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	Quarter	Site	Title	Description
05000445/FIN-2018010-01	2018Q3	Comanche Peak	Failure to Establish Test Program to Verify Residual Heat Removal Suction Valve Capability	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to establish a test program to ensure that residual heat removal suction isolation valves would perform adequately in service.
05000445/FIN-2018010-02	2018Q3	Comanche Peak	Failure to Provide Procedural Guidance for the Failure of a Component Cooling Water Surge Tank Makeup Valve	The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide procedural guidance for the failure of a component cooling water surge tank makeup valve.
05000458/FIN-2018003-01	2018Q3	River Bend	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the design basis for those structures, systems, and components to which Appendix B applies is correctly translated into specifications, drawings, procedures, and instructions. The design basis for the control building air conditioning system, as specified in the updated safety analysis report, requires that the system be capable of performing its safety function in the event of a single failure in any component. Contrary to the above, the licensee failed to assure that the design basis was correctly translated into specifications for the control building air conditioning system. Specifically, while reviewing the control logic for the control building air conditioning system, the licensee discovered that the control logic was designed such that a single failure in a component in the control logic could have prevented the system from performing its specified safety function.
05000382/FIN-2018002-01	2018Q2	Waterford	Failure to Ensure Appropriate Chemistry Controls on the Component Cooling Water Heat Exchangers	The inspectors reviewed a self-revealed, Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred because the licensee did not prescribe procedures for preventing fouling of the component cooling water heat exchangers that were appropriate to the circumstances. Specifically, the licensee did not require in its instructions for adding biocide to the auxiliary component cooling water system that additions be coupled with running the associated auxiliary component cooling water pump or other means of ensuring that the biocide would be sufficiently circulated through the system. As a result, on February 8, 2018, component cooling water heat exchanger B failed a performance test and therefore would not maintain required design basis temperatures under all accident conditions due to biological fouling.
05000382/FIN-2018002-02	2018Q2	Waterford	10 CFR 50.59 Evaluation Associated with Emergency Feedwater Logic Modification	The licensee changed the emergency feedwater logic, as described in the Updated Final Safety Analysis Report (UFSAR), Section 7.3.1.1.6, from flow control mode to level control mode during a safety injection actuation signal. To accomplish this change, the licensee had to modify the following logic system signals and setpoints: steam generator critical level, steam generator lo level, steam generator lo-lo level, safety injection actuation, control board manual control, and the steam generator lo-lo level annunciator. The NRC team questioned whether the emergency feedwater modification required additional information to determine if the 10 CFR 50.59 evaluation was adequate, or if NRC approval was needed for the change. Specifically, the NRC team questioned if the emergency feedwater logic change: used a method of evaluation other than what was described in the UFSAR (e.g. the use of the TRANFLOW program) or would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a system important to safety. Specifically, because the emergency feedwater logic change introduced the potential to overcool the reactor, and substituted a previous automatic action for manual operator action, the NRC team questioned if the change and associated 50.59 evaluation addressed these concerns. Planned Closure Actions: The NRC and the licensee are working to gather more information related to the Final Safety Analysis Report-described methods for steam generator analyses and if the change resulted in a more-than-minimal increase in risk. Specifically, the licensee plans to provide an analysis that demonstrates the emergency feedwater logic change would not result in a more than minimal increase in the likelihood of an overcooling accident. Licensee Actions: The licensee has implemented a compensatory measure to take manual control of the emergency feedwater system during a safety injection signal such that an overcooling event will be prevented. Corrective Action References: CR-WF3-2017-06067, CR-WF3-2017-05882, CR-WF3-2017-05173
05000382/FIN-2018002-03	2018Q2	Waterford	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification 3.6.3, Containment Isolation Valves, requires, in part, that when an isolation valve for containment penetrations associated with an open system are inoperable, the licensee must restore the inoperable valve(s) to operable status within 4 hours, isolate the affected penetration within 4 hours, or be in hot standby within the next 6 hours. Contrary to the above, between December 8, 2017, and December 11, 2017, with containment isolation valves inoperable, the licensee did not restore the inoperable valves to operable status within 4 hours, isolate the affected penetrations within 4 hours, or place the unit in hot standby within the next 6 hours. The licensee restored the valves to operable status on December 20, 2017, exceeding the Technical Specification 3.6.3 allowed outage time by approximately 70 hours. Significance/Severity Level: The finding was of very low safety significance (Green) because the containment isolation valves were maintained closed during the period and did not represent an actual open pathway in the physical integrity of the reactor containment. Corrective Action Reference: CR-WF3-2018-00983
05000416/FIN-2017011-01	2018Q1	Grand Gulf	Failure to Categorize Condition Reports for Significant Conditions Adverse to Quality as Required by Procedures	The inspectors identified five examples of a finding for the licensees failure to categorize and evaluate conditions in accordance with procedural requirements. Specifically, the licensee did not categorize adverse conditions that represented the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 28. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to categorize conditions that represent the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, root cause evaluations, corrective actions to prevent recurrence, and effectiveness reviews are used per licensee Procedure EN-LI-102 to ensure availability and reliability of structures, systems, and components are maintained. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently evaluate the conditions during initial screening led to the incorrect categorization of the condition reports (H.13)
05000416/FIN-2017011-02	2018Q1	Grand Gulf	Failure to Disposition Adverse Conditions as Required by Procedures	The inspectors identified a finding for the licensees failure to disposition conditions as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 30. Specifically, the licensee did not identify 72 conditions as either Adverse Category B, C, or D as required by the procedure. As a result, the licensee failed to perform the required cause evaluations and develop corrective actions to address the conditions. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to disposition conditions as adverse (B, C, or D) as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, condition reports associated with deficiencies or potential deficiencies involving safety-related equipment are required to be categorized as adverse and appropriate corrective actions are assigned including causal analyses appropriate to the circumstances per licensee Procedure EN-LI-102. The inspectors performed an initial screening of the finding in accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition identified conditions as adverse led to the failure to fully evaluate the conditions (H.13).
05000416/FIN-2017011-03	2018Q1	Grand Gulf	Failure to Conduct Common Cause Failure Evaluation in Response to Inoperable Emergency Diesel Generator	The inspectors identified three instances of a non-cited violation of Technical Specification 3.8.1, AC Sources Operating, for the licensees failure to take required actions for an inoperable emergency diesel generator. Specifically, after classifying the Division I or Division II emergency diesel generator as inoperable on the basis of nonconforming conditions, and after failing to either verify that the opposite train emergency diesel generator was not inoperable due to common cause failure within 24 hours or conduct a surveillance run on the opposite train emergency diesel generator within 24 hours, the licensee failed to enter Mode 3 within 12 hours as required by Technical Specification 3.8.1, Actions B.3.1, B.3.2, and G.1, respectively. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11393. The licensee initiated corrective actions to conduct an adverse condition analysis. The failure to take required actions for an inoperable emergency diesel generator was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment reliability attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Actions B.3.1 and B.3.2 of Technical Specification 3.8.1 exist to ensure the availability, reliability, and capability of at least one emergency diesel generator in scenarios where there is a potential for a common cause failure of both emergency diesel generators, and the licensee took neither action. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of either the Division I or Division II emergency diesel generator for greater than its technical specifications allowed outage time. The finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee failed to use a consistent, systematic approach to make decisions. Specifically, the licensee failed to review the required actions of the applicable technical specification to ensure that all of those actions would be properly carried out (H.13).
