UT examinations, conditions that result in the formation of voids in the SI, RHR, and CS systems could go undetected with the potential to adversely affect the systems capability to perform their functions. 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 it was not a design deficiency, it did not represent the loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The inspectors identified a cross-cutting aspect in the work practices component of the human performance area, because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel following procedures. Specifically, on two occasions, the licensee did not perform pre-RO UTs in accordance with their gas management program, as described in PLP-085.
05000261/FIN-2012005-03	2012Q4	Robinson	Questions Regarding Whether GOTHIC is Sufficiently Qualified for Use in Operability Determinations	Information Notice 2011-17, issued July 26, 2011, informed addressees of recent instances of gas accumulation in safety-related systems in which the resulting operability determination of the as-found condition relied on computer models (i.e., GOTHIC) that were not demonstrated to be technically appropriate for the intended application. Specifically, the computer models had not been sufficiently qualified by benchmarking against test or plant data. The inspectors reviewed information related to the licensees response to GL 2008-01 and determined that the licensee had found voids in the SI system, RHR system, and CS piping. In most instances, the licensee had used GOTHIC to evaluate the past operability of the subject systems with voids, and then vented the gas prior to returning the subject systems back to service. The licensee had also evaluated the continued operability of the subject systems with a void left in place until corrective actions were implemented. Specifically, in 2008, the licensee evaluated eight gas voids found following filling and venting of the subject systems that could not be successfully removed during RO-25. The inspectors observed that the licensee used the GOTHIC as part of these evaluations to perform analysis of gas movement to predict how a void volume in piping is translated into a transient void fraction at the entrance of the pumps. The evaluations were the basis for the continued operability until corrective actions could be taken to remove the voids during the following RO-26, approximately 19 months later. While acknowledging the NRCs concerns that the GOTHIC models may not be sufficiently qualified by benchmarking against test or plant data for the particular gas transport scenario and piping configuration being analyzed, the licensee prepared engineering change document EC 86423 to document their justifications for continued use of the GOTHIC models to support operability determinations. The inspectors determined that this issue will remain unresolved pending additional inspection and consultation with a GOTHIC subject matter expert at NRC headquarters to evaluate the licensees use of GOTHIC to support operability determinations. This issue will be identified as URI 05000261/2012005-03, Questions Regarding Whether GOTHIC is Sufficiently Qualified for Use in Operability Determinations.
05000261/FIN-2012005-04	2012Q4	Robinson	Questions Regarding the Adequacy of the Fill and Vent Procedure for the Residual Heat Removal Heat Exchangers	Procedure OP-201-1, RHR System Venting directs system venting by a series of static and dynamic venting evolutions. The inspectors noted that the procedure did not specify the minimum flowrates necessary to ensure an adequate dynamic flush of the HXs. Specifically, the inspectors identified that dynamic venting of the system is performed by establishing flow via both the RHR HXs and its bypass line, which reduces the effective flow available to dynamically vent the HXs. The licensee indicated that following the fill and vent procedure, operations performs a post maintenance test (per OST-253, Comprehensive Flow Test for the RHR Pumps ), before returning the system to service, that establishes full flow through the HXs and would completely vent the HXs if the initial fill and vent was not successful. The inspectors was concerned because establishing full flow through the HXs with a large enough void size inside the HXs could potentially result in a water hammer condition that exceeds the structural design limitations of the system. The licensee is performing an evaluation to determine if any voids could be left in the HXs after fill and vent, and what the potential effects on the system could be. The inspectors determined that this issue will remain unresolved pending additional inspection to evaluate the licensees evaluation. This issue will be identified as URI 05000261/2012005-04, Questions Regarding the Adequacy of the Fill and Vent Procedure for the Residual Heat Removal Heat Exchangers
05000261/FIN-2013007-01	2013Q2	Robinson	Failure to Account for Containment Temperature Measurement Uncertainty	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for instrument uncertainty on the containment bulk temperature instrumentation which was used to verify technical specification (TS) containment operability. This was a performance deficiency. The licensee entered this issue into their corrective action program as Nuclear Condition Report 603294 and performed an evaluation of the temperature instrumentation uncertainty. In addition, the licensee issued Standing Instruction 13-001 which specified the indicated containment temperature for entry into TS Limiting Condition for Operation 3.6.5 was to be 118 degrees Fahrenheit, a value that compensated for the temperature measurement uncertainty. 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, if the licensee did not account for the temperature measurement accuracy, containment temperature could unknowingly exceed the TS operability limit, and the licensee may not declare containment inoperable. The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components and did not involve a reduction in function of hydrogen igniters in the reactor containment. The cause of the finding was indicative of current licensee performance because the licensee failed to consider instrument uncertainty when they performed a containment re-analysis in 2013. The cause of the finding was directly related to the maintaining long term plant safety by maintenance of design margins cross-cutting aspect of the resources component in the area of human performance because when the containment re-analysis was performed, the licensee reduced margin between the analyzed value for containment starting temperature and the TS limit, making the instrument uncertainty of the temperature instruments more significant.
