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	Quarter	Site	Title	Description
05000334/FIN-2018011-01	2018Q3	Beaver Valley	Duties of the Shift Technical Advisor for Control Room Evacuation during a Fire Event.	The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1(a), Procedures, related to the duties of the Shift Technical Advisor (STA) in response to a serious fire requiring control room evacuation. Specifically, procedure 1OM-56C.4.E, Shift Technical Advisors Procedure, Revision 23, directs the STA to perform substantial plant equipment operations outside of the control room (i.e., opening breakers, operating valves, electrical switching, etc.). These duties preclude the STA from maintaining sufficient independence to provide advisory technical support to the Unit 1 and 2 Operating Shift Crews as required by NOP-OP-1002 Conduct of Operations, Revision 12, and Unit 1 TS 5.2.2.f.
05000255/FIN-2018003-01	2018Q3	Palisades	Wire Not Landed on Safety Injection Initiation Relay Circuit	The inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018
05000255/FIN-2018011-03	2018Q2	Palisades	Licensee-Identified Violation	License condition 2.C(3)requires the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program that complies with Title 10of the Code of Federal Regulations(CFR), Part50.48(a) and 10 CFR 50.48(c), NFPA Standard NFPA 805, as approved in the Safety Evaluation Report (SER)dated February 27, 2015. Section 2.4.3.3 of NFPA 805 states, in part, that the Probabilistic Safety Assessment (PSA)(Probabilistic RiskAssessment (PRA))approach, methods, and data shall be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.Contrary to the above, from February 27, 2015, until May 14, 2018, the licensee failed to base the PSA (PRA) approach, methods, and data on the as-built and as-operated and maintained plant.Specifically, the licensees PSA (PRA) model/analysis credited the suppression system located in the cable spreading room to suppress a type 2 fire scenarios, whereas the actual room contained numerous obstructions by the stacked cable trays located near the ceiling that interfered with the water spray pattern discharged from the sprinklers from providing adequate water density pattern to suppress a fire in areas below the cable trays which contained electrical panels.Significance/Severity Level: The performance deficiency was determined to be more-than-minor, and therefore, a finding because the performance deficiency, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to correctly model/analyze the as-built condition of the suppression system located in the cable spreading room in the PRA could potentially affect the risk associated with a fire in the room and could result in inappropriately screening out the effects of otherchanges associated with the fire area.The finding was of very-low safety significance (Green). While there may be a change to the plants baseline risk as a result of this issue, this is a fire modeling issue only; no physical plant fire protection feature was altered by the fire PRA model. Therefore, there was no increase in actual core damage risk to the physical plant.
05000255/FIN-2018011-02	2018Q2	Palisades	Failure to Set Action Levels to Ensure that the Assumptions in the Engineering Analysis Remain Valid	The inspectors reviewed a sample of equipment located in the fire areas selectedfor inspection to determine if the licensee had established a proper method of monitoring that equipment as required by NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required that, A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee utilized Procedure EN-DC-357, NFPA 805 Monitoring Program, Revision 2,to ensure that, the assumptions in the NFPA 805 engineering analyses remain valid by executing an effective and ongoing monitoring program.The inspectors selected the high pressure air compressor (C-6B) and high pressure safety injection pump (P-66B), both of which were located in the West Safeguards Room. The licensee considered these components to be high-safety significant (HSS) structures, systems, or components (SSCs). The licensee chose to monitor the unavailability of these components utilizing the Maintenance Rule (10 CFR 50.65).The licensee set the Maintenance Rule allowable unavailability action level threshold for the high pressure air compressorat 5E-2 (5percent)whereas they assumed in their fire PRA an unavailability of 9.86E-3 (approximately 1percent). For the high pressure safety injection pump the licensee set the Maintenance Rule allowable unavailability at 1.5E-2 (1.5percent) whereas they assumed in their fire PRA an unavailability of 6.32E-3 (approximately 0.6percent). The inspectors believed that by relying on the less conservative action level thresholds in the Maintenance Rule the licensee failed to ensure that the assumptions in the engineering analysis (fire PRA) remained valid.The licensee stated in Procedure EN-DC-357, Section 1.0, Purpose, that, The NFPA 805 Monitoring Program ensures that the assumptions in the NFPA 805 engineering analyses remain valid by executing aneffective and ongoing monitoring program. Under Section 3.0, Definitions, the licensee defined, Action Level Threshold, as, When establishing the action level threshold for reliability and availability, the action level should be no lower than the Fire Probabilistic Safety Analysis (also called fire PRA) assumptions. The licensee stated in Section 5.3.3(c) that, If HSS SSCs have been identified in using the Maintenance Rule guidelines, the associated SSC specific performance criteria may be established as in the Maintenance Rule, provided the criteria are consistent with the Fire Probabilistic safety Analysisassumptions... The inspectors believed that Procedure EN-DC-357 required the licensee set the action level thresholds no lower than the fire PRA assumptions. Procedure section 5.3.4(b)(1) required that HSS equipment that is not sufficiently tracked in the Maintenance Rule be added to the NFPA 805 Monitoring Database. The licensee did not add the high pressure air compressor and the high pressure safety injection pump into the NFPA 805 Monitoring Database. In the SER 2015-2-27 dated February 27, 2015, in which the staff approved the licensee NFPA805 License Amendment Request, the staff noted that the licensee will develop an NFPA 805 Monitoring Program consistent with Frequently Asked Question (FAQ)10-0059. The staff also noted that the stated development of the Monitoring Program would include a review of existing surveillance, inspection, testing, compensatory measures, and oversight 8processes for adequacy. The staff concluded in SER 2015-2-27 that since the final values for availability and reliability, as well as the performance criteria for the SSCs being monitored, have not been established for the Monitoring Program as of the date of this SER, completion of the licensee's NFPA 805 Monitoring Program is an implementation item. Furthermore, the staff concluded that there is reasonable assurance that the licensee will develop a Monitoring Program that meets the requirements specified in Sections 2.6.1, 2.6.2, and 2.6.3 of NFPA 805Section 2.6 of NFPA 805 stated in part that, Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee interpreted this statement to mean that utilizing the existing Maintenance Rule unavailability values is consistent with its commitment in SER 2015-2-27 and would allow the site to appropriately monitor the availability and reliability of fire protection systems and features. The licensee also performed sensitivity studies on the differences in the unavailability values of fire protection systems and features between the Maintenance Rule criteria and the fire PRA values and determined that they were not risk-significant. The inspectors questioned the appropriateness of the licensees interpretation of assumptions as described in Section 2.6 of NFPA 805 above. The inspectors believed that the licensee should monitor the unavailability of fire protection systems and features utilizing the same values as thosedocumented in the fire PRA associated with the NFPA 805 License Amendment Request. The licensee further stated that they were waiting for guidance from the NRCs Office of Nuclear Reactor Regulation and the industry who were working on revising guidance in FAQ10-0059, NFPA 805 Monitoring, to determine if they needed to change their approach. That guidance document was in the process of being revised during the inspection. The inspectors needed to determine if the licensees approach to monitoring the availability and reliability of the fire protection systems and features using the Maintenance Rule monitoring values in order to ensure that the assumptions in the engineering analysis remained valid was an acceptable approach.Planned Closure Action(s): The inspectors will await clarification from the Office of Nuclear Reactor Regulation in order to determine if a performance deficiency exists.Licensee Action(s): The licensee plans to follow the resolution of FAQ 10-0059, Revision 6, and take the appropriate corrective actions based on the guidance provided in that FAQ.
05000255/FIN-2018011-01	2018Q2	Palisades	Failure to Maintain Adequate Fire Protection System Functional Test Procedure	The inspectors identified a finding of very-low safety significance and associated violation of Technical Specification 5.4.1, Procedures,for the licensees failure to maintain fire protection system functional test procedure. Specifically, the licensee failed to maintain Procedure RO-52, Fire Suppression Water System Functional Test and Fire Pump Capacity Test, by failing to include appropriate acceptance criteria in the procedure to demonstrate fire protection system functionality.
