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05000382/FIN-2018002-0130 June 2018 23:59:59WaterfordSelf-revealingFailure to Ensure Appropriate Chemistry Controls on the Component Cooling Water Heat ExchangersThe 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-2018002-0330 June 2018 23:59:59WaterfordLicensee-identifiedLicensee-Identified ViolationThis 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
05000382/FIN-2018002-0230 June 2018 23:59:59WaterfordNRC identified10 CFR 50.59 Evaluation Associated with Emergency Feedwater Logic ModificationThe 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-2018404-0130 June 2018 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2018001-0131 March 2018 23:59:59WaterfordNRC identifiedFailure to Obtain NRC Staff Authorization Prior to Changing a Procedure that Impacts Implementation of Technical SpecificationsThe 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-2017008-0131 December 2017 23:59:59WaterfordNRC identifiedThree examples of Failure to Establish and Maintain Preventive Maintenance Procedures for Safety-Related Electrical EquipmentThe team identified three examples of a Green non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a, for failure to establish, implement, and maintain written procedures for activities referenced in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, prior to November 16, 2017, the licensee failed to establish and maintain procedures covered in Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance, to implement maintenance for safety-related 1600 A, 600 V non-segregated metal-enclosed bus ducts, safety-related 4.16 kV G.E. Magne-Blast circuit breakers, and safety-related 480 V G.E. switchgear AKR breakers.
05000382/FIN-2017008-0231 December 2017 23:59:59WaterfordNRC identifiedFailure to Meet RG 1.9 Emergency Diesel Testing Requirements during Surveillance Test Results in Missed SurveillanceThe team identified a Green non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification Limiting Condition for Operation 3.8.1.1 for failure to maintain operability of two separate independent diesel generators. Specifically, on May 23, 2017, the licensee failed to verify that the train A emergency diesel generator energized all auto-connected shutdown loads through the load sequencer and operated for greater than or equal to five minutes in accordance with Technical Specification Surveillance Requirement 4.8.1.1.2.
05000382/FIN-2017403-0231 December 2017 23:59:59WaterfordLicensee-identifiedLicensee-Identified Violation
05000382/FIN-2017008-0331 December 2017 23:59:59WaterfordNRC identifiedTwo Examples of Failure to Submit and Receive Prior Authorization of Alternatives to ASME OM Code Leak Testing RequirementsThe team identified two examples of a Severity Level IV, non-cited violation of 10 CFR 50.55a(z), for failure to submit and obtain authorization prior to implementation of multiple alternatives to leak testing requirements of the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) of Nuclear Power Plants Code. Specifically, prior to November 16, 2017, the licensee did not submit and receive prior authorization to alternative leak testing requirements for safety injection valves SI-512A and SI-602B.
05000382/FIN-2017403-0131 December 2017 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2017008-0431 December 2017 23:59:59WaterfordNRC identifiedPotential Failure to Obtain a License Amendment for Changes to Diesel Generator Surveillance Test IntervalThe team identified an unresolved item for the licensees failure to perform a 10 CFR 50.59 safety evaluation and subsequently obtain a license amendment for changes to the surveillance testing frequency of the emergency diesel generators. The licensees process for changing surveillance test intervals is controlled by Technical Specification 6.5.18, Surveillance Frequency Control Program. The licensees changes to the surveillance test intervals are made in accordance with NEI 04-10, Risk Informed Method for Control of Surveillance Frequencies, Revision 1, as written in procedure EN-DC-355, Engineering Evaluation of Proposed Surveillance Test Interval Changes, Revision 2. The team reviewed the licensees changes to the surveillance test interval, as required by Technical Specification Surveillance Requirements 4.8.1.1.2.e, for emergency diesel generators. The licensee changed