05000261/FIN-2013007-02	2013Q2	Robinson	Failure to Evaluate SBO Coping Equipment for Environmental Conditions	The team identified a Green finding for the licensees failure to follow NRC Regulatory Guide 1.155, Station Blackout, guidance (to which they are committed in the Updated Final Safety Analysis Report) for evaluating equipment needed to cope with a station blackout for the required duration for associated environmental conditions. This was a performance deficiency. The licensee entered the issue into their corrective action program as Nuclear Condition Report 600522, and established a calculation that determined the maximum expected temperature inside the compartment housing the dedicated shutdown diesel generator (DSDG) and evaluated the equipment to determine its capability to perform its function for the station blackout coping duration. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the capability and reliability of the equipment located in the DSDG compartment was not ensured since a comparison of equipment temperature ratings and expected DSDG compartment temperatures was not performed. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, and the structure, system, or component maintained its functionality. No cross-cutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the installation of the DSDG.
05000261/FIN-2013007-03	2013Q2	Robinson	Failure to Have Adequate Analyses Supporting the Degraded Voltage Relay Setpoints	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to have adequate analyses that supported safety-related load operation during a design basis accident while supplied by offsite power. This was a performance deficiency. The licensee entered the issue into the corrective action program as Nuclear Condition Reports 601201 and 605969, and performed an evaluation that determined the capability of starting the safety-related motors at degraded voltage conditions, as well as the capability of the electrical loads during the degraded grid voltage relay (DGVR) time delay to ensure equipment function was preserved. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of safety related loads to respond to a design basis accident under degraded voltage conditions. Evaluations of the effects of starting motors at the DGVR voltage dropout setpoint and the equipment survivability during the DGVR time delay were not performed. The team determined the finding required a detailed risk analysis, because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, and the team assumed the performance deficiency represented a loss of operability or functionality of the equipment that could be lost during the DGVR time delay. This assumption was made to bound the risk of the finding, because the licensee was still investigating whether or not there would be a loss of function of any equipment during the DGVR time delay period as of the date of this inspection report issuance. The team assumed a recoverable loss of function of all 480V motor control centers and assumed a degraded voltage condition exposure time of one hour per year. The one hour per year assumption is conservative relative to actual plant data which indicated a degraded voltage condition exposure of 44 seconds over the past 3 operating years. The results of the detailed risk analysis indicated an increase in core damage frequency <1E-6/year, which is representative of a finding of very low safety significance (Green). No crosscutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the degraded voltage evaluation.
05000261/FIN-2013007-04	2013Q2	Robinson	Failure to Have Adequate Analyses for the E1 Bus Fast Transfer	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the plant design during fast bus transfers. Specifically, the licensee failed to have an adequate analysis that ensured a successful fast bus transfer of the safety-related E1 bus feeder from the Unit Auxiliary Transformer to the Startup Transformer when required. This was a performance deficiency. The licensee entered the issue into the corrective action program as Nuclear Condition Reports 603357 and 605562, and performed an additional fast bus transfer evaluation of the E1 feeder breaker to ensure that the breaker would not trip under fast bus transfer conditions. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of safety related loads on the E1 bus because the licensee did not verify the E1 feeder 4 breaker would not trip during a fast bus transfer. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability and functionality. No cross-cutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the fast bus transfer evaluation.
05000261/FIN-2013007-05	2013Q2	Robinson	Failure to Have Appropriate Procedure to Verify Degraded Voltage Relay Circuit Status	The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe an adequate procedure that verified DGVR circuit operability following degraded voltage disable switch operation for reactor coolant pump (RCP) starts. This was a performance deficiency. The licensee entered the issue into the corrective action program as Nuclear Condition Report 602516, developed a test procedure, and verified the DGVR operability on both emergency buses. The performance deficiency was more than minor because if left uncorrected, it could become a more significant safety concern. Specifically, by not properly testing the DGVR circuit to ensure continuity following switch manipulation for RCP starts, the circuit could unknowingly become inoperable and non-functional for an entire operating cycle. The finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system function, did not represent an actual loss of function of at least a single train for greater than its technical specification (TS) allowed outage time or two separate safety systems out-of-service for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-TS trains. No cross-cutting aspect was assigned to this finding because the team determined that the cause of the finding was not indicative of current licensee performance due to the age of the modification that added the degraded voltage disable switches.
