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	Quarter	Site	Title	Description
05000269/FIN-2012002-04	2012Q1	Oconee	Failure to Ensure UFSAR described Flood Protection Measures in Place	An NRC-Identified finding was identified for the licensees failure to ensure the Oconee UFSAR-described Auxiliary Building (AB) flood protection measures were maintained. Penetrations below the design basis 796.5 foot mean sea level (msl) elevation were not included in a surveillance program to verify below-grade penetrations would not allow flooding of the AB. The performance deficiency was more than minor because if left uncorrected, it could lead to a more significant safety concern, in that, other onsite activities such as excavation work exterior to the AB walls could provide a pathway for flood waters to enter the AB through the uncontrolled penetrations causing the loss of accident mitigation systems. The finding was of very low safety significance because an actual loss of operability or functionality did not occur. The cause of the finding was directly related to the appropriate corrective actions aspect of the Corrective Action Program component in the area of Problem Identification and Resolution because the licensee failed to correct the O-310 K series to identify that all external AB walls as flood barriers.
05000269/FIN-2012002-05	2012Q1	Oconee	Licensee-Identified Violation	10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews or by the performance of a suitable testing program. Contrary to the above, the licensee failed to perform adequate post modification testing to evaluate the adequacy of a design modification to the RPS/ES system and allowed an improper wiring configuration of the Unit 1 Nuclear Instrumentation (NI) Power Range detector cables. The licensee failed to follow their Modification Test Plan to verify that the Flux/Flow/Imbalance function for the Power Range NIs would work properly. This finding was not greater than very low safety significance (Green) because the RPS could have still fulfilled its safety function with the channels inoperable based on other trip signals available. The licensee entered this issue into their CAP as PIP O-11-7081.
05000280/FIN-2010006-01	2010Q3	Surry	Failure to Identify and Correct Degraded Unit 1 Nuclear Instrument RC Filters	An NRC identified non-cited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified for the licensees failure to identify and correct degraded RC filters associated with Unit 1 Nuclear Instrument (NI) cabinets for N-42 and N-44 based on a similar degraded condition identified on Unit 2 NI cabinet N-43 in November 2009. The issue was entered into the licensees corrective action program as condition report CR383881. All the RC filters in the Surry Unit 1 and 2 NI cabinets have been replaced with new RC filters. The finding was determined to be of more than minor significance because it is associated with the equipment performance attribute of the Initiating Events cornerstone. It adversely affected the cornerstone objective of protection against external events, i.e., fire. The performance deficiency was screened using phase 1 of the Significance Determination Process (SDP) and was determined to be a fire initiator contributor and to have impact on post fire safe shutdown, therefore a phase 2 analysis utilizing Inspection Manual chapter 0609 Appendix F was required. Since the finding involved MCR fire scenarios, a phase 3 analysis was required. A phase 3 risk analysis was performed by a regional SRA in accordance with IMC 0609 Appendix F, NUREG/CR6850, and utilizing the latest Surry SPAR probabilistic risk analysis model. The fire scenarios were determined to impact MCR operator actions but would not credibly require MCR evacuation for either habitability or safe shutdown functional requirements. The dominant sequence was a fire induced reactor trip transient initiator, with failures of auxiliary feedwater, main feedwater and failure to implement feed and bleed leading to core damage. Factors which mitigated the risk of the fire were the minimal fire growth potential and the potential for NI cabinet fires to damage SSD equipment. The risk evaluation result was an increase of <1E-6 for core damage frequency, a finding of very low risk significance (Green). This finding involved the cross cutting area of problem identification and resolution, the component of operating experience (OE), and the aspect of evaluating internal OE (P.2.a), because the licensee did not effectively evaluate the internal operating experience gained from the November 2009 RC filter failure prior to the failure of the RC filters on June 8, 2010.
05000280/FIN-2011004-01	2011Q3	Surry	Failure to Follow Scaffolding Procedure Requirements	The inspectors identified a NCV of Technical Specifications (TS) 6.4.D for failing to follow the requirements of procedure MA-AA-105, Scaffolding. Specifically, the licensee did not adequately implement scaffold evaluation, screening, and risk requirements for multiple scaffolds constructed in the vicinity of safety-related equipment. The inspectors determined that the failure to follow TS required procedure MA-AA- 105, Scaffolding, by not properly identifying scaffolds for safety-related systems and performing the required engineering evaluations, constitutes a performance deficiency. This finding is considered more than minor because it is similar to IMC 0612, Appendix E, Example 4.a in that the licensee routinely failed to perform the required engineering reviews and evaluations for scaffolding. This finding is also associated with the external factors and equipment performance attributes of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors screened this finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance since it was a deficiency determined not to have resulted in the loss of operability or functionality. The cause of this finding involved the cross-cutting area of human performance, the component of resources and the aspect of training (H.2(b)), because the licensee failed to implement training sufficient to ensure that operators were aware of plant equipment which is designated as safety-related.
05000280/FIN-2011004-02	2011Q3	Surry	Failure to Consider Instrument Uncertainty and Establish Calibration Controls for Rotameters Used to Vent Gas from ECCS Systems	An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, (with two examples) was identified for the failure to establish measures to apply rotameter instrument measurement error and appropriate instrument calibration controls or standards when using instruments of this type to determine the size of voids discovered as a result of ECCS system venting. The issue was entered into the licensees corrective action program (CAP) as CR419024 and CR419243. The failure to establish and implement measures (1) to ensure the application of +/- 5% rotameter instrument error to as-found void measurement, and (2) to ensure that rotameters calibrated to standard pressure conditions were used when utilizing those instruments to evaluate the size of as-found voids were performance deficiencies. The performance deficiencies were greater than minor, because, if left uncorrected, they could result in a more significant safety concern. Specifically, the performance deficiencies represented programmatic issues and if instrument error and/or appropriate calibration standards were not applied to instruments used for future void characterization, then sufficient measurement error could reasonably result such that as-found voids, which challenge or exceed established acceptance criteria, may not be identified as intended by post venting evaluations. The finding was screened for significance using the Mitigating Systems cornerstone column of Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because the finding did not represent a design or qualification deficiency, did not represent the loss of a safety system function, did not represent the loss of a train for greater than the allowed outage time, did not represent the loss of risk significant equipment for greater than 24 hours, and was not potentially risk significant due to external events. Because the licensee had failed to implement complete, accurate, and up-to-date controls necessary to ensure that rotameter error and calibration standards were adequately addressed by procedures used to evaluate the impact of voids on emergency core cooling systems, this finding is assigned a cross-cutting aspect in resources component of the human performance area
05000280/FIN-2011012-01	2010Q4	Surry	Inaccurate Fire Watch Records	The licensee identified a violation of 10 CFR 50.48 Fire Protection requirements when it was determined that a laborer failed to conduct a roving fire watch patrol. The licensee took substantial disciplinary actions and entered the deficiency into the corrective action program for resolution as CR 379888. This issue was dispositioned using traditional enforcement due to the deliberate aspects of the performance deficiency. Furthermore, the failure to provide complete and accurate information has the potential to impact the NRCs ability to perform its regulatory function. An individual assigned as a fire watch deliberately documented the completion of fire watch rounds (Fire Watch Tour Documentation Sheet, Attachment 14) for locations in which he did not conduct the fire watches. This issue was considered more than minor due to the deliberate aspects of the performance deficiency. In accordance with the guidance in Supplement VII of the Enforcement Policy, this issue is considered a Severity Level IV violation because it involved information that the NRC required to be maintained by a licensee that was incomplete or inaccurate and of more than minor safety significance. No cross-cutting aspect was identified because this performance deficiency was dispositioned using traditional enforcement.
05000280/FIN-2014007-01	2014Q3	Surry	Failure to Perform Required Preventative Maintenance on Class 1E Molded Case Circuit Breakers	The team identified a Green non-cited violation of Technical Specification 6.4.A.7, Unit Operating Procedures and Programs, for the licensees failure to implement written procedures to perform periodic tests for the Class 1E 125 volt direct current thermal-magnetic molded case circuit breakers (MCCBs). The licensee entered the issue into their corrective action program as condition reports CR558445 and CR560488 and performed an immediate determination of operability, in which they determined that the MCCBs were operable but not fully qualified. The licensees failure to conduct periodic tests to detect the deterioration of the system and to demonstrate that components not exercised during normal operation of the station are operable, as required by IEEE 308-1970, Section 6.3, was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, absent testing to detect deterioration and to demonstrate continued operability, the likelihood that these MCCBs will unpredictably fail when called upon increases with time in service. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and Inspection Manual Chapter 0612, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, which maintained its operability or functionality. The team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance.