05000416/FIN-2017011-04	2018Q1	Grand Gulf	Failure to Install the Residual Heat Removal Pump A Mechanical Seal in Accordance with Procedures	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures. Specifically, on September 22, 2016, maintenance did not install seal assembly drive pins in accordance with Step 7.8.2 of Procedure 07-S-14-279, Revision 007. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2017-08269 and CR-GGN-2017-11009. The licensee implemented immediate corrective actions by declaring the pump inoperable and replacing the mechanical seal. The failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, on September 22, 2016, mechanical maintenance installed the residual heat removal pump A seal drive pins backwards. As a result, the drive pins damaged the seal and on August 23, 2017, caused an unisolable leak from the seal. This resulted in unplanned inoperability and unavailability of the residual heat removal pump A from August 23, 2017, through August 25, 2017, when the mechanical seal was replaced. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, avoid complacency, because the licensee failed to recognize and plan for the possibility of mistakes, and individuals failed to implement appropriate error reduction tools. Specifically, the licensee failed to use appropriate error reductions tools such as self-check or peer checking which resulted in incorrect performance of procedural steps (H.12)
05000416/FIN-2017011-05	2018Q1	Grand Gulf	Failure to Correct Control Room Boundary Door Resulted in Loss of Safety Function	The inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to appropriately correct a condition adverse to quality. Specifically, the control room envelope door had been documented in several condition reports for not consistently working properly. Subsequent to these condition reports, on July 9, 2017, the door was opened and did not close automatically, and therefore the door was left in an unsecured position. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-06705. The licensee restored compliance by securing the door and replacing the hinge bushings to ensure the door would close properly. The failure to correct a condition adverse to quality for a control room envelope boundary door was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the structures, systems, and components and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (functionality of the control room) protect the public from radionuclide releases caused by accidents or events. Specifically, on July 9, 2017, since the licensee had not corrected the adverse conditions identified on the control room envelope door, the door was left in an unsecured position and resulted in the station declaring both trains of standby fresh air inoperable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system, and did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The period of the barrier in the open position was of short duration, approximately 1 minute, and therefore did not result in significant risk input. This finding had a cross-cutting aspect in the area of problem identification and resolution, resolution, because the licensee did not take corrective actions in a timely manner commensurate with their safety significance. Specifically, the licensee did not ensure proper priority of corrective actions on the degraded control room envelope boundary door (P.3).
05000416/FIN-2017011-06	2018Q1	Grand Gulf	Failure to Perform Functionality Assessments as Required by Procedures	The inspectors identified a finding for the licensees failure to follow Procedure EN-OP-104, Operability Determination Process, Revisions 10 through 12. Specifically, the licensee did not perform functionality assessments for adverse conditions on the offgas system as required by the procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11265. The licensee initiated corrective actions to perform functionality assessments for the conditions identified and to evaluate any potential programmatic issues. The failure to perform functionality assessments required by Procedure EN-OP-104 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to perform functionality assessments could affect the availability and reliability of the offgas system to maintain the doses associated with releases to the environment as low as reasonably achievable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it involved the Effluent Release Program, it did not impair the ability to assess dose, and did not exceed the 10 CFR Part 50, Appendix I, or 10 CFR 20.1301(d) limits. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition adverse conditions associated with the offgas system resulted in the station not performing required functionality assessments (H.13)
05000382/FIN-2018001-01	2018Q1	Waterford	Failure to Obtain NRC Staff Authorization Prior to Changing a Procedure that Impacts Implementation of Technical Specifications	The inspectors identified a Severity Level IV, non-cited violation of 10CFR50.59, Changes, Tests, and Experiments, Section (c)(1), for the licensees failure to submit and obtain authorization prior to implementation procedures described in the Final Safety Analysis Report
05000382/FIN-2017002-01	2017Q2	Waterford	Failure to Prepare the Site for Impending Adverse Weather	The inspectors identified multiple examples of a non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, for the licensees failure to follow Licensee Procedure OP-901-521, Severe Weather and Flooding, Revision 323. Specifically, on three occasions, the licensee did not close exterior doors when required by the procedure due to potential severe weather conditions. As a result, plant equipment was at an increased failure risk due to severe weather at the site. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2017-03961 and CR-WF3-2017-04944. The licensee is planning corrective actions to ensure doors do not remain blocked open during conditions that require their closure.The performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to maintain all of the doors required by Licensee Procedure OP-901-521 with all fuel offloaded to the spent fuel pool threatened the licensees ability to maintain the functionality of the spent fuel pool cooling system. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process, and determined that a qualitative analysis by a senior reactor analyst was required. The senior reactor analyst determined that the finding was of very low safety significance (Green). Using Inspection Manual Chapter 0609, Appendix M, Signifiance Determination Process Using Qualitative Criteria, the senior reactor analyst performed a bounding analysis indicated that the total increase in core damage frequency from the failure to close the doors during severe weather was less than 1E-6. The finding had a work management cross-cutting aspect in the area of human performance because the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority and the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups of job activities. Specifically, during the planning and executing of work activities associated with Refueling Outage 21, the licensee did not consider the nuclear safety implications of blocking open exterior watertight and tornado doors and the work process did not include the identification and management of the risk associated with the blocked-open doors (H.5).
05000382/FIN-2017002-02	2017Q2	Waterford	Failure to Ensure Containment Equipment Hatch Closure Prior to RCS Time to Boil	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred because the licensee did not implement instructions for maintaining containment integrity. Specifically, on April 18, 2017, the licensee did not ensure that the containment equipment hatch could be closed within the calculated reactor coolant system time to boil as required by Licensee Procedure OP-010-006, Outage Operations, Revision 330. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-02541. The licensees corrective actions included exiting the applicable condition, re-performing the equipment hatch closure drill to show the equipment hatch could be closed prior to the reactor coolant system time to boil, and performing repairs to the containment equipment hatch. The performance deficiency was more than minor because it was associated with thehuman performance attribute of the Barrier Integrity Cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee must close containment penetrations prior to the reactor coolant system time to boil in order to minimize radionuclide releases under accident conditions. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, instructed the inspectors to use Appendix H, Containment Integrity Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because licensee maintained in-depth shutdown capability and because the duration of the performance deficiency was less than 8 hours. The inspectors concluded that the finding had a teamwork cross-cutting aspect in the area of human performance because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, personnel performed work resulting in a short calculated reactor coolant system time to boil without first communicating their actions to operations or the outage control center, resulting in an unexpected plant condition (H.4).
05000382/FIN-2017002-03	2017Q2	Waterford	Failure to Ensure Appropriate Testing of TSP Baskets Inside Containment	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure that testing required to demonstrate that structures, systems, and components will perform satisfactorily while in service was identified and performed in accordance with written test procedures incorporating the requirements and acceptance limits contained in the applicable design documents. Specifically, prior to performing Licensee Procedure OP-903-027, Inspection of Containment, Attachment 10.3, Trisodium Phosphate Storage Basket Inspection, the licensee routinely performed a preliminary check to fill the trisodium phosphate storagebaskets, thereby ensuring the successful completion of the technical specification-required surveillance. As a result, following unsatisfactory preliminary checks, the trisodium phosphate storage baskets were not evaluated for past operability. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05108. The licensees corrective actions will include performing the surveillance procedure as an as-found check and evaluating failed surveillances for past operability.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, conducting preliminary checks of the trisodium phosphate storage baskets and refilling them prior to performing the technical specification surveillance can mask the as-found condition of the test and preclude an evaluation of past operability if the levels are below the technical specification-required values. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system safety function; (3) did not represent an actual loss of safety function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; (4) with the cavity flooded, it did not represent an actual loss of safety function of one or more nontechnicalspecification trains of equipment during shutdown designated as risk-significant, for greater than 24 hours; (5) did not degrade the reactor coolant system level indication and/or core exit thermal couples when the cavity was not flooded; (6) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event; (7) did not involve fire brigade training and qualification requirements, or brigade staffing; (8) did not involve the response time of the fire brigade to a fire, and; (9) did not involve fire extinguishers, fire hoses, or fire hose stations. The finding had a change management cross-cutting aspect in the area of human performance because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, when the licensee implemented the preliminary check practice in 2012, they did not evaluate the unintended consequences of how that practice would impact the results of the technicalspecification surveillance. Additionally, the licensee performed the preliminary check during each successive refueling outage between 2012 and 2017 giving the licensee an opportunity to identify the improper practice. As a result, the inspectors concluded this performance deficiency was indicative of current performance (H.3).
