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	Quarter	Site	Title	Description
05000338/FIN-2018003-01	2018Q3	North Anna	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2.a of the Enforcement Policy. Violation: TS 5.4.1.a, requires in part, that written procedures shall be established per Revision 2 of Regulatory Guide 1.33, Appendix A, of which part 9.a requires written procedures and documented instructions appropriate to the circumstances for performing maintenance that can affect the performance of safety related equipment. Contrary to the above, on June 12, 2018, the licensee failed to adequately establish a procedure appropriate to the circumstances during maintenance on the safety-related main control chillers. Specifically, licensee mechanical preventative maintenance procedure, 0-MPM-0806-02, Inspection of Control Room Chillers, Revision 0, did not provide a proper method to adequately monitor the Freon level in main control room chillers. Consequently, the licensee discovered a low Freon level condition on main control room chiller 1-HV-3-4B, which rendered the chiller inoperable. Significance: The inspectors reviewed Exhibit 2 Mitigating Systems Screening Questions of IMC 0609 Appendix A, The Significance Determination Process (SDP) for findings at Power and determined this finding was of very low safety significance, Green, because there was no design deficiency, it did not represent a loss of system or function, and did not represent an actual loss of function for greater than its TS allowed outage time. Corrective Action Reference: CR109958
05000400/FIN-2018003-01	2018Q3	Harris	Failure to Implement Adequate Periodic Exercising of Turbine Trip Solenoid Operated Valves	A self-revealing Green finding was identified for the licensees failure to establish and implement adequate preventive maintenance (PM) for exercising the turbine electro-hydraulic auto-stop trip (AST) solenoid operated valves (SOVs) in accordance with procedure AD-EG-ALL-1202, Preventive Maintenance and Surveillance Testing Administration. As a result of the failure to exercise the SOVs at the weekly vendor recommended frequency, three of the four SOVs experienced mechanical binding (sticking) which rendered the turbine emergency trip system incapable of tripping the main turbine within the time response requirements of Technical Specifications.
05000400/FIN-2018002-03	2018Q2	Harris	Failure to Adequately Document Changes to the Emergency Plan	The inspectors identified multiple examples of a Severity Level IV (SL-IV) NCV of 10 CFR 50.54(q)(3), for changes to the licensees radiological emergency plan (E-Plan) associated with protective action recommendation (PAR) procedures and emergency response equipment that failed to demonstrate that the changes would not reduce the effectiveness of the E-Plan. Specifically, the licensee did not provide an adequate analysis to demonstrate that the removal of the sheltering in-place PARs was not a reduction in effectiveness of the E-Plan. Additionally, the licensee did not perform an analysis demonstrating that the removal of a temporary diesel generator providing a backup source of power to the Technical Support Center (TSC) did not reduce the effectiveness of the E-Plan.
05000400/FIN-2018002-01	2018Q2	Harris	Failure to Promptly Identify and Correct a Condition Adverse to Quality For a Through-Wall Leak in the ESW Screen Wash Piping	An NRC-identified Green NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Actions, was identified for the licensees failure to promptly identify and correct a condition adverse to quality involving through-wall leakage in the B train ESW screen wash piping. Specifically, on April 30, 2018, operators failed to initiate a work request or condition report after security personnel reported through-wall leakage in the B train ESW screen wash piping. No further follow-up or corrective actions were taken until May 3, 2018, when NRC inspectors identified the same through-wall piping leakage during a plant walkdown inspection and reported the degraded condition.
05000400/FIN-2018002-02	2018Q2	Harris	Inadequate Fire Brigade Performance Assessment of Announced Fire Drill	An NRC-identified Green NCV of 10 CFR 50.48(c) and National Fire Protection Association (NFPA) Standard 805, Section 3.4.3, Training and Drills, was identified for the licensees failure to adequately assess the fire brigade performance during an announced fire drill conducted March 21, 2018. Specifically, the inspectors identified several fire brigade performance deficiencies, improvement items, and lessons learned that were not identified and documented in the licensees corrective action program during the fire drill critique as required by the licensees fire drill administrative control procedure.
05000400/FIN-2018002-04	2018Q2	Harris	Failure to Implement Adequate Steam Generator Blowdown Demineralizer Control Procedures	A self-revealing Green NCV of Technical Specifications (TS) 6.8.1.a, Procedures and Programs, was identified for licensees failure to establish and implement adequate steam generator blowdown demineralizer control operating procedures resulting in exceeding secondary water chemistry Action Level 3 criteria for impurities in the steam generators. Specifically, the licensee did not implement adequate isolation valve controls between the demineralizer resin regeneration system and the feedwater system during resin regeneration activities. This open path allowed leakage of sulfates and chlorides into the feedwater system. The level of these impurities exceeded the secondary chemistry Action Level 3 threshold and resulted in an unplanned shutdown.
05000400/FIN-2018002-05	2018Q2	Harris	Failure to Follow Secondary Water Chemistry Plan for Elevated Levels of Secondary Water Impurities	An NRC-identified Green NCV of TS 6.8.4.c, Secondary Water Chemistry, was identified for the licensees failure to follow secondary water chemistry control requirements in accordance with procedure CSD-CP-HNP-0002, Harris Secondary Water Chemistry Strategic Plan. . Specifically, the licensee remained at 100% power for approximately 10 hours after entering secondary water chemistry Action Level 3 due to elevated chlorine and sulfates chemical impurity concentrations, which was contrary to the procedure requirements to downpower the unit to below 5% power as quickly as safe plant operation permits. This unit downpower delay allowed additional time for the chemical impurities to adversely affect the steam generators.
05000400/FIN-2018002-06	2018Q2	Harris	Failure to Implement Viable Compensatory Actions with Seismic Monitoring System Out of Service for Planned Preventive Maintenance	An NRC-identified Green NCV of 10 CFR 50.54(q)(2) was identified for the licensees failure to follow and maintain the effectiveness of its emergency plan that meets the requirements of the risk-significant emergency planning standard 10 CFR 50.47(b)(4). Specifically, the licensee failed to implement viable compensatory actions while conducting planned preventive maintenance that rendered both seismic monitoring systems unavailable for 53.5 hours resulting in a loss of emergency assessment capability for declaring a Notification of Unusual Event under Emergency Action Level (EAL) HU2.1 for a seismic event.
05000400/FIN-2018002-07	2018Q2	Harris	Minor Violation	A minor, self-revealing violation of TS 6.8.1.a, Procedures and Programs,was identified for failure to follow procedure AD-OP-ALL-0200, Clearance and Tagging. On April 7, 2018, while the plant was in Mode 3 at 0 percent power, the licensee isolated breaker DP-1A-1 circuit 28 in accordance with clearance OPS-1-18-5015-DEH MODS-0093. Isolating this breaker caused an unexpected auto start signal for both motor driven auxiliary feedwater (MDAFW) pumps for a loss of last running main feed pump despite the 1B main feedwater pump still being in operation. Both MDAFWs started and operators manually secured the 1B main feedwater pump to maintain proper feedwater flow to the steam generators. TS 6.8.1.a, requires, in part, that written procedures be implemented covering activities referenced in Regulatory Guide 1.33, Revision 2, dated February 1978, including safety-related activities carried out during operation of the reactor plant. Procedure AD-OP-ALL-0200, Section 5.5, step 4, states Clearance impacts must be evaluated to ensure that effects on systems and components outside of the boundary are identified and are acceptable, or properly dispositioned. Contrary to this requirement, the licensee did not identify that the isolation of breaker DP-1A-1 circuit 28 would cause the MDAFWs to auto start in Mode 3 when developing clearance OPS-1-18-5015-DEH MODS-0093. Screening: The violation is minor because the impact to the plant was minimal; the unit was in Mode 3 throughout the event, the reactor remained subcritical, and feedwater flow to the steam generators was not lost. Enforcement: Because the performance deficiency is minor, it will not be subject to enforcement action in accordance with the NRCs Enforcement Policy. The licensee entered this issue into their CAP as NCR 02196873. The associated LER is closed.