05000237/FIN-2018001-01	2018Q1	Dresden	Enforcement Action: EA18016: Unanalyzed Condition for Tornado Missiles	On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The discretion applied to Technical Specification (TS) limiting condition for operations (LCOs) that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time due to structures, system, and components (SSCs) declared inoperable because of tornado generated missile issues. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Dresden Station, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed the licensee to re-establish operability when the licensee implemented, prior to the expiration of the time mandated by the affected LCOs, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened followed by more comprehensive compensatory measures within 60 days of issue discovery. The enforcement discretion was also conditional to the comprehensive measures remaining in place until permanent repairs are completed or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Section 3.5 of the Dresden Power Station Updated Final Safety Analysis Report (UFSAR) states in part that SSCs important to safety shall be adequately protected against missiles generated by various causes, including natural phenomena. On February 12, 2018, the licensee initiated IR 04103159, identifying a nonconforming condition of Section 3.5. Specifically, the vent lines for the U2, U2/3, and U3 emergency diesel generator (EDG) fuel oil tanks were not adequately protected from tornado-generated missiles. The licensee declared fuel oil tanks and their associated EDGs inoperable, and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53204 as an unanalyzed condition and potential loss of safety function. Corrective Action(s): The licensee documented the inoperability of the SSCs in the Corrective Action Program (CAP) and in the control room operating log. In addition, the affected TS LCO conditions applicable in the mode of operation at the time of discovery were documented in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: Verifying that procedures were in place and training was current for performing actions in response to a tornado event. Verifying that procedures were in place and training was current to respond to a tornado watch, such as: (1) actions to be taken relating to tornado missile hazards; (2) potential restoration of equipment important to maintaining safe shutdown conditions that is unavailable at the time of the tornado watch; (3) warning and protection strategies for personnel; and (4) damage assessment and restorative actions for equipment that may be damaged during a tornado. Establishing a heightened level of station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify DOA001002, Tornado Warning Severe Winds procedure to include actions for damage assessment and restorative actions for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: IR 04103159
05000237/FIN-2017004-01	2017Q4	Dresden	Failure to Follow Procedure,Results in Non-Functional Fire Door	The inspectors identified a finding of very-low safety significance and associated NCV of Technical Specification 5.4.1.c for the licensees failure to implement the established Fire Protection Program procedures which ensure Fire Barrier Integrity. Specifically, the licensee ran an electrical cable through the doorway of an automatically closing fire door. This was contrary to Procedure DFPP 417501, which requires in part that fire doors must not be blocked open by props or any other material in its closing path. The licensee took immediate actions to restore the fire door, by removing the obstruction and entered the issue into their Corrective Action Program (CAP). The inspectors determined that the performance deficiency was more-than-minor because it affected the Mitigating Systems cornerstone objective since the electrical cable could have prevented the fire door from performing its function. The finding was of very-low safety significance per Task 1.4.3A of IMC 0609, Appendix F. Specifically, the total combustible loading on both sides of the affected fire door was representative of a fire duration less than 1.5 hours. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, associated with the Training component, because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee believed the performance deficiency was caused by the one of the new temporary contractors brought onto the site to work in support of the D2R25 refueling outage. (H.9)
05000461/FIN-2017008-01	2017Q2	Clinton	Failure to Perform Required S urveillances on Multiple Fire Dampers	Green . The inspectors identified a finding of very - low safety significance (Green), and an associated Non - Cited Violation of License Condition 2.C(f ) for the licensee's failure to adequately implement surveillance procedures and work processes associated with fire barrier damper inspections. Specifically, the licensee failed to perform fire barrier damper inspections for 15 fire dampers once every 4 8 months (plus an additional 25 percent grace period) as required by the Fire Protection Program. The licensee entered the issue into their Corrective Action Program , and will inspect the fire barrier dampers during the next refueling outage. The inspectors determined that the performance deficiency was more - than - minor because the licensee's failure to inspect the fire barrier dampers could result in not identifying degraded dampers which could affect their ability to prevent a fire from spreading from one fire area to an other. The finding was of very - low safety significance because the failure to inspect the fire barrier dampers did not impact the plant's ability to reach and maintain safe - shutdown. The finding has a cross - cutting aspect in the area of Human Performance, Work Management because the licensee failed to execute a work order to inspect the fire dampers in accordance with the required frequency in P rocedure CPS 9601.01 and instead improperly extended the frequency of the fire damper inspections.
05000454/FIN-2017009-01	2017Q2	Byron	Failure to Perform 10 CFR 50.59 Evaluation for UFSAR Change	The inspectors identified a Severity Level IV, Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section(d)(1) and an associated finding of very low safety significance (Green) for the licensees failure to provide a written evaluation which provided the basis for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for why a change to the surveillance frequencies of emergency diesel generators described in the Updated Final Safety Analysis Report did not require prior NRC approval.The inspectors determined that the performance deficiency was more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The associated finding screened to Green (very low safety significance) because it did not result in the loss of operability or functionality. The diesel generators passed their most recent surveillances. As a result the violation is categorized as Severity Level IV in accordance with section 6.1.d of the NRC Enforcement Policy. The issue did not have a cross-cutting aspect because it was not reflective of current performance.
05000237/FIN-2017007-02	2017Q1	Dresden	Inadequate Procedure Steps to Ensure Proper Valve Rotation for Cold Shutdown Repair	Green . The inspectors identified a finding of very - low safety significance (Green) and associated NCV of Technical Specification 5.4.1.c for the licensees failure to have appropriate written procedures covering the Fire Protection Program for cold shutdown repairs. Specifically , Procedure DSSP 0200-T8 included inadequate repair instructions for three motor operated valves ( MOVs ) that if implemented as written could result in the valve rotating in undesired safe shutdown position , caused damage to MOVs and prevented manipulating the valve to the desire position and caused a delay in reaching cold shutdown condition. The licensee entered the issue into their Corrective Action Program, revised DSSP 0200-T8 and corrected the cable designations at the Motor Control Center for these MOVs for proper connection and phase rotation. The inspectors determined that the performance deficiency was more-than- minor because the inadequate instruction in the repair procedure could have delayed reaching cold shutdown in the event of a fire and added unnecessary burden for operations personnel during an already challenging fire event. The finding was of very-low safety significance per Task 1.3.1 of IMC 0609, Appendix F , because it only affected the ability to reach and maintain cold shutdown conditions. The finding did not have a cross-cutting aspect associated with it because it was not representative of current performance .
05000237/FIN-2017007-01	2017Q1	Dresden	Inadequate Pre - Fire Plans	Green . The inspectors identified a finding of very-low safety significance (Green) and associated NCV of license conditions 2.E and 3.G for Units 2 and 3, respectively for the licensees failure to include the correct information in pre-fire plans. Specifically, the licensee failed to provide the location of compressed flammable gas cylinders and include d them in the Hazards in Area section of the pre-fire plans for two fire areas as required by Procedure OP - AA - 201 - 008, Pre-Fire Plan Manual. The licensee entered the issue into their Corrective Action Program and updated the pre-fire plans to contain the correct information. The inspectors determined that the performance deficiency was more-than-minor because the lack of information in the pre-fire plans regarding the hazards in the area could complicate firefighting activities by the fire brigade and could either increase the likelihood of a larger fire event or the severity of the fire . The finding was of very-low safety significance because it was associated with pre-fire plans and because the fire brigade members receive extensive training to deal with unexpected contingencies. The finding did not have a cross-cutting aspect associated with it because it was not representative of current performance as the licensee last updated the pre-fire plans in 2010.
05000455/FIN-2016007-01	2016Q3	Byron	Licensee-Identified Violation	The licensee identified a finding of very-low safety significance (Green) and associated Non-Cited Violation of Technical Specification 5.4.1.c which required that written procedures shall be established, implemented, and maintained covering the Fire Protection Program implementation. Procedure MA-BY-EM-1-FP003-002, Diesel Generator and Day Tank Room Low Pressure CO2 System Detection Test, was used by the licensee to test and calibrate the time delay relay that controls the time that carbon dioxide is discharged in the 2A diesel generator room when the carbon dioxide suppression system is actuated. The licensee's Fire Hazards Analysis for the 2A diesel generator room, Fire Zone 9.2-2, stated that the total flooding carbon dioxide system would deliver a sufficient quantity of carbon dioxide to maintain a 34 percent concentration for 10 minutes. Calculation BYR 97-041 established that 70 seconds was required to achieve a 34 percent concentration of carbon dioxide in the diesel generator room, and the preservice test for the carbon dioxide system for that room established that a discharge time of 99 seconds was required to ensure a 34 percent concentration for 10 minutes. Contrary to the above, from January 30, 1987, when the Byron, Unit 2 operating license was issued, until October 19, 2015, when the procedure was revised, the licensee failed to maintain a procedure that verified the capability of the carbon dioxide system. Specifically, Procedure MA-BY-EM-1-FP003-002 directed the maintenance technicians to verify the time delay relay was set between 60 and 80 seconds when the licensee's calculations required 70 seconds to achieve the required carbon dioxide concentration in the room and 99 seconds to maintain that concentration for 10 minutes. The performance deficiency was determined to be more-than-minor because the issue adversely impacted the Mitigating Systems Cornerstone objective to ensure the capability of systems that respond to initiating events and prevent undesirable consequences due to external events such as fire. Specifically, the procedure allowed the carbon dioxide system to be calibrated such that it might not have provided sufficient carbon dioxide to extinguish a fire in the 2A diesel generator room. The inspectors screened the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Fire Protection Significance Determination Process. Since the reactor was still able to reach and maintain a SSD condition, the finding screened as very-low safety significance (Green). The licensee entered the issue into the CAP as Issue Report 2571839, 2A Diesel Generator Room CO2 Discharge Time, and revised the procedure to require a 100 second discharge time.
05000315/FIN-2016009-01	2016Q3	Cook	Inadequate Resolution for Double-Break Circuits Design for Several Valves	The inspectors identified a finding of very-low safety significance (Green) and an associated Non-Cited Violation of license conditions 2.C(4) and 2.C(3)(o) for the licensees failure to implement the approved. Specifically, the licensee failed to analyze the double-break circuits design for valves using risk-informed, performance-based techniques for several fire areas. In the event of a fire in several fire areas, fire induced-circuit failures (i.e., inter-cable shorting) for a double-break design for several valves (i.e., Power Operated Relief Valves) could potentially result in spurious operation of the valves. The circuit analysis for these valves in these areas was analyzed using the deterministic approach instead risk-informed, performance-based techniques. The licensee entered the issue into their Corrective Action Program and took credit for existing fire protection features and controls as compensatory measures and planned to review the multiple spurious operations Expert Panel Report and properly disposition the scenario. The performance deficiency was determined to be more-than-minor because if left uncorrected, it would potentially lead to a more significant safety concern. Specifically, the failure to properly evaluate and disposition all potential fire-induced circuit failures for all cables in a fire area could impair the plants ability to safely shutdown in the event of a fire. The performance deficiency was also associated with the Mitigating Systems cornerstone. The finding was of very-low safety significance because it did not impact the reactors ability to reach and maintain a safe shutdown condition. This finding did not have a cross-cutting aspect because it was not representative of current licensee performance.
