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05000443/FIN-2017008-012017Q3SeabrookFailure to Correct Condition Adverse to Fire Protection Associated with Fire Safe ShutdownThe team identified a Green, non-cited violation of Seabrook Station Unit 1 Facility Operating License Condition 2.F, Fire Protection, for failure to implement and maintain in effect all provisions of the approved Fire Protection Program. Specifically, although NextEra identified that procedure OS1200.00 did not properly implement a mitigating action for a fire in the Switchgear Room A as prescribed in the Appendix R Safe Shutdown Analysis Report on August 30, 2010 (Action Requests (ARs) 576775 and 1638123), corrective actions were delayed due to higher priority work and were not timely commensurate with the potential safety significance. NextEra entered the issue into the corrective action program as AR 2214834 and planned to reprioritize the preparation and submittal of a license amendment request to resolve the issue.The issue was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors (fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by failing to correct the condition in a timely manner, NextEra did not ensure that the associated fire safe shutdown procedure implemented actions to mitigate a fire in the Switchgear Room A as analyzed in the Appendix R Safe Shutdown Report. The team performed a Phase 1 screening in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process. The deficiency affected the post-fire safe shutdown category because NextEras fire response procedures were degraded. The finding was screened to very low safety significance (Green) because it was assigned a low degradation rating because the procedural deficiencies could be compensated by operator experience and system familiarity. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Resources, in that, NextEra did not ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically,action to submit a license amendment request to support a deviation from the 10 CFR Part 50, Appendix R, III.G.2 requirements for cable separation had been rescheduled five times due to higher priority licensing work (H.1).
05000443/FIN-2017008-022017Q3SeabrookFailure to Implement Test Program for Appendix R Emergency Lighting UnitsThe team identified a Green, non-cited violation of Seabrook License Condition 2.F, Fire Protection, because NextEra did not implement the fire protection test program to ensure that the emergency lighting units were in conformance with design requirements. Specifically, NextEra did not implement procedure LS0565.31, 8-Hour Emergency Light Inspections, to verify that the Appendix R emergency lighting units would meet the annual inspection requirements, as well as the 3-year preventive maintenance task for battery replacement and the 8-hour capacity test. Additionally, since the 3-year preventive maintenance task was coded incorrectly, there was no process to ensure that the LS0565.31 would be completed going forward. NextEra entered this issue into the corrective action program as AR 2214652. NextEras planned corrective actions included revising the classification of the emergency lighting unit preventive maintenance task in order to ensure that the task is performed at the appropriate frequency.The team determined that this issue was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, failure to conduct the annual inspection requirements and 3-year preventive maintenance activities could result in the emergency lighting units not meeting the 8-hour battery capacity requirement. The team evaluated this finding using Inspection Manual Chapter 0609, Fire Protection Significance Determination Process. Because safe shutdown conditions could be reached and maintained, this finding screened as having very low safety significance (Green). The team determined this finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the three-year preventive maintenance task to replace the batteries in the emergency lighting units was coded incorrectly in the work management system, which resulted in NextEra not completing the required testing and maintenance on the lighting units to ensure that they would perform their function during safe shutdown operations (H.5).
05000354/FIN-2017008-012017Q1Hope CreekImproper Preventive Maintenance Deletion Results in the Inoperability of the A Control Room HVAC SystemGreen . A self -revealing Green non- cited violation ( NCV ) of Technical Specification ( TS ) 6.8.1, Procedures and Programs, as described in Regulatory Guide (R G) 1.33, Revision 2, February 1978, was identified when PSEG did not maintain an appropriate preventive maintenance ( PM ) schedule for the A control room heating, ventilation and air conditioning (HVAC ) system. Specifically, PSEG inadvertently deactivated a PM activity to perform periodic cleaning of the A control room return air fan (AVH -415) low flow switch pitot tubes that resulted in the A train of the control room emergency filtration ( CREF ) to be 3 unavailable on November 23, 2016 . PSEG performed corrective actions to clean the clogged pitot tubes associated with the AH -415 flow switch, re -activate the inadvertently deleted PM, and identify the extent of condition in other systems . This issue was more than minor because it was associated with the structures, systems and components ( SSC ) and barrier performance attribute of the Barrier Integrity Cornerstone (under the areas to measure associated with the radiological barrier function of the control room); and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Attachment 4 and Appendix A, Exhibit 3, because the finding only represented a degradation of the radiological barrier function for the control room. The inspectors determined that there was no cross -cutting aspect associated with this finding since it was not representative of current PSEG performance. Specifically, the causal factors associated with this finding occurred in 2010, which was outside the nominal three- year period of consideration and were not considered representative of present performance in accordance with IMC 0612
