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	Q	Significance	CCA	Identified by	Title	Description
05000461/FIN-2018003-02	2018Q3	Severity level Minor		NRC identified	Minor Violation	Title 10 CFR 50, Appendix B,Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established procedure CPS 8219.01, Personnel Airlock Maintenance, Revision 19, as the implementing procedure for performing maintenance on a safety-related personnel airlock, an activity affecting quality.Procedure CPS 8219.01, Section 2.1.4 states in part, Personnel Airlock Maintenance Checklist, shall be filed with completed work documents.Contrary to the above, on March 30, 2018, the licensee failed to follow Section 2.1.4 of procedure CPS 8219.01. Specifically, the licensee failed to file the personnel airlock maintenance checklist with the completed work documents. After the inspectors questioned the whereabouts of the checklist, it was discovered that it was not used when performing the repair or post maintenance test on the personnel airlock even though the procedure directs personnel to record pertinent data on the checklist during the maintenance activity. Screening: The inspectors determined the performance deficiency was minor because it was determined to be a documentation issue and the values required to be documented in the checklist were satisfactory, therefore, there was no adverse impact. The licensee documented this issue in AR 4126058, NRC ID: Documentation Deficiency Identified.Violation: This failure to comply with 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.Licensee Event Report 05000461/201800100 is closed.
05000461/FIN-2018003-01	2018Q3	Green	H.3	NRC identified	Failure to Revise an Operability Evaluation When No Longer Meeting a Compensatory Measure	The inspectors identified a Green finding for the failure to revise an operability evaluation when no longer meeting a compensatory measure, in accordance with OPAA115, Operability Determinations, Revision 21. Specifically, the licensee failed to revise the operability evaluation documented in EC 387664 when no longer maintaining the Division 1 and Division 2 safety-related buses in a split bus configuration from November 2017 through June 2018.
05000461/FIN-2018412-01	2018Q3	Green	H.2	NRC identified	Security
05000461/FIN-2018002-01	2018Q2	Green	H.6	NRC identified	Failure to Perform an Operability Determination for Suspected Leakage Past Shutdown Service Water Isolation Valves	The inspectors identified a Green finding for the failure to perform an operability determination in accordance with Procedure OPAA108115, Operability Determinations (CM1). Specifically, the licensee failed to determine and document the operability status of the shutdown service water system and the ultimate heat sink after the discovery of leakage past the 1CC075A and 1CC076A isolation valves.
05000461/FIN-2018002-02	2018Q2	Green		NRC identified	Failure to Establish Adequate Leak Rate Test Procedures for Shutdown Service Water Isolation Valve Testing	The inspectors identified a Green finding and a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure testing of the shutdown service water (SX) isolation valves was performed with procedures which: (1) incorporated the requirements and acceptance limits contained in applicable design documents; and (2) included provisions for assuring that all prerequisites for the given test had been met. Specifically, the licensee failed to establish leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions for ensuring the required differential test pressure was met during testing.
05000461/FIN-2018050-01	2018Q2	TBD	H.2	Self-revealing	Failure to Follow Multiple Procedure	On May 17, 2018, a To-Be-Determined (TBD) finding and an associated Apparent Violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and Technical Specification 3.8.2, Condition B.3,were self-revealed for the licensees failure to follow multiple procedures that affected quality.This resulted in the unavailability and inoperability of the Division 2 Emergency Diesel Generator when it was relied upon for plant safety
05000461/FIN-2018050-02	2018Q2	Green	H.12	Self-revealing	Failure to Promptly Identifya Condition Adverse to Quality	On May 17, 2018,a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were self-revealed for the licensees failure to promptly identify that the safety-related Division 2 EDG had its starting air receivers isolated, which was a condition adverse to quality that rendered the EDG inoperable and unavailable
05000461/FIN-2018050-03	2018Q2	Green	H.6	Self-revealing	Equipment Operator Rounds Points Inadequate Acceptance Criteria	On May 17, 2018,a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self-revealed for the licensees failure to include appropriate quantitative acceptance criteria for the Division 2 EDG parameters to ensure the Division 2 EDG could perform its safety function
05000461/FIN-2018002-03	2018Q2	Severity level Minor		NRC identified	Minor Violation	During the inspection quarter, the inspectors reviewed a significant number of licensee CAP documents to assess the following performance attributes: complete, accurate, and timely documentation of the identified problem in the CAP; evaluation and timely disposition of operability and reportability issues; consideration of extent of condition and cause, generic implications, common cause, and previous occurrences; classification and prioritization of the problems resolution commensurate with the safety significance; and identification of negative trends associated with human or equipment performance that can potentially impact nuclear safety. Minor Performance Deficiency: The inspectors determined that issues which could impact the operability of TS-related equipment were generally entered into the CAP in a timely manner. However, operability determinations were not always performed within the timeframes established in Section 4.1 of Procedure OPAA108115, Operability Determinations (CM1), because some issue reports were not directly routed to the operating shift crew for review. The CAP software program used by the licensee included a standard set of questions which were normally answered by the individual entering the issue into the CAP. Depending on the answers to the questions, the CAP document routing could automatically bypass the operating shift crew for review. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The inspectors did not identify any instance where the failure to perform a timely operability determination had a significant consequence on licensed activities. However, the inspectors discussed the vulnerability between the CAP and the operability determination process with the licensee. The licensee implemented a standing order to require a shift review by the operating crew of condition reports not directly routed to the shift. In addition, the licensee is trending the number of condition reports which are returned by the Station Ownership Committee to the shift for review to determine whether further actions are warranted. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000461/FIN-2018002-04	2018Q2	Severity level Minor		NRC identified	Minor Violation	The inspectors reviewed AR 4082490, Reactor SCRAM from Trip of 1AP07EJ. The inspectors selected this sample for review due to the safety significance of the Division 1 and 2 safety-related transformers, which is the subject of the AR. This review focused on actions associated with newly installed Divisions 1 and 2 4160V to 480V transformers. As appropriate, the inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition report and other related condition reports: classification and prioritization of the resolution of the problem commensurate with safety significance; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. As a result of this review the inspectors identified the following minor violation: Minor Violation: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow procedures associated with the CAP. Specifically, on May 10, 2018, the licensee identified discrepant results while testing safety-related transformers 0AP06E2 and 1AP12E2 but failed to enter this issue into the CAP in accordance with PIAA120, Issue Identification and Screening Process, Revision 8, Step 4.3.4, until prompted by the inspectors. Instead, the licensee evaluated the discrepant results within the work order and found them to be acceptable. The licensee generated AR 4137994, Insulation Power Factor Results For 0AP06E & 1AP12E, dated May 15, 2018, after being challenged by the inspectors regarding the need to enter the discrepant test results into the CAP. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The failure to document the discrepant values in the CAP did not adversely impact the safety-related transformers. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion II, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000461/FIN-2018002-05	2018Q2	Severity level Minor		NRC identified	Minor Violation	The inspectors reviewed AR 4116223, Blown Fuses during CPS 9080.23 8.4 for Fast Transfers. The inspectors selected this sample for review due to repetitive fuse failures within the safety-related Division 3 NUS Modules dating back to 2013. As appropriate, the inspectors verified the following attributes during their review: complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery; consideration of the extent of condition, generic implications, common cause, and previous occurrences; evaluation and disposition of operability/functionality/reportability issues; classification and prioritization of the resolution of the problem commensurate with safety significance; identification of corrective actions, which were appropriately focused to correct the problem; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. Description: While reviewing the historical ARs associated with the NUS fuse failures, the inspectors discovered licensee information indicating the NUS fuse failures were likely caused by voltage/current transients within the upstream, safety-related 480V to 120V regulating transformer. The purpose of the transformer was to regulate voltage and current to the downstream components including the NUS modules. However, degradation in the transformers ability to regulate voltage and current levels could create a condition where the voltage and current levels exceeded the NUS fuse rating causing fuse failure. The licensee documented the potential transformer degradation issue on September 20, 2013, in AR 1561455, Division 3, Group 1 Instruments Found De-energized during CPS 9080.23, Specifically, the licensee stated, The most probable cause of the failure of the NUS modules was the transient voltage overshoot of the regulating transformer causing the transient protection varistors on the five NUS modules to actuate, drawing a near fault current until the individual and line feed fuses blew. Station procedure PI-AA-125, Corrective Action Program, defined equipment failure as, damage to or degradation of a system, structure or component that may cause or contribute to the event. Based on the information documented in AR 1561455, the licensee identified transient voltage overshoots in the 480V to 120V regulating transformer, which was a degraded condition causing the NUS modules to fail. Per the licensee definition this would constitute an equipment failure. No further action was taken to identify and correct the regulating transformer degradation until the transformer failed on March 18, 2018, impacting multiple pieces of safety-related Division 3 equipment. Minor Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to this requirement, on September 20, 2013, the licensee identified a failure of the 480V to 120V regulating transformer, which manifested itself as a voltage overshoot causing the failure of the NUS modules, but failed to take actions to correct the condition. On March 18, 2018, the regulating transformer subsequently degraded further causing it to fail in a manner that tripped the upstream breaker and impacted additional pieces of safety-related Division 3 equipment. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. Specifically, the inspectors determined that although the transformer failure affected Division 3 equipment, the failure would not have impacted the Division 3 equipments ability to respond to a DBE or the capability to shut down the reactor and maintain it in a safe shutdown condition. Enforcement: The failure to comply with 10 CFR 50, Appendix B, Criterion XVI, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000461/FIN-2018001-02	2018Q1	Green		Self-revealing	Failure to Identify a Single Point Vulnerability Results in Manual Reactor Scram	A self-revealed Green finding was identified for the licensees failure to identify a single point vulnerability in accordance with procedure ERAA2004, Revision 1. Specifically, during a site single point vulnerability review of the feedwater system, the licensee failed to identify a single point vulnerability that subsequently resulted in a loss of a feedwater heating string. The loss of the heater string caused a drop in temperature in the reactor of 100 degrees which prompted a manual scrambe initiated by the operators
05000461/FIN-2018001-01	2018Q1	Green		Self-revealing	Failure to Follow Procedure Results in Unplanned Reactor Core Isolation Cooling Unavailability	A self-revealed Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified when the licensee failed to follow station procedure Clinton Power Station (CPS) 9030.01C034, RCIC (Reactor Core Isolation Cooling) Steam Line Flow E31N683A(B), E31N684A(B), Checklist. Specifically, the licensee failed to reset the isolation logic for the RCIC steam line outboard isolation valve prior to turning on the breaker for this valve. This resulted in the isolation of the steam supply to RCIC causing RCIC to become unavailable,and elevating the plant risk to Yellow.
05000461/FIN-2017012-01	2017Q4	Green	H.5	NRC identified	Failure to Perform a Corrective Action to Prevent Recurrence	The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective action to preclude repetition (CAPR) of a significant condition adverse to quality. Specifically, CAPRs developed following a December 8, 2013, 480 Volt transformer failure were not completed on Division 2 equipment even though the licensee recognized the 2013 transformer failure as a significant condition adverse to quality. The licensee entered this issue into their corrective action program (CAP) as action request (AR) 04089480. As corrective actions, the licensee planned to perform the testing, which made up the corrective action to prevent recurrence, at the next available opportunity which will be the 2018 refueling outage. This performance deficiency was determined to be more than minor because it adversely impacted the Equipment Reliability attribute and the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Failure to perform the CAPR commensurate with safety reduced the effectiveness of the CAPR and increased the likelihood of a recurring event. This finding was determined to be of very low safety significance because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event and did not affect mitigation equipment. This finding affected the cross-cutting area of human performance, in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Delaying the performance of the testing because it extended the outage did not demonstrate that nuclear safety was the overriding priority. (H.5)
05000461/FIN-2017012-02	2017Q4	Green	H.13	NRC identified	Failure to Properly Classify Non-Safety Related Auxiliary Transformers as Operationally Critical Components	The inspectors identified a finding of very low safety significance for the licensees failure to follow procedure ERAA2001001, Equipment Classification, Revision 3. Specifically, three non-safety related 4160 volt to 480 volt transformers were not properly classified as operationally critical components. The licensee entered this issue into its CAP as AR 04086449. As corrective actions, the licensee corrected the criticality classifications 3 for 0AP44E 480 VAC Auxiliary Transformer D, 0AP92E 480 VAC Auxiliary Transformer P, and 1AP18E2 480 VAC Auxiliary Transformer 1H. Additionally, the licensee planned to perform a work group evaluation to document the extent of condition to ensure that all dry type transformers onsite have the correct criticality classification. The performance deficiency was determined to be more than minor, because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance of the transformers listed above was not fully evaluated as required by the preventive maintenance program to ensure the likelihood of failure was limited. The inspectors determined this finding was of very low safety significance because although the performance deficiency resulted in a preventive maintenance strategy that may have resulted in lower reliability of the respective 480 volt auxiliary transformers, it would not have resulted in the loss of mitigation equipment relied upon to transition the plant from the onset of the scram to a stable shutdown condition. The inspectors determined this finding affected the cross-cutting area of Human Performance in the aspect of Consistent Process where individuals use a consistent systematic approach to make decisions and risk insights are incorporated as appropriate. Specifically, the licensee failed to use a consistent classification process to reach the conclusion that the 480 VAC auxiliary transformers 0AP44E, 0AP92E, 1AP18E2 were properly classified as operationally critical components. (H.13)
05000461/FIN-2017012-03	2017Q4			NRC identified	Evaluation of RCE 04082490, Reactor Scram from Trip of 1AP07EJ)	Inspection Scope The inspectors held discussions with licensee personnel, reviewed the response of equipment and operations personnel, and reviewed historical corrective action program and maintenance related documents to evaluate whether a higher level of NRC response was needed to review this event. b. Discussion The inspectors did not identify any circumstances of the event that warranted escalation of the inspection to an Augmented Inspection Team. The event itself followed the anticipated sequence according to accident analysis and with a few non-consequential exceptions, plant equipment functioned as designed. While performing the preliminary risk analysis for the MD 8.3 Evaluation to determine the risk criteria, the Senior Reactor Analyst modeled the transient as a Loss of Condenser Heat Sink initiating event due to the manual reactor scram and closure of the inboard MSIVs. Direction to use the steam line drains to maintain the condenser as a heat sink when the MSIVs are closed was contained in site procedures. Procedure CPS (Clinton Power Station) EOP1; RPV Control, listed MSL drains as one of the systems to be used to control RPV pressure and cooldown rate. Procedure CPS 4100.01; Reactor Scram, directed the operator to use an appropriate cooldown method listed in CPS 9000.06, Unit Shutdown. In CPS 9000.06 Section 8.8, Cooldown With Main Condenser, MSL drain valves were one method listed and included a statement that it was OK to shut MSIVs when using this method. In this scenario, the control room supervisor stated that he considered using RCIC for pressure control, but determined that he did not need to because the 13 main condenser remained available and he was able to control the pressure/cooldown rate using the MSL drains to the main condenser. When the final MSIV closed and pressure started to rise, the crew started RCIC in the pressure control mode. The operating crew then continued to cooldown the reactor to Mode 4. The inspectors identified a concern that evaluation of the generic implications of the transformer failure could only be completed when the root cause of the transformer failure was known. Determination of the actual cause of the transformer failure to ground required an inspection of the damaged transformer at the ABB facility. The dry type transformer was built in 1980 and the design worst-case loading was 40 percent of the transformer rating. This type transformer was used in 29 480 VAC substations in the plant (only 5 of the 29 are safety-related). The safety-related transformers are inspected and megger tested at an 8 year frequency aligned with the safety-related bus outage schedule. The non-safety dry type transformers are inspected and megger tested at an 8 year frequency (some have been extended to 16 years based on performance). No degraded condition was found during the past preventative maintenance activities on the dry type transformers. However, operators at Clinton identified noises coming from one of the non-safety related dry type transformers in 2015. The transformer was removed from service and replaced. The transformer vendors evaluation identified degraded insulating material as the cause for the noise. Pending additional information from the inspection of the December 2017 transformer failure and the associated root cause investigation, the extent of condition and related activities were determined to be acceptable. c. Findings No findings were identified. During the review of the reactor scram and transformer failure that occurred on December 9, 2017, inspectors concluded that sufficient information was not available to identify generic implications or potential performance deficiencies with the design, manufacture or maintenance of the dry-type transformers pending completion of the licensees root cause analysis to be documented in RCE 04082490, Reactor Scram from Trip of 1AP07EJ. This issue is an unresolved item (URI) pending NRC evaluation of the additional information being developed by the licensee. (URI 05000461/201701203: Evaluation of RCE 04082490, Reactor Scram from Trip of 1AP07EJ)