05000382/FIN-2017002-04	2017Q2	Waterford	Failure to Perform a Post Maintenance Test on a Main Steam Isolation Valve Solenoid Valve	The inspectors identified a non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, for the licensees failure to perform operability testing on a safety-related component. Specifically, following the coil replacement of main steam isolation valve 2 solenoid valve, a safety-related component, the licensee did not perform a retest of the solenoid valve. As a result, main steam isolation valve 2 was returned to service without the assurance that no new deficiencies had been introduced, calling into question its operability. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05507. The licensees corrective action was to perform a voltage check of the solenoid valve to ensure it would energize in the event that a main steam isolation valve 2 closure was needed.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee restored main steam isolation valve 2 to an operable status without ensuring that its solenoid valve, which is a main steam isolation valve support system, was properly retested following maintenance.The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with licensees maintenance rule program for greater than 24 hours.The finding had a conservative bias cross-cutting aspect in the area of human performance because individuals did not use decision making-practices that emphasized prudent choices over those that were simply allowable. Specifically, the licensee did not make a conservative decision when determining whether the main steam isolation valve or its solenoid valve should be tested prior to proceeding with plant startup (H.14).
05000382/FIN-2017002-05	2017Q2	Waterford	Failure to Perform Maintenance on the Correct Safety-Related Component	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred due to the licensees failure to perform field work on reactor coolant loop 2 shutdown cooling warm-up valve, SI-135A. Specifically, mechanical maintenance technicians, who were assigned work on safety injection train A, erroneously performed work on safety injection train B on reactor coolant loop 1 shutdown cooling warm-up valve, SI-135B. As a result, both trains of emergency core cooling systems were simultaneously inoperable, which placed the plant in a 1-hour technical specification shutdown action statement. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-01433. The licensees corrective actions included a revision of the model work order to require concurrent verification for component identification, and adding the valves to the protected equipment list for when the opposite train is inoperable.The performance deficiency was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the mechanics worked on valve SI-135B instead of valve SI-135A, they simultaneously made both trains of emergency core cooling systems inoperable. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, and component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with licensees maintenance rule program for greater than 24 hours.The finding had an avoid complacency cross-cutting aspect in the area of human performance because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and did notimplement appropriate error reduction tools. Specifically, maintenance technicians repeatedly visited the incorrect work location and didnt properly verify the valve number to ensure they would work on the correct component (H.12).
05000382/FIN-2017002-06	2017Q2	Waterford	Licensee-Identified Violation	Licensee Audit LO-WLO-2016-00037, Bioassay Program, dated November 21, 2016, identified that during Refueling Outage 20, staff reviewing air sample and lapel air sampler results had not been identifying positive results. The audit revealed that two positive lapel air samples from Refueling Outage 20 had not been identified nor had estimated personnel exposures been calculated. In addition, the audit identified seven positive air sample results which had no documented estimated exposures. As a result, dose was not assigned to individuals exposed to airborne radioactivity. As a result of the audit findings, the licensee retroactively assigned dose to three individuals working the October 25, 2015, cavity drain job in the amount of 36 mrem committed effective dose equivalent (CEDE) and 700 mrem committed dose equivalent (CDE) to bone surfaces and to one individual working on a November 8, 2015, decontamination job in theamount of 33 mrem CEDE and 661 mrem CDE to bone surfaces.Title 10 CFR 20.1703 states, in part, the licensee shall implement and maintain a respiratory protection program that includes: (1) air sampling sufficient to identify the potential hazard and estimate doses, and (2) surveys and bioassays, as necessary, to evaluate actual intakes.Contrary to the above, on November 21, 2016, the licensee failed to implement and maintain their respiratory protection program to include air sampling sufficient to identify the potential hazard and estimate doses, and surveys and bioassays, as necessary to evaluate actual intakes. Specifically, for two jobs and four individuals, the licensee failed to identify positive air sample results and assign internal dose to the subject individuals.In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, the inspectors determined that the performance deficiency was more than minor. The finding adversely affected the Occupational Radiation Safety Cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the failure to adequately assess internal exposure affects the licensees ability to control and limit radiation exposure to the worker. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) planning and controls; (2) a radiological overexposure; (3) a substantial potential for an exposure; or (4) a compromised ability to assess the dose.The licensees immediate corrective action was to coach all technicians on surveying airborne areas, ensure all air sample and lapel results were discussed with management, and count all air and lapel samples for alpha and beta to evaluate any potential internal radiation exposure. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2016-07300.
05000382/FIN-2017001-01	2017Q1	Waterford	Failure to Perform Field Changes in Accordance with Design Control Measures	Green . The inspectors reviewed a self -revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to perform field changes in accordance with design control measures. Specifically, following maintenance on reactor coolant pump 1B , the licensee performed unauthorized field changes by not reinstalling two design supports for the differential pressure instrument line. As a result, the instrument line developed a vibration- induced fl aw, which caused an increase in reactor coolant system unidentified leakage, and consequently , an unplanned reactor shutdown. The licensee entered this condition into their corrective action program as Condition Report CR- WF3 -2016 -06698. The licensees corrective actions included replacing the damaged instrument line and installing the missing supports. The performance deficiency was more than minor , and therefore a finding, because it affected the equipment performance attribute of the Initiating Events Cornerstone and its 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 licensees failure to reinstall the required supports on the reactor coolant pump 1B instrumentation line resulted in plant operation with increased reactor coolant system unidentified leakage, requiring an unplanned reactor shutdown to perform repairs. The inspectors screened the finding in accordance wit h NRC Inspection Manual Chapter 0609, Significance Determination Process . Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, the inspectors determined that the finding was of very low safety significance (Green) because the instrument line flaw, after a reasonable assessment of degradation, could not result in exceeding the reactor coolant system leak rate for a small loss -of-coolant accident , and could not likely affect other systems used to mitigate a loss-of-coolant accident , resulting in a total loss of their function. Because the licensees review indicated that no work had been performed in this instrument line within the last three years, and a specific date for the performance deficiency was not identified, the inspectors concluded that the finding does not reflect current licensee performance, and therefore, did not assign a cross -cutting aspect .
05000382/FIN-2016004-01	2016Q4	Waterford	Failure to Ensure Appropriate Post-Maintenance Testing on Essential Chiller B	Green. A self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, occurred because the licensee did not assure that the procedures for post-maintenance testing of activities affecting quality included appropriate quantitative or qualitative acceptance criteria for determining that maintenance activities were satisfactorily accomplished. Specifically, the licensee did not assure that post-maintenance testing of essential chiller B would identify inappropriately assembled guide vanes, following maintenance on April 11, 2016, resulting in the unexpected inoperability of essential chiller B on August 12, 2016. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2016-05155. The corrective action taken to restore compliance was to issue work instructions for post-maintenance testing of the essential chillers that ensures the guide vanes respond to load changes. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform maintenance with procedures appropriate to the circumstances resulted in the inoperability of essential chiller B. The inspectors determined the significance of the finding using NRC Inspection Manual Chapter 0609, Significance Determination Process. Using Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because all the screening questions in Exhibit 2, Mitigating Systems Screening Questions, were answered No. The finding had a cross-cutting aspect in the area of human performance, teamwork, because the licensee did not ensure that individual and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, electrical and mechanical maintenance personnel did not communicate and coordinate work to ensure that the guide vane arm and actuator linkage were assembled appropriately (H.4).