05000400/FIN-2018001-01	2018Q1	Harris	Adequacy of Fire Brigade Response During Fire Dril	The inspectors identified an URI during the March 21, 2018, announced fire drill that was observed. The drill involved an electrical failure inside the A transfer panel located in the RAB 286 elevation A cable spread room. The fire scenario assumed the electrical failure caused an explosion and fire in the room. The inspectors noted several performance weaknesses during the drill:The fire brigade leader directed three fire brigade members into the fire hot zone to fight the fire as the attack team. Since there is a 5-member fire brigade, only two fire brigade members remain, one of which is the fire brigade leader (who also serves as the site incident commander (SIC)), to be part of the designated 2-out rescue team, required when fighting internal building fires. This 2-out rescue team is responsible, if necessary, for providing assistance or rescue for any or all of the attack team members. The inspectors were concerned that this fire brigade strategy could result in challenges with fire brigade leader command and control, and with the effectiveness of conducting rescues. The fire brigade leader could be hampered in his primary role of directing a site fire response while serving as a rescue team member. Adding to this complication, in locations where radios are not allowed inside some buildings with electrical sensitive equipment during firefighting, as was the case for this fire drill, it would be difficult for the fire brigade leader to communicate and coordinate with the control room or others during a rescue situation. Regarding the actual rescue activity, its effectiveness could be challenged since a two-person rescue team would be faced with potentially assisting/removing three attack members out of the hot zone. Based on discussions with licensee fire brigade training personnel following the drill, theinspectors learned that this 3-in, 2-out deployment was the current manner in which all internal building firefighting strategies and fire training was based upon.The fire brigade leader allowed the 3-man attack team to enter the fire hot zone with permission to commence firefighting prior to the 2-man rescue team arriving at the fire scenes pre-established incident command post and available for implementing rescue. The inspectors later learned that the rescue team, including the fire brigade leader, had arrived at the incident command post approximately five minutes after the attack team had entered the fire area. This delay involved the fire brigade leader completing his thermal protective clothing dressout in the locker room. The inspectors were concerned that under actual circumstances, if the 2-man rescue team were not ready and prepared to fulfill their rescue responsibilities upon entry of the attack team into the fire hot zone, the effectiveness of the rescue team could be challenged.The inspectors observed that no fire hose or other form of fire suppression was pulled or readily available for the 2-man rescue team to take with them should they have needed to enter the hot zone to rescue the attack team. When questioned about this, the inspectors were told that on the same fire hose that the attack team was using, a 1-1/2 inch gated wye valve had been connected, and the rescue team could have connected another 50-foot, 1-1/2 inch fire hose to it and used that hose as a rescue hose. However, the inspectors determined this was inadequate since to get to this hose connection, the rescue team would have to enter into the hot zone prior to reaching it. In addition, the inspectors learned that the use of this 1-1/2 inch gatedwye valve to create two hose streams for either attack or rescue that essentially splits the available flow capacity through a single 1-1/2 inch hose station nozzlewas allowed in multiple fire pre-plan strategies. At the conclusion of the inspection, the inspectors were continuing to assess whether the use of these gated wye valves had been formally reviewed by the licensee in the past to ensure that the flow capacity of fire hose streams would not be adversely impacted by their use during a fire.Planned Closure Actions: Pending completion of additional evaluations needed to determine whether the above fire brigade issues of concern represented performance deficiencies and if so, whether the performance deficiencies were of more than minor significance, this issue was identified as an unresolved item.Licensee Actions: The licensee initiated an NCR to address the inspectors concerns. In addition, until a more thorough review of their fire brigade program could be performed against their NFPA 805 fire program requirements, an operator standing instruction (#18-009, Fire Brigade 2-Out Response) was developed and implemented. This standing instruction directed the following specific fire brigade required actions:The brigade attack team will consist of two fire members to ensure the fire brigade SIC is not normally utilized as one of the 2-out members. If a runner is needed based on the fire area, the SIC may serve as a 2-out member, but this should be the exception.The 2-out members will establish a ready method of suppression that is accessible outside the fire zone. This should be the identified backup hose in the fire pre-plan. This hose does not need to be charged but should be flaked out and ready for use.The attack team will not enter the fire area, except when search and rescue is necessary, until the 2-out team is in the area with the suppression method ready for use.The inspectors determined that the licensees interim actions were adequate to ensure the fire brigade response would be effective if called upon pending resolution of the issues. Corrective Action Reference: NCR 02194468NRC Tracking Number: URI 05000400/2018001-01
05000400/FIN-2017003-02	2017Q3	Harris	Review of Removal of the Technical Support Center (TSC) Temporary Diesel Generator	The inspectors conducted a detailed review of NCR 02123373, Emergency Action Level Document Calculation Assumptions. The inspectors chose the sample because the EAL issue initially appeared to be potentially more significant than finally determined. The inspectors evaluated the following attributes of the licensees actions: complete and accurate identification of the problem in a timely manner evaluation and disposition of operability and reportability issues consideration of extent of condition, generic implications, common cause, and previous occurrences classification and prioritization of the problem identification of root and contributing causes of the problem 19 identification of any additional condition reports completion of corrective actions in a timely manner 2. The inspectors conducted a detailed review of NCR 00520918, Loss of Offsite Power Impact on Technical Support Center (TSC). The inspectors chose the sample because it was discovered that on July 17, 2017, the licensee had removed a temporary diesel generator that was intended to provide a back -up reliable power source to the TSC until a permanent solution was implemented. The inspectors evaluated the following attributes of the licensees actions: complete and accurate identification of the problem in a timely manner evaluation and disposition of operability and reportability issues consideration of extent of condition, generic implications, common cause, and previous occurrences classification and prioritization of the problem identification of root and contributing causes of the problem identification of any additional condition reports completion of corrective actions in a timely manner b. Findings 1. Incomplete and Inaccurate Emergency Action Level Submittals Introduction: The NRC identified a Severity Level IV NCV of 10 CFR 50.9 , Completeness and accuracy of information, for failure to provide complete and accurate information for prior approval of a new EAL scheme. The documents submitted to the NRC were, Shearon Harris Nuclear Power Plant, Unit 1 Changes to the Emergency Action Level Scheme, dated April 25, 2010, and License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99- 01, Revision 6, dated April 30, 2015. The first submittal to the NRC in 2010 was not complete and accurate in all material respects , and the submittal in 2015 was a missed opportunity to identify the errors made in the first submittal in 2010. Description : On May 10, 2017, Shearon Harris identified the hot operating mode EAL thresholds were calculated incorrectly using a NUREG -0654 methodology vice the required NEI 99- 01 Rev. 6 method, as specified in the current facility licensing basis. When employing the NUREG -0654 methodology to calculate the EAL threshold values, the reactor coolant system (RCS) inventory was assumed to be released at a 50 gallons per minute (gpm) RCS leak rate and activity of 300 micro -Curies per gram (ci/gm) dose equivalent iodine (DEI), over a six -hour period of time. In comparison, when employing the NEI 99- 01 Rev. 6 methodology, the assumption as part of calculating the EAL threshold values was that the entire RCS inventory was released instantaneously at an activity of 300ci/gm DEI. Both of the licensees submittals to the NRC, specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, contained declaration EAL threshold values for the containment high range radiation monitor for loss of fuel clad barrier and potential loss of containment , that were significantly lower than the correct values , due to use of the improper calculation methodology. The submittal dated April 30, 2015, was submitted to provide a complete change to the EAL scheme. This submittal was a missed opportunity by the licensee to identify that the wrong methodology to calculate the EAL threshold values had been used. 20 These submittals were not correct in material content and impacted the NRC s regulatory processes. The NRC evaluated the licensees failure to provide complete and accurate information to determine if there were any unresolved issues. The inspectors concluded that the incomplete and inaccurate information in the license submittal was material to the NRC because, had the NRC staff known the actual methodology used was inaccurate, the staff would have required the licensee to modify the EAL threshold values . The licensee appropriately revised the EAL threshold values utilizing the correct calculation methodology. The licensee issued NC R 02123373, dated May 10, 2017, for EAL thresholds that were calculated without using the correct methodology described in the facility licensing basis. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision - makers of the proper application of the EAL scheme and revised threshold values to be implemented until a permanent change is made to the license. Additionally, the licensee issued NCR 02155272, dated October 3, 2017, for the incomplete and inaccurate EAL submittal, specifically addressing and resolving the completeness and accuracy issues identified by the inspectors. The final significance determination of the underlying technical issue for the licensees failure to maintain the effectiveness of its emergency plan was documented in NRC Inspection Report 05000400/2017003, Section 4OA7, as a Green LIV. Analysis : The inspectors evaluated the underlying technical issue and determined that the licensees failure to maintain the effectiveness of its emergency plan was a performance deficiency. The issue was documented as a Green LIV in Section 4OA7 of this report. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which impeded the NRCs ability to regulate, using traditional enforcement to adequately deter non- compliance. Using the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report , this issue was determined to be a SL IV violation. Though the NRC would have questioned the issue with a request for additional information, it would not have resulted in substantial further inquiry. Additionally, the associated technical violation was determined to be of very low safety significance. Traditional enforcement violations are not assessed for cross -cutting aspects . Enforcement : Section 50.9 of 10 CFR states, in part, that, information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on April 25, 2010, and on April 30, 2015 , information was submitted by the licensee to the NRC that was not complete and accurate in all material respects. Specifically, the submitted documents specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, contained EAL declaration threshold values for the containment high range radiation monitor , that were lower than the actual correct values , due to use of an improper calculation methodology. This was not in accordance with the license. It was used to calculate the loss of fuel clad barrier and potential loss of containment thresholds values. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017 -017 to inform operators and emergency response organization decision -makers of the proper application of the EAL scheme and appropriate threshold values to be implemented. Additionally, the licensee plans to submit a license amendment request to update the 21 EAL scheme. Because this violation was not repetitive or willful, and was entered into the licensees CAP as NC R 02155272, it is being treated as a SL IV NCV, consistent with Section 2.3.2 .a of the NRC Enforcement Policy. ( NCV 05000400/2017003- 01, Incomplete and Inaccurate Emergency Action Level Submittal s) 2. Adequacy of Process for Removal of the TSC Temporary Diesel Generator Introduction: The inspectors opened an Unresolved Item (URI) to complete a review of the licensees removal of a temporary diesel generator on July 17, 2017, that was previously installed to provide reliable backup power to the TSC in the event of a Loss of Offsite Power (LOOP) coincident with a Loss of Coolant Accident (LOCA) event. This temporary diesel generator was originally intended to be installed until a reliable backup power source could be implemented under a permanent modification. Description : The licensee initiated NCR 00520918 on March 1, 2012, to address the consequences of a LOOP/LOCA event on the T SC functionality. Since the TSC is designed with two sources of electrical power, both from offsite power sources, it was recognized that a complete loss of offsite power to the TSC could result in long term TSC operational concerns. Specifically, with t he loss of both offsite power sources, the TSC emergency ventilation system, which provides required radiation protection for event response personnel, would be non- functional, as well as other critical TSC equipment following the loss of short -term (~1 -2 hour s) back -up battery power supplies. The inspectors noted that the operability/functionality section of NCR 00520918 stated that the TSC was functional based on the (current) availability of both of the offsite power sources; however, should a LOOP event occur, then the TSC would be considered non -functional since offsite power would be rendered non -functional. This statement demonstrated the licensees understanding of the vulnerability of continued TSC functionality during a LOOP event. In recognition of this vulnerability, the NCR implemented a short -term solution for procuring and installing a temporary diesel generator in late 2012 under modification EC 85350. The inspectors noted that an emergency preparedness change review evaluation was conducted in accordance with 10 CFR 50.54(q) under action request 00568695. This change request stated that it was necessary to provide the infrastructure for an additional reliable power source for the TSC habitability systems. NCR 00520918 stated that the long- term solution was to provide a permanent backup power supply to the TSC , at which time the temporary diesel generator would be removed. While an action item was initiated to install this TSC permanent backup power source under modification EC 85145, the modification was later revised, removing the intended implementation of a permanent backup power source to the TSC. The inspectors were concerned that the TSC could have equipment and habitability issues during design basis LOOP/LOCA events when the normal TSC offsite power would be non- functional. In addition, the inspectors determined that the TSC temporary diesel generator was removed from the site on July 17, 2017, without implementing the originally intended reliable permanent backup power to the TSC and without conducting a 10 CFR 50.54(q) evaluation specific to its removal to demonstrate that this action did not reduce the effectiveness of implementing the emergency plan. The inspectors requested additional information from the licensee related to the documentation, basis, and process used for the removal of the TSC temporary diesel generator, and evidence that the TSC facility would still be capable of performing all of its intended functions during a LOOP/LOCA event. This issue of concern requires more information to 22 determine if a performance deficiency exists, and if the performance deficiency potentially constitutes a violation of regulatory requirements . Pending review of additional information from the licensee, this issue is identified a s URI 05000400/2017003 -02, Review of Removal of the Technical Support Center ( TSC ) Temporary Diesel Generator.
