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05000366/FIN-2018003-012018Q3HatchInoperability of 2A EDG Due to Inadequate Acceptance Criteria for Determining Cleaning Requirements of Emergency Diesel Generator Day TanksThe inspectors documented a Green, self-revealing, non-cited violation of Unit 2 Technical Specification 5.4.1(a) for the licensees failure to incorporate preventative maintenance criteria for Emergency Diesel Generator (EDG) day tanks as recommended by Regulatory Guide (RG) 1.33, 9.a. Specifically, procedure 52SV-R43-001-0, Diesel, Alternator, and Accessories Inspection, Ver. 30.4, did not contain deterministic criteria in the visual inspection of the fuel filters to initiate the cleaning of the EDG day tanks and thus prevent EDG inoperability. The EDG day tanks had never been inspected and cleaned.
05000321/FIN-2018002-012018Q2HatchEnforcement Action (EA)-18-100: Unanalyzed Conditions for a Postulated Fire Discovered During NFPA 805 TransitionOn April 3, 2017, the licensee submitted Licensee Event Report (LER) 05000321, 366/2017-001-00: Unanalyzed Conditions for a Postulated Fire Discovered During NFPA 805 Transition documenting the discovery of a condition of non-compliance with the sites fire protection program (FPP). In preparation for transiting the fire protection licensing basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), a weak-link and operator manual action analysis was completed for Information Notice 92-18 type hot shorts on motor operated valves (MOV). The licensees examination of their Appendix R Safe Shutdown Analysis identified circuit configurations in multiple fire areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. The licensee failed to protect MOV cables associated with the RHR and RCIC emergency cooling systems in fire areas 0024 (Main Control Room), 1203F (Unit 1 Reactor Building), 1205F (Unit 1 Reactor Building), and 2203F (Unit 2 Reactor Building). Specifically, the licensee failed to ensure that fire induced cable impacts cannot bypass the limit and torque switches and result in physical damage to the MOVs, thus preventing the MOVs from being operated from the Main Control Room, Remote Shutdown Panel, or locally. This condition could prevent operators from achieving and maintaining safe shutdown (SSD) of the plant in the case of a postulated fire. A licensee-identified non-compliance with 10 CFR Part 50, Appendix R, Section III.G.2, was identified for the licensees failure to protect one of the redundant trains of equipment needed to achieve post-fire SSD from fire damage. Specifically, the licensee failed to use one of the means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. The inspectors performed a detailed review of the information and documents related to the LER and discussed the condition with the licensee to assess the adequacy of the licensees compensatory measures and corrective actions. Corrective Action(s): Hourly fire watches and Fire Action Statements were initiated to address the postulated condition for the identified MOVs. Additionally, the licensee committed to completing physical plant modifications to the impacted MOVs during the next Unit 1 and Unit 2 plant refueling outages to rectify the issue of potential spurious operation of the associated MOVs associated with this LER. Corrective Action Reference(s): The licensee entered this issue into their Corrective Action Program (CAP) as condition reports (CRs) 10326399, 10326401, 10326402, 10326404, and 10326405. Enforcement: Violation: 10 CFR Part 50.48(b)(1) requires that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G. 10 CFR 50, Appendix R, Section III.G.2, states, in part, that where cables or equipment, that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided: (a) separation of cables and equipment by a fire barrier having a 3-hour rating, (b) separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards and with fire detectors and an automatic fire suppression system in the fire area, or (c) enclosure of cables and equipment in a fire barrier having a 1-hour rating and with fire detectors and an automatic fire suppression system in the fire area. Contrary to the above, the licensee failed to use one of the means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. Specifically from October 1974 to April 2017, the licensee had not met the requirements of 10 CFR Part 50.48(b) to identify and protect cabling of 51 Unit 1 and Unit 2 RHR and RCIC emergency cooling system MOVs in fire areas 0024 (Main Control Room), 1203F (Unit 1 Reactor Building), 1205F (Unit 1 Reactor Building), and 2203F (Unit 2 Reactor Building). On April 3, 2017, the licensee identified the failure to protect equipment that was required to mitigate fire events and determined that fire damage could cause mal-operation of the affected MOVs, potentially leading to fire induced cable impacts which bypass the limit and torque switches and result in physical damage to the MOVs, thus preventing the MOVs from being operated from the Main Control Room, Remote Shutdown