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05000255/FIN-2018003-012018Q3PalisadesWire Not Landed on Safety Injection Initiation Relay CircuitThe inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018
05000255/FIN-2018001-032018Q1PalisadesLicensee-Identified ViolationA violation of very low safety significance (Green) was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a Non-Cited Violation consistent with Section 2.3.2 of the Enforcement Policy. Enforcement:Violation: Technical Specification 3.7.6 requires that the combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T81) shall be greater or equal than 100,000 gallons. LCO 3.7.6, Condition A states that if the useable volume is not within this limit then A.1 Verify OPERABILITY of backup water supplies in 4 hours andA.2 Restore condensate volume to within limit in 7 days. Condition B states that if the Required Action and associated Completion Time is not met then B.1 Be in MODE 3 in 6 hours and B.2 Be in MODE 4 without reliance on steam generators for heat removal in 30 hours. Contrary to the above, on December 7, 2017 and March 3, 2016, the licensee failed to enter and comply with the actions required by LCO 3.7.6 Condition A and Condition B when Primary Makeup Tank Makeup Control Valve CV2008 could not be fully opened, resulting in a combined useable volume of the CST and T81 of less than 100,000 gallons.Significance/Severity Level: The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the CST and T81 volume were considered inoperable by the TS requirements, there was not a loss of safety function because credited backup water sources were available and operable.Therefore, the finding screened as Green.Corrective Action References: The licensee entered these issues into their CAP as CRPLP20175589, CRPLP20175554, CRPLP20175551, and CRPLP20161116
05000255/FIN-2018001-022018Q1PalisadesLicensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliances that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The EGM applies specifically to a structure, system, or component (SSC) that is determined to be inoperable for tornado-generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Palisades, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. On December 7, 2017, and as supplemented on January 18, 2018, Palisades submitted a request to the NRC to extend the enforcement discretion from June 10, 2018 to June 10, 2020 (ML17341A415 and ML18018A328, respectively). By letter dated February 16, 2018, the NRC granted the request to extend enforcement discretion until June 10, 2020 (ML18046A675). The EGM permitted NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provide additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within about 60 days of issue discovery. In accordance with the EGM, the comprehensive compensatory measures are toremain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Palisades was licensed prior to issuance of Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC). Specifically, GDC 2, Design Bases for Protection Against Natural Phenomena, and GDC 4, Environmental and Dynamic Effects Design Basis, discuss how SSCs important to safety shall be designed to protect against natural phenomena, such as tornadoes and shall be adequately protected against the dynamic effects of tornadoes, including protection against missiles. Palisades site-specific licensing bases compliance with GDC 2 and GDC 4 are described in the Updated Final Safety Analysis Report (UFSAR) Sections 5.1.2.2 and 5.1.2.4. Palisades protection of SSCs against tornado-generated missiles is also discussed in UFSAR Section 5.5, Missile Protection. On January 31, 2018, the licensee initiated condition report (CR) CRPLP201800556, which identified a nonconforming condition in the Palisades licensing basis. Specifically, the surge line from the component cooling water (CCW) surge tank to the CCW suction line was identified to be potentially vulnerable to a tornado missile through a doorway. The licensee previously identified a CCW system-related vulnerability on March 29, 2017. The March 29, 2017 CCW vulnerability and five additional vulnerabilities of other SSCs, which all received enforcement discretion, are documented in NRC Inspection Report 05000255/2017002 (ML17220A349). The licensee assessed this new vulnerability and concluded that previously established compensatory measures for the CCW system were adequate and no additional comprehensive compensatory actions were required. Therefore, the licensee declared the SSC operable, but nonconforming because no additional compensatory measures designed to reduce the likelihood of tornado-generated missile effects were required and the previously implemented compensatory measures were still in place. Corrective Action: The licensee documented the condition of the SSC in the CAP and documented the SSC as operable but nonconforming.Corrective Action Reference: CRPLP201800556 Enforcement: Violation: Enforcement discretion was applied to the required shutdown actions of the following Technical Specification (TS) Limiting Conditions for Operation (LCOs): TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); andTS 3.7.7, Component Cooling Water (CCW) System.Severity/Significance: The subject of this enforcement discretion associated with tornado missile protection deficiencies was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in EGM 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance (ML16355A286). 11 Basis for Discretion:The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory actions to resolve the nonconforming conditions within the required 60 days. These comprehensive measures were to remain in place until permanent repairs were completed, which for Palisades were required to be completed by June 10, 2020, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed.The disposition of this enforcement discretion closes LER 05000255/201700101, Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions.
05000255/FIN-2018001-012018Q1PalisadesFailure to Maintain an Appropriate Documented Work Instruction for Reassembly of Primary Makeup Tank Makeup Control Valve CV2008A self-revealed Green finding and an associated NCV of Technical Specification 5.4.1, Procedures, was identified for the licensees failure to have an adequate maintenance work instruction for the reassembly of Primary Makeup Tank Makeup Control Valve CV2008. Specifically, because a previous CV2008 maintenance activity failed to properly set the height of the CV2008 jam nuts, the valve guide key fell out of place and in December 2017, CV2008 was unable to be manually stroked during surveillance testing
05000255/FIN-2017004-012017Q4PalisadesImproperly Connected M&TE Leads to Unexpected AFU Fan TripA finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to follow step 5.4.4.b of Technical Specification surveillance procedure RT85DA, Control Room Emergency Ventilation Filtration Testing A Train. Specifically, the licensee failed to properly connect maintenance and test equipment (M&TE) across flow transmitter test taps which caused V26A, the air filter unit (AFU) VF26A fan, to stop 17 seconds after operators started the fan from the control room. The licensee entered this issue into their Corrective Action Program (CAP) as condition report (CR) CRPLP201705234. Corrective actions included coaching the vendor on ensuring M&TE is properly connected to plant equipment and ensuring suitable field oversight of the vendor during re-performance of the surveillance.The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, because the inspectors answered "No" to all screening questions. The finding had across-cutting aspect in the area of Human Performance, in the Field Presence aspect, for the failure to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel (H.2).
05000455/FIN-2017004-012017Q4ByronFire Barrier Impaired without AuthorizationA finding of very low safety significance and an associated NCV of Technical Specification 5.4.1.c, Procedures, was self-revealed when an Operations department supervisor identified that a fire door separating two rooms containing safety-related equipment was impaired and did not meet the requirements specified in fire protection program procedures. Specifically, on October 5, 2017, a fire door was left unattended and unable to latch due to the presence of tape over the door latch assembly. The supervisor promptly removed the tape to restore the fire doors functionality and documented the as-found condition in IR 04059911, Fire Door 0DSD474 Improperly Impaired Tape Over Latch. This issue was determined to be of more than minor significance because it was associated with the Initiating Events Cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as having very low safety significance (Green) using IMC 0609, Appendix F, Fire Protection Significance Determination Process, Question 1.4.3A, since the fire finding category was determined to be Fire Containment, due to the door not being able to latch, and the combustion loading on both sides of the door was determined to result in less than the 1.5 hour threshold. The finding affected the cross-cutting area of Human Performance in the aspect of Avoiding Complacency (H.12) because the individual that impaired the door did not recognize the inherent risk in their actions and use error reduction tools to mitigate that risk.