the surveillance test interval for the train A and B emergency diesel generators from both emergency diesel generators tested every 18 months to each emergency diesel generator tested every 36 months. The team determined that testing the emergency diesel generators once every 36 months was contrary to guidance in Regulatory Guide 1.9, Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants, Revision 4. Specifically, Section 2.3.2.3, Refueling Outage Testing, requires the capability of the overall emergency diesel generator design should be demonstrated during every refueling outage not exceeding a period of 24 months. The team determined that the licensee did not correctly evaluate the change to the surveillance interval in accordance with surveillance frequency control program change process. Specifically, the licensee did not correctly evaluate NEI 04-10, step 1, Check for Prohibitive Commitments, and step 2, Can Commitments be Changed? of the change process. The team determined that this change would require a 10 CFR 50.59 safety evaluation and subsequent license amendment because it would result in more than a minimal increase in likelihood of a malfunction of a component important to safety as previously described in the final safety analysis report. Specifically, the test interval would no longer meet the applicable acceptance standard, Regulatory Guide 1.9, to which the licensee is committed. Planned Closure Action(s): The NRC inspectors will review the final corrective actions, pending NRC resolution of applicability of 10 CFR 50.59 to the surveillance frequency control program. Licensee Action(s): Prior to this inspection, the licensee identified this 10 CFR 50.59 issue in the corrective action program because of industry operating experience. At the time of this inspection, the licensee had not completed the final corrective action and 10 CFR 50.59 activities.These corrective actions will be completed once industry guidance on the NRC resolution of applicability of 10 CFR 50.59 to the surveillance frequency control program was available. Corrective Action Reference(s): Condition Reports CR-WF3-2017-05590 and CR-WF3-2017-5602 NRC Tracking Number: 05000382/2017008-04
05000382/FIN-2017010-0130 September 2017 23:59:59WaterfordNRC identifiedFailure to Evaluate Departures from Approved Methodologies for Reactor Vessel FluenceThe inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(1), which states, in part, that a licensee may make changes in the facility as described in the updated safety analysis report without obtaining a license amendment pursuant to 10 CFR 50.90 only if: (i) a change to the technical specifications incorporated in the license is not required, and (ii) the change, test, or experiment does not meet any of the criteria in paragraph (c)(2). Title 10 CFR 50.59, Section (c)(2)(viii), states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in a departure from a method of evaluation described in the updated safety analysis report used in establishing the design bases or in the safety analyses. Specifically, since January 2017, the licensee revised updated final safety analysis report Section 4.3.3.3 to reflect RAPTOR-M3G as the current licensing basis fluence method without first obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2017-04748. The inspectors determined that the failure to evaluate proposed changes to determine if prior NRC review was required in accordance with 10 CFR 50.59 was a performance deficiency. Using NRC Inspection Manual Chapter 0612, Appendix B, Issue Screening, the inspectors determined that this performance deficiency had minor safety significance. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the Reactor Oversight Process significance determination process because this violation potentially impacted the ability of the NRC to perform its regulatory oversight function. Therefore, this violation was processed through traditional enforcement examples of Section 6.1 of the NRC Enforcement Policy. This violation was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation, similar to the more than minor example of a change in requirements in the NRC Enforcement Manual, Appendix E, Minor Violations Examples, dated September 9, 2013. Since the violation was associated with a performance deficiency of minor significance, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.