05000461/FIN-2017004-03	2017Q4	Green	H.8	NRC identified	Failure to Identify the Extent of Condition for an Inadequate 10 CFR 50.59 Evaluation	The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to follow a procedure that implemented the Quality Assurance Program requirements. Specifically, the licensee failed to follow procedure PIAA1251003, Corrective Action Program Evaluation Manual, and identify the extent of condition for a lack of proficiency in applying the licensing basis when performing 10 CFR 50.59 evaluations. The licensee documented this issue in their CAP as AR 04075581. The licensee planned an Updated Safety Analysis Report (USAR) Upgrade Project which reportedly would include a review of safety evaluations for USAR changes that dated back to 1986 and determined the scope of this project would be adequate to identify the extent of condition for this issue. The inspectors determined that this issue was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, because the extent of condition review was not adequate, there is a potential for other safety systems to have been adversely affected by a lack of proficiency in applying the licensing basis during safety related system changes. As a result, safety-related systems may not be able to perform intended safety functions as defined in the USAR. This issue would also adversely affect the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was screened against all cornerstones and determined to be of very low safety significance because the finding met each of the applicable screening questions to be characterized as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence, which stated individuals follow processes, procedures, and work instructions. Specifically, when the NRC violation was documented in the CAP previously it was not appropriately classified in accordance with PIAA120 and was also incorrectly closed to an unrelated evaluation. This contributed to the failure to appropriately perform an extent of condition. (H.8)
05000461/FIN-2017011-01	2017Q4	White	P.2	Self-revealing	Failure to Correct an Identified Degraded Condition on the Division 3 Shutdown Service Water Pump	A self-revealing finding and an apparent violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, with associated violations of Technical Specification (TS) 3.7.2 and TS 3.5.1 were identified on June 15, 2017, for the licensees failure to correct a degraded condition identified during the evaluation performed as a result of the Division 3 shutdown service water (SX) pump failure in 2014. Specifically, the licensee identified corrosion of the Division 3 SX pump sleeves as a contributing cause of the 2014 pump failure and failed to appropriately evaluate and correct this issue. This resulted in the Division 3 SX pumps failure to start on June 15, 2017, and rendered the Division 3 SX pump inoperable for a time longer than its TS allowed outage time. The licensee entered this issue into the corrective action program and implemented design changes to the pump and motor assembly, including installing a new motor with higher starting torque characteristics and replacing the pump shaft sleeves and packing with parts more resistant to corrosion. The licensee has completed multiple successful runs of the new pump with no abnormalities noted. The inspectors determined that the licensees failure to correct a degraded condition identified during the evaluation performed as a result of the 2014 Division 3 SX pump failure appears to be not in accordance with the requirements of 10 CFR 50, Appendix B, Criterion XVI, and was a performance deficiency. The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that responds to initiating events. Specifically, the performance deficiency resulted in the failure of the Division 3 SX pump, which impacted the operability and functionality of the high pressure core spray system and the Division 3 emergency diesel generator. Using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, a Significance and Enforcement Review Panel preliminarily determined the finding to be of low to moderate safety significance. The inspectors determined that this finding affected the cross-cutting area of problem identification and resolution in the aspect of evaluation, where the organization thoroughly evaluates issues to ensure that resolutions address causes and extent of 3 conditions commensurate with their safety significance. Specifically, the licensee failed to properly evaluate the Division 3 SX pump sleeve corrosion rates when performing the component life evaluation, the component operability evaluation and the evaluation in response to the abnormal noises identified during periodic pump runs. (P.2)
05000461/FIN-2017004-01	2017Q4	Green	H.5	Self-revealing	Reactor Down Power due to Reactor Recirculation Pump Motor Lower Bearing Oil Leak	A self-revealed finding of very low safety significance was identified for the licensees failure to comply with the requirements of station procedure MAAA716004, Compression Fittings Inspection, Installation, Remake and Repair, Revision 3. Specifically, the licensee failed to properly assemble a joint that was part of the reactor recirculation (RR) pump motor B oil level monitoring system that subsequently leaked requiring the plant operators to perform an unplanned power reduction to allow for identification and repairs of the leak. The licensee documented this issue in the corrective action program (CAP) as Action Request (AR) 04029024. As corrective actions, the licensee made repairs to the effected joint, inspected the remaining joints to ensure proper integrity, and filled the lower bearing reservoir. This issue was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and impacted 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, the leak on the reactor recirculation pump motor oil level monitoring system would have eventually resulted in the failure of the reactor recirculation pump causing a transient and upsetting plant stability. This finding was determined to be of very low safety significance because the event did not cause a reactor trip. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the licensee failed to ensure that all personnel installing the compression fittings received an installers brief prior to performing work on the reactor recirculation pump motor oil level monitoring system. (H.5)
05000461/FIN-2017004-02	2017Q4	Green	H.5	NRC identified	Failure to Assess and Manage Risk Associated with the Performance of Control Rod Time Testing	The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, the licensee failed to assess and manage the increase in risk associated with performing control rod scram time testing in Mode 1, prior to performing the activity. As corrective actions, the licensee assessed the increase in risk for performing control rod scram time testing at power and developed a risk mitigation plan that was used to complete the testing. This performance deficiency was determined to be more than minor because the finding was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to assess risk and develop a risk mitigation plan for control rod time testing at power contributed to an automatic reactor scram. Using IMC 0609, Attachment 4, Initial Characterization of Findings, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued May 5, 2005, and Appendix M, Significance Determination Process Using Qualitative Criteria dated April 12, 2002, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance. Specifically, the inspectors and the Region III Senior Reactor Analyst (SRA) determined that Appendix K was not directly applicable to this finding because the licensee performs qualitative evaluations of maintenance activities that have the potential to cause a transient rather than quantitative evaluations. The SRA used insights from Appendix K to support a qualitative SDP evaluation using the principles of Appendix M. The SRA determined that the maintenance activity could only result in an uncomplicated reactor transient event and that the increased risk of a transient compared to the baseline risk of the plant was of very low safety significance. The SRA considered the conditional core damage probability of an uncomplicated transient in this evaluation, which was less than 1E6, to conclude that the finding was Green. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. While performing the work order risk screening for completing the control rod scram time testing while the reactor was shut down, the screener identified that a new screening would be needed if the testing was performed at power. However, no holds were placed on the work order to ensure the risk screening was completed. (H.5)