05000261/FIN-2017007-01	2017Q4	Robinson	Failure to Correctly Determine Qualified Life	The NRC identified a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish a qualified life for the motors covered by Environmental Qualification Documentation Package (EQDP)-0 803 in accordance with their administrative procedure AD-EG-ALL-1612, Environmental Qualification (EQ) Program. Specifically, the licensee did not correctly establish a qualified life for the motors covered by EQDP-0803 due to a calculational error. In response to the issue, Robinson staff placed the issue in their corrective action program as NCRs 2155050 and 2158467, and demonstrated operability by removing conservatisms regarding assumptions for cumulative energized time of the motors. Additionally, the licensee plans to replace the affected motors. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not establishing the correct qualified life for the motors resulted in a reduction in margin that impacted the reliability of the equipment. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The inspectors determined that the finding was indicative of current licensee performance, because the error occurred on June 28, 2017. A cross-cutting aspect of Documentation (H.7) in the Human Performance Area was assigned because the organization did not create and maintain complete, accurate and up to-date documentation.
05000261/FIN-2017007-02	2017Q4	Robinson	Failure to Perform Required O-ring Replacement to Maintain Qualification	The NRC identified a non-cited violation of 10 CFR Part 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, for the licensees failure to correctly identify the maintenance required to maintain the core exit thermocouple reference junction box in a qualified state. Specifically, the licensee did not identify that the qualifying entity required that the cover O-ring be replaced on a 5 year frequency in addition to being replaced any time the junction box cover was removed, and due to this, the O-rings have not been replaced since original installation. In response to the issue, Robinson staff placed the issue in their corrective action program as NCRs 2157897 and 2161580, and demonstrated operability via analysis of the qualification test results. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining the equipment in its qualified configuration affected its reliability. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. A cross-cutting aspect was not assigned because the finding was not indicative of current licensee performance.
05000261/FIN-2017007-03	2017Q4	Robinson	Failure to Determine Most Severe Containment Spray pH	The NRC identified a non-cited violation of 10 CFR Part 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, for the licensees failure to correctly determine the most severe composition of chemicals for containment spray for the purposes of environment al qualification of equipment in containment. Specifically, the licensee did not identify that the pH of the chemical spray could have been more severe than what was identified in the Environmental Qualification zone maps if the Spray Additive Tank (SAT) had been operated at its limits provided in procedures CP-001 and OST- 023. In response to this issue, the licensee placed the issue into their corrective action program as NCR 2162081, demonstrated operability by reviewing current and historical operating conditions of the tank, and implemented administrative controls to prevent exceeding the qualified pH limit. This performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the containment spray pH could have exceeded the pH to which equipment inside containment was qualified, if the SAT had been operated at its procedural limits. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. A cross-cutting aspect was not assigned because the finding was not indicative of current licensee performance.