05000400/FIN-2017003-01	2017Q3	Harris	Incomplete and Inaccurate Emergency Action Level Submittals	The NRC identified a Severity Level (SL ) IV non- cited violation (NCV) of 10 CFR 50.9, Completeness and accuracy of information, for failure to provide complete and accurate information for prior approval of a new emergency action level (EAL) scheme. The documents submitted to the NRC were, Shearon Harris Nuclear Power Plant, Unit 1 Changes to the Emergency Action Level Scheme, dated April 25, 2010, and License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99- 01, Revision 6, dated April 30, 2015. The submit ted documents specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, which contained declaration EAL threshold values for the containment high range radiation monitor that were lower than the correct values due to use of a n improper calculation methodology. The calculation methodology that was used was not in accordance with the license. It was used to calculate the loss of fuel clad barrier and potential loss of containment threshold values. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017 -017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented. Additionally, the licensee plans to submit a license amendment request to update the EAL scheme. The licensee entered this violation into their corrective action program (CAP) as nuclear condition report (NCR) 02155272. The inspectors evaluated the underlying technical issue and determined that the licensees failure to maintain the effectiveness of its emergency plan was a performance deficiency. The issue was documented as a Green licensee- identified violation (LIV) in Section 4OA7 of this report. The reactor oversight process (ROP) , significance determination process , does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which impeded the NRCs ability to regulate, using traditional enforcement to adequately deter non- compliance. Using the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, this issue was determined to be a SL IV violation. Though the NRC would have questioned the issue with a request for additional information, it would not have resulted in substantial further inquiry. 3 Additionally, the associated technical violation was determined to be of very low safety significance. Traditional enforcement violations are not assessed for cross -cutting aspects
05000400/FIN-2017003-03	2017Q3	Harris	Licensee-Identified Violation	Section 50.54(q)(2) of 10 CFR requires, in part, that a licensee shall follow and maintain the effectiveness of an emergency plan which meets the planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR Part 50, Appendix E . Section 50.47(b)(4) of 10 CFR requires that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, from April 2010 to May 2017, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, the licensee's emergency classification scheme action levels for Category F Fission Product Barrier EAL , contained declaration threshold values for the containment high range radiation monitor , which were lower than the correct values due to an improper methodology used in calculating the loss of fuel clad barrier and potential loss of containment barrier threshold values and rendered the EALs ineffective. The licensee implemented compensatory actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented until a permanent change can be made to the license. The issue was entered into the licensees CAP as NCR 02123373. The inspectors evaluated this issue as an ineffective EAL per IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process , Figure 5.4 -1. The inspectors concluded that the violation was of very low safety significance (Green). Although the incorrect EAL would alone render an early EAL classification of a General Emergency (GE) based upon the specific radiation monitor, other EALs would provide a GE classification in an accurate and timely manner aligned with the incorrect threshold values of the containment high range radiation monitor .
05000400/FIN-2017002-02	2017Q2	Harris	B ESCW Chiller Failure to Start	The inspectors opened a URI to facilitate the completion of inspection and determination of whether a performance deficiency was associated with the start failure of the B ESCW chiller on May 13, 2017. Description: On May 13, 2017, while attempting to start the B ESCW chiller, the motor compressor immediately tripped on C phase instantaneous overcurrent relay actuation. The chiller was declared inoperable and immediate troubleshooting was conducted to determine the cause of the trip. The licensees initial investigation did not identify any electrical or mechanical issues with the compressor motor, supply breaker and electrical bus, or other chiller control components. While the calibration of the C phase instantaneous overcurrent relay was checked and found to be within specification, the licensee determined the most probable cause of the trip was an intermittent failure of this relay. The relay was replaced and subsequent post-maintenance testing of the chiller was successfully performed without any other chiller operational problems being identified. The chiller was returned to operability early May 14, 2017, following the completion of this post-maintenance testing. At the end of the inspection period, the licensees investigation into the cause of the start failure had just completed. A URI is being opened for the NRC to review the licensees failure analysis and causal evaluation to determine whether the chiller start failure was reasonably within the licensees ability to predict or prevent and therefore a performance deficiency. This issue is being tracked as URI 05000400/2017002-02, B ESCW Chiller Failure to Start
05000400/FIN-2017002-01	2017Q2	Harris	Evaluate Fire Protection Discrepancies in RHR/CS Pump Rooms	An unresolved item (URI) was identified by the inspectors during the walkdown of the A and B RHR and CS pump rooms, involving the use of unapproved non-fire retardant plastic sheeting to contain contamination on the A RHR piping. Additionally, the inspectors identified that the fire pre-plan for fire brigade response delineates a hose station that did not contain adequate fire hose length.Description: The inspectors identified two issues of concern during the fire protection walkdown of the A and B RHR and CS pump rooms as follows: 1) Use of Unapproved Plastic for Contamination Control: The inspectors noted that an approximately 30 foot section of the A RHR pump suction piping had been wrapped with multiple layers of plastic sheeting materials that included radiation protection yellow Caution Radioactive Materials stamped plastic sheeting overlaid with clear stretch wrap plastic. The section of RHR piping where this plastic was installedincluded the motor-operated RHR suction valves from the containment recirculation sump (valve 1SI-310) and the refueling water storage tank supply (valve 1SI-322). The inspectors were concerned that these valves could be adversely impacted from a potential fire involving this plastic material. The inspectors questioned whether this plastic was fire retardant material or had been evaluated and allowed under the licensees transient combustible control procedure. The licensee subsequently determined that none of the plastic material was fire retardant or met the requirements of National Fire Protection Association (NFPA) 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, and no previous transient combustible evaluation could be found that allowed the use of the non-fire retardant plastic in the RHR pump room. In addition, radiation protection personnel indicated that there could be other areas where this plastic was used since it was a typical practice to use the material to prevent the spread of contamination from leaking piping connections, valves, and valve packing. The licensee subsequently removed the plastic from the A RHR piping and initiated NCR 02132781 to evaluate this issue of concern. 2) Inadequate Fire Hose Length in Hose Station Described in Fire Pre-Plan: During review of the fire pre-plan procedure (FPP-012-02-RAB190-216) for the A and B RHR/CS pump rooms on the RAB 190 elevation, the inspectors noted that the procedure described two fire hose stations intended to be used during fire brigade response for a fire in either of the pump rooms. These two hose stations were the respective hose stations located just inside the access door to each of the two RHR/CS pump rooms. The procedure states that an extra 100 feet of hose would be needed to account for the additional distance for the hose from the opposite train pump room. However, the inspectors identified that even with the extra 100 feet added to the existing 100 feet that is already in each hose station, there would still not be adequate length for this second hose to reach the opposite train pump room with the fire. The inspectors measured the actual distance between the two locations and estimated the hose would have to be over 300 feet in length in order to be effective in fighting a fire in either of the rooms. A separate hose station on the 216 RAB elevation may provide adequate backup coverage. However, the inspectors were concerned that the issue with the fire pre-plan hose station use could cause confusion or pose an unnecessary delay in fire brigade response for a fire in either of the rooms. The licensee subsequently initiated NCR 02134163 to evaluate this issue of concern.Pending completion of additional evaluations needed to determine whether the above issues of concern represented performance deficiencies and if so, whether the performance deficiencies were of more than minor significance, this issue is identified as URI 05000400/2017002-01, Evaluate Fire Protection Discrepancies in RHR/CS Pump Rooms.