05000461/FIN-2017004-04	2017Q4	Severity level IV		NRC identified	Failure to Perform an Evaluation in Accordance with 10 CFR 72.48 for Changes Made to the Time-to-Boil Calculation	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 72.48(d)(1), Changes, Tests, and Experiments, for the licensees failure to perform a written evaluation which provides the bases for the determination that changes do not require a Certificate of Compliance amendment pursuant to 10 CFR 72.48(c)(2). Specifically, the licensee accepted Engineering Change Order ECO501825R0 (1), ECO501848R1 (1), and ECO501848R1 (4) on June 20, 2016, to the time-to-boil calculation as described in the HI-STORM FW Final Safety Analysis Report and incorrectly screened out performing an evaluation of those changes in accordance with 10 CFR 72.48. The licensee documented this issue in its CAP as AR 02714091 and AR 04081583. The licensee is performing a 10 CFR 72.48 evaluation for Engineering Change Order ECO501825R0 (1) and ECO501848R1 (4) while planning to revise the acceptance of ECO501848R1(1). The inspectors determined that the violation was of more than minor significance as the inspectors could not reasonably conclude that the above changes did not require prior NRC approval. The violation screened as a Severity Level IV non-cited violation using example 6.1.d.2 of the NRC Enforcement Policy. No cross-cutting aspect was identified since cross-cutting aspects are not assigned to traditional enforcement violations.
05000461/FIN-2017003-03	2017Q3	Green	H.13	NRC identified	Failure to Perform Engineering Evaluation to Determine the Cause of Failure of Snubbers	The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to demonstrate compliance with the requirement as prescribed in procedure ERCL330, CPS Snubber Program, Revisions 1 and 2. Specifically, the licensee failed to perform engineering evaluations to determine the cause of failure of snubbers that did not satisfy their functional testing acceptance criteria. The licensee entered this issue into their CAP as ARs 04015242 and 04041302. As corrective actions, the licensee evaluated the components affected by the failed snubber and determined that no operability issues existed. The performance deficiency was determined to be more-than-minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability for mitigating systems to respond to initiating events. Specifically, compliance with ERCL330 would ensure the failed snubber wasevaluated for the cause of failure, to ensure the licensee identified other snubbers that may have been vulnerable to the same type of deficiency. This would ensure that any potential undesired loading on the piping system could be avoided and the affected safety-related residual heat removal and reactor water cleanup piping systems could continue to perform their design function of maintaining the pressure boundary and structural integrity following a postulated design basis seismic event. The inspectors determined the finding could be evaluated using the Significance Determination Processin accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone and then Exhibit 4, External Events Screening Question. The finding screened as having very low safety significance because in each instance, the inspectors answered No to Questions 1 and 2 ofExhibit 4. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the licensee failed to establish a systematic approach to evaluating snubbers that did not meet the acceptance criteria to ensure all required aspects were addressed. (H.13)
05000461/FIN-2017009-01	2017Q3	White	H.11	Self-revealing	Failure to Evaluate Replacement Relay Dropout Voltage	Preliminary White. A self-revealed finding preliminarily determined to be of low to moderate safety significance, and an associated apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, was identified on March 9, 2017, for the licensees failure to implement measures for the selection and review for suitability of application replacement relays for the Division 1 Emergency Diesel Generator (EDG) Room Vent Fan, which were components subject to the requirements of 10 CFR Part 50, Appendix B. Specifically, Engineering Changes 330624 and 366624 failed to evaluate the change in the actual drop out voltages for replacement relays on the associated fan circuitry, and instead, introduced new relays into the circuit that resulted in the failure of the fan to operate during an under voltage condition. This rendered the Division 1 EDG inoperable for a time longer than its technical specification allowed outage time, which was a violation of Technical Specification 3.8.1, AC SourcesOperating. The licensee entered this issue into the corrective action program as action request (AR) 03982792. Corrective actions for this issue included restoring the circuit to allow the ventilation fan to operate and returning the emergency diesel generator to an operable condition. The inspectors determined that the licensees failure to verify the suitability of the replacement relays for the Division 1 EDG room vent fan was contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion III and a performance deficiency which was within the licensees ability to foresee and correct. 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 to ensure the availability, reliability, and capability of the systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify the suitability of the replacement relays prior to installation in the Division 1 EDG room vent fan circuitry resulted in the inoperability and unavailability of the Division 1 EDG from May 18, 2016 to March 11, 2017, when one of the unsuitable relays was replaced. Using IMC 0609, Appendix A, Significance Determination Process for 3 Findings At-Power, dated June 19, 2012, a Significance and Enforcement Review Panel preliminarily determined the finding to be of low to moderate safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of challenge the unknown, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, a questioning attitude was not used to understand the consequence of the differences in relay features resulting with installing a relay that was incompatible with the current design. (H.11)
05000461/FIN-2017201-01	2017Q3	Green	H.5	NRC identified	Security
05000461/FIN-2017007-01	2017Q3	Green	H.1	NRC identified	Failure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support Procedure	The inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the change to an emergency operating procedure (EOP) support procedure to incorporate three reactor core isolation cooling (RCIC) system interlock defeats and did not provide a basis for the determination that this change would not create a possibility for a malfunction of a structure, system or component (SSC ) important to safety with a different result than any previously evaluated in the updated safety analysis report. The licensee entered this issue into the CAP as action request ( AR ) 04056394 and planned to perform a screening for the procedure change. 3 This performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the change did not ensure RCIC system reliability and availability during and following design basis accidents because it introduced a new failure mode and added reliance on monitoring activities and manual actions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non- technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV significance because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The team determined that this finding had a cross -cutting aspect of Resources in the area of Human Performance because the licensee did not ensure that procedures, and other resources were available and adequate to support nuclear safety. Specifically, the procedure which required a 50.59 screening for changes to EOP support procedures, was not explicit in requiring the screening. (H.1)
05000461/FIN-2017003-02	2017Q3	Green	H.1	Self-revealing	Failure to Adequately Control Access in Locked High Radiation Area	A finding of very low safety significance and an associated NCV of TS 5.4.1 was self-revealed when individuals failed to adequately control access in locked high radiation areas (LHRAs). Specifically, the failure to meet all of the requirements of Procedure RPAA460, Attachment 5, represented a failure to comply with Radiation Work Permit CL1700518, C1R17 (Drywell) DW Bioshield Inservice Inspection Activities. This resulted in four individuals entering a LHRA that they had not been specifically authorized to enter. These individuals entered the incorrect location and were inside the area for approximately 2-3 minutes before they noticed that they were in the incorrect area. The individuals knew that they were in the incorrect location when they could not find the nozzles that they planned on inspecting. The individuals exited the area and were simultaneously told to exit the area by the radiation protection technician (RPT) providing remote coverage which demonstrated that the four workers were not in the authorized work area. Immediate corrective actions taken by the licensee included immediately suspending the work that was scheduled to take place within the bioshield associated with this job. Electronic dosimeters and dosimeters were immediately collected from the individuals that entered the area so the dose that was received could be known. The licensee also interviewed all the individuals that were involved in this bioshield entry, and the RPT that performed the brief. These interviews were conducted to understand which parts of the process associated with entry into LHRAs failed and led to this event transpiring. The licensee entered this event into their CAP as AR 04012075. As corrective actions the licensee planned to observe high radiation area and locked high radiation area briefs, for both in house and traveling RPTs. The licensee also planned to modify the bioshield as-low-as-reasonably-achievable (ALARA) plan template to label all accessible bioshield doors with elevation and azimuth.The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, because the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers entered an area that required the radiation dosimeter to be relocated to the workers knee, and the workers were wearing them on the head for the intended work location. The finding was determined to be of very-low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised.The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, 6 equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, radiation protection leadership failed to ensure that the RPT was capable of meeting the expectations for performing the LHRA briefing in accordance with station procedure RPAA460, Attachment 5. (H.1)