05000382/FIN-2016003-02	2016Q3	Waterford	Licensee-Identified Violation	Title 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors, section (b)(3)(v), requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) shut down the reactor and maintain it in a safe condition; (B) remove residual heat; (C) control the release of radioactive material; or (D) mitigate the consequences of an accident. Contrary to the above, on August 12, 2016, the licensee experienced a loss of the essential chilled services water safety function, which is needed to mitigate the consequences of an accident, and did not notify the NRC within 8 hours. The licensee identified this issue and entered it into their corrective action program as CR-WF3-2016-05188 and made the required notification on August 15, 2016. This violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy, revised February 4, 2015. Using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72, the issue was determined to be a Severity Level IV violation.
05000382/FIN-2016002-02	2016Q2	Waterford	Failure to Properly Assess and Manage Risk When Performing Dry Cooling Tower Maintenance	The inspectors identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, section (a)(4) because the licensee did not properly assess and manage risk associated with maintenance on the dry cooling tower fans train B. Specifically, the licensee failed to adequately assess risk and take appropriate risk management actions when replacing a logic card associated with the dry cooling tower train B fans. As a result, an electrical transient occurred that caused unexpected valve movements in component cooling water and auxiliary component cooling water train B systems, an unexpected start of the auxiliary component cooling water pump train B, and the unexpected shutdown of essential chiller train AB. The licensee entered this issue into their corrective action program as condition report CR-WF3-2016-04084. Corrective actions included reassessing the risk associated with the maintenance and identifying appropriate risk management actions to use when performing similar maintenance activities in the future. The inspectors determined that the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to take appropriate risk management actions resulted in unexpected valve movements, an unexpected start of auxiliary component cooling water pump B, and an unplanned entry into Technical Specification 3.7.4, Ultimate Heat Sink. The inspectors used Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, Flowchart 2, Assessment RMAs, and determined the need to calculate the incremental core damage probability to determine the significance of this issue. The Waterford probabilistic risk assessment model yielded an incremental core damage probability, or actual increase in risk during this work window, of 1.5x10-8. In accordance with Flowchart 2 in Appendix K, because the incremental core damage probability was less than 1x10-6, the finding screened as having very low safety significance (Green). This finding had a Procedure Adherence cross-cutting aspect in the area of Human Performance because individuals did not follow processes, procedures and work instructions. Specifically, the licensee did not assess and manage the risk associated with the maintenance in accordance with EN-WM-104, On Line Risk Assessment (H.8).
05000382/FIN-2016002-03	2016Q2	Waterford	Failure to Perform Drills Required by the Site Emergency Plan	The inspectors identified a non-cited violation of 10 CFR 50.54(q)(2), which requires a power reactor licensee to follow and maintain the effectiveness of the site emergency plan. Specifically, Waterford Steam Electric Station, Unit 3, failed to conduct two proficiency drills in calendar year 2015 as required by the Site Emergency Plan, Revision 46, Section 8.1.2.4. The licensee has initiated work tracker surveillances to ensure all drills required in 2016 are performed. The issue is more than minor because the finding was associated with the Emergency Response Organization Performance attribute and adversely affected the Emergency Preparedness cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was evaluated using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 23, 2014, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not a risk-significant planning standard function, and was not a lost or degraded planning standard function. The inspectors determined that the finding had a Work Management cross-cutting aspect in the area of Human Performance, because the emergency preparedness department did not properly schedule, oversee, and manage required activities (H.5).
05000382/FIN-2016002-01	2016Q2	Waterford	Failure to Properly Pre-Plan and Perform Maintenance on the Cable Vault and Switchgear Ventilation System	The inspectors identified a non-cited violation of Technical Specification 6.8, Procedures and Programs, associated with the licensees failure to properly pre-plan and perform maintenance on safety-related components in accordance with EN-DC-335, Preventative Maintenance Basis Template. Specifically, the licensee did not follow the required preventive maintenance basis template for the safety-related cable vault and switchgear ventilation system, and was performing vibration monitoring of these components on an 18-month frequency instead of the required 3-month frequency. As a result, the licensee was deviating from the industry standard preventive maintenance recommendations without documented technical bases, and the required preventive maintenance tasks on these safety-related components were not performed. The licensee entered this condition into their corrective action program as condition report CR-WF3-2016-02353. The licensee restored compliance by assigning the proper preventive maintenance activities for the components in this system and instituting the appropriate frequency. In addition, a maintenance scope review is being performed. The performance deficiency was more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone and its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, actions to detect, preclude and address degradation of the safety-related components were delayed. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low significance (Green) because all the screening questions in Exhibit 2 Mitigating Systems Screening Questions were answered No. The finding had an Identification cross-cutting aspect in the area of Problem Identification and Resolution because individuals did not identify issues completely, accurately, and in a timely manner in accordance with the corrective action program. Specifically, during previous vibration tests, the licensee had opportunities to identify the incorrect classification of the preventive maintenance task but did not do so (P.1).
05000382/FIN-2016002-04	2016Q2	Waterford	Failure to Account for Starting Air Design Features in Emergency Diesel Operating Procedures	A self-revealing, Green, non-cited violation of Technical Specification 6.8, Procedures and Programs, occurred because the licensee did not establish adequate procedures for the operation of the emergency diesel generators. Specifically, prior to July 7, 2015, the licensees procedure for operating the emergency diesel generators allowed lube oil pressure to be maintained low enough to activate a design feature of the starting air system that injects starting air into the diesel cylinders, which could damage the emergency diesel generator turbocharger. The licensee entered this issue into their corrective action program as condition report CR-WF3-2015-04459. The corrective action taken to restore compliance was to increase the procedure requirement for operating lube oil pressure from 35 psig to 45 psig. The inspectors concluded that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedural allowance to run the emergency diesel generator lube oil pressure at the starting air injection setpoint could have resulted in the failure of the emergency diesel generators when they were called upon to perform their safety function. The inspectors used NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, to determine the significance of the finding. The inspectors determined that the finding required a detailed risk evaluation because it represented the loss of a system or function. The detailed risk evaluation determined that the finding is of very low safety significance (Green). The senior reactor analyst estimated the increase in core damage frequency to be 4.6E-7/year and the increase in large early release frequency to be 3.9E-8/year. Dominant core damage sequences were medium break losses of coolant accidents and steam generator tube ruptures with associated losses of off-site power. Core damage was mitigated by the remaining emergency diesel generator. This finding had an Evaluation cross-cutting aspect in the area Problem Identification and Resolution, because the licensee did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensees previous evaluation performed for operating the emergency diesel generators with low lube oil pressures did not thoroughly evaluate the risk associated with the starting air system (P.2).
05000382/FIN-2016001-01	2016Q1	Waterford	Failure to Assess and Manage the Increase in Risk from Emergent Maintenance Activities	The inspectors identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, section a(4), for the licensees failure to assess and manage the increase in risk during an auxiliary component cooling water system work window. Specifically, the licensee failed to re-asses risk when a dry cooling tower fan in the component cooling water system was declared unavailable during the ongoing auxiliary component cooling water system work window. As a result, for approximately 6 hours, on-line risk was maintained as Green when it should have been elevated to Orange, which would have required additional risk management actions. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2016-0660. Corrective actions included restoring the dry cooling tower fan to available status such that risk returned to Green and sending a communication to operations supervisors to re-emphasize the requirements to adequately address unavailability of plant components. The inspectors determined that the performance deficiency was more than minor, and therefore a finding, because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by not ensuring the risk assessment was adequate when an additional component was emergently declared unavailable, the licensee proceeded with a maintenance work window with no understanding of the increased risk associated with a different plant configuration. The inspectors used Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, Flowchart 1, Assessment of Risk Deficit, and determined the need to calculate the risk deficit to determine the significance of this issue. The risk deficits were assumed to be equal to the incremental core damage probability (ICDP)actual and incremental large early release probability (ILERP)actual. The Waterford probabilistic risk assessment model yielded an incremental core damage probability (ICDP), or actual increase in risk during this work window, of 6.1x10-8. The regional senior reactor analyst evaluated the licensees risk significance evaluation and agreed with the results from the licensees model. The ILERP, screened out as not risk significant. In accordance with Flowchart 1 in Appendix K, because the ICDP was less than 1x10-6 and the ILERP was less than 1x10-7, the finding screened as having very low safety significance (Green). This finding has a procedure adherence cross-cutting aspect in the area of human performance, because individuals did not follow processes, procedures, and work instructions. Specifically, when the additional dry cooling tower fan was declared unavailable, the licensee did not re-assess risk as soon as practical as specified in site procedures (H.8).