05000261/FIN-2017007-04	2017Q4	Robinson	Crouse-Hinds Qualification and Life Extension	Introduction: The inspectors identified an unresolved item (URI) involving three separate concerns that could affect the qualification of Robinsons Crouse-Hinds (C-H) electrical penetration assemblies (EPAs). First, the inspectors were concerned that a similarity analysis, which fulfilled the requirements of Commission memorandum and Order CLI 80-21, In the matter of Petition for Emergency and Remedial Action, and 10 CFR 50.49, Environmental Qualification for Electric Equipment Important to Safety for Nuclear Power Plants, may not have been completed. Second, the inspectors were concerned that Robinson may not have demonstrated that the penetrations electrical performance specifications were met using appropriate IEEE standards, as stated in the UFSAR. Third, the inspectors were concerned that the licensee may not have used appropriate methods when extending the qualified life of the C-H EPAs. Description: (1) In Robinsons initial Bulletin 79-01 response dated June 1980, to justify the qualification of the C-H EPAs by similarity, Robinson submitted a Westinghouse (WEC) qualification report AB-11/12/73, Qualification Tests for a Modular Penetration 5 dia. (Prototype B1), obtained from Brunswick nuclear station; a record of a phone conversation between Robinson and WEC, CPL-77-550, dated 11/29/1977; and a WEC design specification for the C-H EPAs, CPL-R2-E3, dated 6/26/1968. In the technical evaluation report (TER) dated July 8, 1982, that accompanied the NRC staff safety evaluation report (SER) dated January 5, 1983, 10 regarding the Robinson EQ Program, the C-H EPAs qualification was identified as Category IV Documentation Not Available. In the 1982 TER and NRC SER, these specific submitted documents were listed as reviewed and, the qualification of the C-H EPAs remained Category IV. In a licensee letter, dated March 2, 1984, the licensee documented a meeting with the NRC staff discussing Robinsons proposed methods of resolution for each of the EQ deficiencies identified. Robinson appeared to commit to documenting a similarity analysis between their C-H manufactured EPAs and other similar EPAs found acceptable by the NRC staff. In the 1985 final NRC SER, the staff found Robinsons proposed method of resolution specified in the March 2, 1984 letter, acceptable. However, the 1984 submittal summarized a January 18, 1984 meeting with NRC where it was stated the NRC would not perform any additional equipment review and it was left up to the utility to state the adequacy of the documentation. During the inspection, Robinson provided the documents originally submitted (AB- 11/12/73, CPL-77-550, and CPL-R2-E3) to the inspectors to justify qualification by similarity. The inspectors had concerns with these documents justifying similarity between the WEC and C-H EPAs. a) In a review of AB-11/12/73 and comparing it to what was known about the C-H EPAs, the inspectors identified that the materials used in the WEC EPAs were not identical or sufficiently similar in material composition or performance specifications. The WEC tested EPAs used silicone rubber O-rings, a proprietary WEC composition Q epoxy resin potting material as the internal filler, and had a 5 diameter. The C- H EPAs did not use O-rings, used room temperature vulcanized (RTV) silicone rubber potting material as the internal filler, a thin layer of Sty-Cast epoxy resin to seal the end opening exposed to a DBA, and has an approximately 11 diameter. b) The inspectors noted the performance requirements demonstrated by the WEC pressure tests did not appear to envelope the required Robinson DBA pressure performance. The WEC maximum pressure only developed 1286.9lbf at 105psig, and the C-H EPA would develop 3955.2lbf at 42 psig. The effects of the more substantial forces on the C-H EPAs was not addressed. c) In the review of specification, CPL-R2-E3, the inspectors noted that specification CPL-R2-E3 was actually an EBASCO specification rather than a WEC specification as had been stated, and that C-H had taken exception to the specification due to chemical incompatibilities between the RTV potting material and cable insulations specified by EBASC O. Many of the Robinson documents still specify these incompatible cable insulations for use with the C-H EPAs without justification. d) In the review of CPL-77-550, the inspectors noted that the record of the phone call did not have any suitably specific information that could justify similarity to the C-H in materials, performance specifications, or manufacturing methods. The inspectors are concerned that Robinson was unable to provide an acceptable similarity analysis to address the deviati ons between the tested and installed EPAs. The licensee entered this concern into t heir corrective action program as NCR 2161911, and determined the equipment was operable. 11 (2) Robinsons UFSAR Section 3.8.1.2 stated, in part, that electrical penetrations are designed and demonstrated by test to withstand, without loss of leak tightness, the containment post-accident environment and to meet the National Electric Code, IEEE - Proposed Guide for Electrical Penetration Assemblies in Containment Structures for Stationary Nuclear Power Reactors or subsequent issues of this standard, IEEE Electric Penetration Assemblies in Containment Structure for Nuclear Power Generating Stations (IEEE 317). In accordance with the IEEE 317 versions reviewed from 1971 to 1976, the performance requirements are to be met by test during all conditions from mild plant conditions (normal) to the most limiting environmental conditions produced during DBAs (accident), and post-accident conditions. When asked to provide the test documentation that met these original requirements, Robinson was not able to provide them. In addition, the inspectors noted that electrical calculation RNP-E-5. 