05000461/FIN-2017003-01	2017Q3	Green	H.6	Self-revealing	MSIV TS Leakage Limits Exceeded Due to Condition Based Maintenance Approach	The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS limiting condition for operation (LCO) 3.6.1.3, for the failure to follow station procedure ERAA200, Preventative Maintenance Program, Revision 3. Specifically, the licensee utilized a condition-based maintenance approach on the main steam isolation valves (MSIVs) that failed to monitor and trend equipment performance so that planned maintenance could be performed prior to the MSIVs exceeding the TS leakage limits. The licensee entered this issue into their CAP as AR 04009845. As corrective actions, the licensee repaired and tested the valves prior to returning the unit to the modes of applicability.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the monitoring and trending of local leak rate tests on the MSIVs did not provide performance data that would allow planned maintenance to the valves prior to the valves failing resulting in exceeding TSleakage requirements for the MSIVs. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone Reactor Containment and did represent an actual open pathway in the physical integrity of reactor containment. The inspectors proceeded to Appendix H, Containment Integrity Significance Determination Process, and determined that it was a Type B finding that was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. The inspectors used Figure 6.1, Road Map for LERF based Risk Significance for Evaluation of Type-B Findings at Full Power and determined this finding is of very low safety significance (Green). The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense-in-depth and safety related equipment. Specifically, the procedure for testing the MSIVs utilized an administrative limit that provided no margin to correct performance prior the valves becoming inoperable. (H.6)
05000461/FIN-2017003-04	2017Q3	Green	H.14	Self-revealing	Flow Control Valves Not Locked Out Results in Reactor Recirculation Pump Runback	The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the licensees failure to establish sufficient instructions in station procedure Clinton Power Station (CPS) 3103.01, Feedwater (FW), Revision 31e, for changing modes of operation for the nuclear steam supply system. Specifically, the station procedure did not provide instructions requiring the locking out the flow control valves (FCVs) to prevent a reactor recirculation FCV runback while changing the feedwater pump lineup resulting in an unexpected plant transient and 9.2 percent change in reactor power. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 04007861. As corrective actions, the licensee revised their CPS 3103.01 procedure to require that the FCVs be locked out prior to shifting reactor feed water pumps. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012,because the finding was associated with the procedure quality 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, the failure to have adequate procedures for shifting feedwater pumps during a plant shutdown on May 7, 2017, resulted in an unexpected recirculation pump run back and a 9.2 percent change in reactor power. Using IMC 0609, Attachment 4, Initial Characterization of Findings, andAppendix A, The Significance Determination Process for Findings At-Power, issuedJune 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the event did not cause a reactor scram. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the procedure provided for the option to lockout the reactor 3 recirculation flow control valves if deemed necessary during a shift of the reactor feedwater pumps and the operations crew did not make the prudent choice of locking out the valves before determining that it was safe to proceed. (H.14)
05000461/FIN-2017003-05	2017Q3	Green	H.9	Self-revealing	Failure to Establish Secondary Containment Prior to Entering MODE 2	The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS LCO 3.0.4, for the failure to follow station procedure CCAA201, Plant Barrier Control Program, Revision 11. Specifically, the licensee entered MODE 2 from MODE 4 without meeting the requirements of LCO 3.0.4 for entering a mode when an applicable LCO is not met. The licensee had not met LCO 3.6.4.1 because the doors to the B reactor water cleanup room were both opened instead of being closed to make secondary containment operable as required in MODE 2. The licensee entered this issue into their CAP as AR 04017613. As corrective actions, the licensee planned to conduct training for site personnel.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow the station procedure by not identifying that the open doors required a plant barrier impairment (PBI) permit that would have identified the doors as a constraint to entering MODE 2 resulted in the unit transitioning to MODE 2 with the secondary containment inoperable. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone and determined 5 to be of very low safety significance because the finding only represented a degradation of a radiological barrier function provided for auxiliary building. The inspectors determined that this finding affected the cross-cutting are of human performance in the aspect of training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent work force and instill nuclear safety values. Specifically, station personnel did not know the process for routing a PBI permit and did not know when a PBI permit was required. (H.9)
05000461/FIN-2017002-03	2017Q2	Green	H.12	Self-revealing	Unexpected Start of the Division 3 Emergency Diesel Generato	Green . The inspectors documented a self -revealed finding o f very low safety significance and an associated non- cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow steps in Work Order (WO) 04640788 while performing troubleshooting on blown power transformer fuses in the division 3 emergency diesel start circuitry. Specifically, the electricians opened test switches in the wrong electrical cubicle resulting in the unexpected start of the division 3 emergency diesel generator and a loss of power to the 1C1 bus from an offsite source. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR ) 04012393. As corrective actions, the licensee performed a human performance review to identify the reasons the procedure was not followed and restored power to the 1C1 safety bus . The performance deficiency was determined to be more than minor because it impacted the Initiating Event s cornerstone at tribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure of the electrical maintenance technicians to follow their procedures resulted in a loss of power to the 1C1 electrical bus. T he finding was screened against the Initiating Event s cornerstone and determined to be of very low safety significance because the loss of power to the 1C1 bus occurred while Clinton was in a refueling outage when the high pressure core spray system was removed from service and not being relied upon for shutdown safety defense in depth. The loss of the 1C1 bus did not affect decay heat removal from the core, did not affect reactor coolant inventory, and the event occurred while the refuel cavity was flooded up for refueling operations. The inspectors determined that this finding affected the cross -cutting area of human performance in the aspect of avoid complacency where individuals implement 3 appropriate error reduction tools. Specifically, as documented in the licensees human performance review, the electricians performing the work did not utilize any human performance tools to flag the equipment to be operated and improperly performed the concurrent verification of the component to be manipulated. (H.12)
05000461/FIN-2017002-01	2017Q2	Green		NRC identified	Failure of Operators to Meet Time Critical Operator Actions	Green . The inspectors identified a finding of very low safety significance and an associated non -cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that applicable regulatory requirements and the design basis was correctly translated into specifications, drawings, procedures, and instructions and that design control measures provided for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation al methods, or by the performance of a suitable testing program . Specifically, the licensee failed to assure/validate operators were able to complete the standby liquid control time critical action for an anticipated transient without a scram specified in their licensing documents. The licensee entered this issue into their CAP as AR 03980202. As corrective actions, the licensee determined the scram choreography required to complete the time critical action in the specified time, initiated a standing order to inform the operating crews, processed a procedure change for the anticipated transient without scram choreography and performed an evaluation to determine the impact of initiating the standby liquid control system at 172 seconds. The performance deficiency was determined to be more than minor because the finding was associated with the procedure quality attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the operators initiating standby liquid control at 172 seconds instead of 120 seconds, the accident analysis calculations were required to be re- performed to assure the accident analysis requirements were met. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors were able to answer all of the associated screening questions No. The inspectors determined that this finding is not indicative of current performance and therefore did not assign a cross -cutting aspect.