Panel, or locally. A fire-induced failure could have caused the loss of the required Safe Shutdown components. Severity/Significance: Failure to protect one train of cables and equipment necessary to achieve post-fire SSD from fire damage for fire areas designated in the Fire Protection Program (FPP) as meeting Appendix R, Section III.G.2, was a performance deficiency. This finding was more than minor because it was associated with the reactor safety mitigating system cornerstone attribute of protection against external events (i.e., fire). Specifically, failure to protect safe shutdown cables and equipment from fire damage negatively affected the reactor safety mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this issue relates to fire protection and this non-compliance was identified as a part of the sites transition to NFPA 805, this issue is being dispositioned in accordance with Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) of the NRC Enforcement Policy. The significance of this licensee-identified non-compliance with 10 CFR Part 50, Appendix R, Section III.G.2, was determined by the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. The quantitative screening approach performed by a Region II Senior Risk Analyst resulted in a calculated delta core damage frequency (CDF) of less than 1E-04, which screens this noncompliance to less-than-red significance. Additionally, in order to verify that this noncompliance was not associated with a finding of high safety significance (Red), inspectors reviewed qualitative and quantitative risk analyses performed by the licensee. These risk evaluations took ignition source and target information from the ongoing HNP fire PRA to demonstrate that the significance of the non-compliances were less-thanthan 1E-4/year). The inspectors also performed walk-downs to verify key assumptions were applicable. Based on the ignition frequency of fire sources in the affected areas, inspectors determined that the significance of this non-compliance was less-than-red. The inspectors also noted that the values in the licensees quantitative analysis were conservative, in that they used screening values instead of more detailed values. This provided additional confidence that this non-compliance was not associated with a finding of high safety significance (Red). The inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), a Confirmatory Order (ML16223A467) which extended the period for discretion, and Inspection Manual Chapter 0305. On April 4, 2018 (ML18096A955), the licensee submitted a license amendment request to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c). The inspectors reached this conclusion due to the fact that this issue was licensee-identified and will be addressed during the licensees transition to NFPA 805, it was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance (Red).
05000321/FIN-2017502-012017Q4HatchLicensee-Identified ViolationThe following licensee-identified violation of NRC requirements was determined to be of very low safety significance or Severity Level IV and meet the NRC Enforcement Policy criteria for being dispositioned as a Non-Cited Violation. Because it had the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with Inspection Manual Chapter 0612, Appendix B. This finding was also determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involved the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan was reduced. Title 10 of the Code of Federal Regulations, Part 50.54(q) states, in part, that a licensee may make changes to emergency plans without prior NRC approval only if the changes do not reduce the effectiveness of the plans and the plans, as changed, continue to meet the standards of 50.47(b) and the requirements of Appendix E. Proposed changes that reduce the effectiveness of the approved emergency plans may not be implemented without application to and approval by the NRC. Contrary to the above, on multiple occasions between 2008 and 2014, the licensee implemented changes to their Radiological Emergency Plan and Emergency Action Levels (EALs) which reduced the effectiveness of the Plan. Specifically, the licensee deleted and/or changed EAL threshold values, all of which would have resulted in a change that reduced the effectiveness of the approved Emergency Plan and was implemented without application to and approval by the NRC. Because the violation was entered into the licensees corrective action program as Condition Report 10421212, it is being treated as a Green non-cited licensee-identified SL IV violation consistent with Section 2.3.2 of the Enforcement Policy.
05000400/FIN-2017003-022017Q3HarrisReview of Removal of the Technical Support Center (TSC) Temporary Diesel GeneratorThe 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-2017003-012017Q3HarrisIncomplete and Inaccurate Emergency Action Level SubmittalsThe 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-2017003-032017Q3HarrisLicensee-Identified ViolationSection 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 .