05000382/FIN-2017002-0230 June 2017 23:59:59WaterfordSelf-revealingFailure to Ensure Containment Equipment Hatch Closure Prior to RCS Time to BoilThe 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-2017002-0330 June 2017 23:59:59WaterfordNRC identifiedFailure to Ensure Appropriate Testing of TSP Baskets Inside ContainmentThe 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-2017002-0430 June 2017 23:59:59WaterfordNRC identifiedFailure to Perform a Post Maintenance Test on a Main Steam Isolation Valve Solenoid ValveThe 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-2017002-0530 June 2017 23:59:59WaterfordSelf-revealingFailure to Perform Maintenance on the Correct Safety-Related ComponentThe 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-2017002-0130 June 2017 23:59:59WaterfordNRC identifiedFailure to Prepare the Site for Impending Adverse WeatherThe 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-2017002-0630 June 2017 23:59:59WaterfordLicensee-identifiedLicensee-Identified ViolationLicensee 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-2017001-0131 March 2017 23:59:59WaterfordSelf-revealingFailure to Perform Field Changes in Accordance with Design Control MeasuresGreen . 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-2016008-0131 December 2016 23:59:59WaterfordNRC identifiedFailure to Control Nonconforming PartsThe team identified a Green non-cited violation of 10 CFR Part 50,Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, which occurred when the licensee failed to dedicate commercial-grade relays for use insafety-related applications. After receiving information from a vendor that more than124 relays potentially installed in safety-related applications did not conform to quality assurance standards, the licensee failed to take appropriate steps to accept these Commercial-grade relays as basic components. After discussion with the team, the licensee documented this condition in Condition Report CR-WF3-2016-07710 and initiated actions to ensure compliance with quality assurance requirements. The failure to dedicate commercial-grade relays used asor intended for use asbasic components (in safety-related applications) as required by plant procedures and by10 CFR Part 21 was a performance deficiency. This performance deficiency wasmore-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability,reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system, or component, and operability was maintained.The finding has a conservative bias cross-cutting aspect in the human performance cross cutting area because licensee personnel improperly rationalized the adequacy of the nonconforming components to perform their safety-related functions (H.14). Because this performance deficiency was also a violation that impacted the regulatory process, in that the licensee accepted a change to plant design without appropriate evaluation and notification, it was also evaluated for traditional enforcement. The team determined that the violation was Severity Level IV because it was similar to several examples in Section 6.5.d of the NRC Enforcement Policy.
05000382/FIN-2016004-0131 December 2016 23:59:59WaterfordSelf-revealingFailure to Ensure Appropriate Post-Maintenance Testing on Essential Chiller BGreen. 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-2016008-0431 December 2016 23:59:59WaterfordNRC identifiedDeparture from Approved Method to Determine Steam Generator Internal Loads During Main Steam Line BreakThe team identified a Severity Level IV non-cited violation of 10 CFR 50.59(c)(2),Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment prior to implementing a proposed change, test, or experiment that would result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Specifically, the licensee departed from their approved CEFLASH-4A methodology to determine steam generator internal differential loads caused by a main steam line break to an unapproved TRANFLOW methodology. In response to this issue, the licensee entered the issue into the corrective action program as Condition Report CR-WF3-2016-07639 and initiated actions to prepare a new evaluation under current regulatory guidelines or to submit a license amendment request to the NRC.The licensees failure to obtain a license amendment prior to implementing a change that resulted in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses, as required by 10 CFR 50.59(c)(2) was a violation. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the Reactor Oversight Process significance determination process because this violation potentially impacted the ability of the NRC to perform its regulatory oversight function. Therefore, this violation was processed through traditional enforcement examples of Section 6.1 of the NRC Enforcement Policy. This violation was more-than-minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation, similar to the more-than-minor example of a change in requirements in the NRC Enforcement Manual,Appendix E, Minor Violations Examples, dated September 9, 2013. In accordance with the NRC Enforcement Policy, the significance determination process was used to inform the significance of the failure to obtain a license amendment prior to implementing a proposed change. The departure from the original CEFLASH-4A method to the TRANFLOW method to determine differential loads on steam generator internal structures following a main steam line break event was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012,Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the issue would not result in the complete or partial loss of a support system that contributes to the likelihood of an initiating event, or result in the steam generators violating accident leakage performance criterion. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000382/FIN-2016008-0231 December 2016 23:59:59WaterfordNRC identifiedFailure to Perform Operability Determinations for Nonconforming ConditionsThe team identified a Green non-cited violation of 10 CFR Part 50, Appendix B,Criterion V, Instructions, Procedures, and Drawings, that occurred when the licensee failed on two occasions to perform an operability determination for a nonconforming condition affecting numerous safety-related components. Following receipt of information from a vendor that more than 124 relays potentially installed in safety-related applications did not conform to quality requirements, licensee personnel failed to perform an operability evaluation. Later, during a Part 21 evaluation for the potential defect, the evaluator noted that an operability determination was needed, but failed to initiate the appropriate processes. After discussion with the team, the licensee documented this condition in Condition Report CR-WF3-2016-07710, declared the affected components operable, but degraded, and initiated actions to restore full qualification.Failures to perform an operability determination following identification of a nonconforming condition as required by station procedures were two examples of a performance deficiency.This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012,the team determined that this finding was of very low safety significance (Green) because it did not represent the actual loss of function of any system or train. The finding has an identification cross-cutting aspect in the problem identification and resolution cross-cuttingarea because licensee personnel failed to recognize a nonconforming condition as a condition adverse to quality (P.1).