05000315/FIN-2016009-02	2016Q3	Cook	Licensee-Identified Violation	The licensee identified a finding of very-low safety significance (Green) and associated NCV of License Conditions 2.C.4 and 2.C.3.o for Units 1 and 2 respectively for the licensees failure to establish an appropriate Monitoring Program in accordance with NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required, in part, that monitoring shall ensure that the assumptions in the engineering analysis remain valid. Contrary to the above, the licensee failed to ensure that the assumptions in the engineering analysis remained valid for the availability and reliability of the auxiliary feedwater pumps in the Monitoring Program. The licensee used Maintenance Rule availability criteria to monitor the auxiliary feedwater pumps which did not bound the Fire PRA assumptions for the unavailability of the components. The performance deficiency was determined to be more-than-minor because the issue adversely impacted the Mitigating Systems cornerstone objective to ensure the capability of systems that respond to initiating events and prevent undesirable consequences due to external events such as fire. Specifically, the failure to adequately monitor plant equipment credited for post-fire SSD could result in that equipment being unavailable for longer periods of time than had been analyzed. The inspectors screened the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Fire Protection Significance Determination Process. Since the reactor was still able to reach and maintain a SSD condition, the finding screened as very-low safety significance (Green). The licensee entered the issue into the CAP as AR 2016-7239, NFPA 805 Monitoring Program FPRA\Maintenance Performance, and revised Maintenance Rule administrative procedures to consider the unavailability criteria of components and the impact on the fire PRA.
05000456/FIN-2016002-04	2016Q2	Braidwood	Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance	On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection, focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced enforcement guidance memorandum (EGM) 15002, which was also issued on June 10, 2015. The EGM provided guidance to allow the NRC staff to exercise enforcement discretion when an operating power plant licensee did not in comply with the current license basis for tornado-generated missile protection. Specifically, the discretion would be applied to structure system and components (SSCs) declared inoperable resulting in TS LCOs that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time. The discretion allowed the licensee to reestablish operability through compensatory measures and established criteria for continued operation of the facility as longer term corrective actions were implemented. This allows the licensee to continue operating until final corrective actions are taken in the timelines established in the EGM. The EGM stated that the bounding risk analysis performed for this issue concluded that this issue was of low risk significance and, in Braidwoods case, provided for enforcement discretion of up to 3 years from the date of issue of the EGM. However, the EGM does not provide the licensees enforcement discretion for any related underlying technical violations; and moreover, the EGM specifically requires that any associated underlying technical violation be assessed through the enforcement process. Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC), Criterion 4, Environmental and Dynamic Effects Design Basis, stated in part that SSCs important to safety shall be adequately protected against dynamic effects including missiles. On May 25, 2016, the licensee initiated IR 02673854, identifying a nonconforming condition of Criterion 4. Specifically, multiple locations were identified in the refueling water storage tank (RWST) roof hatches and in the L-line wall above the 451 elevation (separating the turbine building from the Class I auxiliary building) where SSCs were not adequately protected from tornado-generated missiles. The licensee declared multiple SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The inspectors reviewed the licensees compensatory measures that included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados/high winds, and potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The condition was reported to the NRC as Event Notice 51959 as an unanalyzed condition and potential loss of safety function. The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); 19 TS 3.3.7, Control Room Ventilation (VC) Filtration System Actuation Instrumentation; TS 3.5.2, ECCS Operating; TS 3.5.4, Refueling Water Storage Tank (RWST); TS 3.6.6, Containment Spray and Cooling Systems; TS 3.7.5, Auxiliary Feedwater System; TS 3.7.10, Control Room Ventilation (VC) Filtration System; TS 3.7.11, Control Room Ventilation (VC) Temperature Control System; TS 3.8.4, DC Sources Operating; TS 3.8.7, Inverters Operating; and TS 3.8.9, Distribution Systems Operating. The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance documentation for the EGM. The inspectors also evaluated whether the measures as implemented would function as intended and were properly controlled. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions were to remain in place until permanent repairs were completed, which for Braidwood were required to be completed in 3 years, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed. The inspectors did not review the underlying circumstances that resulted in the TS violations. As stated in the EGM guidance, violations of other requirements, including 10 CFR 50 Appendix A Criterion 4, that may have contributed to the TS violations would be evaluated independently of the EGM implementation. This operability inspection constituted a partial sample as defined in IP 71111.1505 since all the corrective actions to support continued operability and resolution of the nonconforming conditions had not been identified. These actions and any underlying technical violations will be addressed with the completion of this inspection sample.
05000456/FIN-2016002-01	2016Q2	Braidwood	Multiple Failure to Follow Procedures Leads to Inadequate Monitoring of Gas Susceptible Locations	The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to follow Revision 3 of procedure ERAA2009, Managing Gas Accumulation. Specifically, 36 gas-susceptible safety-related piping locations were not being monitored in accordance with the procedure. The planned corrective actions included an action to revise the Surveillance Frequency Control Program surveillance frequencies of accessible locations from 18 months to 6 months to align with procedural requirements, and an action to address the monitoring of locations inside the missile barrier (non-accessible locations at power). This issue was entered into the licensees Corrective Action Program (CAP) as Issue Reports (IRs) 2644532 and 2660824. The inspectors determined the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately monitor for gas accumulation in piping did not ensure the availability and reliability of systems required to perform accident mitigating functions because a potential adverse void would not be detected and assessed for operability impact. The inspectors determined that this finding was of very low safety significance because it did not result in the loss of operability or functionality of mitigating systems. Specifically, an engineering evaluation reasonably determined that the non-conforming condition did not result in a loss of operability. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risks, even while expecting successful outcomes. Specifically, the licensee had multiple recent opportunities to discover the non-compliance, but failed to do so because the licensee assumed that the surveillance frequencies were established correctly (H.12)
05000456/FIN-2016002-03	2016Q2	Braidwood	Missed Radiological Environmental Monitoring Program Sampling	The 2015 Braidwood Annual Radiological Environmental Operating Report for 2015 identified missed food samples in three out of four quadrants where food products were required by the licensees ODCM. The inspectors also noted that this issue was identified in the licensees CAP as Action Request 0214924, dated January 20, 2016, but it did not appear that any action was taken in 2015 to identify suitable alternative sampling locations. Discussion: The assessment of the issue could not be completed within this inspection period. In particular, the licensees ODCM is a site-specific document that included the radioactive effluent controls and the associated radiological environmental monitoring activities used to validate those controls. At the end of this inspection, the inspectors had not had the opportunity to review the bases documents for the ODCM to better understand the site-specific dose pathways for airborne and liquid effluent receptors and to assess the impact of these missed samples. Specifically, it was not clear whether the intended food product samples were designed to validate the airborne effluent control program or the liquid effluent control pathway. Each pathway has a different or unique requirement for validating the effluent controls. For example, validation of the airborne effluent control program would frequently measure, throughout the growing season, the radioactive material deposited on the sample surface and validation of the liquid effluent control pathway would measure, at the time of harvest, the radioactive material incorporated into the food product through irrigation. The issue remains under review by the NRC to determine the adequacy of ODCM performance and whether any violation of regulatory requirements occurred. This issue is categorized as an Unresolved Item (URI) pending completion of this review. (URI 05000456/201600203; 05000457/201600203; Missed Radiological Environmental Monitoring Program Sampling)
05000457/FIN-2016002-02	2016Q2	Braidwood	Failure to Manage Gas Accumulation in the 2A SI Train	The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to manage gas accumulation in the safety injection (SI) system in accordance with procedure ERAA2009, Managing Gas Accumulation. Specifically, following identification of a void in the 2A SI train, the licensee failed to increase the monitoring frequency and account for the potential for the void to grow due to active gas mechanisms or planned evolutions, as required by the procedure. This ultimately led to a previously identified void growing beyond the pre-established limit by the next scheduled surveillance. Corrective actions for this issue included a planned action to establish an increased monitoring frequency for the affected line, and an action to remove the void in the upcoming Unit 2 Outage (Spring 2017). This issue was entered into the licensees CAP as IR 2640751. The inspectors determined the performance deficiency was more than minor because, it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to monitor the gas accumulation for the 2A train of SI at the appropriate frequency did not ensure the availability and reliability of the SI system to perform its accident mitigating function. Additionally, this failure led to the 2A SI train exceeding the associated operability limits as established by evaluation BW150100M during the next scheduled surveillance. The inspectors determined that this finding was of very low safety significance because it did not result in the loss of operability or functionality of mitigating systems. Specifically, an engineering evaluation reasonably determined that the non-conforming condition did not result in a loss of operability. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because the licensee did not stop when faced with uncertain conditions. Specifically, the licensee did not reassess the gas accumulation monitoring plan to consider the potential for void growth due to active gas mechanisms or planned evolutions when accepting an unexpected void condition that differed with the initial conditions assumed by the monitoring plan. Ultimately, this led to a monitoring plan not being implemented as required (H.11).