05000382/FIN-2016001-02	2016Q1	Waterford	Licensee-Identified Violation	Technical Specification 6.8.1, states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Quality Assurance Program Requirements, Appendix A, Section 2.l, requires procedures for refueling and core alterations. Step 6.2.1 of Procedure OP-001-005, Revision 309, RCS Drain and Fill Below RCS Hot Leg Centerline, instructs the licensee to verify, in part, that Containment Purge is aligned for Refueling Ventilation with RAB Normal Ventilation, or adequate provisions or controls are in place to acceptably address radiological concerns. Contrary to the above, on November 18, 2015, the licensee failed to verify that Containment Purge was aligned for Refueling Ventilation with RAB Normal Ventilation or that adequate controls were in place to acceptably address radiological concerns. Specifically, the licensee proceeded with RCS fill without radiation protection monitoring for airborne radioactivity in the vicinity of the TRH hoses as required. The alignment for Refueling Ventilation was not completed because the required valve (CAP-201), which allows alignment between containment purge and refuel ventilation, was inoperable. The licensee indicated that this condition had existed since at least Refueling Outage 18 in 2012. This finding adversely affected the Occupational Radiation Safety cornerstone because it had the potential to cause a high airborne condition local to the refuel cavity and cause unplanned exposures. The licensees immediate corrective action was to initiate a work order to complete repairs of the inoperable CAP-201 valve. The licensee entered this issue into their corrective action program as CR-WF3-2015-08474. The significance of the finding was determined to be of very low safety significance (Green) because it was: (1) not an ALARA finding, (2) did not result in an overexposure, (3) did not involve substantial potential for an exposure, and (4) the ability to assess dose was not compromised. Licensee-identified findings do not involve cross-cutting aspects.
05000382/FIN-2015004-01	2015Q4	Waterford	Failure to Properly Pre-Plan and Perform Maintenance on the Core Element Assembly Calculators	The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.8.1.a, associated with the licensees failure to properly pre-plan and perform maintenance in accordance with EN-DC-153, Preventative Maintenance Component Classification. The licensee entered this condition into their corrective action program as condition report CR-WF3-2015-06438. The licensee restored compliance by properly classifying the components as High Critical in accordance with EN-DC-153, Revision 2, and by initiating development of appropriate preventative-maintenance for the control element assembly calculators (CEACs). In addition, the licensee initiated work to improve the reliability of the CEACs, including reviewing card refurbishments to ensure circuit card reliability is enhanced. The performance deficiency was more than minor because it is associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, inappropriate preventative maintenance on the circuit cards associated with the CEACs ultimately resulted in a plant trip on October 3, 2015. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process. Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the finding was of very low significance (Green) because the finding did not cause a trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Because the performance deficiency occurred in 2008, the inspectors concluded that the finding does not reflect current licensee performance and therefore did not assign a cross-cutting aspect.
05000382/FIN-2015003-02	2015Q3	Waterford	Failure to Follow Procedures when Changing Materials Used for Feedwater Heater Level Control Valves	The inspectors reviewed a self-revealing finding of very low safety significance that occurred because the licensee did not follow procedural guidance when changing materials used for feedwater heater level control valves. As a result, a feedwater heater normal level control valve failed unexpectedly, causing a trip of feedwater pump A and ultimately resulted in a plant trip. The licensee entered this issue into their corrective action program for resolution as condition report CR-WF3-2015-03563. The immediate action taken to restore compliance was to replace the valve internals with those of appropriate materials. The performance deficiency was more than minor because it was associated with the Design Control 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. A detailed risk evaluation determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 4E-7/year, and the finding was not significant with respect to the large early release frequency. The dominant core damage sequences included transients with the common-cause failure of the essential chilled water system and the failure of the turbine driven emergency feedwater pump. This finding has a cross-cutting aspect in the Evaluation aspect of the Problem Identification and Resolution area because the licensee did not thoroughly evaluate the causes of several previous feedwater level control valve failures.
05000382/FIN-2015003-01	2015Q3	Waterford	Failure to Establish Design Control Measures for Safety-Related Emergency Feedwater System Valves	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to verify the adequacy of the design of the emergency feedwater system. As a result, on June 3, 2015, following a manual plant trip that occurred due to a loss of the main feedwater system, the emergency feedwater back-up flow control valves oscillated so severely that control room personnel removed the system from automatic operations and manually controlled flow to the steam generators. The licensee entered this condition into their corrective action program as condition report CR-WF3-2015-03565. Long term corrective actions are to develop a modification to the system for better flow control, and complete testing that would demonstrate the automatic function of these valves. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that the safety-related emergency feedwater back-up flow control valves would perform as designed, impacted the systems ability to perform its safety function during the feedwater loss event on June 3, 2015. A bounding detailed risk evaluation determined that the finding was of very low safety significance (Green) and was not significant to the large early release frequency. The dominant sequences included losses of off-site power, failure of the backup essential feedwater valves in the closed direction, and random failures of the primary essential feedwater flow control valves in the closed direction. The primary essential feedwater flow control valves and the diversity of the emergency feedwater system helped to minimize the risk. The finding does not have a cross-cutting aspect because the most significant contributor to the performance deficiency of not identifying the design flaws or the need for a test occurred more than two years ago and did not reflect current licensee performance.
05000382/FIN-2015002-01	2015Q2	Waterford	Failure to Identify and Secure Potential Tornado-Borne Missile Hazards	The inspectors identified a non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the licensees failure to follow procedure OP-901-521, Severe Weather and Flooding, Revision 313. Specifically, on April 24, 2015, the licensee failed to assess and control potential tornado-borne missile hazards on-site as required by the procedure. The licensee entered this condition into their corrective action program as condition report CR-WF3-2015-02556. The licensee restored compliance by securing the identified hazards. This finding was more than minor because it was associated with the protection against external factors 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, in the event of a tornado at the site, the loose items could have become missiles with the potential to initiate a loss of off-site power adversely impacting safety-related equipment and personnel. The inspectors performed the initial significance determination for the finding using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Event Screening Questions, dated June 12, 2012. The finding required a detailed evaluation because it had the potential to degrade at least one train of a system that supports a risk significant system or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 1.1E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado induced losses of off-site power, and random and common cause diesel generator failures. The ability to recover the diesel generators helped to minimize the significance of the event. The finding has a Resolution cross-cutting aspect in the area of Problem Identification and Resolution, because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not take effective corrective actions to address the issue after the inspectors identified it during previous tornado watches in 2013 and 2014.
05000482/FIN-2015009-01	2015Q2	Wolf Creek	Failure to Provide Adequate Instructions for Control of Feedwater Flow in Startup Procedures	The inspectors reviewed a self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not assure the procedures for reactor startup were appropriate to the circumstances. Specifically, prior to May 3, 2015, the licensee failed to include adequate instructions for transferring feedwater flow from the main feedwater regulating valve bypass valves to the main feedwater regulating valves in Procedure GEN 00-003, Hot Standby to Minimum Load. As a result, operations personnel did not properly control feedwater flow during a reactor startup, which led to a plant trip on May 3, 2015. The licensee entered this condition into their corrective action program as Condition Reports 96064 and 100583. The corrective action taken to restore compliance was to revise Procedure GEN 00-003 to update the process for transferring main feedwater control from the main feedwater regulating valve bypass valves to the main feedwater regulating valves, including the monitoring of necessary parameters steam flow and feedwater flow. The failure to assure the procedures for reactor startup were appropriate to the circumstances was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the human performance attribute of the initiating events 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, prior to May 4, 2015, the licensee did not provide adequate guidance for the control of feedwater flow during plant startup, resulting in a plant trip on May 3, 2015. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not cause a trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, following the plant trip, all mitigation equipment responded as designed. The inspectors concluded that the finding reflected current licensee performance and had a cross-cutting aspect in the area of human performance, avoid complacency, in that the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk even while expecting successful outcomes. Specifically, the licensee did not recognize and plan for potential of mistakes when using a procedure that did not contain adequate guidance for minimizing mismatches in steam flow and feedwater flow.