05000461/FIN-2017002-02	2017Q2	Green	P.3	NRC identified	Failure to Perform Adequate Evaluation of Crane Rail Clips	Green . The inspectors identified a finding of very -low safety significance and an associated cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly verify the adequacy of design of the fuel building crane and crane support structure elements. Specifically, calculations involving the crane rail clips and clip bolts had multiple technical errors and failed to adequately demonstrate that the design met the design basis requirements. The licensee initiated corrective actions by documenting the deficiency in A R 4001089 and performed an evaluation demonstrating that the functionality of the crane was maintained. The finding was determined to be more -than -minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of maintaining the functionality of the spent fuel pool (SFP) cooling system. Specifically, crane rail clip bolts were required to ensure structural integrity of structures, systems, and components described in the Updated Safety Analysis Report, 5 when subjected to design loads as part of safe load handling of heavy loads near the SFP and to ensure integrity of the spent fuel cask. In accordance with IMC 0609, Significance Determination Process , Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Barrier Integrity cornerstone because it was associated with SFP/fuel handling activities . Based on answering No to questions A through F in Table 3, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 3, for the Barrier Integrity cornerstone screening questions. Based on the crane remaining functional, the inspectors answered No to Questions D.1 through D.4 because the finding did not adversely affect decay heat removal capabilities, did not result from fuel handling errors, did not result in loss of SFP inventory, and did not affect the SFP neutron absorber or fuel bundle misplacement ; therefore , the finding screened as having very -low safety significance. The finding was cross- cutting in the resolution aspect of the problem identification and resolution area because the licensee failed to take effective corrective actions in a timely manner to address issues identified earlier in the rail clip evaluations. (P.3)
05000461/FIN-2017002-06	2017Q2	Green		NRC identified	Failure to Perform Preventive Maintenance on a Safety - Related Breaker Cubicl	Green . The inspectors identified a finding of very low safety significance for the licensees failure to perform maintenance on a safety -related motor control center cubicle. Specifically, the licensee failed to perform thermography on the division 1 shutdown service water pump room cooler breaker cubicle in accordance with the maintenance strategy/template without providing justification for differing from the template as required by MA AA 716 210, Performance Centered Maintenance Process , Revision 3. This resulted in the division 1 shutdown service water pump room cooler fan failing because of a high resistance connection that went undetected. The licensee entered this issue into their CAP as AR 02667822. As corrective actions, the licensee replaced the thermal overload relays and created a preventative maintenance action to perform thermography on this equipment on a periodic basis. This performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that responds to initiating events. Specifically , the room cooler fan failure directly impacted the operability of the division 1 shutdown service water pump and the 4 division 1 emergency diesel generator which are safety -related, risk significant systems. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors were able to answer all of the associated screening questions No. The inspectors determined that this finding is not indicative of current plant performance and therefore did not assign a cross -cutting aspect.
05000461/FIN-2017002-05	2017Q2	Green	H.11	Self-revealing	Failure to Properly Classify a Shipment per DOT Regulations	Green . A finding of very low safety significance and an associated non -cited violation of Title 10 of CFR 71.5(a) and 49 CFR 173.421(b) was self -revealed when the licensee 6 failed to properly classify a shipment per Department of Transportation (DOT) regulations. The failure to properly classify the shipment per DOT regulations allowed the shipment to proceed in transit with dose rates that were greater than what was stated on the shipping manifest. When the discrepancy in dose rates was noticed by the receiving entity, the shipment was immediately isolated and the licensee was contacted about the survey results. The licensee then dispatched two radiation protection technicians to perform confirmatory surveys. The survey data was confirmed, and the licensee was able to determine that the misclassification of the shipment was caused by dust and debris contained inside of a dust collector shifting during transportation, which created the elevated dose rate. The site implemented immediate corrective actions which included all shipments classified as limited quantity to be approved by a senior manager in the Radiation Protection Department prior to shipping. Another immediate corrective action required that the first 4 shipments conducted by the site shipper after this event be under the direct observation of a fleet independent shipper and a senior manager in the Radiation Protection Department . The licensee entered this event into their CAP as AR 03961544. The inspectors determined that the performance deficiency was more than minor because the finding impacted the program and process attribute of the Public Radiation Safety cornerstone and adversely effected the cornerstone objective of ensuring adequate protection to public health and safety from exposure to radiation from routine civilian nuclear operations. Specifically, the misclassification of the shipment per DOT regulations could have led to individuals in the public domain being exposed to radiation dose that was greater than anticipated if conditions had been slightly altered. The finding was screened against the Public Radiation Safety cornerstone and determined to be of very low safety significance because: (1) the finding did not involve a certificate of compliance issue; (2) the failure to make emergency Notifications; (3) a lo w-level burial issue; or (4) a breach of the transportation package occurring during transit. The finding did involve a radioactive shipment above radiation limits. However, the shipment contained less than a Type A quantity of material (LSA I shipment), and dose rates were <2 millirem per hour on contact. The inspectors determined that this finding affected the cross- cutting area of human performance in the aspect of challenging the unknown, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the risk associated with the content of the dust -collector shifting during transportation and creating an area that would lead to elevated dose rates was not evaluated by Clinton Power Station radiation protection staff. (H.11)
05000461/FIN-2017002-04	2017Q2	Green	H.6	Self-revealing	Failure to Provide Sufficient Work Instructions for Performing Maintenance on the Control Room Ventilation System Charcoal Filte	Green . The inspectors documented a self -revealed finding of very low safety significance and an associated non- cited violation of 10 of CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of the licensee to provide sufficient work instructions for performing maintenance on the control room ventilation charcoal filter bed. Specifically, the work order used to change out the charcoal filter bed ( Work Order 01494189 ) contained only the minimum required amount of charcoal to place in the bed. Sometime after filling the bed April 6, 2015, the charcoal settled, resulting in the B control room ventilation system being declared inoperable after failing a surveillance test. The licensee entered this issue into their CAP as AR 03995612. As corrective actions, the licensee is revising the WO instructions and Clinton Power Station Procedure 9866.03 to require that charcoal be filled completely to the bottom of the deluge piping to allow for settling. The performance deficiency was determined to be more than minor because it impacted the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to provide sufficient guidance in the work order regarding the quantity of charcoal to be installed resulted in the B control room ventilation system failing a surveillance test and being declared inoperable. The finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance because the finding only represents a degradation of a radiological barrier function provided for the control room. The inspectors determined that this finding affected the cross -cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense in depth, and safety -related equipment. Specifically, when performing maintenance on the charcoal bed, the licensee failed to recognize that filling the charcoal to the minimum bed level provided no margin if settling occurred. (H.6)
05000461/FIN-2017002-07	2017Q2	Green	H.1	NRC identified	Root Cause Evaluation Failed to Identify Corrective Action to Preclude Repetition	Green . The inspectors identified a finding of very low safety significance and an associated non -cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement a quality assurance program procedure. Specifically, the licensee failed to document a root cause and develop a corrective action to preclude repetition for the 1A bus transformer failure in accordance with quality assurance procedure PI AA 125 1001, Root Cause Analysis Manual. The licensee entered this issue into their CAP as AR 01594407. The corrective actions in response to this issue were to revise the root cause report with a root cause of insulation degradation of the phase windings over time and develop a corrective action to prevent recurrence by using Doble testing to ensure indication of transformer insulation degradation was discovered prior to failure. The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the root cause and corrective actions to prevent recurrence were not identified until the licensee was prompted by the inspectors. As a result, additional transformer failures could have occurred. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of or cause an initiating event nor did it affect mitigation equipment. The inspectors determined this finding affected the cross -cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensees station procedure did not provide guidance on when a corrective action to preclude repetition is required, regardless of whether a risk assessment was performed. (H.1)