05000456/FIN-2017008-052017Q4BraidwoodInaccurate Analysis of RecordThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to maintain an accurate and up- to-date analysis of record for a postulated HELB in th e MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not resul t in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-042017Q4BraidwoodUntimely Corrective Action for Secondary MissilesThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with t he Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Sc reening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a los s of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-032017Q4BraidwoodFailure to Properly Correct Errors in Design Analysis for Main Steam Line Break in Main Steam TunnelThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly correct errors in the design analysis for a main steam line break in the main steam tunnel. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and completed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-022017Q4BraidwoodInadequate Blow Out Panel Design ControlThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee originall y designed the MSSV blow out panels in a manner that prevented them 3 from functioning properly. The licensee entered this issue into their CAP as A R 4075641 and corrected the design issue in March of 2009 . The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, E xhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-012017Q4BraidwoodFailure to Prevent Secondary Missiles Following a Postulated HELBThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-052017Q4ByronInaccurate Analysis of RecordThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to maintain an accurate and up- to-date analysis of record for a postulated HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-042017Q4ByronUntimely Corrective Action for Secondary MissilesThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-032017Q4ByronFailure to Properly Correct Errors in Design Analysis for Main Steam Line Break in Main Steam TunnelThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-022017Q4ByronInadequate Blow Out Panel Design ControlThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee originally designed the MSSV blow out panels in a manner that prevented them 3 from functioning properly. The licensee entered this issue into their CAP as A R 4075608 and corrected the design issue in March of 2009 . The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000454/FIN-2017010-012017Q4ByronFailure to Prevent Secondary Missiles Following a Postulated HELBThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regualtions (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of system s that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance.
05000483/FIN-2017003-012017Q3CallawaySpurious Containment Spray Pump StartThe inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to implement Preventative Maintenance Basis document IC-LSELS, Load Shed and Emergency Load Sequencer (LSELS), Revision 0. Specifically, the licensee failed to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23, a Consolidated Controls 6N232 relay driver card, within the scheduled periodicity. On June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours, of which all 44 hours w ere unplanned. As immediate corrective actions, the licensee replaced the circuit card under Job 17002747, completed post -maintenance testing, and restored the system to operable status on June 30, 2017. The licensee entered this issue into the corrective action program under Condition Report 20170 3433. The failure to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 within the scheduled periodicity was a performance deficiency. This performance deficiency was more than minor , and therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, o n June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours . Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At - Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; ( 3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non- technical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. Specifically, the total duration of inoperability was 44 hours which is less 3 than the technical specification allowed completion time of 72 hours for this system. The finding had a cross -cutting aspect in the area of problem identification and resolution associated with resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 prior to failure although this issue was documented in corrective actions ranging from April 2008 to January 2017 (P.3).
05000461/FIN-2017007-012017Q3ClintonFailure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support ProcedureThe inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the change to an emergency operating procedure (EOP) support procedure to incorporate three reactor core isolation cooling (RCIC) system interlock defeats and did not provide a basis for the determination that this change would not create a possibility for a malfunction of a structure, system or component (SSC ) important to safety with a different result than any previously evaluated in the updated safety analysis report. The licensee entered this issue into the CAP as action request ( AR ) 04056394 and planned to perform a screening for the procedure change. 3 This performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the change did not ensure RCIC system reliability and availability during and following design basis accidents because it introduced a new failure mode and added reliance on monitoring activities and manual actions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non- technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV significance because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The team determined that this finding had a cross -cutting aspect of Resources in the area of Human Performance because the licensee did not ensure that procedures, and other resources were available and adequate to support nuclear safety. Specifically, the procedure which required a 50.59 screening for changes to EOP support procedures, was not explicit in requiring the screening. (H.1)
05000255/FIN-2017003-042017Q3PalisadesLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV o f TS 5.7.2, which requires, in part, that each entryway into High Radiation Areas ( HRAs) with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation source shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry. Contrary to the above, on May 4, 2017, the licensee failed to lock or continuously guard an entryway into a HRA with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation source. Specifically, an entryway was left unguarded when the individual assigned to guard the entryway left the area prior to another guard being stationed. This issue was identified by a radiation protection technician who immediately stationed another guard. This issue was entered into the licensees CAP as CR PL 2017 02160. The failure to continuously guard the HRA entryway was a performance deficiency that was within the licensees ability to foresee and should have been prevented. The performance deficiency was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring the adequate protect ion of worker health and safety from exposure to radiation. The finding was determined to be of very low safety significance (Green) because it did not involve as -low -as-reasonably -achievable planning or work controls, there was no overexposure or substantial potential for an overexposure, and the licensees ability to assess dose was not compromised.
05000255/FIN-2017003-032017Q3Palisades12 Diesel Generator Trip During Maintenance Resulting in Additional Unavailability of the 12 DGA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self -revealed on March 31, 2017, when the 12 Diesel Generator ( DG ) tripped during performance of monthly TS surveillance procedure MO 7A 2, Emergency Diesel Generator 1 2. Specifically, during conduct of the monthly surveillance procedure, restoration activities associated with maintenance of breaker 152 213, 1 2 DG to Bus 1D, were being performed. When maintenance personnel closed the trip cutouts for the Z -phase of the 1 2 DG differential overcurrent relay, an unbalanced current flow into the differential relay resulted in relay actuation. This actuation resulted in a trip of the output breaker and subsequently the 1 2 DG. The trip caused a delay in the TS surveillance activities and resulted in the extended unavailability and inoperability of the 1 2 DG. The licensee entered this issue into their corrective action program (CAP) as condition report (CR) CR PLP 2017 01291. Corrective actions included retesting the 1 2 DG and updating the work instructions associated with the differential overcurrent relays to include caution statements that opening or closing trip cutouts for the relays while the output breaker s from the DGs to the associated buses were closed could cause the differential relay s to actuate and trip the DG . The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating System s cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, since the inspectors answered No to all screening questions. The finding had a cross- cutting aspect in the area of Human Performance, in the Work Management aspect , for the licensees failure to identify and manage risk commensurate to the work (H.5).