05000482/FIN-2015009-02	2015Q2	Wolf Creek	Failure of the Plant Referenced Simulator to Demonstrate Expected Plant Response	The inspectors reviewed a self-revealing Green non-cited violation of 10 CFR 55.46(c)(1), Plant-referenced Simulators, due to the licensees failure to maintain a plant-referenced simulator used for the administration of the operating test such that it would demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, until June 13, 2015, the licensee failed to maintain the simulator consistent with actual plant response when using the main feed regulating valves in manual control. The licensee entered this condition into their corrective action program as Condition Report 96252. The corrective action taken to restore compliance was to change the simulator modeling of the main feedwater regulating valve controller to match the installed plant controllers. The failure to maintain the plant-referenced simulator such that it would accurately reproduce the operating characteristics of the facility was a performance deficiency. The performance deficiency is more than minor because it adversely affected the human performance attribute of the initiating events 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, prior to June 13, 2015, the licensee failed to maintain the simulator consistent with actual plant response when using the main feed regulating valves in manual control, which impacted operator control of the plant during power operations. Using NRC Inspection Manual Chapter 0609, Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), issued December 6, 2011, the inspectors determined that the finding was of very low safety significance (Green) because the deficient simulator performance did not negatively impact operator personnel performance in the actual plant during a reportable event. Specifically, after the trip occurred the operators took all appropriate required actions. The inspectors concluded that the finding did not have a cross-cutting aspect because the finding was not indicative of current performance. The configuration change that introduced the error occurred more than three years before the event. Specifically, the discrepancy between the simulator and the plant manual controller rates had existed since simulator use began in 1985.
05000382/FIN-2015002-02	2015Q2	Waterford	Failure to Follow Instructions in Painting Procedure while Painting on Safety-Related Equipment	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow procedure PMC-002-007, Maintenance and Construction Painting, while performing work on emergency diesel generator A. Specifically, while conducting painting activities in the emergency diesel generator cubicle between June 2014 and October 2014, the licensee failed to ensure that painting activities would not have an adverse impact on the moving parts and surfaces of the emergency diesel generator. Consequently, emergency diesel generator A failed to operate properly during a surveillance test on March 2, 2015. Immediate corrective actions included replacing the cylinder 7R fuel injector and fuel injection pump. The licensee restored emergency diesel generator A to operable status on March 4, 2015. The licensee entered this issue into their corrective action program as CR-WF3-2015-02626. This finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee conducted painting on and around the emergency diesel generator in such a manner that paint was inadvertently deposited and remained in a location which caused the cylinder 7R fuel metering rod to jam at the full-fuel position, which ultimately caused emergency diesel generator A to fail its surveillance test on March 2, 2015, and be declared inoperable. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that this finding was of very low safety significance (Green) because it did not represent a design or qualification deficiency, did not represent a loss of safety function for a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a Field Presence cross-cutting aspect in the area of Human Performance in that the licensee failed to provide adequate supervisory and management oversight of work activities to ensure deviations from standards and expectations were corrected promptly.
05000482/FIN-2015002-01	2015Q2	Wolf Creek	Class 1E 4kV Feeder Breakers from Station Blackout Diesel Generators Current Transformer Wiring not Installed per Design Drawings	The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, for not installing the current transformer wiring in the Class 1E 4kV alternate feeder breaker cubicles from the station blackout diesel generators per the design drawings. As a result, testing performed seven months after the system was declared operational identified that the connections were unable to power the safety-related buses due to incorrect wiring of the current transformers. The licensee entered this issue into the corrective action program as Condition Report 83379. This finding was more than minor because it was associated with the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, due to the incorrect wiring of the current transformers, the SBO diesel generators were unable to power safety related buses as they were designed. The inspectors performed the initial significance determination for the finding using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding required a detailed evaluation because it had the potential to degrade at least one train of a system that supports a risk significant system or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The finding was of very low safety significance (Green) because the risk assessment programs quantified the change in core damage frequency less than 1.0x10-6. The inspectors determined that the finding had a teamwork cross-cutting aspect in the area of human performance. The licensee individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries. Specifically a drawing revision was not properly attached to the work order which resulted in the incorrect wiring of both trains, and because different groups were completing different components, parts of the wiring were incorrectly installed per a superseded revision.
05000382/FIN-2014007-05	2015Q1	Waterford	Failure to Identify and Correct Through Wall Corrosion on Emergency Diesel Generator A and B Day Tank Vents	The team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, prior to October 22, 2014, the licensee failed to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents, a condition adverse to quality. Prior to discovery by the team, the licensee had been unaware of the corrosion, which was significant enough that a through-wall hole had formed at the base of the each vent pipe where it penetrates the roof. Consequently, any water that collects on the roof of the building would have the potential to drain into the day tanks. The licensee performed an immediate operability determination and concluded that the diesel and its support systems were operable based on no severe weather in the area. While evaluating permanent corrective actions, the licensee installed a temporary repair to the vent pipes using a rubber wrap and installed a small concrete berm to minimize the potential amount of water in the immediate area. This finding was entered in to the licensees corrective action program as CR-WF3-2014-05413. The team determined that the failure to identify and correct through-wall corrosion on the emergency diesel generator A and B day tank vents was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control and equipment performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to identify, evaluate, and correct through-wall corrosion on the emergency diesel generator A and B day tank vents. 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 to Exhibit 4, External Events Screening Questions, because it was potentially risk-significant due to seismic, flooding, or severe weather. Per Exhibit 4 this finding screened to a Detailed Risk Evaluation because if the safety function were assumed completely failed it would degrade two trains of a multi-train system and it would degrade one or more trains of a system that supports a risk-significant system. A Region IV senior reactor analyst performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk-important sequences included heavy-rain-induced losses of off-site power with the consequential failure of both emergency diesel generators. The ability to restore off-site power within four hours was important to avoid core damage. The finding was not significant to the large early release frequency. See Attachment 2, Detailed Risk Evaluation, for a detailed review of qualitative criteria also considered. This finding had a cross-cutting aspect in the area of human performance associated with procedure adherence because the licensee failed to ensure that individuals follow process, procedures, and work instructions (H.8).
05000382/FIN-2015001-01	2015Q1	Waterford	Failure to Identify and Perform Testing of Safety-Related Dry Cooling Tower Tube Bundle Isolation Valves	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not identify and perform testing for safety-related components to demonstrate that they would perform satisfactorily in service. Specifically, prior to February 12, 2015, the licensee did not identify and perform testing to demonstrate that, as described in the licensees design basis, the dry cooling tower tube bundle isolation valves could be used to isolate a dry cooling tower tube bundle following a tornado missile strike on the non-missile-protected portions of the dry cooling tower. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2015-00828. The planned corrective actions are to develop seat leakage criteria for the dry cooling tower tube bundle isolation valves and to perform periodic seat leakage testing. The inspectors determined that the performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to establish a test program for a safety related component to demonstrate that it would perform satisfactorily following a tornado missile strike could impact the systems ability to perform its safety function in the event of a tornado. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Event Screening Questions. The finding required a detailed evaluation because it would degrade one or more trains of a system that supports a risk significant system or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 2.9E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado-induced losses of offsite power, failure of the train B dry cooling tower pressure boundary, random failure of the train A component cooling water system, random failures of the emergency diesel generators, and failure to recover offsite power in 4 hours. Risk was minimized because the diesel generators have air cooled radiators and do not require component cooling water to remain functional. The low tornado frequency also minimized the risk. The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency of not identifying the need for a leak test occurred more than two years ago and did not reflect current licensee performance.