05000280/FIN-2014007-02	2014Q3	Surry	Failure to Evaluate the Range of Conditions that Effect Canal Level Probes	The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate and quantify the system response times and accuracies over the range of conditions under which the service water canal level probes must operate. The licensee entered the issue into their corrective action program as condition report CR558429 and performed an immediate determination of operability, in which they determined the canal level probes to be operable but not fully qualified. The licensees failure to evaluate conditions that affected system response times and accuracy of the canal level probes, as required by IEEE 279-1968, Section 4.1, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, response time delays could allow the canal water level to fall below Technical Specification limits reducing the available heat removal required to mitigate Updated Final Safety Analysis Report chapter 14 design basis accidents. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and Inspection Manual Chapter 0612, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, which maintained its operability or functionality. The team determined that the finding was associated with the Design Margin cross-cutting aspect of the Human Performance area because recent modification designs for the canal probes were completed and approved without evaluating effects on the canal level probe response times and accuracies.
05000280/FIN-2014007-03	2014Q3	Surry	Adequacy of Class 1E 125VDC Branch Circuit Breaker Design	The team identified an Unresolved Item (URI) regarding the adequacy of design of the Class 1E 125VDC power branch circuit breaker for the 1H 4160V Bus controls. The team reviewed the Class 1E 125VDC power distribution design to verify compliance with the licensing basis requirements in IEEE 308-1970, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations. The Surry licensing basis commitment to IEEE 308-1970 required the quality of the Class 1E power system design to be sufficient to ensure that multiple engineered safety features (ESF) would not lose power because of design vulnerabilities. Specifically IEEE 308-1970 stated, in part, The Class IE electric systems shall be designed to assure that any design basis event as listed in Table 1 will not cause: 1) A loss of electric power to a number of engineered safety features, surveillance devices, or protection system devices sufficient to jeopardize the safety of the plant. Table 1 stated, in part, that design basis events include Single act, event, component failure, or circuit fault that can cause multiple equipment malfunctions. The team identified design vulnerabilities in design basis documents and in the sampled branch circuitry. In Calculation EE-0499, DC Vital Bus Short Circuit Current, dated 11/30/1998, the licensee used AC power time current curve (TCC) data for HFB MCCBs (used in the 125VDC distribution system) instead of DC TCC data. In addition, in this calculation, the licensee did not de-rate components for the ambient temperature in the switchgear room. Furthermore, in 2009, the licensee replaced certain HFB MCCBs with model HFDDC MCCBs; however, did not evaluate the DC characteristics of these HFDDC MCCBs, and instead evaluated an AC model HFD MCCB. Because of these vulnerabilities the team questioned the coordination of the installed HFDDC breaker and whether it was adequate to protect the 1H branch circuit in the ambient temperature of the switchgear room. These calculational vulnerabilities were consistent across both trains A & B and for both Units 1 & 2. The licensee captured the inspectors questions in their corrective action program as CR559872 and CR559875. This issue is a URI pending further review of information provided by the licensee on November 4, 2014, and consultation with the Office of Nuclear Reactor Regulation to determine if this issue of concern constitutes a violation.
05000280/FIN-2014007-04	2014Q3	Surry	Qualification Basis for Safety-Related Molded Case Circuit Breakers	The team identified a URI regarding the licensees actions to maintain or extend the qualification basis for safety-related MCCBs installed in mild environments greater than vendor design life specifications. In 2004, the licensee received Westinghouse Electric Technical Bulletin TB-04-13, Replacement Solutions for Obsolete Classic MCCBs, UL (Underwriters Laboratory) Testing Issues, Breaker Design Life and Trip Band Adjustment, which was superseded in 2006 by TB-06-02, Aging Issues and Subsequent Operating Issues for Breakers That are at Their 20-Year Design/Qualified Lives; UL Certification/Testing Issues Update. These bulletins informed the licensee of MCCB aging and operating issues. Specifically, grease and red oil used in these breakers were found to be key limiting factors for continued operability within published specifications. As grease and red oil aged beyond 20 years, their lubrication properties were reduced, resulting in slower trip times beyond the published time-current curves. The bulletins further defined the design life of MCCBs in mild environments as 20 years. However, the inspectors noted that approximately 60 safety-related MCCBs installed in mild environments exceeded 20 years of service, and the licensee had not performed an engineering evaluation to justify continued operation beyond this design life. The affected MCCBs were associated with the Class 1E 125VDC distribution systems (switchgear) on both units. The licensee captured the inspectors questions in their corrective action program as CR558445 and CR560488. This issue is a URI pending further review, including consultation with the Office of Nuclear Reactor Regulation, to determine if this issue of concern constitutes a violation.
05000302/FIN-2008006-01	2008Q1	Crystal River	Failure to Control Transient Combustibles	The team identified a non-cited violation of Crystal River Unit 3 Operating License Condition 2.C.(9), for the licensees failure to properly implement fire protection program procedures for control of transient combustible materials. Specifically, transient combustible materials were left unattended for four days in the 3B 480V ES Switchgear Room after work had been completed, which was a violation of the licensees administrative procedures for control of transient combustibles. Once identified, the licensee removed the combustible materials and initiated a nuclear condition report to address the issue. The finding is more than minor because the transient combustible materials presented a credible fire scenario involving equipment important to safety, which degraded the reactor safety Initiating Events cornerstone objective to limit the likelihood of those events that may upset plant stability and challenge critical safety functions. The amount of unattended transient combustible materials did not violate the licensees transient combustible control limits for the fire area. Therefore, the finding was assigned a low degradation rating against the combustible controls program. The finding was of very low safety significance (Green) based on the low degradation rating. This finding has a cross-cutting aspect in the Work Practices component of the Human Performance area because the licensee failed to effectively communicate expectations regarding procedural compliance and personnel following procedures (NRC Inspection Manual Chapter 0305, H.4(b))
05000302/FIN-2008006-02	2008Q1	Crystal River	Failure to Adequately Protect Cables for Valve DHV-42	The team identified a non-cited violation of 10 CFR 50, Appendix R, Section III.G.2., for failure to protect cables from fire damage for components required for safe shutdown. Specifically, the Mecatiss MTS-3 fire wrap installed around the cables for valve DHV-42 (suction from the reactor building sump to the Train A decay heat pump) was not installed in accordance with the vendors tested configuration. The licensee initiated a nuclear condition report and implemented an hourly roving fire watch to address this issue. Additionally, the licensee implemented repairs during the March 2008 forced outage to upgrade the Mecatiss MTS-3 fire wrap to comply with the vendor tested configuration. This finding is more than minor because it is associated with the external factors attribute, i.e., fire, and it degraded the reactor safety Mitigating Systems cornerstone objective. The inspectors completed a Phase 1 screening of the finding in accordance with IMC 0609, Appendix F, Attachment 1, Step 1.3, Qualitative Screening Approach, and concluded that the finding, when given credit for the fixed automatic suppression system in the area, was of very low safety significance (Green)
05000302/FIN-2008006-03	2008Q1	Crystal River	Evaluate Opening Access Hatch to Cable Spread Room	The team identified an unresolved item (URI) related to the licensees compliance with the CR-3 operating license condition 2.C.(9) and the approved FPP when the access hatch from the MCR floor to the CSR was opened on more than one occasion for maintenance troubleshooting activities. The team reviewed NCR 264494 which the licensee initiated in response to questions from the NRC resident inspectors who observed the access hatch from the MCR floor to the CSR was open and there did not appear to be any compensatory measures in place. The NCR stated that the licensee opened the access hatch between the MCR floor and the CSR to perform battery ground troubleshooting activities. The team questioned if this activity potentially degraded the CSR Halon suppression system. With the hatch open, the team questioned the capability of the Halon suppression system to meet the licensing basis requirement to maintain a 5% Halon concentration for 10 minutes in the event of an Appendix R fire in the CSR. The team also questioned if the licensee performed an evaluation to determine the impact of the hatch being open on the CSR Halon suppression system and to determine if compensatory measures were needed. As a result of questions raised by the team during the inspection, the licensee initiated NCR 266356 to evaluate the impact on the operability of the CSR Halon suppression system with the MCR access door hatch open. The team requested additional information from the licensee regarding the amount of thermoplastic cables in the CSR, how many times and the duration each time the hatch was opened for maintenance troubleshooting in the past year. The licensee provided the requested information to the team and the information is currently being reviewed. The team informed the licensee that this issue will be identified as an URI pending further NRC review of the requested information. This item will be tracked as URI 05000302/2008006-03, Evaluate Opening Access Hatch to Cable Spread Room
05000302/FIN-2008006-04	2008Q1	Crystal River	Reactor Coolant Pump 1B Lube Oil Collection System Leakage	A self-revealing non-cited violation of 10 CFR 50, Appendix R, Section III.O, was identified for failure of the reactor coolant pump (RCP) oil collection system to collect and drain RCP oil leakage to a vented closed container. Specifically, the licensee found an estimated one to two gallons of oil on the reactor building floor beneath RCP-1B. The licensee initiated a nuclear condition report for this issue. This finding is more than minor because it is associated with the external factors attribute, i.e., fire, and it degraded the reactor safety Initiating Events cornerstone objective. The team completed a Phase 1 screening of the finding in accordance with IMC 0609, Appendix F, Attachment 1, Step 1.3, Qualitative Screening Approach, and concluded that the finding was of very low safety significance (Green) because the amount of oil identified in 2008 was bounded by the licensees 2004 analysis (which assumed a 21 gallon oil leak). This finding has a cross-cutting aspect in the Corrective Action Program component of the Problem Identification and Resolution area because the licensee did not take appropriate corrective actions in a timely manner to address the adverse trend related to oil leakage for RCP-1B (NRC Inspection Manual Chapter 0305, P.1(d))
05000302/FIN-2008006-05	2008Q1	Crystal River	Design Oversight Results in 10 CFR 50, Appendix R, Cable Separation Criteria Not Met	The problem described in the LER is a performance deficiency because the licensee failed to protect cables important to safe shutdown as required. The problem is more than minor because it was associated with the external factors attribute, i.e. fire, of the Mitigating Systems cornerstone and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. When the probability of fire starting in the penetration area or inside containment is multiplied by the probability of the multiple cable damage states described above the result indicates the postulated event is lower than high safety significance (Red) and indicative of having very low safety significance.