05000382/FIN-2016403-0131 December 2016 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2016008-0331 December 2016 23:59:59WaterfordNRC identifiedFailure to Include Appropriate Quantitative Acceptance Criteria for the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced VibrationThe team identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include appropriate quantitative accept an cecriteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensees reconstituted feedwater and emergency feedwater system monitoring plan, which was created to monitor both systems vibrations following the sites steam generators replacement, did not include a range for acceptable vibration levels for all The team identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include appropriate quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensees reconstituted feedwater and emergency feedwater system monitoring plan, which was created to monitor both systems vibrations following the sites steam generators replacement, did not include a range for acceptable vibration levels for all
05000382/FIN-2016403-0231 December 2016 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2016003-0230 September 2016 23:59:59WaterfordLicensee-identifiedLicensee-Identified ViolationTitle 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-2016002-0330 June 2016 23:59:59WaterfordNRC identifiedFailure to Perform Drills Required by the Site Emergency PlanThe 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-2016002-0130 June 2016 23:59:59WaterfordNRC identifiedFailure to Properly Pre-Plan and Perform Maintenance on the Cable Vault and Switchgear Ventilation SystemThe 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-2016002-0430 June 2016 23:59:59WaterfordSelf-revealingFailure to Account for Starting Air Design Features in Emergency Diesel Operating ProceduresA 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-2016002-0230 June 2016 23:59:59WaterfordNRC identifiedFailure to Properly Assess and Manage Risk When Performing Dry Cooling Tower MaintenanceThe 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-2015407-0231 March 2016 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2016001-0131 March 2016 23:59:59WaterfordNRC identifiedFailure to Assess and Manage the Increase in Risk from Emergent Maintenance ActivitiesThe 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-2016001-0231 March 2016 23:59:59WaterfordLicensee-identifiedLicensee-Identified ViolationTechnical 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-2015407-0131 March 2016 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2015403-0131 December 2015 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2015004-0131 December 2015 23:59:59WaterfordSelf-revealingFailure to Properly Pre-Plan and Perform Maintenance on the Core Element Assembly CalculatorsThe 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-2015406-0130 September 2015 23:59:59WaterfordLicensee-identifiedLicensee-Identified Violation
05000382/FIN-2015008-0130 September 2015 23:59:59WaterfordNRC identifiedInadequate Procedures for Design Basis Tornado EventThe team identified two examples of a Green, non-cited violation of Technical Specification 6.8.1, which states, in part, Written procedures shall be established, implemented, and maintained, covering the activities including procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A.6.w, Acts of Nature (e.g., tornado, flood, damn failure, earthquakes). Specifically, in the first example, prior to August 27, 2015, the licensee failed to establish adequate procedures to ensure the manual actions required within specified time limits can be completed before full draindown of the ultimate heat sink (wet cooling tower basins) after a tornado event. In the second example, prior to August 27, 2015, the licensee failed to establish adequate procedures to clarify whether the main steam isolation valve area was considered outdoors and therefore subject to the requirements for unmonitored items stored in the protected area. Unsecured scaffold material stored in this area had not been evaluated for potential to become projectiles and endangering nearby safety-related equipment during high winds. In response to this issue, the licensee inspected the area and secured all loose debris. This finding was entered into the licensees corrective action program as Condition Reports CR-WF3-2015-05624 and CR-WF3-2015-05601. The team determined that the failure to maintain adequate procedures to ensure compliance with technical specifications and Regulatory Guide 1.33 was a performance deficiency. This finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and 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 licensee failed to establish adequate procedures to ensure availability of mitigating equipment during and after an event involving acts of nature. 