30, Crouse-Hinds Electrical Penetration Ampacity, Short Circuit, and Heat Generation Calculation, revision 6, indicated that the current plant design exceeded the electr ical performance specification for some of the C-H EPAs, and thus these EPAs would not meet the UFSAR and IEEE 317 specifications. The inspectors requested evidence that Robinson met the required verifications testing specified in the UFSAR Section 3.8.1.2, and that those test conditions are bounding of the current electrical plant design described in RNP-E-5.30. The inspectors are concerned that Robinson may not be in conformance with statements in the UFSAR and 10 CFR 50, Appendix B, Criterion III, Design Control, which required, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. The licensee entered this issue into their corrective action program as NCRs 2159165 and 2164589. (3) The inspectors identified two concerns with the way Robinson extended the qualified life of the C-H EPAs. First, Robinson reverse calculated an activation energy which appears to be outside of known acceptable Arrhenius techniques. Second, Robinson derived activation energies from EPAs with materials that were not the same as in the C-H EPAs. The inspectors noted that the Division of Operating Reactors (DOR) guidelines, Guidelines for Evaluating Qualification of Class 1E Electrical Equipment in Operating Reactors, and NUREG 0588 both accepted Arrhenius techniques as acceptable methods for determining the qualified lives of components, and required that the materials be identical or be justified by analysis. For the first concern, UFSAR Section 3.11.3, Qualification Tests Results, specified the EQDPs contained the qualification justification analysis for EQ components. The EQDP-0900, for the C-H EPA, credited the WEC EQ report AB-11/12/73 for thermal aging life calculation. The WEC EQ report applied Arrhenius techniques in accordance with IEEE 98-1972, IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials, and IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data. The WEC EQ report indicated that they had determined an activation energy and the confidence bounds, but they did not include this information or the data used to derive it. The omitted information would be required to identify the limitations of what WEC had derived for their thermal aging. To derive the pseudo activation energy and extend the life of the C-H EPAs from 40 to 60 years, Robinson applied 12 an Arrhenius equation and discounted the limitations involved with using the Arrhenius extrapolation techniques as specified in known quality standards. For the second concern, the inspectors determined that there were material deviations between the WEC and C-H EPAs that could potentially invalidate the pseudo activation energy Robinson derived. Robinson derived a 1.018eV activation energy, when the silicone RTV known to be used in construction of the C-H EPA had a more limiting activation energy of 0.63eV. The 0.63eV would have significant negative effect on the qualified life of the C-H EPA, invalidating the life extension and current EQ status. In addition, the inspectors noted that in the Robinson license renewal application and safety evaluation report, NUREG 1785, Section 4.4.1.1, Summary of Technical Information in the Application, the licensee appeared to commit to using the Arrhenius method, as described in Electric Power Research Institute (EPRI) NP- 1558, A Review of Equipment Aging Theory and Technology. The inspector noted that NP-1558 was not a quality standard as required by general design criteria 1 and 10 CFR 50.54(jj); however, its use would have likewise invalidated the WEC information for the C-H life extension. The inspectors are concerned that despite the specifications in the IEEE quality standards and the information in EPRI report NP-1558, Robinson extrapolated an invalid qualified life for the EPAs possibly making them unqualified to withstand a DBA. The licensee entered this concern into their corrective action program as NCR 2164567. This URI is opened to determine if a performance deficiency or a violation exists. To resolve the various aspects of this URI, the inspectors need: (1) Actual material and performance specification similarity analysis or confirmation of licensing basis; (2) The documented verification testing that satisfies statements in UFSAR 3.8.1.2, and confirmation that the electrical performance specifications tested are bounding of the current plant design; and 3) Confirmation that the actual penetration materials needed to be used when extending the qualified life, and what is required for appropriate application of Arrhenius techniques. (URI 05000261/2017007-04, Crouse-Hinds Qualification and Life Extension)
05000261/FIN-2017007-05	2017Q4	Robinson	Questions Regarding EQDP-0401 Method Used to Determine Activation Energy and Responsibility for Verification	Introduction: The inspectors identified a URI concerning Robinsons requirement to verify the qualification of components (e.g., Rosemount transmitters) required to meet 10 CFR 50.49. Description: The Rosemount transmitters EQ described by Robinson EQDP-0401, referenced Wyle test report 45592-3 for qualification, which referenced NUREG 0588 Category 1 requirements. The Wyle report, Table III Aging Matrix, identified electronic components along with their respective activation energies (eV) and the references that identified the source of this information. The report specified that thin film metal resistors were the most limiting of these components. The reference for the thin film metal resistor activation energy was an IEEE white paper published in 1965, The Determination and Application of Aging Mechanisms Data in Accelerated Testing of Selected Semiconductors, Capacitors and Resistors. The validity of Wyles determination of activation energies was in question because their methods had not been validated, as stated in the IEEE white paper. The inspectors reviewed the other