05000255/FIN-2017003-022017Q3PalisadesCause of 422/RPS Breaker Failure to OpenIntroduction: The inspectors identified an URI associated with the failure mechanism of the 42 -2/RPS control rod clutch breaker failure to open. Specifically, at the end of the inspection period the licensee was working to understand the cause of the breaker failure and determine the actions required to address the failure mechanism. Description : On May 17, 2017, the licensee conducted a shutdown to complete emergent repairs to a leaking seal identified on control rod drive mechanism 40. In accordance with GOP 8, Power Reduction and Plant Shutdown to Mode 2 or Mode 3 525 F, the operators depressed the reactor trip pushbutton from the EC 06, reactor protection system panel. When the pushbutton was depressed, the reactor did not trip as expected. The operators successfully tripped the reactor using the reactor trip pushbutton on the EC 02, primary process and reactor controls console. The licensee identified that the 42 1/RPS breaker tripped as expected when the reactor trip pushbutton on the EC 06 panel was depressed, however, the 42 2/RPS breaker did not trip as expected. This resulted in the reactor trip not occurring as expected when the reactor trip pushbutton on the EC 06 panel was depressed as both breakers a re required to open to result in a reactor trip. The licensee performed troubleshooting activities to determine the cause of the 42 2/RPS breaker failure. The direct cause of the breaker failure was found to be the 42 2/RPS breaker undervoltage release mechanism failing to provide enough downward force to fully depress the trip plunger. This resulted in a physical failure of the breaker to open. At the end of the inspection period, the cause of this physical failure mode was unknown. The licensees equipment failure evaluation identified that it could be age- related degradation or a physical degradation of the breaker. As a corrective action, a failure analysis of the breaker was planned. Once the failure analysis i s complete, the licensee plans to re-assess the failure mechanism and determine any additional corrective actions that are required to address the issue. This item is considered unresolved, pending the inspectors review of the failure analysis and any changes made to the equipment failure evaluation, to determine if this issue constitutes a performance deficiency and/or violation of NRC requirements. (URI 05000255/2017003 02, Cause of 42 2/Reactor Protection System Breaker Failure to Open)
05000255/FIN-2017003-012017Q3PalisadesLeft Train Emergency Diesel Generator Load Sequencer FailureIntroduction: The inspectors identified an Unresolved Item ( URI ) associated with the failure of the left train emergency DG load sequencer to run its program. Since this sequencer is required for left train DG operability, this condition resulted in an unanticipated entry into a TS shutdown action statement. The cause of this failure is currently unknown, pending the results of a vendor evaluation of a failed load sequencer component. Description : On August 3, 2017, the control room received alarm EK 1145, Sequencer Trouble, unexpectedly. The operators identified that the indication lights were not lit on the left channel load sequencer, MC -34L101; declared the associated DG inoperable; and entered the appropriate TS action statement. The failed sequencer was removed and replaced with a new module that was satisfactorily post -maintenance tested and the left train EDG was subsequently declared operable on August 4, 2017. The failed sequencer was sent to an on -site lab for further troubleshooting. No obvious visual signs of failure were identified and the electrolytic capacitors in the module all tested satisfactorily. The module was then bench tested using a test program, which identified that although it would power up, no program would run. The licensee completed an equipment failure evaluation to review the bench test data, along with information collected in the failure modes analysis, and determined that the direct cause of the failure was a memory fault within the sequencer module that caused the sequencer to lock -up and not run its program. A fault in the memory module, memory processing interface circuitry, or the executive module could have caused the sequencer to lock up. At the end of the inspection period, further examination by t he vendor was required and in progress to determine the exact initiating point of the fault. In addition to replacing the failed sequencer, the licensees immediate corrective actions included inspecting the right train load sequencer and completing the quarterly surveillance test to ensure proper operation; the results of which were satisfactory. A plant operating experience review was conducted and did not identify any prior memory failures on the load sequencers. Once the vendors evaluation is complete, the licensee plans to re-assess the failure mechanism and any additional corrective actions required. This item is considered unresolved, pending the inspectors review of the vendor analysis and any changes made to the equipment failure evaluation, to determine if this issue constitutes a performance deficiency and/or violation of NRC requirements. (URI 05000255/2017003 01, Left Train Emergency Diesel Generator Load Sequencer Failure )
05000483/FIN-2017002-022017Q2CallawayFailure to Analyze the Effect of Changes to Maintaining the Gaitronics SystemSeverity Level IV. The inspectors identified a Severity Level IV non- cited violation for the licensees failure to perform an analysis of a change to processes supporting the emergency preparedness program that demonstrated the change did not reduce the effectiveness of the emergency plan in accordance with the requirements of 10 CFR 50.54(q)(3). There were no immediate safety concerns associated with this violation because less than 10 percent of the public address speakers were determined to be degraded or non- functional. This issue has been placed in the licensees corrective action system as Condition Report 201702343. The failure to perform an analysis of the effect of changes in processes supporting emergency preparedness is a performance deficiency within the licensees ability to foresee and correct. The finding was more than minor because the finding was associated with the Facilities and Equipment Cornerstone attribute and adversely affected the Emergency Preparedness Cornerstone objective. The finding was assessed using traditional enforcement because the licensees failure to perform a required analysis impacted the regulatory process . The finding was evaluated using the NRCs Enforcement Policy, dated November 1, 2016, Section 6.6(d) , and was determined to be a Severity Level IV violation because the violation did not affect radiological assessment or offsite notification. Traditional enforcement violations are not assessed for cross -cutting aspects.
05000483/FIN-2017002-012017Q2CallawayFailure to Follow Motor Control Center ProcedureGreen . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Procedure MPE-ZZ-QS001, Cleaning and Inspection of Motor Control Centers, Revision 34. On May 2, 2017, the licensee failed to ensure contactors operated freely per step 7.6.8 during reassembly of motor control center NG08F for the essential service water cooling tower by pass valve EFHV0066. As a result, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. As immediate corrective actions, the licensee replaced the starter assembly under Job 17001973, completed testing including electrically cycling valve EFHV0066, and restored the system to operable status on May 4, 2017. The licensee entered this issue into the corrective action program under Condition Report 201702418. The failure to follow Procedure MPE-ZZ-QS001 was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it adversely affected the configuration control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non- technical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. Specifically, the total duration of 3 inoperability was approximately 57 hours which is less than the allowed completion time of 72 hours for this system. The finding had a cross-cutting aspect in the area of human performance associated with challenge the unknown because the licensee failed to stop when faced with uncertain conditions. Specifically, the maintenance technician encountered resistance when manually operating the contactors, signed off the step as complete, and later rationalized the decision with the supervisor aft er completing the work (H.11 ).