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 both examples involved a design basis tornado. Per Exhibit 4, the issue screened to a more detailed risk evaluation because: 1) the first issue could starve safety systems of water, failing the safety function, and 2) the second issue could cause a plant trip and a loss of condenser heat sink initiating event. Therefore, the Region IV senior reactor analyst performed a more detailed risk evaluation that included both issues. Given that there was no change in core damage frequency for the first issue, and the change in core damage frequency for the second example was 1.2 x 10-9 per year, combined, the analyst determined that the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of problem identification and resolution, evaluation, because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000382/FIN-2015406-0230 September 2015 23:59:59WaterfordLicensee-identifiedLicensee-Identified Violation
05000382/FIN-2015008-0330 September 2015 23:59:59WaterfordLicensee-identifiedLicensee-Identified ViolationThe licensee failed to perform a full 10 CFR 50.59 evaluation for changes made to the Core Operating Limits Report during Refueling Cycle 20. Title 10 of the Code of Federal Regulations, Part 50.59, states, in part, that licensees may make changes to the facility as described in the final safety analysis report without obtaining a license amendment only if the change does not meet any of the criteria in paragraph (c)(2), which includes a full evaluation of eight criteria supporting the conclusion that the change is not adverse. Several physics assessment checklist exceptions for Cycle 20 inputs were more limiting as compared to those in the licensing basis analysis of record, including the axial shape index limit and azimuthal tilt. Contrary to the above, prior to August 27, 2015, the licensee failed to complete a full 10 CFR 50.59 evaluation for changes to the core operating limits report for Cycle 20. Specifically, the axial shape index limit is an input to various Reload Physics and Accident analyses, and is specifically discussed in the updated final safety analysis report, while the azimuthal tilt had two inputs related to heating of reactor vessel internals that should have been considered adverse and therefore required a full 10 CFR 50.59 evaluation. This finding was assessed using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, and was determined to be of very low safety significance (Green) because it did not result in the mismanagement of reactivity by operators. This issue was entered into the licensees corrective action program as Condition Reports CR-WF3-2015-04040, CR-WF3-2015-04045, and CR-HQN-2015-00684.
05000382/FIN-2015406-0330 September 2015 23:59:59WaterfordLicensee-identifiedLicensee-Identified Violation
05000382/FIN-2015003-0130 September 2015 23:59:59WaterfordSelf-revealingFailure to Establish Design Control Measures for Safety-Related Emergency Feedwater System ValvesThe 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-2015008-0230 September 2015 23:59:59WaterfordLicensee-identifiedLicensee-Identified ViolationThe licensee failed to establish the alarm setpoints for the pressurizer relief valve tailpipe temperature in accordance with their design basis. Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, prior to November 7, 2011, the licensee failed to assure that the design basis for the pressurizer relief valve tailpipe alarm setpoints were correctly translated to specifications and procedures. Specifically, the pressurizer relief valve tailpipe setpoint was set at 280 degrees Fahrenheit in accordance with the setpoint study documented prior to initial startup. The licensee failed to adjust the setpoint to the design basis setting of the actual system temperature plus 25 degrees Fahrenheit as specified. This finding was assessed using Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings at-Power," and was determined to be of very low safety significance (Green) because it did not result in the loss of a system or function. This issue was entered into the licensees corrective action program as Condition Report 2015-05607.