05000266/FIN-2016002-01	2016Q2	Point Beach	Failure to Perform Required Fire Watches in Areas Containing Transient Combustibles	A finding of very low safety significance and associated NCV of license condition 4.F was identified by the inspectors for the licensees failure to conduct required fire watch inspections in accordance with the licensees Fire Protection Program requirements. Specifically, while conducting fire protection walkdowns of both units residual heat removal (RHR) pipeway and heat exchanger rooms, the inspectors discovered numerous transient combustible items in areas that the licensee had controlled using tamper seals on the entrances in lieu of physical entry. The licensees corrective actions included documenting and quantifying the removal of the items from the zones and additional actions to perform additional evaluation of the fire zones. The finding was determined to be more than minor because the failure to conduct the required fire watch inspections was associated with the Initiating Events cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage). Specifically, the failure to conduct the required fire watch inspections or meet the alternate measures specified by the licensees engineers, allowed unanalyzed transient combustibles and ignition sources to be present in fire zones that contained both trains of both units RHR pumps, heat exchangers and associated equipment. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue under the Phase 1 Screening Question 1.3.1A, and determined that determined that the finding was of very low safety-significance (Green), because the inspectors determined that the impact of a fire would not prevent either reactor from reaching and maintaining safe shutdown (hot). This finding has a cross-cutting aspect of Bases for Decisions (H.10), in the area of human performance, because the licensees leadership did not ensure that the bases for operational and organizational decisions are communicated in a timely manner. Specifically, the licensee did not periodically verify the understanding of the individuals assigned to fire watches, in particular, that the relief from physical entry and application of a tamper seal required a thorough tour of the zones following any entry into those fire zones.
05000266/FIN-2016002-04	2016Q2	Point Beach	Violation of Technical Specifications During Mode 4 Entry with LCO 3.6.6 Not Met	A finding of very low safety significance and associated NCV of Technical Specification 3.0.4 was identified by the inspectors for the licensees failure to follow procedure OP 1A, Cold Shutdown to Hot Standby Unit 1 and checklist CL 2C, Mode 5 to Mode 4 Checklist. Specifically, the licensee entered Mode 4 from Mode 5 without meeting the requirements of LCO 3.0.4 for entering a Mode when an applicable LCO is not met. The licensee had not met LCO 3.6.6 because the control switches for two out of the required four containment accident recirculation fans were in their pullout position instead of the required automatic position. Corrective actions for this event included restoration of accident cooler fan control switches to automatic. Additional corrective actions included: performance of an apparent cause evaluation; changes to the licensees ORT 3 test procedures to restore accident fan cooler switches after completion of testing; updating OP 1A to include performance of a control room shift turnover checklist prior to changing modes; and planned enhancements to CL 2 series procedures to strengthen a note on the responsibility of the SRO when ensuring operability of LCOs. The inspectors determined that the finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and affected the cornerstone objective of providing 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 failure to follow procedures OP 1A and CL 2C caused the licensee to unknowingly operate with multiple containment accident recirculation fans inoperable, which were required in Mode 4. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier Integrity Screening Questions, dated May 9, 2014. The inspectors answered no to the Containment Barrier Screening Questions and determined the finding had very low safety significance (Green). This finding has a cross-cutting aspect of Challenge the Unknown (H.11), in the area of Human Performance, for failing to stop when faced with uncertain conditions. Specifically, when the licensee assessed the illuminated Safeguards Equipment Locked Off alarm, during their control board walk down, they confirmed that the safety injection pump control switch was in pullout and was a reason for the alarm to actuate; however, they failed to confirm that other inputs to the alarm were also not valid.
05000266/FIN-2016002-05	2016Q2	Point Beach	Fuel Assembly Move Sequence Planned Incorrectly	A finding of very low safety significance was identified by the inspectors, for the licensees failure to follow procedure REI 26.0, Fuel/Insert/Component Movement Planning. Specifically, the licensee failed to follow procedure REI 26.0, Step 5.5.7.b, which verified that the licensee would not place fuel assemblies with cooling times less than 295 days into spent fuel pool rack foot locations. The licensees corrective actions included completing additional spent fuel moves, which placed the spent fuel pool into an appropriate configuration. The inspectors determined that the finding was more than minor, because, if left uncorrected, it had the potential to become a more significant safety concern. Specifically, if the inspectors had not questioned the licensee about spent fuel pool rack foot locations, the spent fuel pool would have remained in an incorrect configuration. The inspectors concluded this finding was associated with the Barrier Integrity cornerstone. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix L, B.5.b Significance Determination Process, Table 2 Significance Characterization, The inspectors determined that the finding did not meet the criteria in Table 2 for a Greater-Than-Green significance; therefore, the finding was of very low safety significance (Green). This finding has a cross-cutting aspect of Avoid Complacency (H.12), in the area of Human Performance, for failing to implement appropriate error reduction tools. Specifically, the licensee became desensitized to overriding fuel placement constraints and failed to implement effective human performance tools to prevent the error.
05000266/FIN-2016002-06	2016Q2	Point Beach	Incorrect Wiring Causes Transformer Lockout	A finding of very low safety significance and associated NCVs of TS 3.8.1, AC Sources-Operating and TS 3.8.2, AC Sources-Shutdown, were self-revealed for the licensees failure to follow procedure RMP 90569B, 1X03, Protective Relay Calibration and Testing. Specifically, a wiring error in the 1X03 connection box, which occurred in 2013, caused the 1X03 transformers differential protection circuity to lockout the transformer at current levels below the design protection values. The licensees corrective actions included correcting the improper wiring in the 1X03 connection box and evaluating other work performed by the same vendor during that timeframe. The inspectors determined that the finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the lockout of 1X03 caused a loss of one of the licensees offsite power lines and also caused a loss of power to multiple station battery chargers placing Unit 2 into limiting condition for operation (LCO) 3.0.3. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012. The inspectors answered Yes to the Support System Initiators question; therefore, a Detailed Risk Evaluation was required. Based on the conclusions in the Detailed Risk Evaluation, the SRA determined that the finding was of very low safety significance (Green). This finding has a cross-cutting aspect of Avoid Complacency (H.12), in the area of Human Performance, for failing to implement appropriate error reduction tools. Specifically, the incorrectly performed procedure step, in RMP 9056-9B, clearly specified which terminal point to land the wires on, the terminal points were clearly labeled, and the step required a concurrent verification; however, even with those barriers in place, the task performers still landed the wires on the wrong location.
05000266/FIN-2016002-02	2016Q2	Point Beach	Submerged Safety-Related EDG Fuel Oil Transfer Pump Cables	A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors, for the failure to maintain emergency diesel generator (EDG) fuel oil transfer pump safety-related cables in an environment for which they were designed. Specifically, the licensee allowed the safety-related cables to be submerged in water, which was outside of their design, in manhole Z066B. The licensees corrective actions included pumping the water out of the manholes, repairing the failed sump pump, level switch, and alarm circuit; and performing an engineering evaluation to quantify the level of degradation as a result of the submergence. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered "Yes" to the question does the SSC maintain its operability or functionality. Specifically, the submergence of the G01 and G02 EDG fuel oil transfer pump cables did not render the transfer pumps inoperable. This finding has a cross-cutting aspect Evaluation (P.2) in the area of problem identification and resolution, because the licensee did not thoroughly evaluate problems to ensure that resolutions address causes and extent of conditions, commensurate with their safety significance. Specifically the licensee failed to thoroughly investigate and prioritize the failure of the manhole alarm and pumping system according to the safety significance of the cables contained within the manholes which led to prolonged and unevaluated submergence of the cables.
05000266/FIN-2016002-03	2016Q2	Point Beach	Suitability of Reactor Protection System and Engineered Safeguards System Components	During the review of the Reactor Protection System (RPS), the inspectors identified an Unresolved Item (URI) associated with components in both units RPS and engineered safeguards (ESF) system which contained components known to degrade with age, including electrolytic capacitors. In some cases, these components may have been installed as original plant equipment. During the inspectors review of system health reports associated with both Units 1 and 2 RPS, and ESF system as an extent of condition review, the inspectors identified a URI associated with components in hundreds of safety-related RPS and ESF printed circuit boards, power supplies, amplifiers, transmitters, and other related components that potentially exceeded their design criteria for the time period that the components were installed for which no evaluations existed. The inspectors determined that this was an issue of concern in which more information was needed to determine if the issue constituted one or more violations of NRC requirements. Specifically, the inspectors determined that subcomponents, including but not limited to electrolytic capacitors, were installed in both safety trains of both units RPS and ESF components, in some cases for over 40 years without any documented evaluation of age-related degradation mechanisms. The inspectors needed to evaluate the licensees operability determinations that resulted from this inspection activity, any engineering evaluations to provide justification for suitability with respect to design control, recovery plans, a review of the proposed preventative maintenance activities, current failure rates and drift trending, and any other information provided by the licensee that may provide a technically defensible basis for the continued operation. The issue is unresolved pending further NRC review of the licensees evaluation.