05000424/FIN-2017002-022017Q2VogtleFailure to Follow Work Instructions for Implementation of Open Phase Protection System(Green). A self -revealing, Green, non -cited violation of Technical Specifications 5.4.1.a, Procedures, was identified for the licensees failure to redline new wiring installation associated with an open phase protection system modification, as required by work instructions . As result, control circuit wires were not installed per wiring diagrams and caused a loss of the offsite power feed to the B train 4160- volt emergency power bus. The licensee's failure to redline new wiring installation associated with an open phase protection system modification installation, as required by work instruction SNC804606 and 3 maintenance procedure NMP -MA -017 was a performance deficiency. The licensee entered this issue into their corrective action program under condition reports 10343972 and 10344136 and restored offsite power to the emergency bus by correcting the wiring configuration . The performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the in- service train of shutdown cooling (i.e. , 'A' train of the residual heat removal system ) was not affected. The finding was assigned a cross -cutting aspect of Procedure Adherence, in the Human Performance area becaus e individuals did not follow work instructions and redline procedures when installing new wiring for the open phase protection system (H.8)
05000424/FIN-2017002-012017Q2VogtleFailure to Correct a Condition Adverse to Quality involving an MSIV Manufacturing Deficiency(Green). A self -revealing, Green, non -cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to identify and correct a condition adverse to quality (i.e., manufacturing deficiency), which led to a repetitive failure of main steam isolation valve ( MSIV ) 1HV -3006B. The fail ure to determine the cause of a significant condition adverse to quality and take corrective action to preclude repetition was a performance deficiency. Specifically, the licensee failed to identify the root cause of an MSIV actuator failure on April 12, 2014, that resulted in a reactor trip. As a result, appropriate corrective actions were not taken and a repeat failure of the valve actuator caused another reactor trip on February 3, 2017 . The licensee has entered this issue into the corrective action pr ogram as condition report 10326456. This performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective of limiting the likeli hood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance (Green) because the finding did not result in a loss of mitigation equipment use d to transition the reactor to a stable shutdown condition. The finding was not assigned a cross cutting aspect since it was not indicative of current licensee performance due to the root cause evaluation in question being performed greater than three years ago
05000335/FIN-2017001-012017Q1Saint LucieInadequate Procedure Results in Adding an Incorrect Lubrication Oil to the 1B CS Motor Inboard BearingAn NRC-identified Green, non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for the licensees failure to establish, implement, and maintain written procedures covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensees failure to maintain a plant lubrication manual with correct lubrication oil specifications for the 1B containment spray (CS) pump motor resulted in adding unacceptably low viscosity lubrication oil to the inboard bearing of the 1B CS pump motor. Immediate corrective actions included restoring the 1B CS pump inboard bearing with the correct lubrication oil and placing the issue in the licensees corrective action program.The licensees failure to correctly specify the 1B CS pump motor inboard bearing lubrication requirements in licensee general maintenance procedure GMP-22 was a performance deficiency (PD). The PD was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedure resulted in adding the incorrect lubrication oil to the 1B CS pump motor bearing, causing the pump to be declared inoperable for approximately 56.5 hours. The finding screened to Green because the failure did not: (1) affect the design or qualification of the systems, structures and components, (2) represent an actual loss of function, and (3) represent an actual loss of function of at least a single train for greater than its TS allowed outage time. The finding involved the cross-cutting area of human performance, in the aspect of avoid complacency, in that, the individuals involved with the procedure revision did not implement appropriate error reduction tools to ensure the procedure was appropriately changed to reflect the new lubrication oil requirement (H.12).