components in Table III of the Wyle report to verify what components were more limiting and determined that the metal film resistors were not the most limiting. The inspectors identified that the activation energy in the Wyle report for transistors was for metal enclosed transistors, 1.02eV, but the transistors used in the transmitter construction were actually plastic enclosed transistors with activation energies ranging from 0.5eV to 0.66eV. The transmitters used some carbon resistors that were more limiting than metal film resistors and were more sensitive to radiation synergisms. Further, the information in the IEEE white paper seemed to indicate a phase change with an associated more limiting activation energy in the range of the normal plant environmental temperatures. The licensee appeared to not have evaluated this phase change and used the less conservative activation energy from the IEEE white paper throughout their extrapolations. Finally, Robinson may not have reviewed the actual activation energy test data, the test plan and acceptance criteria for the activation energy, or information about the test program, or if any equivalent App. B program supported the informations quality. NUREG 0588 Section 5(2), specified that independent verification of similarity or equivalence must be established, and that it was incumbent on the applicant to have the necessary documentation to justify the adequacy of using data from similar or equivalent equipment. In addition, this Section 5(2) and NUREG 0588, Appendix E, specified, that for electrical equipment that will experience the environmental conditions of design basis accidents for-which it-must function, the licensee must provide: the qualification test plan, test setup, test procedures, acceptance criteria and a summary of test results that demonstrates the adequacy of the qualification program. Additionally, if analysis is used for qualification, justification of all analysis assumptions must be provided. Further, NUREG 0588 Section 4(5) specified that known material phase changes must be addressed; and Section 4(6) specified that the aging acceleration rate used during qualification testing, and the basis upon which the rate was established, should be described and justified. In NUREG 0588 Part II, the comment resolution to Section 4(6), it was specified that the testing of the equipment should be conducted using the most limiting (lowest) activation energy of the components. Standard IEEE 323-1974 Section 5, Principles of Qualification, specified, that principles and procedures for demonstrating qualification include assurance that any extrapolation or inference be justified by allowances for known potential failure modes and the mechanism leading to them. Section 5.1, Type Testing, specified that test alone satisfies qualification only if the equipment to be tested is aged, subjected to all environmental influences, and operated under post-event conditions to provide assurance that all such equipment will be able to perform their intended function for at least the required operating time. The inspectors identified other known failure mechanisms were not considered. For instance, electro-migration of aluminum in diodes, transistors, and Zener diodes present in the electronics has an activation energy between 0.5eV and 0.63eV, which is more limiting than what was used. This failure mechanism was identified in EPRI NP-1558, A Review of Equipment Aging Theory and Technology, and in many IEEE documents that were known at the time of qualification. Robinson used what appeared to be an unvalidated activation energy that also appeared to overlook a phase change that occurs within the licensees service conditions to extend the qualified life. The activation energy value and the method used to arrive at this value are in question. This URI is opened to determine if a performance deficiency or violation exists. To resolve the various aspects of this URI, the inspectors need to: (1) assess the validity of the methods used in the IEEE white paper, which includes addressing the apparent phase change; (2) assess the difference of the more limiting activation energies for the resistors used in the Robinson transmitters compared to the value the licensee is using (including addressing the more limiting activation energies for the other electronics in question); and (3) evaluate the self-heating effects of the junctions in the electronic components and its impact on activation energy. Finally, the inspectors need to assess what responsibilities and to what extent, the licensee has to ensure the activation energies provided by an Appendix B vendor, are accurate and reasonable. The licensee entered this concern into their corrective action program as NCR 2164598. (URI 05000261/2017007-05, Questions Regarding EQDP-0401 Method Used to Determine Activation Energy and Responsibility for Verification)
05000261/FIN-2017007-06	2017Q4	Robinson	Penetration F01 Submergence	Introduction: The inspectors identified a URI concerning the submergence qualification of Robinson EPA F-01. The qualification may not have qualified the EPA in accordance with NUREG-0588, Category 1 requirements. Description: In 1988, the licensee determined that penetration F-01 would become submerged and subsequently contracted testing to demonstrate qualification. The inspectors reviewed Wyle qualification test report 41175-1, and EGS qualification test report, EGS-TR-903200-04-R000. These two reports were credited for submergence in EQDP-1700 for the CONAX penetrations. The inspectors were concerned that the CONAX penetration F-01 was not tested in its most limiting configuration. To place the penetration pigtails in a configuration that could support qualification, the licensee performed a modification, MOD 977, Repairs to Protect Penetration F-01, to re-terminate the pigtails by adding Raychem heat shrink to provide submergence protection. Modification, MOD 