05000282/FIN-2016008-01	2016Q1	Prairie Island	Failure to Maintain Cold Shutdown Repair Procedure	The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Technical Specifications Section 5.4.1.d for the licensees failure to maintain Procedure F5 Appendix B. Specifically, the licensee failed to update the procedure to reflect physical changes made in the plant that resulted in the licensee not being able to perform the procedure as written. The licensee entered the issue into their Corrective Action Program, and planned to update drawings and label components in the field and include the proper tools to accomplish the actions specified in the procedure. The inspectors determined that the performance deficiency was more than minor because the licensees failure to maintain Procedure F5 Appendix B would have resulted in a delay in achieving and maintaining cold shutdown. The finding was of very-low safety significance because it did not impact the licensees ability to reach hot shutdown. The finding did not have a cross-cutting aspect associated with it because it was not reflective of current performance.
05000440/FIN-2016008-02	2016Q1	Perry	Failure to Provide Instructions to Completely Vent Reference Legs	The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to follow fleet procedure NOPSS3001, Procedure Review and Approval, and to ensure that a newly developed hardcard was properly reviewed and approved prior to implementation. Specifically, the licensee characterized the hardcard development and implementation as only an administrative change, and was thereby exempted from the fleet procedure review process for new procedures. The licensee entered this finding into the corrective action program (CAP) as condition report (CR) 201603033 and planned to perform a causal review to ensure that actions taken in response to information provided in operations administrative instruction, OAI1703, Hardcards, have received appropriate review under 10 CFR 50.59. The inspectors determined that the failure to follow the licensees fleet and site-specific procedures to ensure that a newly developed hardcard was properly reviewed and approved prior to implementation was a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing review and approval activities in accordance with established procedures, the licensee might unintentionally challenge the operators by requiring equipment manipulation that impose unnecessary plant transients, which would result in unwarranted challenges to safety related equipment. Additionally, the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, and was therefore a finding. The finding was determined to be of very low safety significance because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding had a cross-cutting aspect of conservative bias in the human performance area where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, when the licensee determined to develop the hardcard procedure as an administrative change, the decision precluded the opportunity for the licensee to properly evaluate that the procedure actions did not adversely impact existing station procedures and equipment (IMC 0310, H.14).
05000440/FIN-2016008-03	2016Q1	Perry	Hardcard Development Failed to Follow Procedure Change Process	A self-revealed finding and an associated NCV of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to prescribe instructions appropriate to the circumstance into procedures for an activity affecting quality. Specifically, the licensee failed to incorporate instructions into procedures to fill and vent all portions of the reactor water level reference leg purge system. This issue has been entered the issue into the CAP as CR 201602716 to provide a process for the activities. The failure to prescribe instructions appropriate to the circumstance into procedures for an activity affecting quality was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown as well as power operations and was therefore a finding. Specifically, gas left in the reactor water level instrument reference leg purge system during maintenance equipment alignment was known to have the potential to interfere with the proper operation of pressure and level indicators relied upon for safety functions, as documented in Generic Letter 9303. The finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident (LOCA), cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined this finding had a cross-cutting aspect of challenge the unknown in the human performance area where individuals stop when faced with uncertain conditions and risks are evaluated and managed before proceeding. Specifically, the technicians involved in the April 18, 2015, system recovery activities did not stop when faced with an uncertain condition, communicate with supervisors, nor consult system experts to resolve the condition prior to continuing work activities. Since this condition was not placed into the corrective action process at the time, no further consideration was given to venting the reference leg portion of the reactor water level reference leg purge system (IMC 0310, H.11).
05000440/FIN-2016008-04	2016Q1	Perry	Failure to Maintain Traceability of Safety Related Fuses	The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 CFR 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components, for the licensees failure to assure that identification of items was maintained by appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. Specifically, the licensee failed to maintain traceability of safety related fuses installed in safety related systems. The licensee has entered this issue into the CAP as CR 201602048 and CR 201602258. Corrective actions being performed by the licensee include evaluating implementation of procedure NOPWM4300 for documenting use of parts in safety related systems and issuing work orders to determine where the potentially defective fuses were installed in the Division 2 and 3 safety related buses for replacement. The inspectors determined that the failure to assure that identification of items was maintained by appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item was a performance deficiency. Specifically, the licensee failed to maintain traceability of safety related fuses installed in safety related systems. The performance deficiency was more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, identification and control measures are designed to prevent the use of incorrect or defective materials, parts or components which could render safety systems inoperable. Additionally, the performance deficiency was more than minor because it is 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 and was, therefore, a finding. The finding was determined to be of very low safety significance because the finding was not a deficiency affecting the design or qualification of a mitigating structure system or component, did not represent a loss of system safety function, did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, and did not represent an actual loss of safety function of one or more non-technical specifications trains of equipment during shutdown for equipment designated as high safety significant for greater than 24 hours. The inspectors determined this finding had a cross-cutting aspect of documentation in the human performance area where the organization creates and maintains complete, accurate and up-to-date documentation. Specifically, a review by the licensee of existing work orders that may have utilized the fuses did not clearly document if the fuses were installed, returned to the warehouse or scrapped (IMC 0310, H.7).
05000456/FIN-2016001-04	2016Q1	Braidwood	Questions Regarding the Implementation of the Gas Accumulation Program	Questions Regarding the Implementation of the Gas Accumulation Progra The inspectors identified an URI regarding the implementation of the Gas Accumulation Program at Braidwood. Specifically, the inspectors were concerned with whether a number of surveillance frequencies that were contained in the Surveillance Frequency Program meet the requirements as specified in procedure ERAA2009, Managing Gas Accumulation. Additionally, the inspectors were concerned with the basis for not increasing the frequency of the UT examinations following the discovery of a void on October 20, 2015. At the end of the inspection period, the licensees investigation on the cause of an unexpected void growth, and the potential surveillance frequency discrepancies was ongoing. Resolution of this issue will be based on the inspectors review of the licensees completed investigation. On March 15, 2016, while performing a semi-annual gas monitoring surveillance on Unit 2 under 2BwOSR 3.2.22, ECCS and Containment Spray Venting and Valve Alignment/UT Verification Surveillance, a gas void was found along line 2SI03BA, which is a SI line that feeds the A and D SI hot leg injection lines. The ultrasonic examination revealed that a 0.960 cubic foot void was present. A void had been previously identified in the same location on October 20, 2015, which had a volume of 0.25 cubic feet. Calculation BRW150100M was performed in October 2015 to justify operability of the SI system. The calculation produced a void size acceptance criteria of 0.389 cubic feet. Upon identification of the void in March 2016, the licensee declared the 2A SI train inoperable due to the previously established acceptance criteria of 0.389 cubic feet not being met, and entered LCO 3.5.2, ECCS Operating, Condition A, which required that the affected train be restored to an operable status within 7 days. The licensee exited the LCO on March 16, 2016 upon completion of a revision to calculation BRW150100M, which documented a revised acceptance criteria of 1.5 cubic feet. During this inspection period, the inspectors reviewed the licensees revision to the aforementioned calculation, and the requirements contained in procedure ERAA2009. Based on their review, the inspectors questioned the basis for not increasing the frequency of the UT examinations following the discovery of the void on October 20, 2015. Additionally, the inspectors were concerned with the frequency of inspection of a number of locations outside the missile barrier (17 for Unit 1 and 19 for Unit 2), which appeared to conflict with what was specified in the procedure. Specifically, the locations in question were examined at an 18 month frequency, although the procedure stated that frequency of once per refueling outage shall be used only for locations that are inaccessible due to actual (not just posted) high radiation conditions. Finally, the inspectors had a concern regarding the means by which gas accumulation was managed for locations inside the missile barrier, since the prescribed locations were only monitored once upon Mode ascension from an outage. The licensee entered the inspectors concerns into their CAP as IR 2644532 and IR 2640751. At the conclusion of the inspection, two work group evaluations were in progress to: 1) address the void growth observed since October 2015, and 2) evaluate the compliance with the program document procedure, ERAA2009. This URI will remain open until the evaluations are completed and the inspectors review the evaluations to determine whether a performance deficiency exists. (URI 05000456/20160104; 05000457/201600104; Questions Regarding the Implementation of the Gas Accumulation Program)