05000382/FIN-2015001-02	2015Q1	Waterford	Failure to Identify and Correct a Condition Adverse to Fire Protection	The inspectors identified a finding of very low safety significance and an associated non-cited violation of Waterford Steam Electric Station, Unit 3, License Condition 2.C.9, and the fire protection program for the licensees failure to identify and correct a condition adverse to fire protection. Specifically, the inspectors identified that the ventilation dampers that are used to maintain the environmental conditions of the No. 2 diesel fire pump room and that are needed for pump protection were damaged and not functional for an extended period of time. As a result, the reliability of the No. 2 diesel fire pump could have been impacted at high environmental temperatures. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2015-00132. The licensee manually opened the dampers and additional planned corrective actions included repairing the broken dampers linkage before the temperatures outside reach 90oF. This 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, if left uncorrected, the licensees failure to repair the damaged ventilation damper in the No. 2 diesel fire pump room would result in an ongoing degraded condition, which could have impacted the capability of the No. 2 diesel fire pump to fulfill its function of providing a water supply to the sites Fire Protection Systems. Using Inspectional Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the use of Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, was required because the finding involved fixed fire protection systems. Using Inspection Manual Chapter 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, the finding screened as Green because the reactor would have been able to reach and maintain a safe shutdown condition. Specifically, since only the No. 2 diesel fire pump was impacted by the performance deficiency, the No. 1 diesel fire pump and the motor driven pump would have been able to supply the fire systems because they are all rated for full flow capacity. This finding had a cross-cutting aspect in the area of human performance, avoid complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, licensee personnel frequently tour the fire pump house for operations and maintenance activities; however, a thorough review of the work site had not been performed (H.12).
05000382/FIN-2015001-04	2015Q1	Waterford	Inadequate Procedure for Tightening Thermal Overload Connections for Safety-Related Components	A self-revealing, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, was identified for the failure to perform maintenance that could affect the performance of safety-related equipment in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Specifically, prior to December 17, 2014, the licensee used a procedure that contained insufficient detail for tightening a thermal overload connection that resulted in a loose connection on a motor starter and eventual trip of a wet cooling tower fan, resulting in the A train of ultimate heat sink being declared inoperable. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2014-04430. The corrective action taken to restore compliance was to add additional detail to the procedure to ensure thermal overload connections are verified secure after their mechanical connections are tightened. The inspectors determined that the performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure successful tightening of the thermal overload connections for the wet cooling tower fans adversely impacted the capability of the system to perform its function. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings. The inspectors determined the finding was of very low safety significance (Green) because it affected one train for less than the allowed outage time. When the A train of ultimate heat sink was declared inoperable, the B train of ultimate heat sink was already inoperable for planned maintenance. As a result, the B train maintenance was unrelated to the performance deficiency. In addition, the finding did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. The inspectors concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency occurred more than two years ago and did not reflect current licensee performance.
05000382/FIN-2015001-03	2015Q1	Waterford	Failure to Develop the Transportation Security Plan	The inspectors identified a non-cited violation of 10 CFR 71.5, Transportation of Licensed Material, and 49 CFR 172, Subpart I, Safety and Security Plans. Specifically, licensee personnel failed to adequately develop their transportation security plan. This resulted in three Category 2 shipments being transported on public highways without security risk assessments being performed. The planned corrective actions were still being evaluated. The inspectors determined that no immediate safety concern existed because the shipments that had been made were received with no issues and the licensee had no pending Category 2 or higher shipments. The licensee documented the issue in its corrective action program as Condition Report CR-W3-2015-00506. The licensees failure to adequately develop their transportation security plan is a performance deficiency. Procedure EN-RW-106, Integrated Transportation Security Plan, did not include all the components required by 49 CFR 172.802, Components of a Security Plan. The performance deficiency is more than minor because it is associated with the program and process attribute of the Public Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined the finding has very low safety significance (Green) because Waterford had an issue involving transportation of radioactive waste, but it did not involve: (1) a radiation limit being exceeded, (2) a breach of package during transport, (3) a certificate of compliance issue, (4) a low level burial ground nonconformance, or (5) a failure to make notifications or provide emergency information. The finding has a resources cross-cutting aspect in the human performance cross-cutting area, because licensee management did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety (H.1).
05000382/FIN-2014005-01	2014Q4	Waterford	Failure to Identify and Control Potential Tornado-Borne Missile Hazards	The inspectors identified a non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for the licensees failure to follow procedure OP-901-521, Severe Weather and Flooding, Revision 312, in two separate instances. Specifically, on both November 16 and December 23, 2014, the licensee entered the offnormal procedure due to a tornado watch, but failed to assess and control potential tornadoborne missile hazards on site as required by the procedure. The licensee entered this condition into their corrective action program as condition reports CR-WF3-2014-05912 and CR-WF3-2014-06453. The immediate corrective action taken to restore compliance was to secure the identified hazards. This finding was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in the event of a tornado at the site, these loose items could have become missiles with the potential to impact safety-related site equipment and personnel. The inspectors determined the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or functions specifically designed to mitigate a seismic, flooding, or severe weather event (e.g. seismic snubbers, flooding barriers, tornado doors). The inspectors concluded that the finding had a cross-cutting aspect in the area of Human Performance, Field Presence, because the licensee did not ensure supervisory and management oversight of work activities (H.2) (Section 1R01).
05000382/FIN-2014005-02	2014Q4	Waterford	Failure to Follow the Operability Determination Process	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to assess immediate operability of safety-related systems in accordance with site procedures, in three separate instances. Specifically, on two occasions, the licensee did not properly assess operability of safety-related relays in the Engineered Safety Features Actuation Signal system, which in turn brought into question the operability of the emergency diesel generators. A third example involved the licensees failure to properly assess operability of safety-related class 3 piping in the dry cooling towers, which brought into question the operability of the component cooling water system. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-06014. The licensee restored compliance by revising the immediate operability determinations to reflect an adequate reason to justify operability of the systems in question. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to follow the Operability Determination procedure caused the licensee to incorrectly assess the capability of the systems impacted by the relays and dry cooling tower tube leak to perform their safety function and there was a reasonable doubt on the operability of the systems. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. This finding had a crosscutting aspect in the area of Human Performance, Consistent Process, because individuals did not use a consistent, systematic approach to make a decision, and risk insights were not incorporated appropriately (H.13) (Section 1R15).
05000382/FIN-2014005-03	2014Q4	Waterford	Failure to Establish an Inspection Schedule of the Dry Cooling Towers	The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.8.1.a and Regulatory Guide 1.33, Revision 2, Appendix A, for failure of the licensee to develop a preventative maintenance schedule for inspections of safety-related equipment. Specifically, the licensee did not develop a preventative maintenance schedule to visually inspect all portions of the dry cooling towers. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04930 and CR-WF3- 2014-06100. The licensee developed preventative maintenance tasks to inspect the dry cooling tower tubes, including disassembly where necessary, to restore compliance. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to inspect portions of the dry cooling towers prevented the licensee from identifying corrosion that eventually degraded the system enough to cause a leak. The inspectors determined the finding had very low safety significance (Green) because it did not affect the design or qualification of the system, did not represent the loss of a safety system or function, did not represent the loss of function of at least a single train for greater than its Technical Specification allowed outage time, and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment. The inspectors concluded that the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee did not implement a corrective action program with a low threshold for identifying issues (P.1) (Section 1R19).