05000261/FIN-2017001-03	2017Q1	Robinson	Licensee-Identified Violation	10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non- confor mances are promptly identified. Contrary to the above, in March 2014, while performing examinations in steam generator C during forced shutdown RFO229F3, the licensee failed to identify a loose part lodged in contact with tube R37C22. The licensee identified the loose part in March 2017 during refueling outage RO30. The licensee verified that indications of the part were detectable during RFO229F3, retrieved the part, verified that degradation caused by the part met all structural integrity requirement s, plugged the tube, and removed it from service. This issue was identified in the licensees CAP as NCR 0210725. The inspectors evaluated this violation using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings At -Power, and determined that the violation was of very low safety significance (Green) because evaluations demonstrated that the tube could sustain three times the differential pressure across it during normal full power steady state operation and that the steam generator did not violate the accident leakage performance criterion
05000261/FIN-2017001-01	2017Q1	Robinson	Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak -Chase Test Connection	Green . An NRC- identified Green non -cited violation ( NCV ) of 10 CFR Part 50.55a, Codes and Standards, was identified for the failure to perform general visual examinations of moisture barriers in the containment leak -chase channel test connections in accordance with the American Societ y of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC), Section XI, Subsection IWE , Requirements for Class MC and Metallic Liners of Class CC Components of Light -Water Cooled Plants . Following the inspectors identification of this issue, t he licensee initiated actions to conduct the re quired visual examinations during the March 2017 refueling outage and initiated actions to revise the containment inservice inspection (ISI) plan such that the required examinations will be performed in the future . This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 02109909. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME B PVC , Section XI, Subsection IWE , was a performance deficiency (PD) . The finding was of more than minor significance because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, visual examinations of mois ture barriers associated with the containment leak -chase channel test connections provide assurance that the containment metal liner and liner seam welds remain capable of performing its intended safety function. In the absence of such examinations, corro sive conditions at the moisture barrier (concrete -to-tubing interface) could go undetected. As a result, degradation of inaccessible portions of the containment liner could progress to challenge the containment operational capability. Using IMC 0609, A ttachment 4, Initial Characterization of Findings, the finding was determined to affect the Barrier Integrity Cornerstone because it involved ISI program examinations designed to identify degradation of the containment metal liner. The inspectors screen ed the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At -Power, Exhibit 3 Barrier Integrity Screening Questions, and determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the containment. The inspectors reviewed this performance deficiency for cross -cutting aspects as required by IMC 0310, Components With Cross -Cutting Aspects. The finding was determined to be reflective of present licensee performance because in 2014, the licensee did not take effective corrective actions to implement the ASME BPVC 3 requirements in the Subsection IWE P rogram , when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014- 07, which highlighted this industry -wide problem. Therefore, the finding was assigned a cross - cutting aspect in the resolution component of the problem identification and resolution cross -cutting area (P.3)
05000261/FIN-2017001-02	2017Q1	Robinson	Failure to Submit Complete and Accurate Information for a Requested License Amendment	Severity Level IV. An NRC -identified severity level IV (SL IV) NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, was identified for the licensees failure to provide complete and accurate information in a license amendment request (LAR), dated November 19, 2015, requesting extension of the containment leak rate test frequencies required by various containment technical specifications (TS s). In this LAR, the licensee incorrectly stated that they had revised their ASME BPVC, Section XI, Subsection IWE program to include visual examinations of the test connections in the leak -chase channel penetration pressurization system ( PPS) , when in fact, the program had not been revised and the examinations had not been performed . This information was material to the NRC because it was used, in part, as the basis for the approval and issuance of License Amendment 247, dated October 11, 2016, extending the TS containment leak rate test frequencies. The licensees corrective actions included conducting the visual examinations of the test connections in the leak -chase channel PPS during the ongoing refueling outage in March 2017 and initiating actions to add the visual examination requirements to their Subsection IWE program. This issue was entered into the licensees CAP as NCR 02110516. The failure to provide complete and accurate information in accordance with 10 CFR 50.9(a) for the LAR associated with License Amendment 247 is a violation of NRC requirements . This violation was screened against the ROP guidance in IMC 0612, Appendix B, Issue Screening, and no associated ROP finding was identified. The inspectors evaluated this issue using the Traditional Enforcement process because it had the potential to impact the NRCs ability to perform its regulatory function. Specifically, the violation impacted the regulatory process, in that the inaccurate information was material to the NRCs review and acceptance of licensee actions to address the industry -wide operating experience discussed in NRC IN 2014- 07. Based on licensee inaccurate information that they had addressed IN 2014 -07 by revising their containment ISI program to perform visual inspections of accessible tubing in the containment leak -chase channel PPS system, the NRC staff concluded that the licensee was properly implementing the ASME BPVC, Section XI, Subsection IWE program. In accordance with the guidance in Sections 2.2 and 6.9 of the NRC Enforcement Policy, the inspectors determined this is an SL IV violation, because had the information been complete and accurate at the time provided, it likely would have resulted in the need for further clarification of the licensees actions to address NRC IN 2014- 07 , but would not have caused the NRC to change its decision to issue the license amendment or resulted in substantial further inquiry . Also, on March 23, 2017, the licensee completed the visual examinations of the subject tubing in the leak -chase channel system and did not identify any significant degradation. In accordance with IMC 0612, Appendix B, traditional enforcement issues are not assigned a cross -cutting aspect.
05000400/FIN-2016003-01	2016Q3	Harris	Subsequent Loss of Safety-Related Chilled Water System Results in a Loss of Safety Function	The inspectors opened a URI to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, associated with the A ESCW chiller failures. On July 15, 2016, the A ESCW chiller tripped on low oil pressure. Licensee investigation identified that oil was leaking from the threaded portion of a brass fitting located between a pressure switch and needle valve associated with PDS-01CY-9428ASA-HI. Upon removal, it was observed that significant radial cracking occurred in the threaded portion of the brass fitting. A like-for-like replacement was installed and the A ESCW chiller was returned to service. One week later, on July 22, 2016, the A ESCW chiller tripped again on low oil pressure. The investigation revealed that the same brass fitting had failed and the A ESCW chiller could not meet its mission time of 30 days of continuous operation in the event of a loss of cooling accident. During this 7-day period, the B ESCW chiller was inoperable for a period of time, which means the ESCW system would not have been able to meet its safety function. The licensees investigation into the cause of the subsequent failure is ongoing. A URI is being opened to determine whether the subsequent failure of the brass fitting was reasonably within the licensees ability to predict and therefore a performance deficiency. This issue is being tracked as URI 05000400/2016003-01, Subsequent Loss of Safety-Related Chilled Water System Results in a Loss of Safety Function.
05000369/FIN-2016003-01	2016Q3	McGuire	Licensee-Identified Violation	Technical Specifications 5.4.1.a, Procedures, requires, in part, that procedures for certain activities recommended in Regulatory Guide 1.33, Rev. 2, Appendix A, be established, implemented, and maintained. Administrative procedures for shift and relief turnover is one of the identified activities. Administrative procedure AD-OP-ALL-1000, Conduct of Operations, Rev. 4, implements the licensees shift and relief turnover standards. This procedure requires shift turnovers to contain detailed information on equipment and system status, alignments, and activities, to ensure watchstanders have a complete understanding of plant status. Contrary to the above, from August 10 to August 13, 2015, operators were not aware of the required nuclear service water system alignment which required a continuous vent (passing water flow) to be maintained in the condenser cooling water (RC) suction supply to the Unit 1 turbine driven auxiliary feedwater pump. The continuous vent mitigates the potential for air entrainment in the RC piping high point and is needed in order for the standby shutdown system to be functional during an Appendix R fire event when the suction of the turbine driven auxiliary feedwater pump is transferred from the auxiliary feedwater storage tank to the long term water supply provided by the RC system. This lack of operator awareness stemmed from a misunderstanding in the operator turnovers that the nuclear service water system was in a standby nuclear service water pond cooling alignment, which does not require the continuous vent to be maintained. The discrepancy was subsequently identified by oncoming shift operations personnel and the continuous vent was re-established on August 16, 2015, after removing material that obstructed the continuous vent line. As a result of not maintaining the continuous vent at the suction of the turbine driven auxiliary feedwater pump, the standby shutdown system was rendered non-functional for a period of eleven days, which was in excess of the 7-day limit allowed by Selected Licensee Commitments 16.9.7. This violation was determined to be of very low safety significance (Green) because it only affected the non-safety related Appendix R water supply to the turbine driven auxiliary feedwater pump. This violation was entered into the licensees corrective action program as NCR 01943414.