05000440/FIN-2016008-01	2016Q1	Perry	Coincidence Logic to Preclude Spurious Trips of the Offsite Power Source	The inspectors identified an unresolved item (URI) concerning the installed designed of the safety-related 4-kilovolt under voltage protection scheme. On February 11, 2016, while the plant was shut down, a fuse failure in a bus voltage detection scheme resulted in the actuation of associated under voltage relays and a trip of the safety related EH11 bus. With the under voltage condition locked in, the Division 1 safety-related equipment remained unavailable when the Division 1 EDG started and powered up the EH11 bus. As a consequence of the under voltage scheme, the ESW pump for the EDG was not available to provide cooling water and the EDG was manually shut down by operators to prevent damage to the diesel engine. By letter dated June 3, 1977, Statement of Staff Positions Relative to Emergency Power Systems for Operating Reactors (Agencywide Document Access and Management System (ADAMS) Accession No. 8111230342), the NRC requested all licensees, including Perry Nuclear Power Station, to assess the susceptibility of Class 1E electrical equipment to sustained degraded voltage conditions from offsite power sources and to the interaction between the offsite and onsite emergency power systems. In this same letter, the NRC requested that licensees compare the current design of the emergency power systems at plant facilities with the NRC staff positions and that licensees propose plant modifications, as necessary, to meet the NRC staff positions, or provide a detailed analysis which shows that the facility design has equivalent capabilities and protective features. The NRC staff subsequently issued BTP PSB1, Adequacy of Station Electric Distribution System Voltages, Revision 0, dated July 1981 (Appendix A to Standard Review Plan Chapter 8), which provided additional guidance on technical requirements of the degraded voltage scheme. The NRC staff in addressing Perrys design for conformance with BTP PSB1, stated in Section 8.2.4.1 of the Safety Evaluation Report (SER) for Perry (ADAMS Accession No. 8211120305 dated May 31, 1982) that the applicant states in the FSAR that degraded voltage conditions for the Class 1E power system are detected by an under voltage protection system on each division... The system includes specific coincident logic for each bus. The staff also stated that the applicant has committed to provide the final design of the first and second level under voltage protection of the safety equipment in conformance with the staff position, before plant startup. Therefore, the staff finds this to be acceptable pending confirmation of the set point values and analysis. The staff identified this item as Confirmatory Issue (34) in the Section 1.0 of the SER. In Section 8.2.5 of this SER the staff also stated that on the basis of this review..., the staff concludes that the offsite power system for Perry meets the requirements of GDC 17 and 18 and is, therefore, acceptable. By letters dated June 8, 1982, and August 26, 1982, (ADAMS Accession No. 8206170298 and 820310416) the licensee provided details of the degraded voltage relay scheme to be utilized at Perry. In the August 26, 1982, letter, the licensee provided details associated with set point for the first level of under voltage protection. Specifically the licensee stated that the first level of voltage protection had been changed to trip at 75 percent of motor rated voltage instead of 86 percent and the three second fixed time delay would still remain in effect. The licensee also stated the diesel generator start signal after the 15 second time delay for the second level of under voltage had been eliminated. The licensee attached sketches to clarify the logic for a typical bus under voltage protection scheme. However, the licensee did not provide a detailed description of how the attached sketch satisfied the requirements of BTP PSB1. The August 1982 letter again states that the final design and set points will be established after a review of onsite preoperational test results. In a Supplemental Safety Evaluation Report (SSER 2) (ADAMS Accession No. 8301280169 dated January 13, 1983), the NRC staff acknowledged that based on information provided in the August 26, 1982 letter, the relays in the loss of power protection relays were arranged on a two-out-of-two coincident logic to initiate a timer with a 3-sec time delay. The SSER also acknowledged that in the degraded grid voltage protection, the relays were arranged in a two-out-of-two coincident logic to initiate two separate time-delay relays. The staff stated that the applicants final design of the first and second level under voltage protection of safety equipment was acceptable; thus Confirmatory Issue (34) listed in Section 1.10 of the SER, was satisfactorily resolved. The licensee revised the USAR to document conformance with BTP PSB1 and the requirements of Institute of Electrical and Electronics Engineers (IEEE) Std. 2791971, Criteria for Protection Systems for Nuclear Power Generating Stations. Specifically, USAR Section 8.3.1.1.2.9.2 states that under voltage protection shall include coincidence logic on a per bus basis to preclude spurious trips of the offsite power source. This is in conformance with staff guidance stating that improper voltage protection logic can itself cause adverse effects on the Class 1E systems and equipment such as spurious load shedding of Class 1E loads from the standby diesel generators and spurious separation of Class 1E systems from offsite power. The inspection team noted that the licensees installed design utilizes a single voltage sensor that is shared between two relays in the under voltage trip logic. As a consequence of a shared voltage sensor between both under voltage relays, the overall under voltage protection did not constitute a coincidence logic as stated in BTP PSB1. A single malfunction in the voltage sensing circuit resulted in a trip of the offsite power source and precluded the onsite power source from performing its safety related functions. As evidence by the February 11, 2016, event, a secondary potential transformer fuse failure caused the EH11 bus to separate from offsite power (even though there were no deficiencies in offsite power voltage) and resulted in shutdown of the division 1 EDG bus due to ESW being unavailable. The inspectors believe that this design is deficient in that there is no coincident logic to ensure that spurious trips are precluded as delineated in BTP PSB1 Section B.1.c. The licensee agrees that the issue relative to the failure of a single fuse resulting in actuation of both relays, satisfying the under voltage protection scheme logic, represented a design vulnerability. However, the licensee contended that this original design had been maintained, and was approved by the NRC during the initial licensing. Specifically, the licensee concluded, in a white paper addressing the SIT concerns, that The current design of the potential transformers fusing and the under voltage relays for the division 1 4160 V ESF bus is reflective of the original design and licensing basis. The NRC Staff reviewers clearly used PSB1 in their review of the PNPP licensing and design bases. They were aided by the schematics contained in the following letter (First Energy Nuclear Operating Company (FENOC) letter docket Nos. (50440; 50441), dated August 26, 1982). They recognized and approved of the relays arrangement in a two-out-of-two coincident logic. (This is identified on the second schematic - reference Sheet 2 of 2 on lines 19 and 21, which address the time delay relays.) This design is reflective of PNPP's current design and is consistent with the original PNPP licensing basis. Additionally, the NRC stated in the SER that Perry meets the requirements of GDC 17 and 18. This issue is considered an unresolved item (URI 05000440/20160801) pending receipt of clarification of the design basis with respect to 4 Kilovolt safety bus under voltage protection scheme.
05000440/FIN-2015008-03	2015Q4	Perry	Failure to Provide Adequate Guidance to Override Spurious CO2 Initiation Signal in the Diesel Generator Rooms	The inspectors identified a finding of very low safety significance (Green), and an associated NCV of TS Section 5.4.1.a for the licensees failure to have adequate procedural guidance in their fire response procedure. Specifically, Procedure ONI-P54, Fire, Revision 19 did not list all the fire areas where a potential fire induced spurious carbon dioxide (CO2) initiation in the emergency diesel generator (EDG) room could occur. The licensee entered this issue into their CAP, and established hourly fire watches for the affected areas. The inspectors determined that the performance deficiency was more than minor because a fire in any of the affected fire zones could damage circuits for the nonsafety-related CO2 systems for the EDG rooms causing a potential spurious CO2 initiation in the diesel rooms and affecting the operation of the ventilation fans and dampers in the diesel rooms. The finding was of very low safety significance because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The finding did not have a cross-cutting aspect associated with it because it was not reflective of current performance.
05000440/FIN-2015008-02	2015Q4	Perry	Failure to Inspect Penetration Seals Within the Required Time Frequency	The inspectors identified a finding of very low safety significance (Green), and an associated NCV of license condition 2.C(6) for the licensees failure to adequately implement and maintain surveillance procedures and work processes associated with fire barrier and penetration seal inspections. Specifically, the licensee failed to perform fire barrier penetration seal inspections for 42 penetration seals at least once per 15 years (plus an additional 25 percent grace period) as required by the Fire Protection Program. The licensee entered the issue into their CAP, and will inspect the accessible portions of the barriers and will perform a full inspection at the next available opportunity. The inspectors determined that the performance deficiency was more than minor because the licensees failure to inspect the fire barrier penetrations could result in not identifying degraded seals which could affect their ability to prevent a fire from spreading from one fire area to another. The finding was of very low safety significance because the failure to inspect a portion of fire barrier penetration seals did not impact the plants ability to reach and maintain safe shutdown. The finding has a cross-cutting aspect in the area of Human Performance, Work Management because the licensee improperly closed a notification to track the inspection of fire barrier penetrations without creating a work order.
05000440/FIN-2015008-01	2015Q4	Perry	Failure to Ensure that Systems, Structures, and Components Necessary to Achieve and Maintain Hot Shutdown Conditions were Free of Fire Damage without Repair Actions	The inspectors identified a finding of very low safety significance (Green), and associated NCV of license condition 2.C(6) for the licensees failure to ensure that systems, structures, and components necessary to achieve and maintain hot shutdown conditions were free of fire damage. Specifically, the licensee did not ensure that circuits associated with the emergency closed cooling (ECC) heat exchanger A temperature control valve 1P42-F665A were free of fire damage for a fire in the control room and instead relied on lifting leads and replacing fuses to take manual control of the valve. The licensee entered the issue into their CAP, and credited the existing repair activities in the procedure. The inspectors determined that the performance deficiency was more than minor because a fire in the control room could result in the licensee losing the ability to remotely control the ECC heat exchanger A temperature control valve and needing to take manual control of the valve. The finding was of very low safety significance because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The finding did not have a cross-cutting aspect associated with it because it was not reflective of current performance.
05000440/FIN-2015008-04	2015Q4	Perry	Failure to Perform Fire Watches	The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specifications (TS) Section 5.4.1.a for the licensees failure to perform fire watches in two fire areas for a non-functional fire barrier. Specifically, the licensee failed to perform fire watches as required by Section 16.D(1)a.(1) of Attachment 3 to procedure PAP-1910, Fire Protection Program. The licensee entered the issue into their Corrective Action Program (CAP), and added the two fire areas to the fire watch list. The inspectors determined that the performance deficiency was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, by failing to perform fire watches the licensee may not have been able to identify transient combustible materials that could have impacted the unprotected circuits associated with this deficiency in the event of a fire. This finding was of very low safety significance because it only impacted one train of equipment important to safety. This finding has a cross-cutting aspect in the area of Human Performance, Documentation because the licensee did not create and maintain complete, accurate, and up-to-date documentation. Specifically, when the licensee developed the fire watch list they did not include all impacted fire zones as listed in the initial impairment.