05000321/FIN-2009005-01	2009Q4	Hatch	Submerged safety-related medium voltage cables	An unresolved item (URI) was opened related to underground pull box inspections which revealed a safety-related 4160 volt cable located in two pull boxes was submerged under water. The determination of a performance deficiency cannot be made until further information is provided by the licensee to support that the cables are designed, qualified, and acceptable for operation in a wetted and/or submerged environment. On December 10, 2009 during inspection of underground bunkers subject to flooding, the inspectors identified that safety-related 4160 volt cable, R22-S005-ES1- M08, located in pull boxes PB1-BF and PB1-BB was submerged. This issue was captured in the licensees corrective action program as CR 2009111808. The inspectors require documentation supporting the cables design, qualification, and testing history to evaluate whether this issue constitutes a performance deficiency. URI 05000321,366/2009005-01, Submerged safety-related medium voltage cable was identified to track this issue
05000321/FIN-2010003-01	2010Q2	Hatch	Failure to maintain safety related cables in a nonsubmerged environment	The NRC identified a NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure implement measures to assure that safety-related cables remained in an environment for which they were designed. Safety-related cables purchased and installed in underground electrical pull boxes at Hatch Nuclear Plant have been subjected to submergence, a condition for which they are not designed. To address this issue the licensee has performed the immediate corrective action of increasing the frequency of measuring water level and pump down of the pull boxes. The licensee initiated CR 2010104298 to address this issue. This performance deficiency is more than minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it is reasonable to conclude the cables may be in a degraded condition where the continued reliability of the cable cannot be ensured because: 1) the licensee does not have a cable testing/monitoring program to detect degradation of inaccessible or underground power cables; 2) the cables have been subject to a submerged physical environment which is outside the cables design parameters; and 3) there have been documented failures of cables throughout the nuclear industry due to degradation caused by submergence in water. Because the finding affects the safety of an operating reactor, the significance of this finding was screened using the Phase 1 of the SDP in accordance with NRC IMC 0609, Attachment 4, Table 4a. The finding screened as Green, because the finding is a design or qualification deficiency confirmed not to result in loss of operability or functionality. This finding has a cross-cutting aspect in the Work Control component of the Human Performance area, because the licensee did not appropriately coordinate activities by incorporating actions where maintenance scheduling is more preventive than reactive. Specifically, the licensee did not schedule performance of procedure 52PM-Y46-001-0, Inground Pull Box and Cable Duct Inspection for Water, at a frequency that prevented safety related cable submersion
05000321/FIN-2010003-02	2010Q2	Hatch	Failure to follow corrective action program procedure and prevent recurrence of severity level 2 root cause	A self-revealing NCV of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to follow their corrective action program procedure, NMP-GM-002, Ver. 4.0, that required severity level 1 and 2 condition reports (CR) to have corrective actions that prevent recurrence. From May 2006 to April 2010 licensee procedure NMP-GM-002, Corrective Action Program, Ver. 4.0, was not followed because corrective actions to prevent recurrence were not implemented prior to failure of Analog Transmitter Trip System (ATTS) card 1B21-N641C. The licensees immediate corrective actions were to replace the failed card, 1B21-N641C, the adjacent card 1B21-N690C and the high drywell pressure trip cards 1E11-N694A and C. The licensee initiated CR 2010105161 to address this issue. The performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the Initiating Events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure of ATTS card 1B21-N641C resulted in a spurious Loss of Coolant Accident (LOCA) signal that started Emergency Core Cooling System (ECCS) equipment and resulted in a power reduction to approximately 85%. Due to this finding affecting the safety of an operating reactor, the significance of this finding was screened using NRC IMC 0609, Attachment 4, Table 4a. Because the finding contributed to the likelihood of a reactor scram, but did not affect mitigation equipment availability, the finding screened as Green. The inspectors concluded that the performance deficiency does not have an associated cross-cutting aspect because the performance deficiency occurred in 2006 and is not indicative of the licensees current performance in the area of root cause investigations.
05000321/FIN-2010003-03	2010Q2	Hatch	Failure to follow procedure while in shutdown cooling to record corrected reactor water level	The NRC identified a NCV of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe in procedure 34GO-OPS-015-2, Maintaining Cold Shutdown or Refueling Condition, appropriate documented instructions for recording and verifying reactor water level when reactor vessel level is greater than 60 inches and instrument 2B21-R605 is unavailable. To address this issue the licensee performed the immediate corrective action of initiating CR 2010104615 and has generated an action item to upgrade procedure 34GO-OPS-015-2. This performance deficiency is more than minor because it is associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability of systems (ability of operators to monitor, trend, and maintain reactor water level) to prevent undesirable consequences. Because this finding is associated with the safety of a reactor while the unit was in cold shutdown and on residual heat removal shutdown cooling, NRC IMC 0609, Attachment 4, directs using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, to determine the significance of this finding. In Appendix G, Attachment 1, Checklist 6 was used because during the time period of this finding the unit was in cold shutdown, with a time to boil < 2 hours, and reactor coolant system level < 23 feet above the top of the reactor vessel flange. Each item in Appendix G, Attachment 1, Checklist 6 was determined to have been met, therefore per Figure 1 of Appendix G this finding screened as GREEN significance because a Qualitative Assessment was not required by Checklist 6. This finding has a cross-cutting aspect in the Work Control component of the Human Performance area, because the licensee did not plan and coordinate work activities consistent with nuclear safety including planned contingencies, compensatory actions, or abort criteria. Specifically, the licensee did not plan and coordinate the activity of transitioning the reference leg for reactor water level instrument 2B21-R605 with contingencies, compensatory actions, or abort criteria addressed to ensure measurable reactor water level was available to control room operators.
05000321/FIN-2010003-04	2010Q2	Hatch	1A EDG fuel oil return line failure	An unresolved item (URI) was opened related to CR 2010104391, fuel oil return line fitting failure on the 1A EDG. As of the end of this inspection period the licensee had not completed their investigation into this issue. The determination of a performance deficiency cannot be made until the licensee completes and documents their inspection efforts in this area On April 1, 2010 a fitting leak on the 1A EDG fuel oil return line was identified by the licensee during a monthly surveillance test run. CR 2010104391 documents that an attempt was made to tighten the fitting but the leak continued and that the leak needed to be repaired after the monthly surveillance test run was complete. The leak on this fitting was added to existing WO 1092436001 and scheduled for the 1A EDG system outage in May 2011. On June 3, 2010 during the monthly surveillance test run, the tubing associated with this fitting failed and diesel fuel oil was identified spraying onto the 1A EDG exhaust. The surveillance test run was terminated and the 1A EDG was secured by local operators in order to prevent a fire from starting. The failure of the 1A EDG fuel oil tubing was documented in CR 2010107248. URI 05000321/2010003- 04, 1A EDG fuel oil return line failure, will be identified to track this issue pending review of the investigation conducted under CR 2010107248 to evaluate whether this issue constitutes a performance deficiency.
05000321/FIN-2010003-05	2010Q2	Hatch	Licensee-Identified Violation	TS 5.4.1.a requires written procedures be established, implemented and, maintained covering the activities specified in Regulatory Guide 1.33, Appendix A. Items 2g and 4a of Appendix A requires procedures for general power operation and operation of the reactors recirculation system to be established and implemented. Contrary to the above, Unit 2 operated at a core flow higher than that allowed on the power/flow map described in licensee procedure 34GO-OPS-005-2, Power Changes. This issue was documented in the licensees corrective action program as CR 2009108237. Because the finding is associated with the fuel barrier and sufficient fuel thermal limit margin was maintained during the time core flow was outside the bounds of the power/flow map, this finding is of very low safety significance.
05000321/FIN-2010003-06	2010Q2	Hatch	Licensee-Identified Violation	TS 3.4.3 requires 10 of 11 SRVs to be operable during Modes 1, 2 and 3. Contrary to the above, on March 11, 2010 on Unit 1 it was identified during bench testing that five safety relief valves failed to lift at the required TS setpoint. The cause was found to be corrosion induced bonding between the pilot disc and seating surface. This condition was documented in CR 2010103338. This finding is of very low safety significance because a previous evaluation performed by the licensee bounds this condition and RCS pressure would be maintained below the TS safety limit.