05000261/FIN-2016003-01	2016Q3	Robinson	Failure to Scope Tainter Gate Flood Protection Features in Maintenance Rule Resulting in Degraded Performance	A self-revealing Green NCV of 10 CFR 50.65(b)(2)(ii) was identified for the failure to scope the external flood protection function of the Robinson Lake Dam spillway (Tainter) gates in the maintenance rule (MR) monitoring program. The failure to include the Tainter gates in the MR program resulted in ineffective maintenance being performed and subsequent degraded opening capability which challenged the availability of safety-related equipment during design basis rainfall events due to site flooding. The licensee took immediate corrective actions to replace/refurbish the chains to both gates and completed full open testing to restore their functionality. In addition, the licensee has developed and initiated implementation of an action plan to improve and ensure reliability of the gates, and initiated actions to revise the MR scoping program to include the Tainter gates. The issue was entered into the licensees CAP as CR 2035500. The failure to scope the flood protection function of the Lake Robinson Dam Tainter gates in the maintenance rule monitoring program was a PD. The finding is more than minor because it is associated with the protection against external factors (i.e., flood hazard) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to monitor flood protection features associated with the Tainter gates resulted in degraded gate opening performance that could have resulted in site flooding during design basis rainfall events and adversely impact multiple trains of safety-related equipment due to water intrusion. Using IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding involved the degradation of equipment specifically designed to mitigate flooding events. In accordance with Exhibit 4, External Events Screening Questions, the inspectors determined that the finding represented a degradation of two or more trains of a multi-train system or function during an external flooding event, therefore it required a detailed risk evaluation. A regional senior reactor analyst completed a detailed risk evaluation in accordance with NRC IMC 0609 Appendix A, and Appendix M, Significance Determination Process Using Qualitative Criteria, using the latest NRC Robinson Standardized Plant Analysis Risk model. The high uncertainty associated with estimating flood frequencies was the reason for using the NRC IMC Appendix M approach. The major analysis assumptions included a one-year exposure interval, recovery credit for opening the Tainter gates subsequent to binding of the chain, and limited credit for FLEX flooding mitigation strategies. If the rainfall produced a water surface elevation which would overtop the dam, the dam was considered failed and the ultimate heat sink lost. The rainfall frequencies requiring gate operation were estimated using a combination of National Oceanographic and Atmospheric Administration rainfall data and a probabilistic technique to establish precipitation frequency estimates performed by the licensee. The dominant sequence was a flood event inducing a non-recoverable loss of offsite power and loss of the emergency buses with a failure of the operators to manually recover the Tainter gates and failure of the operators to depressurize the steam generators to facilitate FLEX injection leading to a loss of core heat removal and core damage. The risk was mitigated by the low flood frequency, and the likely recovery of the Tainter gates prior to site flooding. There were additional conservatisms which were not applied to the result but would reduce the risk. These included the fact that the plant would be shutdown prior to flooding impacting safety-related equipment, which would reduce decay heat cooling required, and additional FLEX flooding strategies which could provide cooling even if the dam was lost. The risk increase due to the performance deficiency was < 1.0E-6/year, a Green finding of very low safety significance. The licensees analysis and full scope probabilistic risk assessment model produced a similar result. The inspectors determined that since the scoping of plant systems had occurred more than three years in the past, the finding did not represent current plant performance and therefore did not have a cross-cutting aspect associated with it.
05000261/FIN-2016003-02	2016Q3	Robinson	Failure to Assess and Manage Risk for Main Turbine Trip Maintenance Resulting in Turbine/Reactor Trip	A self-revealing Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) was identified for the failure to adequately assess and manage the increase in risk associated with online maintenance activities involving the removal of the cover to the main turbine trip mechanism in order to perform visual inspections. During removal of the cover, the turbine trip mechanism lever was contacted causing an automatic turbine/reactor trip. The licensee took immediate corrective actions to reemphasize the need to enter all applicable types of work activities into the work management process and to conduct formal risk assessments in accordance with the risk management program. The licensee entered this issue into the corrective action program (CAP) as condition report (CR) 2056554. The licensees failure to adequately assess and manage the risk of maintenance associated with visual inspection of the turbine trip mechanism was a performance deficiency (PD). The inspectors evaluated the PD in accordance with IMC 0612, Appendix B, Issue Screening, and determined it to be more than minor because it impacted the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to assess and manage the risk associated with removing the turbine trip mechanism cover to conduct visual inspections resulted in a turbine/reactor trip. The inspectors evaluated the finding in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K, the inspectors requested that a regional Senior Reactor Analyst (SRA) independently evaluate the risk. A Region II SRA performed an analysis of the risk deficit for the unevaluated condition associated with the work activity on the turbine trip mechanism. The latest Robinson Standardized Plant Analysis Risk (SPAR) model was used to calculate an incremental core damage probability deficit (ICDPD). The result was an ICDPD of 3.74E-7 and represented the increase in core damage probability associated with a turbine/reactor trip coincident with the dedicated shutdown diesel generator being out of service at the time of the event. In accordance with IMC 0609, Appendix K, because the calculated ICDPD was not greater than 1E-6, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of work management in the cross-cutting area of human performance because the licensee failed to adequately implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. Specifically, the licensee failed to adequately assess, manage, and implement risk management actions for activities associated with trip sensitive equipment.
05000369/FIN-2016002-02	2016Q2	McGuire	Failure to Ensure Containment Equipment Hatch Was Properly Closed During Fuel Movements	An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for the licensees failure to adequately implement the commitments in Selective Licensee Commitment (SLC) 16.9.25, Refueling Operations Containment Equipment Hatch, which required the containment equipment hatch to be closed during the movement of non-recently irradiated fuel inside containment. Specifically, during reactor vessel fuel reload activities, the inspectors identified that the equipment hatch was left partially open due to the failure to properly tighten the bolts evenly around the hatch resulting in direct communication of the containment atmosphere with the environment. The licensee took immediate corrective action to suspend fuel movements and properly tighten the equipment hatch bolts prior to resuming fuel movements and entered the issue into their corrective action program as ARs 02018605 and 02018701. The PD was more than minor because it impacted the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that containment protects the public from radionuclide releases caused by accidents or events. Additionally, if left uncorrected, the PD would have the potential to lead to a more significant safety concern. Specifically, the radiological barrier functionality of the containment equipment hatch was degraded due to the gap opening which could have allowed direct access of radiological releases from the containment atmosphere to the outside environment during a potential fuel handling accident inside containment. The inspectors screened the finding in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Because the finding degraded the ability to close or isolate the containment, it required review using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. While the containment boundary function was considered degraded, the incident occurred eight days after the beginning of the refueling outage when short lived volatile radioisotopes had decayed sufficiently such that the potential radiological releases to the public would not likely contribute to the large early release frequency (LERF). Based on this, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of procedure adherence in the cross-cutting area of human performance because the licensee failed to follow containment equipment hatch closing procedures which explicitly required performing a visual inspection that the containment equipment hatch was sealed and secured with metal-to-metal contact with the containment hatch flange and had no visual gaps.
05000369/FIN-2016002-01	2016Q2	Mcguire McGuire	Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak Chase Test Connections	An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50.55a, Codes and Standards, was identified for the licensees failure to perform general visual examinations of moisture barrier material in the reactor containment leak-chase channel test connections in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV Code), Section XI, Subsection IWE. The licensee performed the required examinations in Unit 1 during the March 2016 refueling outage and initiated corrective actions to revise the Containment Inservice Inspection (ISI) Plan. The licensee also planned to perform similar examinations in Unit 2 prior to the end of the first containment ISI period. Additionally, the licensee performed a containment operability determination to justify continuous operation of the Unit 1 and Unit 2 containment based on the results of all visual examinations, extent of condition activities, and the results of containment integrated leak rate tests. The licensee entered this issue into their corrective action program as action request (AR) 02038505. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME BPV Code was a performance deficiency (PD). The PD was of more than minor significance per IMC-0612, Appendix B, Issue Screening, because the current Containment ISI Plan did not adequately implement the ASME BPV Code requirements for the examination of moisture barriers, and if left uncorrected, it had the potential to lead to a more significant concern. The finding was of very low safety significance (Green) per IMC-0609 because it did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect of resolution in the problem identification and resolution cross-cutting area because the licensee did not take effective corrective actions to implement the ASME BPV code requirements in the Containment ISI Plan when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014-07.
05000369/FIN-2016001-01	2016Q1	Mcguire McGuire	Failure to Maintain Fire Extinguishers in Contaminated Radiation Control Zones in Accordance with the Fire Protection Program	An NRC-identified Green non-cited violation (NCV) of the McGuire Nuclear Station Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for failure to perform annual maintenance on fire extinguishers located in contaminated radiation control zones (RCZs). The licensee took immediate corrective action to replace the past due fire extinguishers and entered the issue into their corrective action program as action request (AR) 02009794. The performance deficiency (PD) was more than minor because if left uncorrected the PD could have the potential to lead to a more significant safety concern, in that, fire extinguishers located in any contaminated RCZs may not be functional for firefighting purposes due to lack of maintenance. Every fire extinguisher, five total, located in a contaminated RCZ, did not have its annual maintenance up-to-date. The longest duration without annual maintenance was six years for two of the five extinguishers. The finding was determined to be of very low safety significance (Green) within the mitigating system cornerstone because it would not affect the ability to reach and maintain a safe shutdown condition, in that, for each of the fire areas where the out-of-date extinguishers were present, there were also properly maintained fire extinguishers and hose stations outside of the RCZ. The out-of-date extinguishers were weighed and it was determined that they would have performed their function, if needed. The cause of the PD was directly related to the cross-cutting aspect of field presence in the cross-cutting area of human performance because the licensee failed to correct deviations from the FPP and ensure proper oversight of the vendor contracted to perform fire extinguisher maintenance.
05000370/FIN-2015004-01	2015Q4	Mcguire	Failure to Report Unit 2 Unplanned Valid Auxiliary Feedwater Actuation in Mode 4	An NRC identified Severity Level (SL) IV non-cited violation (NCV) of 10 CFR 50.72(b)(3)(iv)(A) was identified for the licensees failure to make a required NRC event notification within eight hours for an unplanned valid actuation of the auxiliary feedwater (CA) system. The unplanned valid actuation occurred during main turbine and main feedwater pump safety injection (SI) train trip function testing with Unit 2 in Mode 4 on October 7, 2015. The licensee entered this issue into their corrective action program and subsequently reported this CA actuation to the NRC on October 15, 2015. The failure to submit an event notification to the NRC within eight hours of occurrence of an unplanned valid CA system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A) was a performance deficiency (PD). Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this PD was dispositioned under the traditional enforcement process and was determined to be a SL IV violation. Because this SL IV violation was not repetitive or willful, and did not have an underlying technical violation that would be considered more-than-minor, a cross-cutting aspect was not assigned to this violation.