05000373/FIN-2015010-01	2015Q3	LaSalle	Severity Level (SL)-III Violation (Enforcement Discretion	A SL-III violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, Criterion 17, Electric Power Systems, was self-revealed on April 17, 2013, when a lightning strike and corresponding C phase fault on 138kV line 0112 occurred in the 138kV switchyard and caused a dual unit loss of offsite power (LOOP). Specifically, the lightning strike and corresponding phase to ground fault caused flashovers due to improperly installed ground segments in the 138kV switchyard grounding system installation. The licensee performed a Root Cause Investigation under report number 1503409-04, and identified several deficiencies in the quality of construction of the 138kV switchyard grounding system that allowed a lightning induced fault to flash over onto the DC protective system. The degradation was due to inadequate workmanship during initial installation of the 138kV ground system in the 138kV switchyard. In accordance with Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that this issue did not meet the definition of a performance deficiency since it was not reasonably within the licensees ability to foresee and correct. However, this issue was a violation of 10 CFR 50, Appendix A, Criterion 17, Electric Power Systems. Therefore, the traditional enforcement process was used in accordance with the NRC Enforcement Policy. The violation was similar to one example in the Enforcement Policy under SL-III violations, Section 6.1(c) (5), Equipment failures caused by inadequate or improper maintenance substantially complicate recovery from a plant transient. Although the improper grounding system was a result of improper installation rather than improper maintenance, the inspectors determined that the violation met the intent of the example. The NRC determined that this violation resulted from matters not reasonably within the licensees control; that is, the failure to meet the requirements could not be readily identified and addressed. Therefore, in accordance with the Enforcement Policy, and after consultation with the Director of the Office of Enforcement, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and to refrain from issuing enforcement action for the violation. In accordance with the NRCs Reactor Oversight Process, this condition will not be considered in the assessment process or the NRCs Action Matrix.
05000456/FIN-2015007-01	2015Q3	Braidwood	Failure to Ensure that Circuits Associated with Pressurizer PORVs and Block Valves Were Free of Fire Damage	The inspectors identified a finding of very low safety significance, and an associated NCV of the Braidwood Station facility operating license condition 2.E associated with the Fire Protection Program for the licensees failure to ensure that the safe shutdown capability was independent of the fire area and thus free of fire damage. Specifically, in the event of a fire in the control room, cable spreading rooms, or electrical cable penetration areas the circuits associated with the Pressurizer Power Operated Relief Valve (PORV) block valves, which are relied upon to safely shutdown the plant, could be affected and may not be available due to fire-induced failures. The licensee entered this issue into their Corrective Action Program, established fire watches, and intended to perform plant modifications to correct the issue. The inspectors determined that the issue was more than minor because fire-induced circuit failures could impair the operation of the PORV block valves and complicate shutdown of the plant in the event of a fire in the control room, cable spreading rooms, or electrical cable penetration areas. The finding affected the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance based on a detailed risk-evaluation by a Region III Senior Reactor Analyst. This finding was not associated with a cross-cutting aspect because the finding was not representative of the licensees current performance.
05000237/FIN-2015007-01	2015Q2	Dresden	Procedure Revisions Resulted in Isolation Condenser Unable to Meet Design Basis	The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that applicable regulatory requirements and the isolation condensers (ICs) design bases were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the ICs design bases. The licensee entered the issue into their Corrective Action Program (CAP) as Action Request 02506445, NRC MOD/5059 Inspection: ISCO (Isolation Condenser) Operating Procedures, dated May 28, 2015. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality, and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedures would drive the operators to stop the IC during a design bases event and prevent the IC from performing its design function of removing decay heat from the reactor. The finding has a cross-cutting aspect in the area of Human Performance; Teamwork, because the licensee did not communicate and coordinate activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Operations Department failed to communicate and coordinate with the Engineering Department when developing the procedural changes. (H.4)
05000255/FIN-2015008-01	2015Q2	Palisades	Failure to Correctly Assess the Suppression System in the Cable Spreading Room in the Probabilistic Risk Assessment for NFPA 805	The inspectors identified a finding of very-low safety significance, and an associated NCV of Title 10, Code of Federal Regulations (CFR) 50.48(c), and National Fire Protection Association Standard 805, Section 2.4.3.3 for the licensees failure to correctly model the as-built plant in the Fire Probabilistic Risk Assessment (PRA). Specifically, the licensee credited the suppression system located in the cable spreading room in the PRA to suppress type 2 fire scenarios, whereas the actual room contained numerous obstructions due to the stacked cable trays located near the ceiling that interfered with the water spray pattern discharged from the sprinklers. These obstructions could have prevented the suppression system from providing an adequate water density pattern to suppress a fire below the cable trays in areas which contained electrical panels. The licensee entered this issue into their Corrective Action Program, and already had compensatory measures in place in the cable spreading room, including hourly fire tours and a standing order for an immediate call out for the fire brigade for a fire alarm in the room. The inspectors determined that the performance deficiency was more than minor because the finding, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to correctly model/analyze the as-built condition of the suppression system located in the cable spreading room in the PRA could potentially affect the risk associated with a fire in the room, and could result in inappropriately screening out the effects of other changes associated with the fire area. Appendix M was used because the existing SDP Appendices do not adequately address the risk of performance deficiencies associated with licensees PRAs. The Senior Reactor Analyst concluded that the finding was of very-low safety significance (Green) because while there may be a change to the plants baseline risk as a result of this issue, there is no delta plant risk due to a deficiency in the licensees PRA model/analysis. This finding has a cross-cutting aspect in the area of Human Performance associated with Team Work because the licensee did not communicate and coordinate activities between the PRA and the fire protection groups.
05000237/FIN-2015007-02	2015Q2	Dresden	EDG Usable Fuel Calculations Failed to Consider Appropriate EDG Frequency Variations	The inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to account for increased fuel oil consumption during the development of the Emergency Diesel Generator (EDG) Calculation 10553-CALC-07, Dresden Station Emergency Diesel Generators Endurance Calculations, Revision 2, which resulted in non-conservative Technical Specifications (TS). Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hertz (Hz), and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour. The licensee entered the issue into their CAP as Action Request 02506869, NRC MOD/5059 Inspection: Emergency Diesel Generator Fuel Consumption, dated May 28, 2015. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to account for the increased fuel oil consumption resulting from operation at a higher EDG frequency. Therefore, the licensee did not ensure that the minimum fuel oil level in the day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution; Identification, because the licensee did not did not thoroughly evaluate the EDG fuel oil consumption when considering EDG frequency variation. Specifically, the licensee failed to translate applicable design bases into specifications which resulted in non-conservative TS. (P.1)
05000331/FIN-2015001-03	2015Q1	Duane Arnold	Licensee-Identified Violation	Duane Arnold TS 5.4, Procedures, Section 5.4.1.a requires, 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, Revision 2, Appendix A, contains, in part under Section 8.b(2)(t), surveillance test procedures for inspection of the reactor coolant system pressure boundary. Contrary to the above, on November 8, 2014, the licensee failed to properly implement surveillance test procedure (STP) 3.10.102, Non Nuclear Heat Class 1 Ten Year System Leakage Pressure Test, Revision 32. Specifically, during the Fall 2014 refueling outage, licensee personnel identified leakage during visual undervessel inspections per STP 3.10.102. Although several CRs were generated to capture the identified leakage locations and approximate leakage rates from control rod drive mechanism (CRDM) flanges, the personnel failed to fully implement STP 3.10.102, Attachment 3 requirements to perform a detailed inspection of the associated CRDM flanges to identify the leakage source and to verify pressure boundary integrity. Had this identification/verification been performed, STP 3.10.1-02, Attachment 3, further required implementation of GMPTEST66, CRD (-) Troubleshooting Procedure, Revision 8, for CRDM flange leakage. Because CRs were written, the licensee personnel considered the under-vessel inspection results satisfactory and moved forward in the STP. Upon further review of the completed STP, the licensee identified that required detailed inspections were not performed for the CRDM flange leaks. The licensee entered the issue into the CAP and successfully re-performed STP 3.10.102 after resolving the leakage issues. Because the inspectors answered No to all questions under Exhibit 4 of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02006364. Corrective actions included a revision to STP 3.10.102 to more clearly define under-vessel visual inspection requirements.
05000373/FIN-2014008-01	2015Q1	LaSalle	Failure to Ensure Circuits associated with Alternate Shutdown Capability Free of Fire-induced Da	The inspectors identified a finding of very-low safety significance (Green) and associated NCV of the LaSalle County Station Operating License for the licensees failure to ensure that the alternate shutdown capability was independent of the fire area. Specifically, in the event of a fire in the control room, the alternate shutdown capability for 16 motor operated valves (MOVs) associated with the Reactor Core Isolation Cooling (RCIC) may be affected, and may not be available due to lack of breaker fuse coordination. Fire-induced failures could result in tripping valve power supply breakers prior to tripping the control power fuses for several motor operated valves, thereby, potentially imparing the operation of RCIC from the Remote Shutdown Panel (RSP). The licensee entered this issue into their Corrective Action Program and established compensatory measures, and added steps to the safe shutdown procedures to reset the affected breakers if needed. In addition, the licensee intended to perform plant modifications to replace or revise existing breakers settings to correct the issue. The inspectors determined that the issue was more than minor, because fire induced circuits could impair the operation of RCIC and complicated shutdown of the plant in the event of a fire in the control room. The finding affected the Mitigating Systems Cornerstone. The finding was determined to be of very-low safety significance based on a detailed risk-evaluation. This finding was not associated with a cross-cutting aspect because the finding was not representative of the licensees current performance.