05000321/FIN-2016010-02	2016Q2	Hatch	Failure to Identify N2E Nozzle Weld Through-Wall Flaw	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to promptly identify a condition adverse to quality regarding a through-wall flaw in the safe end-to-nozzle weld of the reactor coolant system N2E nozzle. The licensee has since repaired the flaw, completed all required postrepair examinations, and entered this issue entered this into their corrective action program as CR 10247856. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors screened this finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012. Because after a reasonable assessment of degradation, the finding could neither result in exceeding the RCS leak rate for a small LOCA, nor likely affected other systems used to mitigate a LOCA resulting in a total loss of their function, the finding screened as Green. This finding has a cross-cutting aspect of Challenge the Unknown in the area of Human Performance (H.11) because upon discovery of a less robust configuration of the N2E nozzle overlay, the licensee failed to consider the implications on the flaw that had existed in that component since 1988.
05000324/FIN-2011004-01	2011Q3	Brunswick	Inadequate Configuration Control Resulted in Rainwater Intrusion into the Unit 2 Reactor Building	A self-revealing Green non-cited violation of TS 5.4.1, Procedures, was identified for failure to implement procedural requirements of the equipment configuration control program to ensure that temporary power cables routed through an open manhole and into the reactor building north RHR (NRHR) room did not adversely impact the flood mitigation function of the storm drain system. This finding resulted in rainwater intrusion into the unit 2 reactor building. Upon discovery of this condition, the licensee resealed the manhole. The condition was entered into the licensees CAP as AR #483473. The failure to implement the requirements of the equipment configuration control program to ensure that the temporary cable routing did not adversely impact external flood protection features was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors - Flood Hazards and adversely affected the cornerstone objective in that the temporary change impacted the storm drain system which was credited for external flood protection. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Screening Worksheet, the finding screened as very low safety significance (Green) because it: (1) was not a design or qualification deficiency that was confirmed not to affect equipment operability; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; (4) did not represent a loss of risk significant non-Technical Specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event per table 4b of the worksheet because the leakage did not degrade the RHR system. The cause of the finding was directly related to the appropriately planning work activities cross-cutting aspect in the Work Control component of the Human Performance area because the licensee failed to incorporate environmental conditions which may impact plant structures, systems, and components into the temporary change.
05000324/FIN-2011004-02	2011Q3	Brunswick	Inadequate Corrective Actions for Control Building Air Conditioning Failures	The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criteria XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality related to the Control Room Air Conditioning (AC) system. Specifically, the licensee failed to identify and correct repetitive failures of nonconforming low ambient temperature damper actuators for the 2D control building air cooled condenser unit. This resulted in multiple control building AC refrigerant circuit failures. Upon discovery of the issue, the licensee placed the control building AC system in a safe condition for summer operation and initiated actions to procure acceptable damper actuators prior to the onset of low seasonal temperatures. The condition was entered into the licensees CAP as AR #462873. The inspectors determined that the licensees failure to promptly identify and correct the failures of the 2D control room AC system low ambient temperature damper actuators was a performance deficiency. This finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the finding reduced the reliability of the control building AC system and its ability to maintain control building equipment within specified temperature limits. The significance of the finding was evaluated using Phase 1 of the significance determination process in accordance with the Inspection Manual Chapter 0609 Attachment 4. The finding was determined to be of very low safety significance (Green) because the finding was a design or qualification deficiency that was confirmed not to affect equipment operability. The cause of this finding was directly related to the cross cutting aspect of thorough evaluation of problems in the Corrective Action Program component of the Problem Identification and Resolution area, because the licensee failed to promptly evaluate the failures of the low ambient temperature damper actuators and eliminate the adverse condition.
05000324/FIN-2011004-03	2011Q3	Brunswick	Inadequate Maintenance Results in Containment Isolation Valve Failure	A self-revealing Green finding was identified for inadequate maintenance on the overload relay of the unit 2 reactor water cleanup (RWCU) system inlet isolation valve 2-G31-F001. As a result of the inadequate maintenance, the overload relay actuated during operation of the valve under normal conditions, and the valve failed to shut. This was revealed while operators were attempting to isolate the RWCU system on August 2, 2011. After the valve failed to fully shut on August 2, 2011, the licensee shut the valve in series with 2-G31-F001 (2-G31-F004), repaired the overload relay for the 2-G31-F001 valve by installing the correct fasteners, returned the 2-G31-F001 valve to service, and entered the issue into their corrective action program (AR #480063). The inadequate maintenance on the 2-G31-F001 valve overload relay was a performance deficiency. The finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of structure, system, and component (SSC) and Barrier Performance, and it affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the finding prevented a primary containment isolation valve from shutting. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for Containment Barriers. The finding was determined to be of very low safety significance (Green) because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. The cause of this finding has no cross-cutting aspect because the maintenance took place in 1992 and is not indicative of current licensee performance.
05000324/FIN-2011004-04	2011Q3	Brunswick	Licensee-Identified Violation	Technical Specification 5.4.1, Procedures, requires that written procedures shall be implemented covering applicable procedures recommended in Regulatory Guide 1.33, Appendix A, November 1972 (Safety Guide 33, November 1972). Regulatory Guide 1.33, section I (Safety Guide 33, November 1972) requires written procedures for performing maintenance. Contrary to the above, the licensee identified that maintenance procedure 0CM-VFC500, Instructions for Repair, Reassembly, and Adjustment of the RCIC Terry Turbine Governor Valve, did not contain adequate guidance for assembling the unit 2 RCIC turbine governor valve. As a result, inadequate maintenance was performed on the unit 2 RCIC governor valve in 2009 in that proper spacing of the valve stem packing spacers was not maintained. This inadequate maintenance on the RCIC governor valve led to failure of the valve during quarterly surveillance testing on April 15, 2011. This finding was evaluated by the Regional Senior Reactor Analyst performing a Phase 3 significance analysis. The finding was determined to have a risk lower than 1E-6, and is GREEN. The short exposure time, and the availability of the severe accident mitigation alternative (SAMA) diesels for battery charging contributed to the low impact of the finding. The results were dominated by loss of the DC bus that powers HPCI, combined with automatic depressurization system (ADS) failures that could lead to high pressure core melt. External Events and Large Early Release Probability were found not to be major contributors to the risk of the finding. As corrective actions, the licensee revised the maintenance procedure and repaired the valve. This issue is in the licensees CAP as NCR #468283.
05000324/FIN-2011004-05	2011Q3	Brunswick	Licensee-Identified Violation	Technical Specification (TS) 3.3.6.1, Primary Containment Isolation Instrumentation, requires that the RWCU high differential flow instrumentation be operable in modes 1, 2, or 3. If the instrumentation is not operable, then TS 3.3.6.1 requires that the RWCU penetration flow path be isolated within 1 hour. Contrary to the above, the licensee identified that the RWCU high differential flow instrumentation was not operable and the penetration flow path was not isolated when the unit entered mode 1 on April 16, 2011 until August 2, 2011, because the RWCU inlet flow sensing element was installed backwards, causing the flow sensing element to be inaccurate. The resulting inaccuracy caused the instrumentation to be unable to isolate within the required TS limit of less than or equal to 73 gallons per minute differential flow. The finding was determined to be of very low safety significance per Appendix A of Inspection Manual Chapter 0609, Significance Determination Process, because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. Upon discovery of the condition, the licensee isolated the affected penetration flow path and installed the flow sensing element correctly. The issue is in the licensees CAP as NCR #479248.
05000327/FIN-2013007-01	2013Q3	Sequoyah	Failure to Evaluate a Potential Condition Adverse to Quality Prior to Mode Change	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow a test control procedure to evaluate indications of excessive check valve leakage prior to changing modes. Specifically, the licensee failed to evaluate the potential inoperability of residual heat removal check valve 2-63-563, which exhibited indications of excessive leakage, as required by procedure NPG-SPP- 06.9.1, Conduct of Testing, prior to transitioning to Mode 3, during startup. This was a performance deficiency. After conducting interviews with operations staff and performing a prompt determination of operability, the licensee concluded that the valve was never inoperable, since the valve subsequently passed its leak rate test in Mode 3 with no maintenance being performed. The operability determination was documented in PER 757559. This performance deficiency was determined to be more than minor because it affected the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, failing to evaluate indications of excessive check valve leakage while performing procedure 2-SISXV- 063-206.0, ECCS Check Valve Leak Testing section 6.3.2, adversely affected the cornerstone objective of limiting the likelihood of events that challenge the critical safety function of maintaining the RCS pressure boundary. 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 would not have affected other systems used to mitigate a LOCA resulting in a total loss of their functions. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area of Human Performance, component of Decision-Making because the licensee did not use conservative assumptions in their decision making when they failed to evaluate the potential inoperability of check valve 2-63-563 prior to transitioning to Mode 3.