05000461/FIN-2017008-01	2017Q2	Green	H.5	NRC identified	Failure to Perform Required S urveillances on Multiple Fire Dampers	Green . The inspectors identified a finding of very - low safety significance (Green), and an associated Non - Cited Violation of License Condition 2.C(f ) for the licensee's failure to adequately implement surveillance procedures and work processes associated with fire barrier damper inspections. Specifically, the licensee failed to perform fire barrier damper inspections for 15 fire dampers once every 4 8 months (plus an additional 25 percent grace period) as required by the Fire Protection Program. The licensee entered the issue into their Corrective Action Program , and will inspect the fire barrier dampers during the next refueling outage. The inspectors determined that the performance deficiency was more - than - minor because the licensee's failure to inspect the fire barrier dampers could result in not identifying degraded dampers which could affect their ability to prevent a fire from spreading from one fire area to an other. The finding was of very - low safety significance because the failure to inspect the fire barrier dampers did not impact the plant's ability to reach and maintain safe - shutdown. The finding has a cross - cutting aspect in the area of Human Performance, Work Management because the licensee failed to execute a work order to inspect the fire dampers in accordance with the required frequency in P rocedure CPS 9601.01 and instead improperly extended the frequency of the fire damper inspections.
05000461/FIN-2017001-04	2017Q1	Green	H.1	NRC identified	Failure to Perform Maintenance on Residual Heat Removal Pump C Breaker in Accordance with Procedures	Green. The inspectors documented a self-revealed finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to perform maintenance on a safety related breaker in accordance with station procedure Clinton Power Station (CPS) 8410.12C001, Westinghouse DHP Circuit Breaker Checklist, Revision 7. Specifically, the licensee failed to ensure the remaining travel on the latch check switch for the RHR C pump breaker was within the acceptable range resulting in the RHR C pump failing to start. The licensee entered this issue into their corrective action program as AR 03949655. The corrective actions taken by the licensee included providing coaching to the involved individuals as well as changing the procedure to include a block to record the latch check switch over travel. The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability, and reliability of equipment that responds to initiating events. Specifically, the performance deficiency adversely impacted the operability of the RHR C pump. The inspectors reviewed the Mitigating Systems screening questions and determined a detailed risk evaluation was required because question A.3 was answered yes. The SRA performed the detailed risk evaluation and concluded the finding was of very low safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the organization failed to ensure the procedure step included a block for recording the latch check switch over travel value, which led to confusion on whether the value was required to be recorded and ultimately resulted in a failure to perform the step as written. (H.1)
05000461/FIN-2017001-03	2017Q1	Green	H.12	NRC identified	Failure to Develop and Review a Worker Tag Out	Green. The inspectors documented a self-revealed finding of very low safety significance and associated non-cited violation of Technical Specification 5.4.1, Procedures, for the licensees failure to develop and review a worker tag out in accordance with station procedure OPAA10910, Clearance and Tagging, Revision 12. Specifically, the licensee failed to identify the effect of a worker tag out on the in-service steam jet air ejector suction valve, which caused condenser vacuum to degrade resulting in the operators entering the off normal procedure for loss of condenser vacuum. The licensee entered this issue into their corrective action program as action request (AR) 03980495. As corrective actions, the operations department issued a standing order to require worker tag outs to be challenged by a second senior reactor operator. The performance deficiency was determined to be more than minor because it impacted the Initiating Events cornerstone attribute of configuration control and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to properly develop the worker tag out caused the condenser vacuum to degrade, challenging the operators to quickly diagnose the issue and take action to avoid a turbine trip. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because it did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of avoid complacency, where individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reductions tools. Specifically, the operations department failed to implement appropriate error reduction tools such as questioning attitude and thorough work product reviews to ensure the worker tag out considered all potential effects to other plant equipment. (H.12)
05000461/FIN-2017001-01	2017Q1	Green	H.14	NRC identified	Plant Barrier Control Program Failed to Compensate for an Impacted Flood Barrier (	Green. The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to implement the plant barrier control program for an impacted flood barrier. Specifically, the plant barrier impairment (PBI) permit, PBI201702003, for work on watertight door 1SD124, failed to identify the door as a flood barrier and that appropriate compensatory measures for 1SD124 being open for an extended period were identified or implemented in accordance with station procedure CCAA201, Plant Barrier Control Program, Revision 11. The licensee entered this issue into their corrective action program as AR 03980495. The corrective actions in response to this violation were to identify appropriate compensatory measures for impairment of 1SD124 and incorporate them into the PBI log. The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of protection against external events and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. With the flood barrier nonfunctional and without compensatory actions in place the residual heat removal (RHR) B and RHR C pumps were inoperable. The finding was screened against the Mitigating Systems cornerstone and the inspectors determined that the finding involved the loss or degradation of equipment or function specifically designated to mitigate a seismic, flooding or severe weather initiating event. The inspectors determined that the loss of this equipment or function by itself during the external initiating event would degrade one or more trains of a system that supports a risk significant system or function and would require a detailed risk evaluation. The senior reactor analyst (SRA) performed the detailed risk evaluation and concluded the finding was of very low safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, during preparation of the PBI permit, the station PBI log was reviewed and actions for previous work associated with the watertight door were deemed acceptable even though the work on the door in those instances was different than the work being performed this time. (H.14)
05000461/FIN-2017001-02	2017Q1	Green	H.14	NRC identified	Failed to Verify an Appropriate Alternate Method of Decay Heat Removal	Green. The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.36(c)(2)(i), Limiting conditions for operation, for failing to meet/follow the required actions for limiting condition for operation 3.9.9 and 3.4.10. Specifically, the operators failed to verify a credited alternate decay heat removal method that would satisfy the required action for the limiting condition for operation. The licensee entered this issue into their corrective action program as AR 03987440. The corrective actions in response to this violation were to identify appropriate alternate methods of decay heat removal and incorporate them into the shutdown safety management program utilized during plant outages. The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance 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, with the operators failing to identify a credited alternate method of decay heat removal and taking credit for the inoperable but in service RHR shutdown cooling train, the actual available methods that could have been credited were not verified to ensure their availability to provide the required function. The finding was screened against the Mitigating Systems Screening questions and determined to be of very low safety significance because the answer to all of the applicable screening questions was No. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the senior reactor operators at the station had historically credited inoperable RHR shutdown cooling subsystems as their own alternate decay heat remove method because they believed it was allowable without determining that it was safe in order to proceed. (H.14
05000461/FIN-2016009-06	2016Q4	Green	H.9	NRC identified	Failure to Follow the Operability Determination Process Following the Identification of a Control Room HVAC System Design Issue	The team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, for the licensee failure to follow the operability evaluation procedure after the identification of a significant design error associated with the control room HVAC system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. The licensee captured these issues into the CAP as AR 03948266 and performed a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room HVAC system remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a condition where reasonable doubt on the operability of the control room HVAC system remained following the identification of a significant design error. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue.