05000369/FIN-2015003-01	2015Q3	Mcguire McGuire	Failure to Adequately Implement a Temporary Modification for a Leak Enclosure	A self-revealing Green finding (FIN) was identified for failure to adequately implement the modification procedural requirements of engineering directives manual (EDM)-601, Engineering Change Manual, for a temporary modification that installed a valve leak seal enclosure on main steam drain valve 2SM-27. Specifically, EDM-601 required the weight and vibration response of the enclosure to be evaluated as part of the installation. The failure to consider this resulted in vibration induced piping failure upstream of the valve and an unexpected rapid plant down power. The failure to adequately implement a temporary modification in accordance with EDM- 601 was a performance deficiency (PD). The PD was more than minor because it was associated with the design control attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability during power operations. Specifically, the performance deficiency resulted in a rapid down power to approximately 20 percent and subsequent actions to take the Unit 2 turbine generator offline to repair the leak. Using NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance because the it did not contribute to both the cause of a reactor trip and affect mitigation equipment. The finding had a cross cutting aspect of consistent process, as described in the human performance crosscutting area because the licensee failed to use a consistent, systematic approach to make de.cisions during implementation of a temporary modification.
05000369/FIN-2015002-01	2015Q2	Mcguire McGuire	Failure to Establish Compensatory Actions for Obstructed Fire Sprinkler Spray Nozzle	An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for failure to evaluate and establish adequate compensatory measures for an impaired fire protection automatic water sprinkler system. Specifically, a solid deck scaffold platform was erected below a sprinkler system spray nozzle that would have obstructed the nozzle spray pattern protecting safe shutdown equipment involving the 2B2 component cooling water pump/motor. The licensee entered the issue into the corrective action program (CAP) as nuclear condition report (NCR) 01931412 and implemented immediate corrective actions to remove the scaffolding obstructing the sprinkler nozzle. The failure to evaluate scaffolding obstruction of a sprinkler system spray nozzle and implement required fire protection compensatory actions was a performance deficiency (PD). The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to provide adequate compensatory actions for an obstructed sprinkler nozzle would have reduced the licensees ability to quickly extinguish fires in the area. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings. Using the guidance in IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, the finding was assigned a category of fixed fire protection systems. The inspectors determined the finding to be of very low safety significance (Green), because it was assigned a low degradation rating that was based upon meeting the criteria described in IMC 0609, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements. Specifically, less than ten percent of the sprinkler nozzles were nonfunctional, there were functional nozzles within five feet of the combustibles of concern, and the system was nominally code compliant. The finding had a cross-cutting aspect of procedure adherence in the human performance area, because the licensee failed to follow scaffolding erection procedures which explicitly required not erecting scaffolding that could obstruct sprinkler nozzles unless approved by a fire protection engineer and necessary compensatory actions were implemented.
05000369/FIN-2014005-01	2014Q4	Mcguire McGuire	Failure to Adequately Control Transient Combustible Materials and Ignition Sources in Accordance with the Fire Protection Program	An NRC-identified Green NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately control fire ignition sources in the Unit 1 and Unit 2 exterior doghouses in accordance with the FPP requirements of Nuclear System Directive (NSD)- 313, Control of Transient Fire Loads. Specifically, temporary electric portable heaters were energized for several days without implementing required hourly fire watches, locating the energized heaters greater than prescribed separation distances from safety-related equipment, and preventing other transient combustible materials from being located near the heaters. The licensee placed this issue into their corrective action program (CAP) and took corrective actions to de-energize the heaters, distance the heaters away from safety related feedwater isolation valve electrical cables, and remove unnecessary transient combustibles from the area. The failure to control fire ignition sources in accordance with NSD-313 was a performance deficiency (PD) . The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective in that, a fire could have affected nearby safety-related feedwater isolation valve electrical cables which provide a shutdown mitigation function. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition. This finding had a cross cutting aspect of teamwork in the human performance area because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions.
05000369/FIN-2014005-02	2014Q4	McGuire	Failure to Adequately Implement Containment Closeout Resulting in Loose Debris and Unanalyzed Materials Left in Containment	An NRC-identified Green NCV of Technical Specification 5.4.1.a, Procedures, was identified for the failure to properly implement containment cleanliness and material control closeout procedures in accordance with procedure PT/1A/4600/003F, Containment Cleanliness and ECCS Operability Inspection, prior to entering Mode 4, following the Unit 1 refueling outage. Specifically, a large amount of unanalyzed general loose debris, as well as scaffolding with aluminum walkboards and fibrous lead blankets, were left in containment that could either contribute to emergency core cooling system (ECCS) recirculation sump screen blockage or containment hydrogen generation during design basis accidents. The licensee placed this issue into their CAP and took corrective actions to remove the loose debris and unanalyzed materials and performed re-inspections of containment to identify any additional loose debris or unanalyzed materials left in containment. The failure to perform an adequate containment cleanliness and material control closeout following the Unit 1 refueling outage in accordance with procedure PT/1/A/4600/003F was a PD. The PD was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective in that, loose debris in containment could result in the debris being transported to the ECCS recirculation sump screens in the event of design basis accident and adversely affect the sump performance. In addition, the PD was associated with the configuration control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective in that, the failure to control scaffolding that contained unanalyzed amounts of aluminum in containment challenged the existing analysis for containment aluminum inventory limitations. The finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of safety function of the ECCS sumps, was not safety significant due to external events, and no actual open pathway in the physical integrity of containment occurred. The finding had a cross-cutting aspect of field presence in the human performance area because the licensee failed to ensure that adequate supervisory and management oversight of the containment closeout process was conducted to ensure proper performance of procedure PT/1/A/4600/003F prior to entering Mode 4.
05000369/FIN-2014004-01	2014Q3	McGuire	1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws	The licensee identified flaws with ultrasonic testing in the 1B and 1C cold leg safety injection pipe welds as part of their extent of condition from Unit 2 for MRP- 146, Thermal Fatigue. Further evaluation determined these flaws were a circumferential flaw with an axial component on the nozzle side for 1B and an axial flaw from the centerline of the weld into the base metal for 1C. The licensee completed examinations on all welds included in the MRP-146 program and found them to be within the acceptance criteria. The licensee also removed and repaired the 1B and 1C nozzles. Welding of the new components have been examined and have passed all quality assurance examinations. The licensee determined that the flaws were a result of thermal fatigue. The licensee has performed all required examinations and repairs and is completing a metallurgical analysis of the flaws. This is an unresolved item pending review of the licensees metallurgical analysis of the flaws to determine if there is a performance deficiency. This issue will be tracked as URI 05000369/2014004-01, 1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws.
05000369/FIN-2014004-02	2014Q3	McGuire	Review NOED 14-2-002 Granting Exercise of Enforcement Discretion to Complete 1B EDG Repairs	The inspectors reviewed NOED 14-2-002 and related documents to determine the accuracy and consistency with the licensees assertions and implementation of the licensees compensatory measures and commitments. The inspectors independently verified the proper implementation of these compensatory measures which included deferring non-essential surveillances and other maintenance activities on the 1A EDG, TDCA pump, SSF, switchyard, and posting dedicated fire watches in selected risk significant areas. Additional inspection of this issue will be conducted as part of the NRCs review of the subsequent Licensee Event Report (LER) to be submitted by the licensee within 90 days. This LER will describe the circumstances of the 1B EDG failure, the root cause, and planned licensee corrective actions. This URI is identified as URI 05000369/2014004-02, Review NOED 14-2-02 Granting Exercise of Enforcement Discretion to Complete 1B EDG Repairs.
05000369/FIN-2014003-01	2014Q2	Mcguire McGuire	Licensee-Identified Violation	10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, the licensee to establish measures to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, in March 2008 and September 2012, the licensee failed to promptly identify a condition adverse to quality. A flaw in the Cold Leg 2D Nozzle 4-1 weld was missed during UT examinations of this component. During the most recent outage, the licensee reexamined this weld and identified an 85 percent through-wall flaw originating from the inner diameter in the weld. Destructive testing and analysis established that this flaw most likely existed since 2005. This violation was determined not to be greater than very low safety significance (Green) because it could not result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) and could not have likely affected other systems used to mitigate a LOCA. This violation was documented in PIP M-14-03544.
05000369/FIN-2014003-02	2014Q2	Mcguire McGuire	Licensee-Identified Violation	TS 5.4.1.a requires that written procedures shall be established, implemented, and maintained as recommended in Regulatory Guide (RG) 1.33, Rev. 2, Appendix A, February 1978. Section 3.d of RG 1.33 recommends that appropriate procedures be prepared for energizing, filling, venting, draining, startup, shutdown, and changing modes of operation for the ECCS. Contrary to the above, the licensee failed to develop an adequate procedure to fill and vent the Unit 2 NV system following system draining and argon gas introduction associated with NV 2A mixed bed demineralizer valve modification work conducted December 1-12, 2013. On December 14, 2013, following this modification work, a large gas void was identified in the ECCS piping at high point vent valve 2NV-1056, located in the suction of the ECCS pumps during design basis accident conditions involving cold-leg recirculation. The licensee determined that the use of an inadequate fill and vent procedure during the system restoration from the modification work resulted in the accumulation of a significant amount of the gas at this location. This violation was determined to be of very low safety significance (Green) because the licensee provided reasonable evidence that the ECCS pumps would have been capable of performing their intended safety function had the gas void been ingested into the suction of the pumps. This violation was documented in the licensees CAP as PIP M-13-11181.