05000327/FIN-2013007-02	2013Q3	Sequoyah	Failure to Evaluate Impact for Full Range of EDG Frequency	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate the entire range of allowable emergency diesel generator (EDG) frequencies into design basis documents. The failure to analyze the effects of the technical specification allowable EDG frequency range on the safety-related components powered by the EDGs was a performance deficiency. The licensee entered this issue in their corrective action program as PER 758761 and performed a prompt operability evaluation to determine that the safety-related equipment powered by the EDGs with a limited frequency range variation of 59.9 to 60.1 Hz, would be able to perform their design basis functions under accident conditions. In addition, a review of the results of the EDGs surveillances indicates no history of being outside the range of 59.9 to 60.1 Hz for the last three years. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to account for the allowable range of the EDG frequency and not evaluating the impact on safety related components powered by the EDGs did not ensure the availability and capability of safety-related components to respond to initiating events. 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 determined that this finding represented present licensee performance and directly involved the cross-cutting area of Human Performance, component of Resources because the licensee failed to ensure that design calculations affected by EDG frequency were complete and accurate.
05000327/FIN-2013007-03	2013Q3	Sequoyah	Failure to Properly Translate the Design and Licensing Bases for the 125 VDC System Into Design Calculations	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly translate the design and licensing bases for the 125 VDC system into design calculations. The licensee inappropriately credited the battery chargers for voltage support during accident scenarios in their voltage drop calculations, and failed to include vital inverters in the battery load profile. This was a performance deficiency. In response to the teams inquiries, the licensee initiated PER 758465 that provided reasonable expectation of operability by demonstrating that the required voltages would be available. This was based on interpolation of the vendor battery curves considering the maximum loading on the battery for the applicable portions of the duty cycle. This performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to properly evaluate the 125 VDC system under accident conditions to ensure the capability and availability of 125V control circuits to operate during design basis events. 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 was not indicative of present licensee performance.
05000327/FIN-2013007-04	2013Q3	Sequoyah	Inadequate Basis for AFW MOV Motor Brake Alternate Voltage Criteria	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to check the adequacy of the design of the steam generator feedwater isolation valve motor brakes. Specifically, the licensee based voltage acceptance criterion of 74% of 460V for motor brakes used in a design basis calculation on inadequate testing and calculational methods. This was a performance deficiency. In response to the teams concerns, the licensee initiated PER 763818 and provided reasonable expectation of operability of the motor brakes, by use of administratively controlled voltage, pending restoration of full qualification. This performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inadequate design criteria did not ensure the availability, reliability, and capability of the steam generator feedwater isolation valve motor brakes to operate under design basis degraded voltage 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. A cross-cutting aspect was not identified because this performance deficiency was not indicative of present licensee performance.
05000327/FIN-2013007-05	2013Q3	Sequoyah	Failure to Document Deficiencies Discovered During Receipt Inspections in the Corrective Action Program	The team identified a non-cited violation of TS 6.8.1, Procedures and Programs, the licensees failure to properly implement maintenance procedures for performing receipt inspection of new 480V circuit breakers. Specifically, the licensees failure to evaluate the need to report defects and deficiencies, identified on new safety-related 480V circuit breakers, in the corrective action program as prescribed by procedure was a performance deficiency. The licensee corrected the deficiencies prior to putting the breakers in service. This issue was entered into the licensees corrective action program as PERs 763834 and 759238. This performance deficiency was determined to be more than minor because if left uncorrected could lead to a more significant safety concern. Specifically, not documenting deficiencies that could adversely affect the breakers in the corrective action program, would not ensure breaker issues were being properly trended and that the issues have been adequately corrected and are not recurring. 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 it was not a design deficiency resulting in the loss of functionality or operability. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area Human Performance, component of Work Practices because the licensee failed to meet expectations regarding procedural compliance and did not follow procedures related to 480V safety-related breaker receipt inspections.
05000327/FIN-2013007-06	2013Q3	Sequoyah	Failure to Adequately Translate Design Basis Into Procedure Acceptance Criteria Time to Perform Operator Action	The team identified an Apparent Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate design basis requirements into emergency sub-procedure, ES- 1.3, Transfer to Residual Heat Removal Containment Sump, Rev. 19. Specifically, the time allotted for operators to perform time critical actions to swap emergency core cooling system (ECCS) pump suction from the refueling water storage tank (RWST) to the containment sump during a small break loss of coolant accident (SBLOCA) did not properly account for the instrument uncertainty and the design basis requirement in Updated Final Safety Analysis Report 15.3.1, to ensure the recovery of the core was demonstrated and to ensure continuous operation of the ECCS. This was a performance deficiency. As immediate corrective action, the licensee performed an operability review and documented the results in the corrective action program as PERs 760336 and 758761. The licensee concluded that there were no current operability concerns, and created Standing Order SO-13-025 to reinforce operator time performance requirements. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected objective of ensuring the availability, reliability, and capability of containment spray pumps, safety injection pumps, and charging pumps during a SBLOCA. Specifically, the licensee failed to demonstrate that operators would be able to successfully complete the time critical actions prior to reaching 8% RWST tank level, which required operators to secure all pumps taking suction from the RWST, because they did not consider the worst case allowable RWST level instrument uncertainty acceptance criteria along with the design pump flow rates. This action would result in the momentary loss of all ECCS high pressure injection during a SBLOCA and did not ensure the availability, reliability, and capability of the ECCS to respond to initiating events. 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. The safety significance is to be determined pending review and analysis of additional information from the licensee to determine if this finding is representative of an actual loss of the ECCS safety function. As a result, this finding is characterized as TBD. The finding did not represent an immediate safety concern because a review of past results indicated that operators were consistently performing the actions in times less than required, as documented by simulator testing. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of present licensee performance.
05000327/FIN-2013007-07	2013Q3	Sequoyah	Failure to Perform 50.59 Screens for Scaffolds and Clearances	The team identified a non-cited violation of TS 6.8.1, Procedures and Programs, for the licensees failure to implement procedures for equipment and maintenance control. The licensees failure to perform 10 CFR 50.59 reviews of temporary plant changes (e.g., scaffolding and clearances) that existed for greater than 90 days of plant operation was a performance deficiency. The licensee implemented corrective actions to review all of the temporary plant changes. The licensee generated PERs 756276, 753175, and 756308. This performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the team identified multiple examples where the licensee failed to evaluate temporary plant changes to ensure those changes did not affect the availability, reliability, and capability of systems that respond to events. 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 it was not a design deficiency resulting in the loss of functionality or operability. The team determined that this finding represented present licensee performance and directly involved the cross-cutting area of Human Performance, component of Work Practices because licensee failed to meet expectations regarding procedural compliance and did not follow procedures related to performing 50.59 reviews of temporary plant changes that existed for greater than 90 days of plant operation.
05000327/FIN-2013007-08	2013Q3	Sequoyah	Inadequate Corrective Action for 2010 Degraded Voltage Issues	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 correct deficiencies in electrical calculations for the safety-related AC electrical distribution system identified during the 2010 CDBI. Specifically, the licensees failure to identify that safety-related motor operated valve (MOVs) needed to be evaluated for new lower calculated available voltage (degraded voltage) to ensure their operability was a performance deficiency. The licensee initiated PER 753504 and performed a prompt determination of operability (PDO). The team concluded that the evaluations and compensatory measures described in the PDO provided reasonable expectation of operability. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to identify and evaluate that safety-related MOVs could be affected by degraded voltage conditions did not ensure the availability, reliability, and capability of the MOVs to respond to initiating events. 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 determined that this finding represented present licensee performance and directly involved the cross-cutting area of Problem Identification and Resolution, component of Corrective Action Program because the licensee failed to identify that safety-related MOVs needed to be evaluated for new lower calculated available voltage (degraded voltage) to ensure their operability.