05000321/FIN-2016004-012016Q4HatchFailure to Establish Icing Controls on CAD SubsystemAn NRC-identified non-cited violation (NCV) of Hatch Unit 1 Technical Specification 5.4, Procedures, was identified when procedures did not include inspection criteria for ice buildup of the Unit 1 nitrogen storage tank piping. The licensees failure to establish controls to ensure that ice buildup on the Unit 1 Containment Atmospheric Dilution (CAD) subsystem piping did not exceed ten inches was a performance deficiency. The licensee entered the condition into their corrective action plan as CR10296584, and performed de-icing activities to remove the ice buildup. This performance deficiency was more than minor, because ice buildup on the CAD system may lead to CAD subsystem inoperability if left uncorrected. The finding screened as Green because the CAD subsystem remained operable. The inspectors determined that this finding had a cross-cutting aspect in the Initiation aspect of the problem identification and resolution area, because the licensee did not initiate a condition report upon initially identifying the issue. (P.1)
05000395/FIN-2016003-032016Q3SummerLicensee-Identified ViolationV.C. Summer Operating License condition 2.c(18) states in part that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(c), National Fire Protection Association (NFPA) 805 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on September 14, 2016, the licensee failed to implement the requirements of procedure, Fire Protection Procedure, FPP-025, Fire Containment, Rev. 4, to ensure that fire door and SPB, DRAB/514, remained operable/functional. The inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and performed a Phase 1 analysis to determine the finding was of very low significance or Green. The fire confinement program element was not of low degradation, the non-suppression probability was 0.1, the fire frequencies related to the affected fire zones AB01.21.02 and FH01.04 were 2.79E-3 and 3.98E-4 respectfully, and the duration of the component inoperability was approximately 1 hour or 0.000114, which resulted in screening check frequency of 3.63E-8 that was less than the screening criteria of 1E-6. The licensee has documented this problem in their CAP as CR-16-05073.
05000395/FIN-2016003-022016Q3SummerFailure to Prescribe Work Instructions for a Temporary Repair on a Safety-Related ComponentThe inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to prescribe instructions for a temporary repair of the safety-related C component cooling water (CCW) pump outboard bearing. The licensee entered condition report, CR-16-04576, in their corrective action program for appropriate response. The inspectors determined that the failure to prescribe documented work instructions of a type appropriate to the circumstances for the temporary repair of the C CCW pump outboard bearing was a performance deficiency (PD). The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone by adversely affecting the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design control attribute was impacted because not prescribing instructions that follow vendor instructions for temporary repairs on the safety-related pump resulted in improper repairs causing reasonable doubt in operability. The inspectors evaluated the finding in accordance with IMC 0609, Significant Determination Process, Attachment 4 and Appendix A, and determined that the finding was of very low safety significance, Green, because it did not represent an actual loss of a safety-related train since the C CCW pump was operable but degraded. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of Human Performance and the aspect of resources, H.1, because the licensee failed to ensure instructions were adequate and available to support nuclear safety-related work.
05000395/FIN-2016003-012016Q3SummerFailure to Meet HRA Entry Requirements (Two Examples)The inspectors identified two examples of a Green, self-revealing, non-cited violation (NCV) of Technical Specification (TS) 6.12.1, High Radiation Area. TS 6.12.1 requires that entries into high radiation areas (HRAs) be controlled with issuance of a radiation work permit (RWP) and that individuals entering these areas be made knowledgeable of the dose rates. Contrary to that, on two separate occasions, workers made entries into HRAs without being issued an appropriate RWP and without being knowledgeable of area dose rates. Specifically, on March 28, 2016, a worker tagging a pump on the auxiliary building (AB) 400-01 slab entered a HRA without the required radiological briefing and appropriate RWP. Also, on April 18, 2016, a worker performing dry cask welding operations in the fuel handling building entered a HRA without the required radiological briefing and appropriate RWP. The licensee entered these events into their corrective action program as condition reports CR-16-01528 and CR-16-01863. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure and the ability to assess dose was not compromised. Therefore, the finding was determined to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Avoid Complacency (H.12) because in both examples there were repostings, radiation areas were upgraded to HRAs due to changing radiological conditions, and prior to entry the workers failed to stop and get updated conditions and to adhere to the postings.