977, specifically figure 1, drawing number C20482, and feed through detail drawing number B190670 revision 1, appears to allow 36 conductors to be bundled together in a single pass through. The EGS and Wiley test reports did not test the 36 conductor configuration or demonstrate that the signals passing through these bundles would remain operable for the duration of submergence as required by NUREG-0588, Category 1 requirements. The inspectors were also concerned that while the termination procedures in MOD 977 required a two inch Raychem overlap, it also allowed a one-half inch overlap during Raychem installation. A one-half inch overlap may not ensure submergence qualification in accordance with EGS qualification report EGS-TR-903200-04-R000. In addition, the EGS qualification used an 8.3 pH caustic solution during submergence testing, which is less than what was required for Robinsons harsh environment design basis (10.5 pH). Title 10 CFR 50.49(d)(3) and (e)(6), RG 1.89 revision 1, C.d.3.a, and NUREG 0588 Section 2.2(5) Qualification by Test, required that equipment that could be submerged must be qualified by testing in a submerged condition to demonstrate operability for the duration required. The inspectors are concerned that F-01 is not qualified for submergence and the pigtails may not meet the requirements for submergence qualification. The licensee entered this concern into their corrective action program as NCR 2167136. This URI is opened to determine if a performance deficiency or violation exists. To resolve this URI, the inspectors need the licensee to address the apparent lack of qualification required by NUREG-0588, Category 1 EQ requirements. (URI 05000261/2017007-06, Penetration F01 Submergence)
05000261/FIN-2017007-07	2017Q4	Robinson	Justification of Activation Energy of ASCO Solenoid Coil Assemblies	Introduction: The inspectors identified a URI concerning the qualified life of ASCO solenoid operated valves. The qualified life determined by the licensee utilized unvalidated information provided by a third-party, non-Appendix B vendor and discounted other critical materials in their weak-link analysis without providing justification in accordance with Regulatory Guide 1.89, Rev. 1. Description: In 2006, the Nuclear Utility Group for Environmental Qualification (NUGEQ) provided a letter suggesting methods to extend the qualified lives of the solenoid operated valves. The licensee modified the qualified life of their ASCO valves as described by NUGEQ and failed to validate and justify the informations acceptability for use. Inspectors determined that the use of MW-35 magnet wires activation energy in place of MW-16 was not appropriate as activation energies are material and failure specific, and are not transferrable between different material compositions. Furthermore, the inspectors determined that the licensee (and NUGEQ) failed to adequately justify the discounting of the other materials in the ASCO solenoid coils, which had lower activation energies than the MW-16 magnet wire as reported by ASCO in their qualification test reports. The failure to justify the discounting of MW-16 magnet wire and other identified limiting component of the ASCO coil assembly was a performance deficiency and a violation of 10 CFR 50.49. Regulatory Guide 1.89, Rev. 1, Regulatory Position 5.c requires, in part, that the basis upon which the rate and activation energy were established should be defined, justified, and documented. Contrary to the above, the licensee failed to justify and document their use of the MW-35 activation energy in place of all other identified limiting activation energies in the ASCO solenoid coil assembly. Additionally, 10 CFR 50.49(e)(5) requires, in part, that equipment be replaced before the expiration of its qualified life unless ongoing testing can demonstrate that the equipment has additional life. Contrary to the above, the licensee failed to demonstrate that the ASCO solendoid coil assemblies have additional life when they failed to justify their departure from ASCOs limiting activation energies. This URI is being opened to determine if this performance deficiency is more than minor. To resolve this URI, the inspectors need to review the licensees response to proposed questions regarding the validation and justification of the appropriate activation energy that will be used in determining the qualified life. (URI 05000261/2017007-07, Justification of Activation Energy of ASCO Solenoid Coil Assemblies)
05000269/FIN-2011017-04	2011Q3	Oconee	Failure to Maintain SSF Pressurizer Heater Breakers as Safety-Related Components	An NRC-identified non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to maintain the Standby Shutdown Facility pressurizer heater breakers and associated electrical components as safety-related components or seismically-qualified as specified in the SSF licensing basis documents. The failure to maintain SSF systems, structures, and components as safety-related and seismically-qualified as required by the SSF licensing basis was a performance deficiency. This PD was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Configuration Control and adversely affects the cornerstone objective in that failure to maintain equipment qualification did not provide reasonable assurance that the SSF Auxiliary Service Water subsystem would perform its safety function. The finding was of very low safety significance because the finding involved a design or qualification deficiency confirmed not to result in loss of operability or functionality. The PD directly involved the cross-cutting aspect of thoroughly evaluates problems such that the resolutions address causes and extent of conditions, as necessary including evaluating for operability in the Corrective Action Program component of the Problem Identification and Resolution cross-cutting area for not properly evaluating an immediate determination of operability.