05000255/FIN-2017002-012017Q2PalisadesInadequate Protection from Tornado Missiles Identified Due to Non- Conforming Design ConditionsA finding and an associated violation of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified based upon the lack of adequate tornado missile protection to the safety -related equipment listed above. The finding was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado- generated missile non -compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15 002, Revision 1, Enforcement Discretion for Tornado- Generated Missile Protection N on- Compliance, and can be found in ADAMS Accession No. ML16355A286. Because this finding and violation was identified during the discretionary period covered by Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado Missile Protection Non-Compliance and because the licensee, prior to the expiration of the associated LCO, took initial compensatory measures that provided additional protection such that the likelihood of tonado missile effects were lessoned, followed by more comprehensive compensatory measures that w ere completed within approximately 60 days of issue discovery , and has final corrective actions planned, the NRC is exercising enforcement discretion by not issuing an enforcement action, as discussed in Section 1R15.2 of this report.
05000255/FIN-2017001-012017Q1PalisadesLicensee-Identified ViolationThe licensee Identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix R, Section III.G.2, which requires, in part, that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shut down conditions are located within the same fire area outside of primary containment, one means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, as of October 1, 2010, the licensee failed to ensure that one of the redundant trains was free of fire damage in areas where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment. Specifically, the licensee failed to analyze a fire scenario in the 1C switch gear room, screen- house room, and component cooling water pump room that could potentially damage the control cable before the load cable, and therefor e result in the loss of safety -related 2400 volt alternating current (VAC) bus 1C and/or 1D, with subsequent loss of equipment credited for Appendix R compliance to support safe shutdown in the event of such a fire. The licensees failure to analyze an Appendix R fire scenario for the three fire areas described above w as a performance deficiency . 21 The performance deficiency 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was determined to be of very low safety significance (Green) because it did not impact the licensees ability to reach hot shutdown because operator manual actions would have allowed operators to shut down the plant following a fire. The licensee identified this issue during the transition to NFPA 805, entered the issue into their CAP as CR PLP 2010 04255, and implemented compensatory measures, including fire watches. The violation was not willful and routine licensee efforts, such as normal surveillance or quality assurance activities, were not likely to have previously identified the violation due to the specific sequence of fire cable damage required for such an Appendix R fire scenario. As a result, the inspectors concluded that the violation met all four criteria for exercising enforcement discretion established by Section 9.1 of the NRCs Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues; therefore, the NRC is exercising enforcement discretion to not cite this violation
05000454/FIN-2017001-012017Q1ByronLicensee-Identified Violation

On March 11, 2017 , with Unit 1 shutdown and in a refueling outage, pipefitters as signed to cut out and replace service water valve 1WS413 discovered that piping was blocked upstream of the valve and the work scope was appropriately changed to remove the blocked piping. Taking action they believed was allowed by the work instructions, the pipefitters opened a pipe union and removed the pipe. They then set the removed section containing valve 1WS023C on a nearby tripod to continue work. A system engineer performing a walkdown in the area identified that the removed valve had a clearance (danger) tag on it and immediately stopped work and contacted the operations department. Technical Specification 5.4.1 requires , in part , that written procedures be established, implemented and maintained covering the procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. One administrative procedure recommended in Appendix A is , Equipment Control ( e.g. locking and tagging). OP AA 109 101, Clearance and Tagging, accomplished the locking and tagging requirement for Byron Station. Section 5.2, Danger Tags, established standards for implementation of the tagging process. Step 5.2.2 stated , A component with a Danger Tag attached to it shall not be physically removed from the system. Contrary to the requirements stated above, a component with a danger tag attached was physically removed from the system on March 11, 2017. Specifically, pipefitters disconnected a pipe union and removed associated service water piping from the system that contained valve 1WS023C which had a clearance (danger) tag attached.

The licensee immediately verified that the cooler the piping served was out -of-service on both the supply and return sides with a clearance boundary in place and drained so that the workers were not exposed to a pressurized sourc e. The workers immediately acknowledged their error stating they did not see the tag because they were focused on the demolition activities. The issue was entered into the licensees CAP as IR 03984215 , and the maintenance organization conducted a stand down to reinforce the station standards for compliance with the clearance procedure. The inspectors determined that this issue was more than minor because the performance deficiency adversely impacted the Configuration Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown operations. The inspectors determined the issue was of very low safety significance , or Green by answering No to all screening questions in IMC 0609, Appendix G, Shutdown Operations Significant Determination Process, Exhibit 2, Initiating Events Screening Questions.

05000255/FIN-2016004-012016Q4PalisadesFailure to Have Appropriate Controls in Place for Combustible MaterialsGreen. A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 48(c) was identified by the inspectors for the licensees failure to appropriately implement the requirements of procedure ENDC161, Control of Combustibles. Specifically, between January 1, 2016 and October 22, 2016, the inspectors identified several examples of the licensees failure to have appropriate controls in place for the storage of combustible materials in excess of the limits required for those respective areas without a completed transient combustible evaluation (TCE). Also, on several occasions from October 19, 2016 to October 22, 2016, the required compensatory actions for a TCE related to the dry fuel storage cask transporter vehicle were not appropriately implemented as required by procedure ENDC161. The licensee entered these issues in their corrective action program (CAP) as condition reports (CRs) CRPLP201603633, CRPLP201605148, and CRPLP20160564. Corrective actions for these issues included completing the required TCEs, ensuring the combustible materials in the areas were addressed by the combustible loading calculations, and ensuring appropriate compensatory measures were implemented. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Protection Against External Factors attribute, in the area of Fire, 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 (i.e., core damage). Specifically, transient combustible materials without required TCEs were stored in the charging pump cubicles and in the refueling and spent fuel pool areas. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, since none of the stored materials were self-igniting, low flashpoint liquids, or heat sources and was therefore assigned a Low degradation rating. The finding had a cross-cutting aspect of Training in the Human Performance cross-cutting area due to the common element of a lack of knowledge of the individuals with the control of combustibles process and understanding their roles in that process (H.9).
05000255/FIN-2016004-022016Q4PalisadesFailure to Correct an Adverse Condition Associated with Diesel Generator Load Sequencer ModuleGreen. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to correct an adverse condition associated with the emergency diesel generator (DG) load sequencer and power supply module as revealed when the electrolytic capacitor failed two days after installation. The 12 DG was declared inoperable, the licensee replaced the failed module, and an equipment apparent cause evaluation was completed for the equipment failure. An internal operating experience review revealed that a similar issue occurred in 2005 and corrective actions to address that failure, which included establishing shelf life and age requirements for electrolytic capacitors that were part of power supply modules, were not applied to this module. The licensee entered this issue into their CAP as CRPLP201603260. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, because the performance deficiency was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to correct a condition adverse to quality, which rendered the 12 DG inoperable. This condition would have prevented the DG from automatically starting and loading on the prescribed signal. The finding was screened in accordance with IMC 0609, Appendix A, and was determined to have very low safety significance (Green) based on answering No to all the screening questions under the Mitigating Structure, System and Components, and Functionality section. The inspectors concluded that the corrective actions for the adverse condition of the aging electrolytic capacitors should have been implemented greater than three years ago, so the finding was not reflective of current licensee performance. Therefore, no cross-cutting aspect was identified.