05000369/FIN-2014002-03	2014Q1	McGuire	Failure to Implement Adequate Design Control Measures for Rod Control Power Supply Replacement Resulting in Reactor Trip	A self-revealing finding (FIN) was identified for the licensees failure to implement adequate design control measures for the rod control power supply modification which resulted in the loss of 24VDC power in the 1AC rod control power cabinet. The inspectors determined that the licensees failure to implement adequate design control measures was more than minor because it affected the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective, in that, the insufficient margin in the rod control power supply OVP function caused a multiple drop rod event which resulted in a reactor trip. This finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance.
05000370/FIN-2014002-01	2014Q1	Mcguire	Failure to Adequately Control Transient Combustible Materials in Accordance with the Fire Protection Program	An NRC-identified NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately control the storage of transient combustibles in the 2A residual heat removal (ND)/containment spray (NS) heat exchanger room near safe shutdown equipment in accordance with the FPP requirements. The licensee initiated immediate corrective actions to evaluate the transient combustible fire loading and remove all the unapproved transient combustibles from the area. This condition was placed in the licensees corrective action program (CAP). The licensees failure to control the storage of transient combustibles in accordance with procedure NSD 313 was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective in that a fire involving transient combustibles could have affected nearby power cables and motor operator for valve 2ND-58A which provides a safe shutdown mitigation function. The finding was determined to have very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition. This finding had a cross cutting aspect of Teamwork in the Human Performance area because multiple groups were responsible for bringing the transient combustibles into the area and the individuals failed to effectively communicate and coordinate their activities to ensure that transient combustible control processes were appropriately implemented.
05000369/FIN-2014002-02	2014Q1	Mcguire McGuire	Failure to Adequately Control the Use of Self- Extinguishing Fire Lids	An NRC-identified NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, FPP, was identified for the licensees failure to adequately control the storage of transient combustibles in waste receptacles equipped with self-extinguishing fire lids in accordance with the FPP requirements. The licensee took actions to correct all waste receptacles in the plant that were filled beyond the manufacturers specification or had loosely fitted lids. This condition was placed in the licensees corrective action program. The licensees failure to control the storage of transient combustibles in accordance with the requirements of NSD-313 was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective in that the self-extinguishing function was not retained which could allow the spread of the fire and adversely affect mitigating system equipment in the area. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown conditions. A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance.
05000370/FIN-2013005-01	2013Q4	Mcguire	Evaluation of Gas Void Identified in Unit 2 ECCS Piping	During the performance of Unit 2 ECCS pipe gas void inspections using ultrasonic test (UT) equipment, a large gas void was found in a 5 foot section of 8 inch diameter piping at high point vent valve 2NV-1056. 2NV-1056 was located on the suction side of both trains of the NI and NV pumps downstream of valve 2ND-58A, which is opened during design basis accident conditions involving cold-leg recirculation to provide the piggyback alignment from the residual heat removal (ND) system. Excessive gas accumulation at 2NV-1056 could result in gas being drawn into the NI/NV pumps causing pump degradation or failure. The licensee vented the piping by opening 2NV-1056, which returned the ECCS piping to water solid conditions. Additional ECCS piping locations were checked for possible gas accumulation and none were identified. The licensee implemented increased frequency UT monitoring for gas accumulation at 2NV-1056 (every 6 hours and subsequently every 12 hours) to ensure timely detection of abnormal gas accumulation until the source was determined. Based on the UT measurements, the licensee determined the size of the gas void to be approximately 2 ft3, which exceeded the existing 0.35 ft3 maximum allowable void volume for this location. The licensee initiated a past operability evaluation to determine if the NI/NV pumps would have been capable of performing their safety function during design basis accident conditions with the void in the piping. In addition, on December 18, the inspectors observed how licensee personnel were conducting the increased frequency UT measurements at location 2NV-1056 using Enclosure 13.7, Supplemental Venting, of procedure PT/2/A/4200/019, ECCS Pumps and Piping Vent. The inspectors noted that personnel were conducting the UT measurement on the 1.5 inch diameter vent piping associated with 2NV-1056 versus the 8 inch ECCS header piping that the vent valve is connected to. The procedure contained a note stating that UT measurement is performed at piping adjacent to valve due to flow being limited by 1/8 inch diameter hole in piping header. The 2NV-1056 vent piping was previously added via a modification to enhance the licensees ECCS piping gas management program. It was installed using a wet tap with a 1/8 inch drilled hole into the top of the header piping with a coupling welded over the hole to connect the vent piping. Due to the small 1/8 inch opening, water tension and/or small trash/debris can inhibit the proper communication of water between the ECCS header pipe and the vent piping. It appeared to the inspectors that the note was directing that the UT measurement needed to be conducted on the ECCS header piping and not the vent piping due to concerns that the vent piping might remain water solid while the ECCS header piping could be voiding. Following discussions with the licensee regarding this note, personnel were directed to conduct the UT measurement in the ECCS header piping. The licensee initiated PIP M-13-11297 to address this issue and to investigate how prevalent past UT measurements were conducted in the vent piping versus the header piping. This issue remains unresolved pending completion of the licensees evaluation of the impact that the gas void would have on the operation of the NI/NV pumps during design basis accident conditions and investigation into the mechanism that resulted in the excessive gas voiding not being identified during routine surveillances designed to identify such conditions. This issue is identified as URI 05000370/2013005-01, Evaluation of Gas Void Identified in Unit 2 ECCS Piping.
05000369/FIN-2013502-01	2013Q4	McGuire	Licensee-Identified Violation	The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy for being disposition as a NCV.TS 3.3.2, Function 6.f, requires that all four instrumentation channels of the TDCA pump suction transfer function be operable in Modes 1, 2, and 3. TS 3.3.2, Condition N, specifies if one or more of the pressure switch instrumentation channels are inoperable, the channel must be restored to operability within 48 hours or the associated TDCA pump must be declared inoperable. Contrary to this requirement, from September 1993 to May 30, 2013, the channel associated with pressure switch 1CAPS5390 was inoperable and the licensee failed to declare the Unit 1 TDCA inoperable within the required TS completion time. This violation was determined to be of very low safety significance (Green) because the channel would still have been capable of actuating and aligning the TDCA to its assured water source within the timeframe necessary for the pump to perform its intended safety function. This violation was documented in the licensees CAP as PIP M-13-05935.
05000324/FIN-2013004-01	2013Q3	Brunswick	Failure to Identify and Correct Nuclear Service Water Pump Shaft Degradation	An NRC identified Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality (CAQ) on the 1B nuclear service water pump (NSWP). Specifically, between June 26, 2012, and January 12, 2013, the licensee failed to identify or correct the pump shaft degradation on the 1B Nuclear Service Water Pump (NSWP) pump. This resulted in the shaft bearing delaminating and bearing material becoming dislodged and trapped in the pump strainer which caused the 1B NSWP to become inoperable. The licensee replaced the pump shaft and returned the pump to operable. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 582584. The inspectors determined that the failure of the licensee to identify and correct the 1B NSWP shaft degradation before the pump failed was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the shaft degradation resulted in the 1B NSWP being inoperable. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system and component (SSC), the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the technical specifications (TS) allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the CAP attribute because the licensee failed to implement a CAP with a low threshold for identifying issues, specifically the licensee did not enter this issue into the CAP in June 2012.
05000324/FIN-2013004-02	2013Q3	Brunswick	Inadequate Preventative Maintenance Procedure for the Service Water Pump Breakers	A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to have an adequate preventative maintenance procedure for the service water pump breakers. Specifically, from December 1, 2004, through the end of this inspection period (September 30, 2013), the licensee failed to have an adequate preventative maintenance procedure to ensure the 52S mechanism was securely bolted to the breaker for the 2C conventional service water pump (CSWP). This resulted in both discharge valves failing to open when the 2C CSWP was started, and the inoperability of the 2C CSWP. The licensee securely bolted and tightened the 52S mechanism to the breaker. The licensee entered this issue into the CAP as NCR 604452. The inspectors determined the failure to have an adequate preventative maintenance procedure for the service water pump breakers was a performance deficiency. The finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the 52S mechanism was securely bolted to the 2C CSWP breaker resulted in the failure of both 2C CSWP discharge valves to open. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. The 2C CSWP breaker was refurbished in December 2004 and installed in the plant in January 2005.
05000369/FIN-2013003-01	2013Q2	Mcguire McGuire	Failure to Implement Adequate Venting Instructions for Condensate Booster Pump Trip Instrumentation Resulting in Reactor Trip	A self-revealing finding was identified for the licensees failure to implement adequate instructions for venting condensate booster pump (CBP) emergency low suction pressure trip instrumentation which resulted in air entrainment causing a non-conservative shift in the trip setpoint. During a subsequent secondary side transient involving a heater drain tank pump trip, the non-conservative trip setpoint resulted in a premature trip of all three CBPs ultimately causing a reactor trip. The performance deficiency was more than minor because it affected the Procedure Quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective, in that, the inadequate venting allowed air entrainment in the instrumentation lines resulting in a reactor trip. This finding was determined to have very low safety significance (Green) because it did not contribute to the likelihood of both a reactor trip and that mitigation equipment or functions would not be available. No cross cutting aspect was identified.