05000369/FIN-2016003-012016Q3McGuireLicensee-Identified ViolationTechnical 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.
05000424/FIN-2016502-012016Q2VogtleFailure to Adequately Maintain Emergency Response FacilitiesThe inspectors identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(2), for the licensees failure to maintain the effectiveness of its emergency plan by ensuring that adequate emergency facilities and equipment to support emergency response are provided and maintained as required by 10 CFR 50.47(b)(8). Specifically, the effectiveness of the emergency plan was reduced by a change to the Technical Support Center (TSC) functionality requirements in Technical Requirements Manual (TRM) TR 13.13.1, Emergency Response Facilities, Revision 1. The requirement to maintain climate control was removed without an adequate basis to support removal. The procedure change had been in place since September 2013, and until a corrected revision is issued, a Standing Order has been put in place. The licensee entered this finding into the corrective action program (CAP) as condition report (CR) 10221041. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Emergency Preparedness (EP) cornerstone, adversely affected the associated cornerstone objective, and would have affected the emergency response organizations ability to effectively perform their duties had an emergency been declared and TSC climate control non-functional. The finding was evaluated using the EP significance determination process and was identified as having very low safety significance (Green) because it was a failure to comply with NRC requirements and was not a loss of the planning standard function or the overall function of the TSC. The finding was associated with a cross-cutting aspect in the Change Management component of the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000400/FIN-2016002-012016Q2HarrisLicensee-Identified ViolationSection 50.48 of 10 CFR, Fire Protection, states that a fire protection program that is maintained to the requirements of National Fire Protection Association (NFPA) standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, is an acceptable method for complying with the requirements of Section 50.48. Section 3.8.1 of NFPA 805 states, in part, that alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble alarms to the control room or other constantly attended location from which required notifications and response can be initiated. Contrary to the above, from December 2015, to May 17, 2016, neither the licensees design reviews nor post-modification tests identified that the fire protection system installed on the 286-ft elevation of the turbine building did not transmit trouble alarms to the Harris main control room. Following installation and testing, the newly-installed Protecta WireTM system and fire detection panel, 1-SFD-E144, were placed in service in late December 2015. On May 17, 2016, while performing maintenance periodic test, MPT-I0052, Turbine Building Local Fire Detection Control Panel LFDCP-10 Test and 1-SFD-E144 Test of the fire detection system, the technicians performing the test recognized that the remote trouble alarm function would not cause an alarm in the control room. The licensee entered the issue concerning the inadequate remote alarm function into the corrective action program via AR 2030427 and implemented actions to incorporate and test the remote trouble alarm function into the EC package. The licensee also initiated corrective actions via AR 2033716 and AR 2038682 to address issues in the design review process. Using IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green) since the reactor would still be able to achieve and maintain safe shutdown.
05000400/FIN-2016002-022016Q2HarrisLicensee-Identified ViolationSection 50.48(c) of 10 CFR and NFPA 805, 2001 Edition, Section 2.4.2.2.2(b), Common Enclosure Circuits, require that those circuits which share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required component, shall be identified to prevent propagating fires outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables. Contrary to the above, from October 1986 to September 2014, the licensee failed to meet the requirements of 10 CFR 50.48(c) and NFPA 805, Section 2.4.2.2.2(b), in that, the licensee failed to identify and provide adequate electrical fault protection for the turbine emergency oil pump control cables 11376C and 11376D. The cables could have created a common enclosure fire hazard under postulated situations which could have resulted in a secondary fire in other fire areas and could have adversely affected the capability to achieve safe and stable plant conditions. A fire-induced failure could have caused the loss of the required safe shutdown components. This violation was determined to be of very low safety significance (Green) based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. A detailed risk evaluation was performed in accordance with NRC IMC 0609 Appendix F, and NUREG/CR6850 Rev. 0 and 1, using inputs from the licensees NFPA 805 Fire PRA. The major analysis assumptions included a one-year exposure interval, and secondary fires occurring between the power supply and the fire induced hot short. The dominant sequence was a fire in the main control board causing a secondary fire in the B cable spreading room which if unsuppressed could result in the inability to achieve safe shutdown resulting in core damage. The quantitative screening approach resulted in a calculated delta core damage frequency of less than 1E-06, which screened this violation to Green (very low safety significance). This violation was documented in the licensees corrective action program as Condition Report 692766.