05000255/FIN-2016004-032016Q4PalisadesFailure to Translate Design Analysis Stack-up Configuration into Specifications, Drawings, Procedures, and InstructionsGreen. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to establish measures to assure that the applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to provide instructions in procedures to construct the spent fuel dry cask loading stack-up, in the safety-related auxiliary building, in the configuration that had been analyzed for in the stack-up seismic design basis calculation. In addition, the licensee failed to provide instructions in revised procedures to construct the stack-up without certain gaps as 4 specified in the stack-up seismic design basis document. The licensee documented these issues in their CAP as CRPLP201600646, CRPLP201601308, CRPLP201601558, CRPLP201604497, and CRPLP201604826; revised the stack-up seismic analysis to address the identified issues; and translated the analyzed stack-up design configuration into stack-up installation procedures prior to performing stack-up operations with spent nuclear fuel in the multi-purpose canister. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Barrier Integrity cornerstone and adversely 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 performance deficiency resulted in a stack-up configuration that did not ensure stack-up dynamic stability or Auxiliary Building structural integrity to maintain radiological barrier functionality during a design basis seismic event. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of the Auxiliary Building. The finding had a cross-cutting aspect of Field Presence in the Human Performance cross-cutting area, because licensee senior managers failed to ensure effective supervisory and management oversight of contractor activities related to the seismic analysis and installation of the stack-up configuration (H.2).
05000440/FIN-2016009-012016Q3PerryFailure to Implement a Periodic Replacement Program for FLEX HosesA finding of very low safety significance was identified by the inspectors for failing to establish a periodic replacement program for the high-temperature rated hoses used during a mitigating strategy for suppression pool cooling. Specifically, the licensee failed to create a periodic replacement program for high temperature FLEX hoses based on the vendor recommendation of a six year shelf-life or justify deviation from the recommendation. The licensee entered this issue into the corrective action program as CR201609776 with an action to generate the appropriate repetitive task for periodic replacement of the high-temperature rated hose. No violation of NRC requirements were identified. This performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects 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), and is therefore a finding. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, appendix M, Significance Determination Process using Qualitative Criteria, informed by draft appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA-12-049 and EA-12-051). The finding screened as very low safety significance, Green, because the inspectors answered no to all Appendix O questions. This finding had a cross-cutting aspect of Procedure Adherence in the area of Human Performance because the licensee failed to follow procedural guidance to replace hoses based on vendor recommendations.
05000440/FIN-2016009-022016Q3PerryFailure to Establish a Periodic Maintenance Program for Communications Equipment Associated with FLEXA finding of very low safety significance was identified by the inspectors for failing to establish period tasks to check the operation of recently installed FLEX related communications equipment in accordance with the Perry Nuclear Power Plant FLEX Final Integrated Plan Report. The licensee entered this issue into the corrective action program as CR201609746 and 201609747 to develop the appropriate periodic maintenance tasks. The finding was determined to be more than minor because it was associated with the Emergency Preparedness Cornerstone Attribute of Facilities and Equipment which includes Maintenance Surveillance and Testing of Facilities, Equipment and Communications Systems. Specifically, communications equipment, particularly batteries, degrade over time and without periodic checks to verify functionality, the equipment might not be available for response to a potential accident. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, appendix M, Significance Determination Process using Qualitative Criteria, informed by draft appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA-12-049 and EA-12-051). The finding screened as very low safety significance, Green, because the inspectors answered no to all Appendix O questions. This finding has a cross-cutting aspect in the area of Human Performance, Work Management because a task to create the activities was initiated, but the completion date was postponed well past the date at which the licensee declared compliance with mitigating systems orders.
05000461/FIN-2016007-012016Q3ClintonFailure to have Hose Configurations that were Verified to be able to Ensure a Timely and Successful Implementation of a FLEX StrategyTwo examples of a finding of very low safety significance was identified by the inspectors for the licensees failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a flexible response (FLEX) strategy. Specifically, the licensee did not ensure through evaluations, calculations, analyses or any other means that the strategy for maintaining core cooling, containment heat removal and Spent Fuel Pool (SFP) cooling during a Beyond-Design-Basis External Event (BDBEE) flooding scenario would be capable of fulfilling its function. No violation of NRC requirements were identified. The performance deficiency is more than minor because it was associated with the mitigating systems cornerstone objective attribute of protection against external factors, specifically the BDBEE flood hazard, and it adversely affected the cornerstone attribute of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Issues identified through TI191 are evaluated through a cross-regional panel using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, as informed by draft Appendix O, Post Fukushima Mitigation Strategies Significance Determination Process. The finding was determined to be of very low safety significance (Green). The inspectors concluded that the cause of the finding involved a cross-cutting component in the Human Performance area of Design Margins because the organization did not ensure the selected strategy contained the required verification that it could be successfully implemented.
05000255/FIN-2016003-012016Q3PalisadesFailure to Appropriately Select and Review for Suitability of Application the Control Switch and Circuit Design of the Engineered Safeguards Room Cooler FansA self-revealed finding of very low safety significance and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, was identified for the failure to appropriately select and review for suitability of application the control switch and circuit design of the engineered safeguards room cooler fans. Specifically, on July 27, 2016, when the licensee was conducting troubleshooting activities for the tripping of engineered safeguards room cooler fan V27B, it was revealed that the control switch design was break before make and as the hand switch was transitioned from one position to the next, the supply voltage and the motor became out of phase and caused an overcurrent trip of the breaker. This resulted in an unplanned entry into a 72 hour limiting condition for operation (LCO) for the right train of the emergency core cooling system (ECCS). In the apparent cause evaluation (ACE) for this issue, the licensee determined that the contributing cause had not previously addressed this particular failure mode (i.e. the control switch and circuit design) when similar overcurrent events occurred in the past. Prior corrective actions included adding guidance to system operating procedures to pause between hand switch movements and replacing other components within those systems. These actions were not successful in eliminating this failure mode. The licensee documented the issue in their CAP, planned to revise the control circuit and switch design, and added specific procedural steps on how to operate these fans until the design change was implemented. The finding was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Reliability and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, as a result of the overcurrent trip of its breaker, V27B was declared non-functional and unavailable and the equipment in the room it cooled was declared inoperable, which included the A high pressure safety injection (HPSI) pump and the A containment spray (CS) pump. This led to an unplanned entry into a 72 hour LCO for the right train of ECCS. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution and was related to the cross-cutting component of Evaluation, which required that the licensee thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. As discussed above, in the ACE for this issue the licensee determined that the corrective actions associated with the identified contributing cause following similar overcurrent events that occurred in the past had not addressed or been successful in eliminating this failure mode (PI.2).