05000327/FIN-2013007-09	2013Q3	Sequoyah	Insufficient EDG Starting Air Pressure Following SBO Coping Period	The team identified an unresolved item (URI) associated with licensee?s capability to meet their station blackout (SBO) mitigation strategy. Specifically, based on the allowable air start check valve leakage and the amount of air used during start attempts of the EDGs, the team found that the licensee did not ensure if adequate starting air pressure would exist to reliably start the EDGs following a SBO. Title 10 CFR 50.2, Definitions, defines a SBO as the complete loss of ac power to the essential and nonessential switchgear buses in a nuclear power plant, concurrent with turbine trip and unavailability of the onsite emergency power system. Essentially, this would involve the loss of the offsite power sources as well as the loss of emergency onsite AC power sources. The licensee is committed to coping with an SBO event for a duration of four hours, after which the licensee will recover AC power. The EDG air start system provides compressed air to start the EDGs. The compressed air is provided by non-safety related air compressors, and is stored in two safety-related air receiver tanks. Receiver tank ?A? is designed to maintain the air between 250 and 300 psig; tank ?B? is designed to maintain between 185 and 200 psig. The EDG air start system is equipped with check valves to maintain the integrity of the safety-related portion of the air start system. The licensee declares the EDG degraded if the receiver tank ?A? is less than 200 psig, due to the inability to meet the five start design basis requirement as described in UFSAR, Section 9.5.6, Diesel Generator Starting System. The EDG is declared inoperable at pressures below 150 psig on receiver ?B? due to the loss of start capability. This is based on the manufacturer?s value at which EDG starting and achieving rated speed and voltage has been demonstrated by testing. The team noted that the leak rate acceptance criterion outlined in procedure 0-PI-SXV- 082-203, Diesel Starting Air Valve Test, was 5 psig/minute for the EDG air start check valves. At this allowable leak rate, the EDG air start pressure could fall below 150 psig within 1 hour after an SBO and completely depressurize the air receiver within 3 hours after an SBO. This would not support the capability of the EDGs to start at the end of the 4-hour SBO coping period. In addition to concerns regarding check valve leak rate acceptance criteria, the team noted that postulated failed start attempts during an SBO event would also adversely impact the amount of air that would be available at the end of the 4- hour coping period. Specifically, in a SBO event, the initial failure of the onsite power sources would be followed by a failure of both onsite EDGs to start. The licensee?s procedures direct operators to attempt to start the EDGs a second time in the first few minutes of the SBO. The first and second start attempts are postulated to be unsuccessful during an SBO. The loss of offsite and onsite emergency ac power would prevent the air start compressors from recharging the tanks after the failed start attempts. Based on allowable check valve leakage and the amount of air used during two failed start attempts of the EDGs, the team found that the licensee did not ensure if adequate starting air pressure would exist to reliably start the EDGs in order to recover from an SBO after the 4 hour coping period. The team also found that the licensee had not developed procedural guidance to provide adequate air pressure to reliably start the EDG in order to recover from a SBO after the 4-hour coping period. The licensee captured these concerns in PER 763335. This issue remains unresolved pending inspector consultation with NRC headquarters technical staff for clarification of the licensee?s current license basis design requirements (with respect to 10 CFR 50.63 compliance), to determine if a performance deficiency exists. This issue is being identified as URI 05000327, 328/2013007-09, Insufficient EDG Starting Air Pressure following SBO Coping Period.
05000327/FIN-2014002-01	2014Q1	Sequoyah	Inadequate Clearance Causes Control Air System Transient	A self-revealing non-cited violation of Units 1 and 2 Technical Specification 6.8.1.a, Administrative Controls (Procedures), was documented for the licensees failure to establish an adequate clearance in preparation for maintenance activities on the B station air compressor. Implementation of this inadequate clearance on February 21, 2014, resulted in a reduction of control air pressure and a plant transient which challenged control room operators. Immediate corrective action was to revise the clearance to establish an adequate boundary. The licensee entered the issue into the corrective action program (CAP) for resolution as PER 850331. The performance deficiency was more than minor because it was associated with the configuration control and human performance attributes of the initiating events cornerstone and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inadequate clearance caused a plant transient during power operations that without operator action would have resulted in a loss of air operated plant components and ultimately require the operators to trip both units. The finding was determined to be of very low (green) safety significance based on Exhibit 1, Initiating Events Screening Questions, found in Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination Process for Findings At-Power, because the finding did not result in a complete or partial loss of a support system that contributed to the likelihood of, or cause, an initiating event and affected mitigation equipment. The inspectors determined the cause of this finding was associated with a cross cutting aspect of Work Management in the Human Performance area. Specifically, the licensee failed to implement their clearance process such that nuclear safety was the overriding priority.
05000327/FIN-2014002-02	2014Q1	Sequoyah	Licensee-Identified Violation	Failure to comply with technical specifications during refueling operations Unit 1 technical specification (TS) 3.3.9.4.c requires that during refueling operations, each penetration providing direct access from the containment atmosphere to the outside atmosphere be closed by a manual valve (if so equipped). Contrary to the above, between October 19 and 22, 2013, there were several instances where a Unit 1 containment penetration, X-108, to the additional equipment building was open (including its associated manual valve) during movement of irradiated fuel. This problem was entered into the licensees corrective action program as PER 800432, 806293, and 824224. Using Inspection Manual Chapter 0609, Appendix G, Shut-down Operations Significance Determination Process dated February 28, 2005 the inspectors determined that, the finding was Green because it did not: 1) involve a loss of reactor coolant system (RCS) inventory; 2) degrade ability to terminate a leak path or add RCS inventory as needed; or 3) degrade the ability to recover RHR once it was lost. This issue is also discussed under Section 4OA3 of this report.
05000327/FIN-2014002-03	2014Q1	Sequoyah	Licensee-Identified Violation	Failure to perform adequate post maintenance testing of the 1B EDG 10 CFR 50, Criterion XI, Test Control requires in part that an established testing program shall require that all testing of SSCs ensure that the SSC can perform its intended function. Contrary to the above, on February 23, 2014, adequate testing to ensure the EDG air start motors could fulfill their required functions was not performed. An adequate test was not performed until March 16 which was part of an annual testing program. This problem was entered into the licensees corrective action program as PER 859633. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the 1B EDG retained the capability to automatically start despite the improper air hose configuration of the air start motors.
05000327/FIN-2017008-01	2017Q4	Sequoyah	Unjustified Qualified Life for Target Rock Power-Operated Relief Valves	The team identified a Green NCV of Title 10 Code of Federal Regulations 50.49(e)(5) Aging when the licensee failed to replace, refurbish, or demonstrate additional life for components that exceeded their qualified life. The licensee failed to justify changes to the accelerated aging calculations used for power operated relief valve harsh environmental qualification. The licensee entered this issue into their corrective action program as CRs 1365730 and 1366082, and performed operability determinations, which determined the systems were operable but non-conforming with 10 CFR 50.49. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that Target Rock power-operated relief valves were qualified for the duration that they were required to operate reduced the reliability of reactor coolant system in the harsh environments of design basis accidents. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the team determined that the finding was of very low significance (Green) because it was a design deficiency that affected the design or qualification of a mitigating system, however, the mitigating system maintained its operability. The team determined there was no cross-cutting aspect associated with this finding since it was not indicative of current licensee performance.
05000327/FIN-2017008-02	2017Q4	Sequoyah	Inadequate Qualification for Unit 1 Reactor Lower Compartment Cooler Motors	The team identified a Green NCV of Title 10 Code of Federal Regulations 50.49(f) Electrical Equipment Qualification when the licensee failed to perform an adequate similarity analysis for the environmental qualification of their Reliance 75 horsepower reactor lower compartment cooling fan motors. The licensee entered this issue into their corrective action program as CR1366056 and performed an operability determination, which determined the reactor lower compartment cooling fan motors were operable but non-conforming in accordance with 10 CFR 50.49. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to ensure the qualification of the reactor lower compartment cooling fan motors adversely affected their reliability and capability in the harsh environment of a design basis accident, which in turn adversely affected the reliability and capability of other environmentally qualified components that rely on the containment cooling system. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the team determined that the finding was of very low significance (Green) because it was a design deficiency that potentially affected the design or qualification of a mitigating system; however, the mitigating system maintained its operability. The team determined there was no cross-cutting aspect associated with this finding since it was not indicative of current licensee performance.