05000331/FIN-2015001-04	2015Q1	Duane Arnold	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to the above, on September 27 and 29, 2014, the licensee failed to prescribe an instruction appropriate to the circumstances associated with the replacement of shielded cables between the A and C RHRSW pump motors and the associated 4kV supply breakers. Specifically, SPECE512, Cable and Wire Installation, Revision 14, did not ensure that shielded cables be grounded only at the switchgear end, and that the cables be routed back through ground fault (ring) current transformers in the cabinet before being grounded. This resulted in the improper development of work instructions used in the installation of replacement cables for the A and C RHRSW pumps and a resultant non-conforming condition which was discovered by the licensee during an extent of condition review in March of 2015. Because the SSCs maintained operability based on the deficiency affecting the design of the SSCs, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02023605. Immediate corrective actions included a determination of operability (the ground fault protection had no required safety supporting function for the RHRSW pumps and switchgear), equipment configuration control until resolution was taken, re-routing of the affected cables to restore full design, and a revision to SPEC-E512 to clearly describe shielded cable installation requirements.
05000331/FIN-2015404-01	2015Q1	Duane Arnold	Security
05000331/FIN-2015008-01	2015Q1	Duane Arnold	Failure to Identify and Evaluate the Effects of Vessel OverfillScenario	The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR) 50.48(c), and National Fire Protection Association Standard 805, Section 2.4.3.2 for the licensees failure to address in the Fire Probabilistic Risk Assessment (PRA) the risk contribution with all potentially risk-significant fire scenarios. Specifically, the licensee did not address potential damage to safety relief valves (SRVs), or the SRV tailpipes as a result from fire induced overfill of the reactor pressure vessel. The licensee entered this issue into their Corrective Action Program to review the multiple spurious operations Expert Panel report, and properly disposition the scenario. The inspectors determined that the performance deficiency was more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors (i.e., fire), and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the missed failure mechanism for the SRVs had the potential to impact the ability to achieve safe and stable conditions. In accordance IMC 0609, Appendix F, Fire Protection SDP, Attachment 1, Step 1.6.1, Screen by Licensee PRA-Based Safety Evaluation, the inspectors were able to use the Licensees PRA to evaluate the safety significance of the finding. The increase in core damage frequency (CDF) as a result of the identified scenario was found to be approximately 2.6E-7 per year; therefore, the inspectors concluded that this finding was of very-low safety significance (Green). This finding did not have a cross-cutting aspect because it was not representative of current licensee performance.
05000331/FIN-2015001-01	2015Q1	Duane Arnold	Failure to Classify and Declare a Notification of Unusual Event	The inspectors identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50.54(q)(2), and 10 CFR 50.47(b)(4) for the failure of the licensee to classify and declare a Notification of Unusual Event. Specifically, on June 30, 2014, the licensee failed to classify and declare a Notification of Unusual Event after a control room instrument peaked at a wind speed that exceeded the Unusual Event Emergency Classification Level threshold for 4 seconds. The licensee entered the issue into the corrective action program (CAP) as condition report (CR) 01975495. Corrective actions included procedure changes to ensure available indications for wind speed are monitored during high wind events. The failure to classify and declare a Notice of Unusual Event when conditions warranted was a performance deficiency. The finding was more than minor because it adversely affected the emergency response organization (ERO) performance attribute of the Emergency Preparedness (EP) cornerstone objective to ensure that licensees are capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Because the finding only involved a failure to declare a Notification of Unusual Event, the finding screened as being of very low safety significance (Green). This finding was associated with the cross-cutting aspect of avoid complacency in the area of Human Performance, because control room operators did not walk-down instrumentation that was available to them in the control room. (H.12)
05000331/FIN-2015001-02	2015Q1	Duane Arnold	Failure to Report Required Monitoring Results to the NRC	The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 20.2206 for the licensees failure to report results of individual radiation exposure monitoring for individuals required to be monitored by 10 CFR 20.1502. Specifically, on or before April 30, 2014, the licensee failed to report results for all individuals requiring monitoring for the calendar year 2013 to the NRCs Radiation Exposure Information and Reporting System (REIRS) database. The issue was entered into the licensees CAP as CR 02028468. Immediate corrective actions included the resubmittal of radiation exposure data to the REIRS database, which included radiation exposure for all individuals that were required to be monitored. The violation of 10 CFR 20.2206 was assessed in accordance with the traditional enforcement path in IMC 0612, Appendix B, Issue Screening. The inspectors determined that traditional enforcement did apply because reporting failures impact the regulatory process. In accordance with the NRC Enforcement Policy, Section 6.9(d)(2), failures to make a timely written report as required by 10 CFR 20.2206 are categorized as SL IV violations. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000341/FIN-2014007-01	2014Q3	Fermi	Incorrect Valve Location in Procedure	The inspectors identified a finding of very low safety significance (Green) and associated NCV of Technical Specifications (TS) Section 5.4.1.a for the licensees failure to maintain Procedure 20.000.23, High RPV (Reactor Pressure Vessel) Water Level to address an RPV overfill event. Specifically, the licensee provided an incorrect location of a manual valve in the Standby Feedwater (SBFW) system. The procedure described the valve as being located in the turbine building basement, while the valve was actually located in a locked high radiation area in the north heater room. The licensee revised the procedure to include the correct location of the valve. The inspectors determined that the issue was more than minor because a reactor overfill event could impair the RCIC and HPCI systems during a fire in fire zone RW. The finding affected the Mitigating Systems cornerstone. The finding was determined to be of very low safety significance based on a detailed risk-evaluation. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not take effective corrective actions to address a potential reactor pressure vessel overfill event.
05000282/FIN-2014008-02	2014Q2	Prairie Island	Licensee-Identified Violation	The licensee identified a Severity Level IV violation of 10 CFR 50.59, (Changes, Tests, and Experiments, for the failure to demonstrate in a written evaluation that prior NRC-approval was not required for changes made to an accident analysis. Specifically, the licensee incorrectly concluded in written Evaluation 1102, Waste Gas Tank Rupture Dose Analysis, Revision 0 that higher activity levels and dose rates at the Exclusion Area Boundary and Low Population Zone associated with extended plant life due to license extension did not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR. The performance deficiency was determined to be more than minor because it was associated with the Radiation Safety cornerstone attribute of program and process and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors determined the violation was of Severity Level IV because the associated finding was of very low safety significance (Green) as there was no actual radioactive material release. The licensee entered this issue into their Corrective Action Program as AR 1417573 and AR 1427150 and intended to submit a license amendment request for review by the NRC.
05000237/FIN-2014008-02	2014Q2	Dresden	Failure to Seismically Secure Nitrogen Bottles	The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specifications (TS) Section 5.4.1.a, for the licensees failure to seismically restrain nitrogen bottles located near safety-related motor control centers (MCCs). Specifically, the licensee failed to seismically restrain a cart with two nitrogen bottles located near safety-related MCCs per their procedures for the handling and storage of compressed gas cylinders and restraint of portable equipment. The licensee entered this issue into their corrective action program, moved the cart with the nitrogen bottles away from the MCCs, and secured it to a column nearby. The inspectors determined that the finding was more than minor because during a seismic event the bottles could have tipped over and impacted the MCCs, thereby causing a loss of safety-related equipment, such as the Unit 2/3 emergency diesel generator. The finding was determined to be of very low safety significance based on a detailed risk-evaluation. The finding has a cross-cutting aspect in the area of Human performance because maintenance and operations personnel did not coordinate during a change out of nitrogen bottles which resulted in the bottles being left unsecured. (H.5)
05000237/FIN-2014008-01	2014Q2	Dresden	Inadequate Applicability Reviews of Configuration Changes for De-Energizing Safety-Related Valves	The inspectors identified a finding of very low safety significance (Green) related to inadequate applicability reviews of operational configuration changes that were implemented as a result of the licensee's Multiple Spurious Operation (MSO) evaluations. Specifically, the licensee failed to follow procedural requirements for determining the applicability for performing 10 CFR 50.59 screening and evaluations for changes made to the facility which de-energized several safety-related motor operated valves (MOVs). The procedural action required that the configuration changes be screened for applicability for a specific 10 CFR Part 50.59 evaluation since aspects of the changes were not completely controlled under the licensee's Fire Protection Program. The licensee entered this issue into their Corrective Action Program to perform a 10 CFR 50.59 screening of changes for each affected system to ensure that all aspects of component design were evaluated. The performance deficiency was determined to be more than minor because the issue, if left uncorrected, would have become a more significant safety concern. Specifically, by not individually evaluating the scope and applicability of plant configuration changes, the licensee lost the ability to ensure that all aspects of component design were appropriately evaluated against the plant's design and licensing basis. Such changes have the potential to adversely affect design or operation of systems. Failure to evaluate such aspects allows the potential for adverse changes to go undetected. This finding has a cross-cutting aspect in the area of Human Performance because the licensee became complacent during the conduct of performing applicability reviews that were related to the facility's Fire Protection Program, and failed to recognize changes that included elements outside of the scope of fire protection. (H.12).

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