05000255/FIN-2016003-022016Q3PalisadesHourly Fire Tour DiscrepanciesThe inspectors identified an unresolved item (URI) related to discrepancies found during fire tour daily log sheet and corresponding badge record reviews. Specifically, the NRC is in the process of reviewing the licensees evaluation of the root and contributing causes of the issue, as well as the corrective actions to prevent recurrence. Also, the NRC will verify that the licensees actions taken to address the issue are sustainable. On May 24 and 25, 2016, while the inspectors were observing a maintenance activity on a service water pump in the screenhouse, they noted that hourly fire tours were not being conducted consistently by security personnel. The inspectors requested plant room badging records and copies of the hourly fire tour daily log sheets from the licensee for hourly fire tours completed on May 24 and 25, 2016. The inspectors identified that some areas on the fire tour log sheets were annotated as complete, yet there were no corresponding badge records for these areas. The inspectors requested additional fire tour daily log sheets and badge records for May 31 and June 1, 2016 for an extent of condition review. Additional issues were identified with the fire tour log sheets not corresponding with badge records for certain plant areas required to be covered by the hourly fire tours. On June 8, 2016, the inspectors discussed these discrepancies with the licensee. The licensee entered this issue into the CAP and promptly began an extent of condition review of the fire tour daily log sheets and plant room badging records for the period of March 1, 2016 through June 8, 2016. The condition report included actions to conduct a root cause evaluation to determine the root and contributing causes of the discrepancies identified in the fire tour and badging records and formulating corrective actions to prevent recurrence. The licensees immediate interim corrective actions included direct supervisor observation of all hourly fire tours being conducted, newly formatted fire tour log sheets with additional detail added, and re-training of personnel conducting the tours on the requirements and expectations for completion of the activity. Pending NRC review of the licensees evaluation of the issue, subsequent corrective actions to prevent recurrence, and verification that the actions are sustainable, this issue is unresolved.
05000255/FIN-2016001-022016Q1PalisadesMovement of Radioactive Material Results in an Unposted and Un-Barricaded High-Radiation AreaA self-revealed finding of very low safety significance and an associated NCV of Technical Specification 5.7.1 was identified when movement of a bag of radioactive material caused an area to become a high radiation area without the proper posting and barricades. The licensee immediately moved this bag of radioactive material to a posted locked high-radiation area and entered this issue into their CAP as CRPLP201505019. The performance deficiency was determined to be more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the movement of the bag from an area that was a high-radiation area to an area that was not posted and barricaded as a high-radiation area removed a barrier that was intended to prevent workers from receiving unexpected dose. The finding was determined to be of very low safety significance in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The violation was of very low safety significance because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding had a cross-cutting aspect of Teamwork in the Human Performance cross-cutting area because the individuals and work groups involved did not communicate or coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4).
05000255/FIN-2016001-032016Q1PalisadesFailure to Meet the Minimum Staffing Requirements of the Fire BrigadeAn NRC-identified finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 48(c) and the National Fire Protection Association (NFPA) Standard 805 Section 3.4.1 was identified for the failure to meet the minimum staffing requirements for the Fire Brigade on January 4 and 5, 2016. Specifically, two nuclear plant operators (NPOs) who had their Fire Brigade qualifications suspended, stood watch as Fire Brigade members during day shift on January 4, 2016 and approximately one half of day shift on January 5, 2016. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PLP-2016-00198, performed an apparent cause evaluation, successfully performed a fire drill to requalify the Fire Brigade members with suspended qualifications on January 6, 2016, and planned to update the tracking method used to validate drill completion for Fire Brigade qualifications. The performance deficiency was determined to be more than minor because it was associated with the Protection Against External Factors 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 (i.e., core damage). The finding screened as having very low safety significance based on using qualitative criteria located in IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding had a cross-cutting aspect of Documentation in the Human Performance cross-cutting area because the licensee informally tracked drill completion and this information was not accessible to each individual Fire Brigade member to validate their qualifications (H.7).
05000255/FIN-2016001-042016Q1PalisadesLicensee-Identified ViolationTitle 10 CFR 50.54(m)(2)(iii), Condition of Licenses, states that when a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the units technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. TS 5.2.1 states in part, that during any absence of the Shift Supervisor from the control room while the plant is in Mode 1, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. Contrary to the above, at approximately 2:00 a.m. on September 2, 2015, with the unit in Mode 1, the Command SRO left the control room without another SRO being present in the control room and without turning over the command function. A few minutes prior to the event, the shift Command SRO turned over to the Shift Technical Advisor (STA) the Command SRO function of the control room so that the shift Command SRO could take a break outside the control room boundary. A minute or so after the STA (who had the Unit Command SRO function at the time) left the control room, a control room reactor operator observed that there were no SROs in the control room and summoned the Shift Manager from an office across the hall to the control room. The Shift Manager then assumed the Command SRO function and the STA was called back to the control room. This issue was identified by the licensee on September 2, 2015, and documented in CRPLP201503637, The SRO with Command and Control Momentarily Left the Control Room. There were no risk-significant plant evolutions in progress and no adverse reactor plant operations occurred during the SROs absence. The STA was relieved from shift responsibilities until corrective actions were taken. The inspectors screened the issue using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power. The inspectors reviewed the screening questions under all three Cornerstones and all of the logic questions did not apply, therefore the finding screened as having a very low safety significance (Green).
05000255/FIN-2016001-052016Q1PalisadesLicensee-Identified ViolationTS Limiting Condition for Operation (LCO) 3.0.6 states, in part, that when a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered; only the support system LCO actions are required to be entered. TS LCO 3.0.6 further specifies that an evaluation shall be performed in accordance with TS 5.5.13, Safety Function Determination Program. Palisades Administrative Procedure 4.11, Safety Function Determination Program, step 5.4.3 requires documentation of entry into TS LCO 3.0.6 for the inoperable supported system in the Operations Log. Contrary to the above, on January 19, 2016, the licensee failed to document entry into TS LCO 3.0.6 in the operations log when work was commenced on breaker 521214, Motor Control Center (MCC) 22 and MCC24 480 Volt feeder breaker. The licensee identified this issue when a similar condition was entered on January 22, 2016 and documented the missed entry into TS LCO 3.0.6 in CRPLP201600413, Operations Failed to Log Entry into LCO 3.8.1B and LCO 3.5.2B or LCO 3.0.6. The licensee provided coaching to the individuals involved. The inspectors screened the issue using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating System Screening Questions, and answered No to all the questions. Therefore, the finding screened as having very low safety significance (Green).