05000387/FIN-2016004-012016Q4SusquehannaFailure Rates Exceed (20%) Twenty Percent for Biennial Requalification ExamGreen. A self-revealing finding was identified associated with inadequate licensed operator performance during the annual licensed operator requalification operating test and biennial written examination. Specifically, 17 of 71 operators (23.9%) failed at least one portion of the requalification examinations. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, 17 of 71 licensed operators failed to demonstrate a satisfactory understanding of the required knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors evaluated this performance deficiency using IMC 0609, SDP, Appendix I, Licensed Operator Requalification SDP. This finding is of very low safety significance (Green) because the finding is related to requalification exam results, did not result in a failure rate of greater than 40 percent and all 17 operators were remediated and successfully retested prior to returning to licensed duties. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Susquehanna did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce. (H.7)
05000387/FIN-2016004-022016Q4SusquehannaFailure to Promptly Correct a Condition Adverse to Quality with LPCI Swing Bus Automatic Transfer SwitchesGreen. A finding of very low safety significance (Green) and associated NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed when Susquehanna failed to assure that conditions adverse to quality were promptly identified and corrected on two separate occasions. Both examples resulted in the failures of safety-related automatic transfer switches (ATSs) associated with the low pressure coolant injection (LPCI) swing buses. Corrective actions included enhancing the work instructions for all applicable ATSs based off original equipment manufacturer (OEM) input and scheduling the enhanced work instructions to be performed on the four swing bus ATSs during their next scheduled bus outages. Inspectors determined that the finding was more than minor because it was associated with the Equipment Performance attribute of the Reactor Safety 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 (i.e., core damage). In both examples, the failure to correct conditions adverse to quality resulted in the loss of power to the LPCI swing bus and inoperability of the respective division of LPCI. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, inspectors and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green). Specifically, though a single train was inoperable for greater than its technical specification (TS) allowed outage time, in consultation with regional senior reactor analysts, inspectors determined it did not represent an actual loss of function. The finding is related to the cross-cutting area of Problem Identification and Resolution, Evaluation, because Susquehanna did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Susquehanna either failed to evaluate deficiencies encountered during maintenance or failed to ensure that corrective actions aligned with and corrected the identified causes. (P.2)
05000388/FIN-2016004-032016Q4SusquehannaRefuel Floor Radiation Monitor Inoperable Due to being Improperly CalibratedGreen. A finding of very low safety significance (Green) and NCV of TS 5.4.1, Procedures was self-revealed when Susquehanna incorrectly calibrated the Unit 1 B refuel floor high exhaust duct high radiation monitor on November 15, 2014. This impacted the initiation capability of secondary containment isolation and control room emergency outside air supply system (CREOASS) and resulted in Susquehanna exceeding the allowed outage time for TSs 3.3.6.2, Secondary Containment Isolation, and 3.3.7.1, CREOASS Instrumentation. Upon identification of the issue, Susquehanna properly calibrated the radiation monitor to restore its operability. This finding is more than minor because it is associated with the Human Performance (Routine OPS/Maintenance Performance) attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment and Control Room Ventilation) protect the public from radionuclide releases caused by accidents or events. Specifically, incorrectly calibrating the radiation monitor resulted in both systems being inoperable for almost two years. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At-Power, both dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was only associated with the radiological barrier function of the Control Room and Secondary Containment. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency because Susquehanna did not recognize and plan for the possibility of mistakes, latent problems, or inherent risk, even while expecting successful outcomes. Specifically, Susquehanna personnel did not consider the potential undesired consequences of their actions before performing work and implement appropriate error-reduction tools (e.g. self-check, peer-check). (H.12)
05000387/FIN-2016004-042016Q4SusquehannaAuxiliary Bus Load Shed when a Daisy Chained Neutral was Interrupted during MaintenanceGreen. A finding of very low safety significance (Green) for failure to develop an adequate work plan for replacement of a voltage potential indicating light on a breaker on the Unit 2 B auxiliary bus was self-revealed when the Unit 2 B reactor recirculation pump (RRP) tripped, along with other non-safety related loads on November 14, 2016, resulting in a rapid unplanned power change and transition to single loop operation. Specifically, operations and maintenance personnel did not recognize that disconnecting the neutral wires from the light socket would interrupt power to all of the degraded voltage relays for the auxiliary bus. Therefore, the relays de-energized when the maintenance was performed, tripping all the breakers on the bus. Susquehannas immediate corrective actions included stabilizing the plant, entering single loop operations, and entering the issue into their corrective action program (CAP). Additionally, Susquehanna performed a maintenance department stand down to communicate immediate lessons learned from the event while a more thorough causal analysis was conducted. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected 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, implementation of work instructions resulted in the trip of the Unit 2 B RRP, B and D circulating water (CW) pumps, B and D condensate pumps, and the B service water (SW) pump, which caused an automatic trip of the C reactor feed pump and runback of the A RRP, resulting in a rapid power reduction to 32 percent rated thermal power (RTP). The inspectors evaluated the finding in accordance with IMC 0609, Appendix A "The SDP for Findings At-Power," dated June 19, 2012, Exhibit 1 for the Initiating Events cornerstone and determined the finding was of very low safety significance (Green) because it did not cause a reactor trip. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management because Susquehanna did not implement a process of planning work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate with the work. Specifically, Susquehanna did not recognize the risk of interrupting a daisy chained neutral when planning a minor maintenance work order and did not recognize the impact of the work activity in the field. (H.5)
05000387/FIN-2016004-052016Q4SusquehannaLERs Associated with Reactor Coolant Pressure Boundary LeakageEnforcement. TS 3.4.4, "RCS" requires RCS leakage be limited to no pressure boundary leakage in Mode 1. Contrary to this, pressure boundary leakage from a LPRM instrument housing and from socket weld #8 occurred between plant start-up in December 2015 and plant shutdown on June 6, 2016, and existed while in Mode 1. The inspectors determined that these violations of TS 3.4.4 are more than minor, but not the result of performance deficiencies. Specifically, for the first event, though leakage likely existed during the previous refueling outage when personnel were performing unrelated maintenance and inspection activities, it was likely too small to reasonably identify and correct. Similarly, for the 2016 leak identified in weld #8, the leakage causes were not within Susquehannas ability to foresee as they had replaced the weld with the industry recommended 2 x 1 taper configuration and used qualified procedures and personnel. The Susquehanna staff had also measured the susceptibility of the attached piping for vibrational inputs. In accordance with the NRC Enforcement Policy guidance and IMC 0612, these violations are being treated under the traditional enforcement process and best characterized as a Severity Level (SL) IV (very low safety significance) violation, similar to example d.1 in NRC Enforcement Policy, Section 6.1, Reactor Operations. Although a performance deficiency was not identified, to verify that the issue was of very low safety significance, the inspectors considered risk insights obtained by using IMC 0609, SDP, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that these TS violations would screen to Green (very low safety significance) because the boundary leakage would not have exceeded the leak rate for a small loss of coolant accident (LOCA) and would not affect any LOCA accident mitigating systems or components. Therefore, the inspectors considered that the SL IV characterization was appropriate. The licensee entered these issues into the Susquehannas CAP as CR-2016-14544 and CR-2016-14366. Because these issues are of very low safety significance, it has been determined that it was not reasonable for Susquehanna to be able to foresee and prevent, and as such no performance deficiencies exist. The NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-16-283). Further, because Susquehanna's actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix.
05000382/FIN-2009005-012009Q4WaterfordFailure to follow radiation protection procedural requirementsThe inspectors reviewed a self-revealing noncited violation of Technical Specification 6.8.1 which resulted from a worker failing to follow radiation protection procedures. A contract radiation worker went to work near steam generator 1 rather than the area for which he/she was briefed and received multiple electronic dosimeter dose rate alarms, but did not leave the area until receiving a continuous dose alarm. In response, the licensee investigated the occurrence and restricted the individuals access. Additional actions were being evaluated. This issue was entered into the licensees corrective action program as Condition Reports CR-WF3-2009-05648 and WF3-2009-06852. This finding is greater than minor because it involved the program attribute of exposure control and affected the cornerstone objective in that the failure of the worker to follow procedural guidance resulted in the worker being unknowledgeable to the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not: (1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a crosscutting aspect in the area of human performance, work practices component, because the worker failed to use human error prevention techniques such as self and peer checking (H.4(a)) (Section 2OS1)
05000382/FIN-2009005-022009Q4WaterfordReactor Coolant Pump Vapor Seal LeakageA self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee did not promptly correct reactor coolant pump vapor seal leakage that resulted in boric acid accumulation on the component cooling water heat exchanger and cover areas of three reactor coolant pumps. Corrective actions for this condition were implemented during Refueling Outage 15, but these corrective actions failed to correct the condition and the vapor seal leakage continued through operating Cycle 16. This resulted in some additional boric acid corrosion and degradation to reactor coolant pump covers and carbon steel component cooling water flanges. The licensee implemented a design modification to correct the condition and documented the condition in Condition Report CR-WF3-2009-5501. The licensees failure to promptly correct a condition adverse to quality is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment was still available. This finding had a crosscutting aspect in the area of human performance associated with work control in that the licensee did not effectively plan for the resources necessary to implement the postmaintenance testing associated with the corrective actions implemented during Refueling Outage 15, and therefore failed to discover that those corrective actions were inadequate to correct the condition (H.3(a)) (Section 4OA2)