05000327/FIN-2017008-03	2017Q4	Sequoyah	Potential Inadequate Use of Thermal Aging and the Arrhenius Methodology	Introduction: The inspectors identified a URI for the licensees use of the Arrhenius methodology without consideration for the limits of extrapolation and confidence bounds for statistical uncertainties. Description: The licensee did not consider the limits of extrapolation specified for Category 1 qualification in NUREG 0588 Section 4 and IEEE 323-1974 Section 6.5, Determination of Qualification. NUREG-0588, Section 4(5) required, in part, that known material phase changes and reactions should be defined to insure that no known changes occur within the extrapolation limits, (staff position: claims that conservative extrapolation limits have been implemented must be supported). Standard IEEE 323-1974, Section 6.5, specified, in part, that the qualified life shall be based upon the known limits of extrapolation of the time dependent environmental effects if an accelerated aging test was used to determine the mathematical model. Ancillary quality standards to IEEE 323-1974 and nuclear industry EPRI reports specified that 9 extrapolating beyond the extrapolation limits could invalidate the results of the Arrhenius methodology. The ancillary standards used for qualification of the various examples specified the limits of extrapolation to be no greater than 30 oC from the test data used to determine activation energies. In addition, the licensee did not consider adequate confidence bounds to account for the statistical uncertainties present when using the Arrhenius methodology. The inspectors noted that the uncertainties grow exponentially when exceeding the extrapolation limits. The Limitorque MOV motor life line appeared to have been extrapolated from test data at 240 C to 50 C, which is 190 C from the test data. The silicone rubber cable, life line appeared to have been extrapolated from test data at 210 C to 51.67 C, which is 188.3 C from the test data. The ASCO Valves, life line appeared to have been extrapolated from test data at 266 C to 40 C, which is 226 C from the test data. The Target Rock Valves, life line appeared to have been extrapolated upward from the test data. The Westinghouse RHR Motor rewind, life line appeared to have been extrapolated from test data at 180 C to 58.6 C, which is 121 C from the test data. 275 C is 216 C from test data The ancillary quality standards used in the qualification of these examples included IEEE 98-1972, IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials; IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data; IEEE 117-1974, IEEE Standard Test Procedure for Evaluation of Systems of Insulating Materials for Random-Wound AC Electric Machinery, and other quality standards. IEEE 98-1972, Section 10, Temperature Exposures, specified, in part, that the lowest test temperature shall be chosen so that the extrapolation necessary to establish the temperature index will not be more than 25 C. IEEE 101-1972, Section 1.3, Extrapolation, specified, in part, that extrapolation of the (qualified life) line below the range of test temperatures may cause erroneous predictions if the chemical reactions controlling the insulation aging are different at lower temperatures or if other conditions affecting the aging or the mode of failure are different. Therefore, the methods outlined in this guide are applicable only if all of the assumptions behind the use of the Arrhenius equation are met (identified in references). IEEE 117-1974, Section 3.3.1, Thermal Aging, specified, in part, for any system being evaluated, tests are made for at least three different temperatures. The lowest test temperature should be no more than 25 C above the system temperature rating. The highest temperature test should be at least 40 C above lowest temperature test, and temperature points should be selected to give approximately equal temperature intervals. The average life at the highest temperature shall be no less than 100 hours. The inspectors are concerned that the licensee did not meet the aforementioned Category 1 requirements in their licensing basis. The licensee has captured these concerns in their corrective action program as CR 1366022. The inspectors need further information from the licensee and NRC technical staff to evaluate the concerns. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-03, 05000328/2017008-03, Potential Inadequate use of thermal aging and the Arrhenius methodology)
05000327/FIN-2017008-04	2017Q4	Sequoyah	Potential Inadequate Determination of Failure Modes for Qualified Life for Foxboro/Weed Instrument	Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for failure modes and the degradation leading to them in the determination of the qualified life for the Foxboro/Weed Instrument transmitters documented in the licensees EQ Binder IPT-002. Description: The inspectors reviewed EQ Binder IPT-002 for Foxboro/Weed Instrument flow transmitter qualification. In reviewing qualification test report QOAACIO, Rev. A, the inspectors identified two concerns with the qualification. a. The inspectors noted that the qualification test report identified that polysulfone had the most limiting activation energy, 0.72 eV, for the Weed instrument assembly. However, the 0.72 eV was not being used by the licensee to determine the qualified life in the Arrhenius calculations in accordance with NUREG-0588 Section 4(6) and RG 1.89, Rev. 1, C.5.c. When inspectors questioned the licensees use of a less limiting activation energy (0.78 eV for resistors), the licensee determined that the 0.72 eV was based on the degradation and failure modes associated with the tensile strength material property for polysulfone. The licensee consulted the component manufacturer, and determined that creep was the correct material property to be evaluated for end of life. The activation energy for polysulfone creep was identified as 3.81 eV. The inspectors have challenged the licensees determination that creep is the only material property that can produce a failure of the sealing function for polysulfone. The inspectors noted that the requirements in IEEE Std. 323-1974, Section 5 require, in part, that assurance be provided that any extrapolation or inference be justified by allowances for known potential failure modes (i.e. loss of sealing function) and the mechanisms leading to them (i.e. the degradation in various material properties). If the degradation associated with creep is not the only degradation mechanism that could lead to a loss of sealing function over time, inspectors question if the degradation of other material properties would have more limiting activation energies than the licensees current activation energy for the Foxboro/Weed Instrument transmitter (0.78 eV). b. The 0.78 eV activation energy used by the licensee for qualified life was derived from an academic white paper that documented experiments performed in the early space program. The white paper specified that its experimental methods were not validated. The vendor that qualified the transmitter subsequently used this experimental information to determine the qualified life of the transmitters. This activation energy appeared not to be valid in the range of service temperatures that the transmitters are expected to age in prior to a DBA. The inspectors identified that these experimental tests did not follow any identifiable quality standard. The tests were conducted as early as 1963, the white paper was published in 1968, and the inspectors could not identify any subsequent verification of these experimental methods. Although no failure modes and effects analysis was evident, the table of 11 components in the qualification appeared to identify other components that could have much more limiting activation energies that were identified by qualification, as low as 0.5 eV. The licensee has captured these concerns in their corrective action program as CR1366039, CR 1363427. The inspectors need further information from the licensee and NRC technical staff to evaluate the concerns. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-04, 05000328/2017008-04, Potential Inadequate Determination of Failure Modes for Qualified Life for Foxboro/Weed Instrument)
05000327/FIN-2017008-05	2017Q4	Sequoyah	Potential Inadequate Justification for Eliminating Preventative Maintenance for ASCO Valves	Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for eliminating the replacement of components that have a shorter life than the qualified life of the ASCO NP-1 valves assemblies. Description: The inspectors reviewed records in environmental qualification data package (EQDP), SQNEQ-SOL-005, Revision 47. After the valve manufacturer (ASCO) stopped providing rebuild kits, the licensee eliminated the replacement schedule for subcomponents that had a shorter life than the valve assembly. The licensee changed the inputs to the accelerated aging calculation and recalculated the life of these subcomponents from the approximate eight-year replacement schedule to approximately 32.5 years. The licensee changed the activation energies from 0.94eV and 0.96eV for ethylene-propylene-diene-monomer (EPDM) and Viton-A, respectively, to 1.1eV for both. Both, EPDM and Viton-A are rubber elastomers used within the ASCO valve assemblies. The licensees written justification in the EQDP referenced a review of several studies for each elastomers, which identified less limiting activation energies than the activation energies ASCO selected in their qualification test reports. Each of these studies used different material degradation mechanisms and end of life failure mechanisms to derive different activation energies. The conclusion in the EQDP, for the change justification analysis, stated in part, that these studies show that Sequoyah's original values were, in many cases, very overly conservative. The inspectors identified that the qualification of record, report AQR-67368, selected the qualification testing criteria based on the maintenance requirements (replacement schedule) specified in Appendix C of the report. The activation energies determined applicable in the ASCO test reports, (EPDM 0.94 eV and for Viton 0.96 eV) were determined by material testing and do not appear to inspectors to be very overly conservative (unrealistically low) or lacking in technical merit. In addition, the licensee did not consider the effect the various formulations for EPDM and Viton-A elastomers. The different formulations could non-conservatively affect the activation energies reviewed in their justification. The licensee subsequently replaced the accelerated aging rate used in ASCO qualification test report AQR-67368 with the more thermally severe acceleration rate used in AQS-21678/TR, ASCO Qualification Test Report, dated 7/1/1979. AQS-21678/TR does not appear to meet Category 1 requirements, yet its accelerated aging rate was used to replace the Category 1 qualification-aging rate in AQR-67368. The AQS-21678/TR report, specified, in part, that coils and elastomeric components shall be replaced every 4 years as noted in the Valve Design Specification Sheets. The thermal 12 aging in AQR-67368 simulated a minimum of 2,000 cycles (~4.5 cycles/hr.), which was more limiting than the once every 6 hours (96 cycles) specified in AQS-21678/TR. The inspector noted that the test program specified in AQS-21678/TR used IEEE 382-1972, which did not meet Category 1 qualification requirements. The test program in AQR-67368 used IEEE 382-1980, which did meet Category 1 requirements. The forward to IEEE 382-1980, stated, in part, that the testing in the report satisfies the latest issued requirements and standards, which were the NUREG-0588 Category 1 requirements issued for comment December 1979 and published in July 1981. The licensee also used a lower self-heating temperature than which was specified in ASCO letter ASCO Solenoid Valve Coil Heat Rise Data, dated 5/8/1986. The licensee determined that the ASCO heat-rise data in the above document was too conservative and used heat-rise data from the first testing done by the Franklin Research Center (FRC). Later, FRC completed NUREG/CR-5141 RV, Aging and Qualification Research on Solenoid Operated Valves, dated 4/1/1988, which was referenced in the EQDP. The NUREG specified that: Aging in forced air ovens significantly limited heat rise from self-heating. The NUREG further specified that the qualified lives of the subcomponents were significantly reduced after accounting for the differences between forced air vs less turbulent air flow and the actual