05000255/FIN-2016001-012016Q1PalisadesDesign Review of Modification to Track Alley Wall for Dry Fuel Storage ActivitiesThe inspectors identified a unresolved item (URI) associated with the design review of a modification to the Track Alley wall for dry fuel storage (DFS) campaign activities. Specifically, the licensee is currently revising the process applicability determination (50.59 and 72.48 screenings), and reviewing any necessary actions, associated with altering the newly modified wall in support of upcoming DFS campaign activities. The wall, a protective barrier with safety functions per the UFSAR, in its newly modified condition, will be altered when the steel plate covering the opening cut into it will be raised to accommodate the DFS transporter. The DFS campaign is currently on hold pending resolution of other issues. In January 2016, the licensee began work on an engineering change to permanently modify the west wall of Track Alley in order to accommodate the new transporter used for moving the casks associated with the dry fuel storage campaign. This modification removed a section of the reinforced concrete wall by cutting out an opening approximately 9 feet wide by 4 feet high by 18 inches deep into the existing wall. A three inch thick steel plate was mounted onto vertical rails which can slide down to cover the window cut into the wall and raised to open the window for when the transporter is brought into Track Alley. The west wall of Track Alley is also the east wall of the Technical Support Center (TSC). This wall is designed to withstand seismic, high wind, and tornado missile loads. It also serves as a radiation protection barrier for personnel in the TSC during emergency situations. The permanent modification of cutting the opening in the wall and installing the steel plate, to provide equivalent protection of the 18 inches of concrete that were cut out, was evaluated in Engineering Change 59170 and calculation EAEC5917001. The inspectors reviewed these documents, the supporting process applicability determination (50.59 screening), and risk assessment of implementing the design change. During this review, the inspectors identified that the licensee did not assess the alteration of the wall, a protective barrier with safety functions per the UFSAR, when the steel plate covering the window would need to be raised to accommodate the DFS transporter. The inspectors questioned this condition and the licensee subsequently completed a process applicability determination (PAD) form (72.48 and 50.59 screening). When reviewing the PAD, the inspectors questioned the licensees underlying assumption that moving the steel plate to uncover the window was considered to be in support of a maintenance activity and, hence, screened out of the 50.59 process, including not requiring certain compensatory actions for the walls safety functions during the period of time in which the opening was exposed. At the end of the inspection period the licensee was reviewing their assessment. Once their review is completed, including any changes that may be made, the inspectors will re-assess their evaluation and determine what actions, if any, will need to be accomplished in support of the DFS campaign. Since the campaign is on hold, a URI is being opened to track resolution of this issue.
05000266/FIN-2015004-032015Q4Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 5.4.1, Procedures for the failure to maintain the emergency operating procedures (EOPs). The licensees TS 5.4.1 required, in part, that written procedures shall be maintained including the EOPs required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. During design reviews, the licensee discovered that following a 2012 calculation update, the licensee inconsistently applied pre and post-modification uncertainties that had resulted from a 2010 modification associated with the sensitivity and calibration of both units Subcooling Margin Monitors. Ultimately the calculative errors resulted in 19 EOP Subcooling setpoints being incorrectly calculated. These Subcooling setpoints are used throughout the licensees EOPs network to provide operators with discrete indications for key EOP decision making. Contrary to the above, from April 12, 2012 through November 5, 2015, the licensees EOP network of procedures for both Unit 1 and 2, contained the incorrect setpoints for decision points with respect to subcooling. The licensee entered this issue into the CAP as AR 02089011 and AR 02099152. The inspectors consulted the Region III Senior Reactor Analysts and determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 1, 2012 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors determined that the issue was a design or qualification deficiency confirmed not to result in a loss of operability; therefore, answered yes to question A.1 in Exhibit 2, Section A, Mitigating SSCs and Functionality. This resulted in the finding screening as Green.
05000301/FIN-2015004-022015Q4Point BeachInadequate Evaluation of Non-Conforming Auxiliary Feedwater System Pipe DefectsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to maintain a Unit 2 auxiliary feedwater system (AFW) pipe segment containing linear defects in accordance with the design and material specifications. As a corrective action, the licensee performed light filing to remove the defects from this pipe segment. The licensee entered the failure to maintain the AFW pipe segment in accordance with the design into the corrective action program (CAP) as action request (AR) 02084077, and was evaluating additional corrective actions. This finding was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to maintain the Unit 2 AFW pipe segment containing linear defects in accordance with the design and material specifications could result in an increase in the possibility of pipe leakage or failure. In addition, the failure to maintain the AFW pipe segment containing linear defects in accordance with the design and material specification adversely affected the Mitigating System Cornerstone attribute of Equipment Performance because it could result in failure of AFW piping which would reduce the availability and reliability of the this mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors answered Yes to screening question A.1 of Exhibit 2. Although this finding adversely affected the design or qualification of the AFW pipe segments, the finding screened as very low safety significance (Green), because it did not result in the loss of operability or functionality of the affected pipe segment. This finding has a cross-cutting aspect in the Teamwork (H.4) component of the human performance cross-cutting area. Specifically, the licensees Projects Team responsible for the AFW modifications did not effectively communicate and coordinate with the licensees Programs Engineering Group for resolution of the AFW pipe nonconforming conditions to ensure nuclear safety was maintained.
05000237/FIN-2015004-012015Q4DresdenFailure to Maintain Design Control of Secondary Containment Interlock DoorsA finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed on September 4, 2015, when the integrity of the Secondary Containment for Units 2 and 3 was not maintained for 39 minutes when interlock features designed to prevent both doors of a Secondary Containment interlock from being simultaneously open prevented the closure of Reactor Building to Turbine Building doors 47 and 48 following simultaneous operation during routine access of the interlock by plant personnel. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of design control, and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance (Green) because the inspectors answered yes to the Barrier Integrity Screening Question C.1, Exhibit 3 of IMC 0609, Appendix A. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee failed to implement a modification which addressed a known design deficiency in the 570 foot elevation Secondary Containment interlock in 2013. The licensee reasoned that the interlock was a low traffic area and that it would be unlikely that the doors would be open simultaneously. (H.14)
05000266/FIN-2015004-012015Q4Point BeachFailure to Follow Fire Protection Program Requirements for Care, Use and Maintenance of Fire HoseThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of license condition 4.F for the licensees failure to have procedures or instructions to prevent firefighting booster hoses from being kinked and/or twisted on hose reels. Specifically, booster hoses were installed on hose reels in both units containments and in the turbine building (TB), which were twisted and kinked. The licensees corrective actions included rewinding hoses in the Unit 2 containment, four hoses in the TB, and creating compensatory measures for hose reels for the Unit 1 containment. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee failed to ensure that activities such as inspection, testing, and maintenance of fire protection systems were prescribed and accomplished in accordance with documented instructions, procedures, and drawings. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1A, because the inspectors determined that the impact of a fire would be limited to one train/division of equipment for the affected fire areas and at least one credited safe shutdown path would be unaffected. This finding has a cross-cutting aspect of Training (H.9), in the area of human performance, because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce, and instill nuclear safety values. Specifically, the inspectors determined that operations personnel were not adequately trained to recognize deficiencies associated with firefighting equipment standards, such as kinked and twisted hoses on hose reels, and subsequently failed to initiate actions to remedy such conditions.