05000369/FIN-2013002-01	2013Q1	Mcguire McGuire	Failure to Revise Turbine Inlet Pressure Calibration Procedures During Implementation of High Pressure Turbine Replacement Design Modific	A self-revealing finding was identified for the licensees failure to follow the requirements of the station modification program manual EDM 601 during implementation of a high pressure turbine replacement modification revision. This resulted in Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC) calibration procedures not being revised with the proper setpoints. The performance deficiency (PD) was more than minor because it affected the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective in that AMSAC actuated causing a turbine trip. The finding was determined to have very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The cause of this finding was related to the cross-cutting aspect of the need for work groups to maintain appropriate interfaces and communicate, coordinate with each other during important work activities as described in the Work Control component of the Human Performance cross-cutting area because necessary revisions to the AMSAC input device calibration procedures were not adequately communicated.
05000369/FIN-2013002-02	2013Q1	Mcguire McGuire	Licensee-Identified Violation	TS 3.6.3 required that each containment isolation valve be operable in Modes 1, 2, 3, and 4. TS 3.6.3, Condition A, specified if one containment isolation valve is inoperable, the flow path must be isolated within 4 hours and verified isolated once per 31 days. Contrary to the above, from November 2, 2012, to November 4, 2012, with Unit 2 in Mode 4, manual containment isolation valve 2NV-1053 was inoperable and the licensee failed to isolate the flow path within 4 hours. This violation was determined to be of very low safety significance (Green) due to the small size of the piping and that a control room air-operated valve (i.e., 2NV-840) located downstream of 2NV-1053 could have been used to isolate the penetration. This violation was documented in the licensees CAP as PIP M-12-09347.
05000400/FIN-2012009-01	2012Q4	Harris	Technical Specification Inoperability of MSIVs Due to Failure to Conduct Diagnostic Testing	The inspectors identified a non-cited violation of Technical Specification (TS) 3.7.1.5, Main Steam Line Isolation Valves, due to one or more MSIVs being inoperable for a time greater than the allowed outage time and a plant shutdown was not completed in accordance with the action statement of TS 3.7.1.5. MSIV diagnostic testing in accordance with EGR-NGGC-0205, Air Operated Valve (AOV) Reliability Program, had not been conducted by the licensee. This contributed to the licensee not identifying long-term corrosion/oxidation of the valve piston rings that resulted in the B and C MSIV failure to initially close during stroke time testing on April 21, 2012. The licensee conducted repairs of all three MSIVs and restored them to an operable condition prior to entering Mode 4 following the completion of an ongoing refueling outage. The licensee entered this condition into their corrective action program (CAP) as Nuclear Condition Report (NCR) 531773. The failure to properly classify the MSIVs as risk significant and implement MSIV diagnostic testing in accordance with the AOV program procedure EGR-NGGC-0205 was a performance deficiency (PD). The PD is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objectives of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is also associated with the containment isolation barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to conduct periodic diagnostic testing that would have identified long-term internal valve degradation due to unexpected corrosion/oxidation of the valve piston rings in all three MSIVs resulted in two MSIVs failing to initially close during TS stroke time testing on April 21, 2012, and excessive internal friction in all three MSIVs such that they may not have been capable of performing their safety-related closure function during certain design basis events. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At- Power, the inspectors determined there was an actual loss of safety function greater than the TS allowed outage time associated with the finding which required a more detailed risk evaluation. A detailed risk evaluation was performed by a regional senior reactor analyst. The result of the analysis of the risk of the PD was a delta core damage frequency (CDF) of <1E-6/year and a delta Large Early Release Fraction (LERF) of <1E- 7/year, a GREEN finding. No cross-cutting aspect was assigned to this finding because licensee decisions made in regard to classifying the MSIVs in the AOV program were made more than three years ago and therefore, not reflective of current plant performance.
05000369/FIN-2012005-03	2012Q4	Mcguire McGuire	Licensee-Identified Violation	Technical Specification 5.7, High Radiation Area, required areas with radiation levels greater than 1,000 millirem (mrem) per hour at 30 centimeters (cm) from the radiation source or from any surface which the radiation penetrates to be provided with locked or continuously guarded doors to prevent unauthorized entry. Contrary to the above, on September 23, 2011, an area with radiation levels greater than 1,000 mrem per hour at 30 cm from the radiation source or from any surface which the radiation penetrates was not locked or continuously guarded to prevent unauthorized entry. The locking method for a LHRA door leading to the reactor head stand did not prevent unauthorized entry. The padlock used to secure retaining bolts on the doors was supposed to be installed through openings in the bolts preventing them from being removed. Instead, the padlock was installed around the bolts allowing them to be removed. Corrective actions included identifying other HRA, LHRA, and VHRA barriers with the unique locking mechanism, photographing the proper locking method, providing proper instructions to individuals during key issuance, and clarifying procedural guidance on the proper use of the locking mechanism. The corrective actions were documented under PIP M-11-07009. The violation was evaluated using the Occupational Radiation Safety Significance Determination Process and was determined to be not more than very low safety significance (Green) because this finding did not have a substantial potential for over-exposure because of additional controls and warnings present such as personal ED alarming devices and LHRA posting.
05000369/FIN-2012005-02	2012Q4	Mcguire McGuire	Evaluation of the Occupational Radiation Dose Assigned to a Worker from a Piece of Contaminated Wire	While working in the reactor building an individual received a puncture wound in their hand from a piece of contaminated wire. Licensee attempts to decontaminate the wound were unsuccessful and the radioactive material from the contaminated wire remained inside the individuals hand. The licensee was reviewing that data and determining what dose to assign to the individual. The NRC will review the methodologies used once the licensee has completed its assessment to determine if a violation of regulatory requirements existed. This issue is identified as URI 05000369,370/2012005-02, Evaluation of the Occupational Radiation Dose Assigned to a Worker from a Piece of Contaminated Wire.
05000369/FIN-2012005-01	2012Q4	Mcguire McGuire	Failure to Maintain Complete and Accurate Pre-Fire Plans	An NRC-identified Green non-cited violation (NCV) of the Unit 2 Facility Operating License, Condition 2.C.4, Fire Protection Program, was identified for failure to maintain prefire plans in areas that contain safety-related equipment. The inspectors identified that all copies of fire strategy plan view for the Unit 2 lower annulus and containment were missing from their pre-fire plans and unavailable to the Fire Brigade Leader and Operations personnel in the event of a fire in the Unit 2 reactor building. Corrective actions included replacement of the missing fire strategy plan views and additional review of the fire strategy books located in the Fire Brigade Leaders Kit, Control Room, and Emergency Preparedness office. This violation was entered into the licensees corrective action program (CAP) as Problem Investigation Program (PIP) M-12-08270. The performance deficiency (PD) was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events (Fire) and adversely affected the cornerstone objective, in that, it degraded the manual fire suppression capability. The finding was determined to be of very low safety significance (Green) because the fire brigade consisted of plant personnel familiar with the plant layout and associated fire hazards and appropriate fire-fighting equipment was available. The cause of the PD was directly related to the aspect of complete, accurate, and up-to-date procedures of the Resources Component in the cross-cutting area of Human Performance because the Fire Brigade Program Administrator failed to include all approved plan view updates into the fire brigade response strategies.
05000369/FIN-2012004-01	2012Q3	Mcguire McGuire	Failure to Correctly Implement Technical Specifications Adversely Affects Requalification Operating Test Quality	An NRC-identified finding was identified associated with the quality of the simulator scenarios developed by the licensee for the licensed operator requalification annual operating test. The licensee failed to follow the Technical Specification (TS) rules of usage for concurrent inoperability as shown in TS Example 1.3-3. The licensee entered this issue into their corrective action program (CAP) as PIP M-12-4157. The performance deficiency (PD) was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective in that it impacted the licensees ability to evaluate and ensure operator performance. The significance determination was performed in accordance with Manual Chapter 0609, Appendix I, and determined to be of very low safety significance (Green). The cause of the finding was directly related to the crosscutting aspect of personnel training and qualifications in the Resources component of the cross-cutting area of Human Performance, in that the licensee failed to ensure the quality of the operating tests used to evaluate the knowledge, skills, abilities, and training provided to operators to assure nuclear safety.
05000400/FIN-2012008-01	2012Q2	Harris	B AND C MSIVs FAIL TO CLOSE DURING SURVEILLANCE TESTING	The inspectors identified an URI associated with issues in the licensees MSIV maintenance and testing. These issues were potential contributing causes to the April 21, 2012, B and C MSIV failure to stroke close. Description: Several issues were identified regarding the licensees MSIV maintenance and testing. Some of the issues identified were: FnIn the last two refueling intervals, maintenance was making minor adjustments to the actuator hydraulic speed control system to decrease the time needed to shut the valves as a result of increasing stroke test closure time results. FnBeginning in 2001, work deficiency documents were initiated due to the MSIVs experiencing difficulty in opening during refueling outage cycling. There had not been any corrective maintenance conducted requiring valve internal disassembly and the licensee had not developed any periodic PMs to visually inspect the condition of valve internals. FnThe valve vendor manual recommended weekly valve partial exercising ten percent of its total stroke in order to assure that the actuator and valve was properly functioning. Prior to 2000, this partial exercising was being performed quarterly. In 2000, the licensee revised their IST program requirements to discontinue quarterly exercising in lieu of the 18-month cold shutdown TS stroke testing that was currently being conducted. FnPrior to the current MSIV failures; the MSIVs had never been tested as part of the licensees AOV program. Summary: The licensees root cause investigation was not completed at the conclusion of the special inspection; the determination as to whether these issues represented performance deficiencies was not completed. Pending completion of the licensees root cause evaluation (RCE) and subsequent NRC review to determine if a performance deficiency exists, disposition of these issues will be tracked via Unresolved Item (URI) 05000400/2012008-01, B and C MSIVs Fail to Close During Surveillance Testing.

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