temperature measurements made by ASCO vs approximate temperature measurements made by other testing. Additional difference between ASCO heat rise testing and other testing including FRCs was that ASCO drilled holes in the valves to measure the actual subcomponent temperature while others only measured externally near the subcomponents to avoid damaging the valves. This produced lower temperature readings than ASCOs. The inspectors noted that even applying the 1.1 eV currently used, the life of the elastomers appeared to be approximately 4 to 8 years, not the 32.5 years identified in the EQDP The ASCO qualification testing used nitrogen during the qualification testing. The NUREG/CR 5141 RV specified, that oxygen exposure from plant compressed air systems produced more degradation than did the nitrogen used in the ASCO qualification testing. The licensee determined that radiation margin from the qualification tests could be used to mitigate the differences between these two gases. The inspectors determined that both ASCO qualification reports, AQS-21678/TR and AQR-67368, used nitrogen instead of air, which limited the aging degradation. The inspectors question whether radiation margin can be applied to account for the difference between oxidizing gases and inert gases. In addition, the ASCO report AQR-67368, specified, that Viton elastomers significantly degraded above 18E6 test dose not 200E6 test dose used for the margin. The inspectors are concerned that the licensee failed to meet the Category 1 requirements as specified in NUREG-0588 and IEEE 323-1974. Category 1 specified proof of conservative extrapolations, and use of the most limiting activation energy. Additionally, that users of IEEE 382-1972 must meet Category 1 requirements. Furthermore, the inspectors noted that the licensee was made aware of similar deficiencies and NRC staff positions in Technical Evaluation Report (TER)-C5257-532, Implementation Guidance for New and Corrective Equipment Environmental Qualification, dated 4/22/1983. The licensee has captured these concerns in their corrective action program as CR 1366024. The inspectors need to evaluate: (1) the licensees justification for changing the activation energy; (2) the licensees assessment of how AQS-21678/TR met Category 1 requirements; (3) the adequacy of the licensees heat rise data; and (4) the licensees evaluation of elastomer degradation from oxygen vs nitrogen gas, and their use of apparent radiation dose margin to account for these differences. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-05, 05000328/2017008-05, Potential Inadequate Justification for Eliminating Preventative Maintenance for ASCO Valves)
05000327/FIN-2017008-06	2017Q4	Sequoyah	Potential Unjustified Qualified Life for ASCO Solenoid Operated Valves	Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for changing the activation energy and calculating a new qualified life for ASCO NP-1 valves assemblies. Description: The manufacturer, ASCO, conservatively established a 1.0 eV activation energy for the valve coil assemblies. The activation energy appeared to be determined by test and realistic coil failure modes. The conservative methodology used by ASCO, that used the most limiting activation energy, met the requirements in 10 CFR Part 50. By memorandum dated 8/19/2004, the nuclear utility user group for environmental qualification (NUGEQ), to which the licensee was a member, provided information supporting the use of revised activation energy values from 1.0 eV to a less limiting 1.37 eV. The memorandum (memo) specified that NUGEQ was tasked to revise the activation energy values for ASCO NP series SOVs to a less limiting one. The inspectors determined that the data and conclusions reported by NUGEQ did not appear to be justified by design control measures in accordance with 10 CFR Part 50 Appendix B Criterion III and 50.49. Adequate design control measures were specified in the Category 1 specifications established in NUREG-0588 Section 4, Aging and IEEE 323-1974 Section 6.3.3, Aging, as supplemented by RG 1.89 revision 1, Regulatory Position 5 Aging. The NUGEQ memo specified that they obtained their data through the research of information acquired from various sources. The use of the 1.37 eV value was for significantly increasing the qualified life of the ASCO coils. The inspectors are concerned that this did not meet the requirement to prove conservative extrapolations and use of the most limiting activation energies. Based on the inspectors review, NUGEQ did not demonstrate that the more limiting activation energies were unrealistic and could be discounted. The memo specified that the information NUGEQ used to derive 1.37 eV was based on emailed recollections of past DuPont testing. The DuPont email appeared to be supported by some identifiable test data, but was not quality related, was not commercial grade dedicated, and performed without any identifiable design control measures. In addition, the memo disregarded other coil components with more limiting activation energies by discounting the failure modes associated with them and the coil. The manufacturer ASCO found these discounted failure modes relevant to the coil safety functions. The inspectors are concerned that the licensee disregarded realistic, more limiting, failure modes without proper justification. The design control requirements in NUREG-0588 Section 4(5) specified, in part, that known material phase changes and reactions should be defined to insure that no known changes occur within the extrapolation limits, (staff position: claims that conservative extrapolation limits have been implemented must be supported), and Section 4(6) required, the aging acceleration rate used during qualification testing and the basis upon which the rate was established should be described and justified, (staff position: testing of the equipment should be conducted using the most limiting (lowest) activation energy of the components). Additionally, RG 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Revision 1, Regulatory Position 5.c, Section 6.3.3, Aging, of IEEE Std. 323-1974, specified, in part, that the aging acceleration rate and the basis upon which it was established be described, documented, and justified. The licensee has captured these concerns in their corrective action program as CR 1366020. The inspectors need to review the licensees analysis and justification for discounting realistic failure modes, changing the activation energy, and calculating a new qualified life for ASCO NP-1 valves assemblies. This URI is opened to determine if the performance deficiency for not providing adequate justification for changing the activation energy, is more than minor. (URI 05000327/2017008-06, 05000328/2017008-06, Potential Unjustified Qualified Life for ASCO Solenoid Operated Valves)
05000335/FIN-2010005-01	2010Q4	Saint Lucie	Failure to Identify and Correct a Condition Adverse to Quality that Resulted in the 1C-AFW Pump Being Out of Service for Greater Than Its Allowed Outage Time	A self-revealing Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to promptly identify and correct a condition adverse to quality (CAQ) that resulted in the 1C Auxiliary Feedwater (AFW) pump being inoperable for greater than its Technical Specifications (TS) allowed outage time (ACT). Specifically, in December 2009, the licensee identified a concern with housekeeping in both Unit 1 and Unit 2 AFW pump areas that could affect the pump motor, bearings, seals, and turbine controls and linkages. Then in June 2010, these same housekeeping issues combined with extended operation of the atmospheric dump valves (ADV5) caused failure of the 1 C AFW pump to reach rated speed during its scheduled surveillance test. The finding was determined to be more than minor because it is similar to Example 4.f in IMC 0612, Appendix E, in that the failure to adequately correct a CAQ affected the 1C-AFW pumps operability and affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated in accordance with IMC 0609.04, Significance Determination Process (SDP) Phase 1 screening worksheets. Because it represented an actual loss of safety function of a single train for greater than its TS ACT, SDP Phase 2 worksheets were evaluated. The phase 2 notebook produced an overly conservative result for a short exposure time (less than 2 week duration), and consequently a phase 3 SDP evaluation was performed. The resultant core damage frequency (CDF) was <1E-6 Green. The inspectors determined that the cause of this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
05000335/FIN-2010005-02	2010Q4	Saint Lucie	Failure to Take Timely and Effective Corrective Actions for ECCS Fan Damper Failures	The inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failing to take timely and effective corrective actions for Emergency Core Cooling System (ECCS) area exhaust fan damper louver failures resulting in TS Limiting Conditions for Operation (LCO) entries for an inoperable ECCS area exhaust air filter train. Specifically, multiple damper failures occurred over at least a two year period where the root cause of the failures was not identified and corrected to prevent recurrence. The finding was more than minor because it is similar to Example 4.f in IMC 0612, Appendix E, in that the failure to adequately correct a condition adverse to quality affected the 1-HVE-9A ECCS area exhaust fans operability. The finding was evaluated in accordance with IMC 0609.04, Significance Determination Process (SDP) Phase 1 screening worksheets and determined to be of very low safety significance because the finding did not represent a degradation of the radiological barrier function provided for the auxiliary building, or represent a degradation of the control room barrier function, or an actual open pathway of containment, or a reduction in function of containment hydrogen ignitors. The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not thoroughl evaluate the problem such that the resolution addressed causes, as necessary.
05000335/FIN-2010005-03	2010Q4	Saint Lucie	Licensee-Identified Violation	TS 3.0.3 requires that when a limiting condition of operation (LCO) is not met, except as provided in the associated action requirements, within 1 hour, action shall be initiated to place the unit in a mode which the specification does not apply. Contrary to this, from June 17-24, 2010, two auxiliary feed water pumps were not operable, and actions were not taken to place the unit in the required mode of operation. This was identified in the licensees CAP as condition report 2010-1 6485 and Unit 1 LER 02010-007-00. The analyst determined the finding was of very low safety significance (<1E-6) Green.
05000335/FIN-2012004-01	2012Q3	Saint Lucie	Failure to Ship Radioactive Material in Accordance with DOT Regulations	A self-revealing, Green non-cited violation (NCV) of 10 CFR 71.5 was identified for the licensees failure to ship radioactive material in accordance with Department of Transportation (DOT) requirements as specified in 49 CFR Parts 171-180. Specifically, upon receipt at its destination, a radioactive shipment classified as an excepted package for limited quantities was found to have external surface package dose rates exceeding the limit of 0.5 millirem per hour (mrem/h) as specified in 49 CFR 173.421)(a)(2). The package recipient identified a maximum dose rate of 3.95 mrem/h on the exterior surface of the package and notified the licensee of the discrepancy. The licensee entered the event into their corrective action program as Action Request (AR)- 01628106. The performance deficiency was more than minor because it was associated with the Program & Process Procedures attribute (DOT package limits) of the Public Radiation Safety Cornerstone. The inspectors determined the cornerstones objective was adversely affected based on the fact that shipment of radioactive material in excess of DOT limits in the public domain is contrary to NRC and DOT regulations. Assurance that the public will not receive unnecessary dose is decreased if packages are not prepared so that dose rates in accessible areas remain below regulatory limits during transit. The finding is of very low safety significance (Green) because there was little to no risk to members of the public. This finding involved the cross-cutting area of Human Performance with the aspect of conservative decision-making, in that the licensee assumptions failed to ensure that equipment packaged for shipment would not exceed DOT limits during transport.

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