05000266/FIN-2015003-042015Q3Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 3.8.9; Distribution SystemsOperating, Condition A, which required the licensee to immediately declare associated supported features inoperable for the 4.16 kV safeguards busses. Failure to implement this action subsequently required the licensee to place both units in mode 5 within 36 hours. Contrary to the above, the licensee discovered that numerous occasions existed over the past three years where safetyrelated 4.16kV switchgear associated with B Train EDGs was inoperable due to the inoperability of the W-185A and W-185B, 1A-06 and 2A-06 Switchgear room fans, which were required support systems for the EDGs and associated switchgear. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, and determined that the finding required a detailed risk evaluation which was performed by Region III SRAs. The SRAs gathered data from licensee GOTHIC model calculations, licensee engineering evaluations associated with the POR of the condition and the NRCs Standard Plant Analysis Risk model. Based on the SSCs being available for their respective 24-hour mission time(s), the SRAs determined that the increase in CDF for this issue was negligible and the delta risk is of very low safety significance (i.e., Green). The licensee reported this condition in LER 2015-004-00, which was closed in Section 4OA3 of this report. The licensees corrective actions included improving administrative and procedural controls for removing these fans from service and used lessons learned from this condition to implement corrective actions to improve procedural guidance for similar activities where ventilation systems may cause support system inoperabilities.
05000266/FIN-2015003-012015Q3Point BeachIncomplete Functionality Assessment for Flooding in the Diesel Generator BuildingThe inspectors identified a finding of very low safety significance for the licensees failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 19. Specifically, when the licensee identified that internal flood sources in the diesel generator building (DGB) were larger than the drain capacity, they failed to identify all affected structures, systems, and components (SSCs). The DGB contains predominately Train B emergency power systems; however, the fuel oil transfer pumps for the Train A emergency diesel generators are located in the southeast corner of the building. The licensee failed to assess the effects of flooding on the Train A fuel oil transfer pumps. The licensees corrective actions included the creation of an adverse condition monitoring plan, which implemented an hourly flood watch in the DGB when the fire pump was manually started. The inspectors determined that the finding was more than minor, because if left uncorrected, it would potentially result in a more safety significant issue. Specifically, the failure to evaluate the effects of flooding on all SSCs resulted in inadequate compensatory measures. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. For the time period in question, May 17, 2015 to September 17, 2015, the inspectors reviewed the security door card reader reports and starting sump levels for the DGB and found that during times when the fire pumps were running, station personnel had toured the DGB at a frequency that would have identified flooding conditions before a loss of system function. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Evaluation (P.2), in the area of Problem Identification and Resolution (PI&R), for failing to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000266/FIN-2015003-022015Q3Point BeachPotential Failure of Multiple Safety-Related Trains During Flooding EventsThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensees failure to ensure that a non-Category I (seismic) component failure, that results in flooding, would not adversely affect safety-related equipment needed to get the plant to safe shutdown (SSD) or to limit the consequences of an accident. Specifically, the design of Point Beach did not ensure that the Residual Heat Removal (RHR) pumps would be protected from all credible non-Category I (seismic) system failures. The licensees corrective actions included an extensive internal flooding design review, which will result in an updated Final Safety Analysis Report (FSAR) with a more detailed description of the stations flooding licensing basis; modifications to multiple flood barriers to bring them into compliance with the licensees flooding licensing basis; installation of additional flood level alarms where necessary, and evaluation or modification of service water (SW) piping to properly qualify it as seismic. The inspectors determined that the finding was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design resulted in an unanalyzed condition and loss of safety function of the RHR system while the plants were in Modes 4, 5, and 6, when relying on the RHR system for decay heat removal. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors answered yes to question 2 of the screening questions because the finding represented a loss of safety function. Thus the inspectors consulted the Region III Senior Risk Analysts (SRAs) who performed a detailed risk evaluation and determined that the finding was of very low safety significance (Green). The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015003-032015Q3Point BeachFailure to Perform a Written Safety Evaluation for FSAR ChangesThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance for the licensees failure to perform a safety evaluation to demonstrate that the removal of statements from the FSAR did not require a license amendment. Specifically, the licensee failed to perform a safety evaluation to determine whether removing an FSAR statement, which defined the RHR pump cubicle design flood height as seven feet, could be performed without a license amendment. The licensee entered the deficiency in their CAP as Action Request (AR) 02069425 by which the licensee intends on re-evaluating the 1996 FSAR change. The inspectors determined that the finding was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, inappropriately removing the information from the FSAR allowed the licensee to decrease the design basis flood protection height of the RHR compartments and significantly reduced the available time to isolate the leaking RHR pump seal. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015002-042015Q2Point BeachLicensee-Identified ViolationSection (b) of TS 5.7.1 requires, in part, that access toand activities ina high-radiation area be controlled by a radiation work permit or equivalent. Section (e) of TS 5.7.1 requires, in part, that entry into HRAs be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Contrary to the above, on April 14, 2015, an individual entered an HRA without being on a radiation work permit that allowed for HRA entry and was not made knowledgeable of the dose rates in the HRA. The licensee entered this issue into the CAP as AR 0204280. This violation is considered to be of very low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls; (2) there were no overexposures; (3) there was not a substantial potential for overexposures; and (4) the ability to assess dose was not compromised.
05000266/FIN-2015002-032015Q2Point BeachInadequate Measures to Control Spare Firing Card AssembliesA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, was self-revealed for the licensees failure to establish measures to ensure non-conforming tantalum electrolytic capacitors that were part of an assembly and that were beyond their recommended shelf-life would not be installed in safety-related equipment in the plant. The licensees corrective actions included repair of the D-107 battery charger, and updating maintenance and procurement requirements with component shelf-life information. The inspectors determined the finding was more than minor since the failure to ensure the quality of spare parts, if left uncorrected, could lead to a more significant safety concern. Specifically, the failure to control circuit boards which contained tantalum electrolytic capacitors that were beyond their shelf-life was self-revealed when the D-107 safety-related battery charger failed three days after the circuit boards were installed. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered "No" to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Change Management (H.3), in the area of Human Performance, for the licensees failure to use a systematic process for implementing changes so that nuclear safety remained the overriding priority.
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