05000382/FIN-2014005-04	2014Q4	Waterford	Failure to Establish Design Control Measures for the Suitability of Safety-Related Relays	The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee did not have an adequate replacement frequency for safety-related relays associated with engineered safety features equipment to ensure that all required equipment operated in the time sequence assumed by the safety analysis. The licensee entered this condition into their corrective action program as condition report CRWF3- 2013-05091. The licensee replaced the affected relays and reduced their replacement frequency from 18 years to 3 years to restore compliance. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to develop an adequate replacement frequency for the relays used to monitor for under-voltage conditions on the safety-related emergency busses could have prevented the equipment from performing its safety function. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating system component and the affected equipment maintained its operability. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, Challenging the Unknown, because the licensee did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding (H.11) (Section 1R19).
05000382/FIN-2014005-06	2014Q4	Waterford	Failure to Correct a Condition Adverse to Quality in a Timely Manner	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to correct a condition adverse to quality in a time commensurate with the safety significance of the issue. Specifically, the licensee failed to repair degraded conduit that had been identified as corroded since 2008. As a result, conduits that were housing cables for safety-related components were degraded to the point where water entered the conduit and submerged cables that were not designed for submergence for an extended period of time. The licensee entered this condition into their corrective action program as condition report CR-WF3-2014-04951. The licensee repaired the degraded conduit associated with the impacted safety-related cables to restore compliance, and also initiated an extent of condition review to identify other cables that could potentially be impacted by degraded conduits. The inspectors determined that the performance deficiency was more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, safety-related cables that were not rated for full submergence were submerged in water since at least 2008, potentially affecting the integrity of the cable and potentially impacting the safety-related equipments ability to perform their safety function in the event of an accident. The inspectors determined that the finding had very low safety significance (Green) because the finding impacted the qualification of mitigating components, but the components maintained operability. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensees decision-making practices did not emphasize prudent choices over those that are simply allowable. Specifically, when evaluating condition reports written through several years that document the degraded conduit, the licensee elected to defer needed maintenance instead of placing the adequate priority on the issue.
05000382/FIN-2014005-07	2014Q4	Waterford	Licensee-Identified Violation	Title 10 of the Code of Federal Regulations, Appendix E to Part 50, Section V, states that licensees who are authorized to operate a nuclear power facility shall submit any changes to the emergency plan or procedures to the Commission, as specified in 50.4, within 30 days of such changes. Title 10 of the Code of Federal Regulations, Section 50.54(q)(5) states, in part, that licensees shall submit a report of changes made after February 21, 2012, that includes a summary of its analysis, within 30 days after the change is put into effect. Contrary to the above, Waterford 3 Steam Electric Station did not submit changes to emergency plan implementing procedures within 30 days of such changes, and did not submit a summary of its analysis of the changes within 30 days after the changes were put into effect. Specifically, the licensee did not submit changes to the following procedures: EN-EP-305, Emergency Planning 10CFR50.54(Q) Review Program, Revision 3; EN-EP-306, Drills and Exercises, Revisions 4 and 5; EN-EP-308, Emergency Planning Critiques, Revision 2; EN-EP-310, Emergency Response Organization Notification System, Revisions 1 through 3; EN-EP-311, Emergency Response Data System (ERDS) Activation Via the Virtual Private Network (VPN), Revision 2; EN-EP-313; Offsite Dose Assessment Using the Unified RASCAL Interface, Revision 0; EN-EP-801, Emergency Response Organization, Revision 8; EN-TQ-110, Emergency Response Organization Training, Revision 7, and EN-TQ-110-01, Fleet E-Plan Training Course Summary, Revision 10. The licensee did not have a process to ensure that fleet procedures necessary to implement the site emergency plan were submitted to the NRC in accordance with the requirements of Appendix E to 10 CFR Part 50. This violation was evaluated using the NRC Enforcement Policy because the licensees failure to submit required procedures affected the NRCs ability to perform adequate regulatory oversight, and was evaluated as a Severity Level IV violation because the violation was not related to the licensees ability to perform notification or assessment during an emergency. This issue has been entered into the licensees corrective action program as Condition Reports CR-HQN-2014-00380, CR-HQN-2014-00597, and CR-WF3-2014-05727.
05000382/FIN-2014005-05	2014Q4	Waterford	Failure to Adequately Plan and Control Work Activities Related to Alloy 600 Pipe Weld Inspections to Ensure Doses were ALARA	The inspectors identified a finding associated with the licensees failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations during Refueling Outage 19. Specifically, the inspectors concluded that, had the licensee appropriately evaluated the Alloy 600 pipe weld conditions/locations during the ALARA planning process and appropriately performed in-progress ALARA reviews, they could have reasonably planned for the full scope of work and provided a better estimate and/or adequately justified revising the estimate for the job. These failures to plan and control the job activities led to unplanned, unintended collective dose. The licensee evaluated the procedures used during this work, including their process for planning and estimating doses, and documented the issue in the corrective action program. The failure to adequately plan and control work activities associated with Alloy 600 ultrasonic examinations is a performance deficiency. This performance deficiency is more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety cornerstone. It adversely affects the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, it caused the collective radiation dose for the work to be greater than 5 man-rem and exceed the planned dose estimate by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding has very low safety significance because: (1) it was associated with ALARA planning and (2) the licensees three-year rolling average collective dose of 121.7 man-rem was less than 135 man-rem. The finding has a Work Management cross-cutting aspect, associated with the Human Performance cross-cutting area, because the licensee did not adequately plan or control work activities such that nuclear safety is the overriding safety priority. Specifically, the ALARA plan did not reflect the time needed to complete the work activities, thus underestimating the dose requirements, and the administrative control of reviewing the work-in-progress at appropriate completion points failed.
05000382/FIN-2014008-01	2014Q2	Waterford	Inadequate Procedures for Securing Dry Cooling Tower Fans	A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III occurred when the licensee did not assure that design basis information was translated into specifications, drawings, procedures, and instructions. Specifically, after a failure revealed new design basis information regarding the need to place a train of dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet, the licensee failed to incorporate this information into procedures. As a result, the failure recurred. The licensee entered this condition into its corrective action program as Condition Reports CR-WF3-2012-05680 and -06908 and updated procedure OP-006-005, Inverters and Distribution, to incorporate the new design basis information into procedures. The licensee documented its failure to timely update design basis information in Condition Report CR-WF3-2014-02981. The failure to assure that design basis information was translated into specifications, drawings, procedures, and instructions as required by 10 CFR Part 50, Appendix B, Criterion III was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to incorporate design basis information regarding the need to place the dry cooling tower fan controllers to the off position prior to de-energizing the associated control cabinet into specifications, drawings, procedures, and instructions impacted the capability, availability, and reliability of both trains of dry cooling towers. Using NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because the required number of dry cooling towers in the protected train maintained their operability. This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee had not taken effective corrective actions to address an issue in a timely manner commensurate with its safety significance.
05000382/FIN-2014008-02	2014Q2	Waterford	Failure to Identify and Correct Condition Adversely Affecting Flooding Mitigation Design	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. On May 19, 2014, the team identified a significant amount of debris on the floor of one of the main steam isolation valve areas. In a probable maximum precipitation event, this debris could have prevented sufficient water removal by the floor drains to meet design basis assumptions. Following identification, the licensee entered this condition into its corrective action program as Condition Report CR-WF3-2014-03037 and removed the debris from the area. Excessive debris in the main steam isolation valve A area that could challenge the water removal capability of safety-related drain systems was a condition adverse to quality. The licensees failure to promptly identify and correct this condition adverse to quality as required by 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. The lead inspector performed the initial significance determination for performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because it involved the degradation of equipment specifically designed to mitigate a flooding event. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The bounding change to the core damage frequency was 4.7x10-8 per year (Green). The dominant core damage sequences included extremely heavy rainfall, a loss of offsite power initiating event, failure of the train B 4.16kV bus, and failure of the pressurizer safety relief valves to close. The low initiating event frequency reduced the risk significance. This finding has a resolution cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensees corrective actions from the previous non-cited violation did not fully address the issue (P.3).

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