05000261/FIN-2016002-012016Q2RobinsonLicensee-Identified ViolationSection 50.55a(h)(2) of 10 CFR states in part, for nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of Institute of Electrical and Electronic Engineers (IEEE) Std. 6031991 and the correction sheet dated January 30, 1995. The Robinson FSAR (current licensing basis) Section 3.1.2.20, states in part that, reactor protection is designed to meet all presently defined reactor protection criteria and is in accordance with the proposed Institute of Electrical and Electronic Engineers (IEEE) 279 Standard for Nuclear Plant Protection Systems, August 1968. IEEE-279, Section 4.2, requires that any single failure within the protection system shall not prevent proper protection system action when required. Contrary to this requirement, from initial startup, until April 13, 2016, when using a FRBV (i.e., FRBV in the open position in Modes 1, 2, and 3), and a MSLB occurred, the protection system would not provide the proper system protection action. Specifically, with a single failure of the FRBV to close, the protective system action to isolate feedwater could not be accomplished. This would cause an increase in secondary mass available for release in containment structure, resulting in a higher peak containment pressure that would challenge the containment design pressure. As corrective actions, the licensee implemented a standing instruction and placed caution tags on the FRBVs to ensure the valves remain closed/isolated while operating in Modes 1, 2, and 3. Additionally, the licensee completed an engineering change to update the containment analysis and licensing basis. The licensee entered this issue into the CAP as CRs 2012658, 2020495, and 2018710. The failure to meet the single failure criterion for feedwater isolation following a main steam line break inside containment was a performance deficiency (PD). Significance Determination Process (SDP) screening in accordance with NRC IMC 0609.04 determined that the PD affected the secondary short term heat removal safety function of the mitigating systems cornerstone. The finding was determined to represent a loss of function and a detailed risk assessment was performed per NRC IMC 0609 Appendix A. The bounding analysis assumed a conditional core damage probability of 1.0, a 14 day exposure period estimated from surveillance and outage schedules, and main steam line break inside containment (MSLBIC) initiating event probability and main feedwater regulating valve bypass (MFWRVBV) failure to close probabilities from the NRC SPAR model data. The dominant sequence was an MSLBIC with a failure to close of the MFWRVBV which was assumed to lead to core damage and large early release. The risk was mitigated by short exposure period and the low likelihood of the MSLBIC and the failure to close of the MFWRVBV. The bounding analysis determined that the PD represented a risk increase of < 1.0E-7/year, a GREEN finding of very low safety significance for both core damage frequency and large early release frequency.
05000250/FIN-2016001-012016Q1Turkey PointFailure to Fully Implement Procedure QI3-PTN-1, Design ControlA self-revealing finding was identified for the licensees failure to provide complete instructions in Maintenance Support Package (MSP) 06-053 for the Isophase Bus Enclosure Collar replacement modification in the Turkey Point switchyard. Specifically, the control power circuitry termination points in the 8W43 switchyard breaker were not identified and documented in the associated MSP for removal as required by procedure QI 3-PTN-1, Design Control. As a result, a direct current (DC) ground was introduced to the back-up protection relay by a b contact when the 8W43 breaker was opened during a planned bus switching sequence. The DC ground on the back-up protection circuitry actuated the protection relay and caused both the supply breakers for the Unit 3 startup transformer (SUT) to open resulting in a loss of off-site power (LOOP) for Unit 3. The licensee entered this performance deficiency in their corrective action program (CAP) as action request (AR) 02092653 The performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to apply procedure QI 3-PTN-1 in its entirety allowed for a DC ground to be introduced to the DC back-up protection relay circuit resulting in a LOOP. Because this finding caused a LOOP and a resultant loss of residual heat removal (RHR), a detailed risk evaluation was required per IMC-0609, Appendix G, Shutdown Operations Significance Determination Process. A Senior Reactor Analyst assessed the risk significance and concluded it was of very low safety significance (Green). The risk of the event was mitigated by the multiple means that the licensee had available to them to either: 1) restore electrical power to the safety related buses, or; 2) establish alternate means of heat removal either via the steam generators or via primary feed and bleed. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not indicative of current performance since the modification package was implemented greater than three years ago.