05000382/FIN-2009005-032009Q4WaterfordFailure to Update Drawings after Design ChangeA self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for the licensees failure to prescribe an activity affecting quality by documented instructions, procedures, or drawings appropriate to the circumstance. Specifically, for all reactor coolant pump heat exchanger to pump cover bolted connection gasket replacements between the refueling outage of 1986 (Refueling Outage 1) and the refueling outage of 2009 (Refueling Outage 16), the licensee prescribed the wrong gasket material, gasket size, and fastener preload because they had failed to incorporate a design change implemented during Refueling Outage 1 into their instructions, procedures, or drawings. Station Modification Package SMP-1427, an engineering change implemented during Refueling Outage 1 in response to industry operating experience, called for a thicker gasket, different gasket material, and an increased bolt preload in order to increase gasket compression and reduce the probability of leakage. As a consequence of failing to incorporate Station Modification Package SMP-1427 changes into procedures, all heat exchanger gasket replacements since Refueling Outage 1, four gasket replacements in total, have utilized thinner gaskets with less than the vendor recommended compression. The licensee documented this condition in Condition Report CR-WF3-2009-5501. The licensees failure to prescribe appropriate gasket replacement requirements is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available. This finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee did not institutionalize operating experience through changes to the station procedures (P.2(b)) (Section 4OA2)
05000382/FIN-2009004-012009Q4WaterfordFailure to Follow Technical Specification Requirements for Reactor Protective InstrumentationThe inspectors identified a Green non-cited violation of technical specification 3.3.1, Reactor Protective Instrumentation. The technical specifications require all four channels (A, B, C, and D) of local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments to be operable when in Mode 1. These Channel B instruments require an input from the Channel B log power instrument, which was previously declared inoperable. With the Channel B log power instrument inoperable, the Channel B local power density, departure from nucleate boiling ratio, and reactor coolant flow instruments should also have been declared inoperable. The licensee entered this finding in their corrective action program as condition reports CR-WF3-2009-4401 and CR-WF3-2009-4407. The failure to either trip or bypass the inoperable channels within one hour was more than minor because it affected the configuration control attribute of the mitigating systems cornerstone. Specifically, deliberate operator action was required to ensure that proper reactor protection system coincidence and reliability were maintained. Also, if left uncorrected, the potential existed for Channel B reactor protective trips to be inadvertently removed while at power. The failure to meet the technical specifications was considered to be of very low safety significance (Green), since there was no actual loss of safety function. This finding has a cross-cutting aspect in the decision-making component of the human performance area because the licensee failed to verify the validity of underlying assumptions and identify unintended consequences of failing to comply with technical specification 3.3.1 by declaring the log power Channel B inoperable and not placing local power density, departure from nucleate boiling ratio, and reactor coolant flow instrument channels in either bypass or trip condition (H.1.b).
05000382/FIN-2009005-042009Q4WaterfordLicensee-Identified ViolationTechnical Specification 6.8.1 requires written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Appendix A lists procedures for access control to radiation areas. Procedure EN-RP-100, Radworker Expectations, Revision 3, Section 5.3(9) requires the radiation work permit to be read, understood, and obeyed as a condition of radiologically controlled area access. Procedure EN-RP-100, Radworker Expectations, Revision 3, Section 5.4(3)(h) requires the worker know where to properly perform his/her task. Section 5.3(17) requires the worker be briefed and sign on the appropriate radiation work permit. Section 5.3(11) requires the worker know the radiological conditions in the work area. The licensee identified an example of a worker entering a high radiation area using an inappropriate radiation work permit and without knowing the dose rates in the area. On October 24, 2009, a security officer entered shutdown heat exchanger Room B and received an electronic dosimeter dose rate alarm. The room was posted as a high radiation area and dose rates within the area were as high as 140 millirem per hour. The officer entered the radiological controlled area using Radiation Work Permit 2009005, Tours and Inspection in All Radiological Controlled Areas, Except High Radiation Areas, Locked High Radiation Areas, Very High Radiation Areas, and the Reactor Containment Building. Because the radiation work permit did not allow entry into high radiation areas, radiation protection personnel did not anticipate the officer would enter the room and did not brief the officer on the dose rates in the area. In response, the licensee conducted a human performance error review and counseled the officer. This finding was of very low safety significance because it did not involve an actual or substantial potential of an overexposure. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2009-05648.