05000382/FIN-2015003-0230 September 2015 23:59:59WaterfordSelf-revealingFailure to Follow Procedures when Changing Materials Used for Feedwater Heater Level Control ValvesThe 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-2015007-0530 June 2015 23:59:59WaterfordNRC identifiedViolation of Technical Specification 6.8.1.f for the failure to implement and maintain adequate written procedures covering fire protection program implementationThe team identified a violation of Technical Specification 6.8.1.f for the failure to implement and maintain adequate written procedures covering fire protection program implementation. Specifically, the team identified four examples where the licensee failed to maintain an alternative shutdown procedure that successfully mitigated all postulated alternative shutdown scenarios. This finding affects 10 CFR 50.48 and has been screened and determined to warrant enforcement discretion per the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48). The licensee used Procedure OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown, Revision 28, to shut down the reactor from the remote shutdown panel in the event a control room or cable vault fire required evacuation of the control room. This alternative shutdown procedure provided steps for operators to transfer control of the credited safe shutdown equipment away from the control room to the remote shutdown panel and to achieve and maintain safe shutdown conditions from the remote shutdown panel. The team performed a timed walkdown of the alternative shutdown procedure. Based on the walkdown results, the team determined that the alternative shutdown procedure was not adequate to ensure that operators could successfully mitigate all postulated alternative shutdown scenarios. In particular, the team identified the following four scenarios where operators may not be able to achieve and maintain a safe shutdown: Example 1: Potential Loss of Credited Safe Shutdown Pumps The first scenario involved fire damage resulting in blown fuses for either the component cooling water or emergency feedwater pumps. In this scenario, the team determined the operators would be unable to control the affected pump from the remote shutdown panel, but the operators would be able to control the affected pump by manually operating the breakers that supplied power to the motors. The team noted that the alternative shutdown procedure did not provide any steps for operators to manually operate the breakers to control these pumps, which were required for safe shutdown. Example 2: Potential Spurious Opening of the Atmospheric Dump Valves The second scenario involved the spurious actuation of two atmospheric dump valves. The team noted that the licensee previously had a 10-minute requirement for operators to mitigate the spurious actuation of two atmospheric dump valves by taking manual control of an open atmospheric dump valve locally and then manually closing the valve. The team determined that the alternative shutdown procedure provided steps for operators to manually close an open atmospheric dump valve; however, the licensee removed the 10-minute requirement for operators to be able to perform this action. The licensee removed the 10-minute requirement based on its understanding that the spurious actuation of only one atmospheric dump valve was required to be analyzed and mitigated. The team referred to guidance in Regulatory Guide 1.189, Revision 2, which stated, in part, after control of the plant is achieved by the alternative or dedicated shutdown system, single or multiple spurious actuations that could occur in the fireaffected area should be considered... The team reviewed the licensees method for isolating the atmospheric dump valves from the effects of a control room or cable vault fire. The team determined that the circuits responsible for isolating the atmospheric dump valves were located within the control room complex and, therefore, could not be relied upon to isolate the atmospheric dump valves in the event of a control room fire. The team concluded that the licensee should have maintained the 10-minute requirement in the alternative shutdown procedure for operators to manually close a spuriously open atmospheric dump valve. During the timed walkdown of the alternative shutdown procedure, the team determined that it would take operators approximately 13 minutes to close a spuriously open atmospheric dump valve. Example 3: Potential Spurious Opening of a Pressurizer Spray Valve The third scenario involved the spurious opening of a pressurizer spray valve. In this scenario, the open pressurizer spray valve results in a rapid depressurization of the reactor coolant system, which could negatively impact the ability to achieve and maintain natural circulation. The licensee considered this scenario in the safe shutdown analysis. The licensee did not perform an analysis or calculation to determine the amount of time operators had available to mitigate this scenario. Instead, the licensee used engineering judgment to specify that operators had 10 minutes available to secure the spurious spray flow. The team was concerned that the 10-minute limit may not be sufficient to ensure that operators could achieve and maintain natural circulation. In response to the teams concern, the licensee modeled this scenario on the simulator. The team noted that the use of the simulator was not a preferred method; however, it provided a reasonable estimate for the amount of time available. The results of the simulator run