05000461/FIN-2016009-04	2016Q4	Green	H.4	NRC identified	Failure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator Actions	The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to verify the adequacy of design assumptions related to time critical operator actions made in calculations associated with the control room HVAC and RHR emergency SFP cooling functions. Subsequently, it was determined that operators did not fully understand the control room HVAC system operational demands and that the operational assumptions of the RHR emergency SFP cooling design were unrealistic. The licensee captured these issues into the CAP as AR 02739012, AR 03943566, and AR 02741909; reasonably demonstrated that SFP makeup sources would be available to cope with a prolonged loss of SFP cooling; conducted operator training; and provided refined procedural guidance to ensure the control room HVAC system would be operated consistent with the design assumptions. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the pilot validations of the control room HVAC system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions. Additionally, a preliminary review of procedures associated with SFP cooling and RHR determined the operational assumptions of the calculation related to RHR emergency SFP cooling were not bounding. The team determined that this finding was of very low safety significance (Green). Specifically, the control room HVAC system finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency SFP cooling did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing the control room HVAC system validation in a manner that supported nuclear safety.
05000461/FIN-2016004-01	2016Q4	Green	P.1	NRC identified	Failure to Demonstrate the Condition of Flood Seals Was Being Effectively Controlled	Green. The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50.65(a)(2), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to demonstrate that the condition of flood seals was being effectively controlled through the performance of appropriate preventive maintenance such that the seals remained capable of performing their intended function. Specifically, the licensee failed to visually inspect more than 700 flood seals per procedure ERAA450, Structures Monitoring, and procedure ERCL4501007, Clinton Surveillance Inspection Program for Seals. As corrective actions, the licensee visually inspected all accessible flood seals and evaluated the acceptability of the inaccessible seals. In addition, the licensee planned to modify ERAA450 to clarify the frequency of the flood seal inspection. The inspectors determined the licensees failure to demonstrate that the condition of flood seals was being effectively controlled through the performance of appropriate preventive maintenance was a performance deficiency. The performance deficiency was determined to be more than minor because if left uncorrected it could become a more significant safety concern. Specifically, the failure to monitor the condition of the flood seals could result in unrecognized flood seal degradation and result in seals being incapable of performing their intended function. The inspectors screened this finding against the Mitigating Systems cornerstone and concluded the finding was of very low safety significance because it did not involve the loss or degradation of equipment or a function specifically designed to mitigate a seismic, flooding or severe weather initiating event. This finding affected the cross-cutting area of Problem Identification and Resolution, in the aspect of identification, where the organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify the flood seals still had not been inspected when they performed the Maintenance Rule ProgramStructures Monitoring Assessment which credited ERCL4501007 in early 2014. (P.1)
05000461/FIN-2016009-02	2016Q4	Green	H.3	NRC identified	Failure to Scope SFP Temperature and Level Instruments into the Maintenance Rule Program	The team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)(2)(i) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 2015.
05000461/FIN-2016009-01	2016Q4	Green		NRC identified	Non Conservative Control Room Radiological Habitability Assessment	The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air-conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their CAP as AR 02742442, completed an operability evaluation, and issued an NRC event notification. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago.
05000461/FIN-2016009-03	2016Q4	Severity level IV		NRC identified	Failure to Amend the UFSAR Indicating Choice to Comply with 10 CFR 50.68(b)	The team identified a Severity Level-IV NCV of 10 CFR 50.68, Criticality Accident Requirements, Paragraph (b)(8), for the licensee failure to amend the Updated Final Safety Analysis Report (UFSAR) to indicate they chose to comply with 10 CFR 50.68(b). Specifically, in 2005, the licensee chose to comply with 10 CFR 50.68(b) but did not amend the UFSAR following the issuance of the associated license amendment. The licensee captured this issue in their CAP as AR 02741851, reasonably confirmed compliance with 10 CFR 50.68(b) requirements (1) through (7) was maintained, and initiated plans to update the UFSAR to specifically indicate that Clinton Power Station chose to comply with 10 CFR 50.68(b). The Significance Determination Process does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation, which potentially impacts the NRCs ability to regulate, using traditional enforcement to adequately deter non-compliance. Specifically, failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating License Amendment Request, and in preparation for and conduct of inspection activities. The team determined the traditional enforcement violation was a Severity Level-IV violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated UFSAR had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the Significance Determination Process. However, it had a material impact on safety or licensed activities. Specifically, the un-updated UFSAR could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000461/FIN-2016009-05	2016Q4	Green		NRC identified	Failure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQ	The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS requirements with the design capabilities. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out-of-service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross-cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago.
05000461/FIN-2016004-02	2016Q4	Severity level IV		Self-revealing	Dry Cask Storage Procedures Were Not Adequate to Ensure Correct Field Configuration	Severity Level IV: A self-revealed violation of 10 CFR 72.150, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to ensure that ISFSI procedures contained the appropriate level of detail for the circumstances such that important loading activities would be satisfactorily accomplished. Specifically, procedure HPP2226200, Revision 0, MPC Loading at Clinton, was not adequate to ensure that the Multi-Purpose Canister (MPC) was correctly oriented in the transfer cask (HI-TRAC) and procedure HPP2226300, Revision 4, MPC Sealing at Clinton, was not adequate to ensure that two thermocouples were appropriately installed during the hydrostatic test of the MPC. The licensee documented these issues in its corrective action program and took timely corrective actions. The violation was determined to be of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. Example 4e is applicable to this violation in that the MPC was incorrectly oriented in the transfer cask and then loaded with spent fuel in this incorrect configuration. Example 4b is also applicable to this violation in that unexpected leakage occurred during the hydrostatic test as a result of the failure to install the thermocouples. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000461/FIN-2016003-03	2016Q3	Green	H.13	NRC identified	Failure to Perform a 50.59 Screening for Changing the Frequency of Exercising the Turbine Bypass Valves	The inspectors identified a Severity Level IV NCV of 10 CFR 50.59 4(d)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform a written evaluation which provided the bases for determining that changing the turbine bypass valve surveillance testing frequency from every 31 days, as specified in the Updated Safety Analysis Report, to once a year did not require a license amendment. The licensee has entered this issue into their corrective action program as AR 02720163. The licensee is currently evaluating the issue in accordance with their procedure for changes to the facility. The inspectors determined that the licensees failure to perform a written evaluation to provide the basis for the determination that a change to the facility, a change to a procedure, or a change to a test or experiment did not require a license amendment was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the procedure quality 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. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. The licensee made a decision to proceed with implementation of a change to the turbine bypass valve surveillance testing frequency after a plant oversight committee review in lieu of following their consistent, systematic process for evaluating changes to the USAR. (H.13)

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