05000269/FIN-2012002-01	2012Q1	Oconee	Evaluation of Probable Maximum Flood Event	During a walkdown of Manhole 7 on February 1, 2012, inspectors noted that two conduit penetrations used to route PSW cabling into the AB were not sealed and provided a direct flooding pathway into the AB. These penetrations were identified as requiring seals whenever not being used for cable pulls or sealed immediately following cable pulling activities. Flooding from these penetrations would exceed the capacity of the AB sump pumps and fill the high pressure injection (HPI), low pressure injection (LPI) and reactor building spray (RBS) pump rooms rendering the pumps inoperable. The inspectors also identified that a field change rerouted the internal drainage system from the yard drain system to the adjacent radwaste trench. Rainwater accumulating in Manhole 7 would flow through the internal drains to the radwaste trench and into the AB through a non-isolable line which drained into the low activity waste tanks. These tanks would eventually overflow flooding the HPI, LPI and RBS pump rooms rendering the pumps inoperable. The design change in the original design package for Manhole 7 and the field change for rerouting the drain did not evaluate the impact they would have on the AB features to mitigate external floods. Consequently, the currently described Updated Final Safety Analysis Report (UFSAR) described PMF event would result in rendering safety-related/risk significant equipment inoperable. Additionally, the licensee recently completed a site inundation study which projected site water levels to be greater than the maximum flood protection measures for a PMF event as described in UFSAR Section 3.4.1.1 even with a fully functional yard drain system. Changes in site topography and construction of new buildings since initial construction appear to be contributing to the increased water levels. The licensee is currently evaluating the impact of the new site inundation level and has implemented interim actions to provide protection from increased water levels on-site. The NRC will perform additional inspection to ensure the impact to the AB from a PMF event is understood, an accurate timeline on Manhole 7 construction activities is developed, and that the extent of condition is fully defined. This issue is identified as URI 05000269, 270, 287/2012002-01, Evaluation of Probable Maximum Flood Event.
05000269/FIN-2012002-02	2012Q1	Oconee	Failure to Adequately Test Safety-Significant Medium Voltage Cables	An NRC-Identified finding was identified for the licensees failure to develop an adequate procedure for performing cable degradation testing on medium voltage cables. Consequently, a degraded condition of one of the conductors from CT-5 to the standby buses was not addressed for approximately 18 months and subsequently failed accruing approximately 30 days of unavailability to replace the cable. The performance deficiency (PD) was determined to be more than minor as it affected the Mitigating Systems cornerstone attribute of equipment performance in that failure to identify the degraded condition resulted in unplanned unavailability of the CT-5 power path. The finding was of very low safety significance because the Y phase cable from CT-5 was capable of performing its function from June 2010 until December 22, 2011. The cause of this finding was directly related to the implementation of operating experience aspect of the Operating Experience component of the Problem Identification and Resolution cross-cutting area, in that, the licensee failed to incorporate industry guidance to establish test acceptance criteria for degradation of power cables insulation.
05000269/FIN-2012002-03	2012Q1	Oconee	Inadequate Procedure for Installation of Safetyrelated Control Cables	An NRC-Identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to develop adequate procedures governing the installation of safety related control cables. The work package did not contain the maximum tension limits and the specified testing method was inadequate to demonstrate that control cables had not been damaged during the cable pull. The licensee revised TI/0/A/3000/030, PSW Cable Pulling in Duct Banks Using Mechanical Device, and re-tested the control cable ensure its functional integrity. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective in that it could represent an indeterminate functional condition for proper control functions for safety-related equipment operation in the protected service water system and the standby shutdown facility. The finding was of very low safety significance because it did not result in the loss of any system safety function. The cause of the finding directly involved the cross-cutting aspect of appropriate planning of work activities in the Work Control component of the Human Performance area, in that the licensee failed to implement procedures which established planned contingencies, compensatory actions, and abort criteria.

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