05000327/FIN-2015004-012015Q4SequoyahFailure to Recognize and Submit for Approval a Reduction in Effectiveness of the Emergency PlanThe inspectors identified a Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations, Part 50.54(q), for changes to the licensees radiological emergency plan, effective December 18, 2014, that reduced the effectiveness of the plan and therefore, should have received NRC approval prior to making the change. Specifically, the effectiveness of TVAs Radiological Emergency Plan (Generic Part), Revision 104, was reduced by the inadvertent removal of the offsite telephone communications description for the Health Physics Network and Emergency Notification System communication tools, as well as the monthly testing of those devices. The licensees failure to recognize that Revision 104 reduced the effectiveness of the emergency plan was a performance deficiency. The licensee entered this issue into their corrective action program (CAP) as Condition Report (CR) 1093684 This finding is more than minor because it brings into question the thoroughness of the licensees review process when making changes to the emergency plan and adversely affects the procedure quality attribute of the emergency preparedness cornerstone objective. This finding is a violation of NRC requirements and because it has the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with IMC 0612, Appendix B. This finding is determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involves the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan is reduced.
05000251/FIN-2015007-032015Q1Turkey PointRequired Appendix R Instrumentation Not Functional on Unit 4 Alternate Shutdown PanelThe inspectors identified an URI regarding the processes and procedures used to evaluate the impacts of Appendix R steam generator (S/G) pressure indicators (PIs), when two of three PIs used for the Unit 4 alternate shutdown panel (ASP) 4C264 were designated as non-functional for approximately ten months. Specifically, the lack of the two S/G PIs during a fire event that requires main control evacuation may have adverse impacts on the ability to safely shutdown the plant and the effects of this condition may not have been evaluated. The USFAR, Revision C26, Appendix 9.6A, Fire Protection Program Report, Section 5.0, Alternate Shutdown Capability, stated, in part, that instrumentation and controls to achieve and maintain hot standby are provided on the ASP and supplemented by manual actions at local stations for achieving cold shutdown. Table 9.6A-2 in Section 5.0, lists components, instruments, and controls required for alternate shutdown. This table included PI-3(4)-1606/-1607/-1608, S/G pressure for A/B/C respectively. On January 29, 2015, the inspectors identified there were three ARs/works requests (WRs) on 4-PI-1606 and 4-PI-1607. These were two of the three required Appendix R S/G PIs on the Unit 4 ASP. The ARs were initiated on October 16, 2013; April 24, 2014; and July 13, 2014. At the time of discovery, the licensee did not have compensatory actions in place for this condition. The licensee captured the inspectors concerns in their corrective action program as AR 02027171, and initiated an apparent cause evaluation. As a result, the licensee performed a calibration check on 4-PI-1607 on February 21, 2015, under work order (WO) 40316782-01 and identified that the surveillance was satisfactory. In addition, they performed corrective maintenance on 4-PI-1606 on February 25, 2015, under WO 40262270-02 and returned the PI to functional status. Based upon the two non-functional S/G PIs on Unit 4 ASP for approximately ten months, the inspectors requested additional information, including the completed apparent cause evaluation, to determine if the licensee followed their processes and procedures required for Appendix R equipment. This issue is unresolved pending further licensee analysis to resolve the issue and to determine if a performance deficiency exists. This issue is identified as URI 5000251/2015007-02, Required Appendix R Instrumentation Not Functional on Unit 4 Alternate Shutdown Panel.
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