indicated that the reactor coolant system would reach saturation pressure in less than 8 minutes. Once the reactor coolant system reaches saturation pressure, voiding begins in the reactor coolant system. This voiding could then negatively impact the ability to achieve and maintain natural circulation. The team determined that the alternative shutdown procedure provided steps for operators to trip the reactor coolant pumps, which would mitigate this scenario by eliminating flow through the pressurizer spray valves. During the timed walkdown of the alternative shutdown procedure, the team determined that it would take operators approximately 9 minutes and 15 seconds to trip all of the reactor coolant pumps. Scenario 4: Potential Overfilling of the Steam Generators The fourth scenario involved the potential overfilling of the steam generators. In this scenario, the open main steam isolation valves continue to provide steam to the turbinedriven main feedwater pumps, which continue to inject feedwater into the steam generators until they overfill. The team noted that the action to close the main steam isolation valves prior to evacuating the control room was an operator action within the fire area. The team determined that this action was not credited in the plants approved fire protection program; therefore, the operators must take action outside of the control room to ensure that the main steam isolation valves were closed. Because the licensee did not have an analysis establishing a time limit, the team was concerned the operators may not perform this action prior to main feedwater overfilling the steam generators. In response to the teams concern, the licensee modeled this scenario on the simulator. The team noted that the use of the simulator was not a preferred method; however, it provided a reasonable estimate for the amount of time available. The results of the simulator run indicated that the continued injection of main feedwater at full flow could overfill the steam generators in approximately 2 minutes and 30 seconds. The team noted that overfilling the steam generators would negatively impact the ability to remove decay heat. The team determined that the alternative shutdown procedure provided steps for operators to close the main steam isolation valves from outside the control room, which would mitigate this scenario by eliminating steam flow to the turbine-driven main feedwater pumps. During the timed walkdown of the alternative shutdown procedure, the team determined that it would take operators approximately 4 minutes and 30 seconds to close all of the main steam isolation valves.
05000382/FIN-2015007-0230 June 2015 23:59:59WaterfordNRC identifiedFailure to Provide a Bounding Calculation for Time Critical ActionsThe team identified a non-cited violation of License Condition 2.C.9, "Fire Protection," for the failure to adequately correct a previous violation. Specifically, the licensee failed to provide a bounding calculation for the amount of time available for operators to establish component cooling water during an alternative shutdown. The licensee developed this calculation in response to Non-cited Violation 2012007-02. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2015-0859 and implemented a fire impairment as a compensatory measure. The failure to provide a bounding calculation for the amount of time available for operators to establish component cooling water during an alternative shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a postulated control room fire that led to control room evacuation and determined this violation was of very low safety significance. This finding had a cross-cutting aspect associated with resolution within the problem identification and resolution area since the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the team determined that the licensees corrective actions were not effective since the licensee failed to provide a bounding calculation for the amount of time available for operators to establish component cooling water during an alternative shutdown (P.3).
05000382/FIN-2015201-0130 June 2015 23:59:59WaterfordNRC identifiedSecurity
05000382/FIN-2015007-0330 June 2015 23:59:59WaterfordNRC identifiedFailure to Periodically Test Emergency Lighting UnitsThe team identified a non-cited violation of License Condition 2.C.9, Fire Protection, for the failure to periodically test and demonstrate the 8-hour capacity of the Appendix R emergency lighting units. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2015-00856 and operators had flashlights available as a compensatory measure. The failure to periodically test and demonstrate the 8-hour capacity of the Appendix R emergency lighting units was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013. The team assigned the finding a low degradation rating because it would not prevent reaching and maintaining safe shutdown conditions in the event of a control room fire. Specifically, the team had reasonable assurance that the emergency lighting units would provide adequate illumination for a sufficient amount of time for operators to perform the most time critical actions. In addition, the team determined that operators performing an alternative shutdown had flashlights available in the Appendix R equipment lockers. Because the team assigned a low degradation rating, this finding screened as having very low safety significance. This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago.
05000382/FIN-2015201-0230 June 2015 23:59:59WaterfordNRC identifiedSecurity