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05000255/FIN-2015001-012015Q1PalisadesInadequate Procedure Results in Failure of Component Cooling Water PumpA finding of very low safety significance and an associated NCV of TS 5.4.1(a) was self-revealed on January 6, 2015, after the licensee identified smoke coming from the C component cooling water (CCW) pump (P52C) as a result of incorrect assembly of the inboard pump bearing in December 2014, due to an inadequate maintenance procedure. This issue was entered into the licensees CAP as CRPLP201500063, Workers Reported Smoke Coming from Shaft of P52C, dated January 6, 2015. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality 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. Based on a detailed risk evaluation performed by a Region III SRA using SAPHIRE Version 8.20 and the Events and Conditions Assessment Feature of the SPAR model (Version 8.1.2), the inspectors determined the finding was of very low safety significance. This finding had a cross-cutting aspect in the Avoid Complacency component of the Human Performance cross-cutting area. Specifically, plant staff accepted the practice of bending the C CCW pump oiler nipple to achieve proper level when the oiler could not be properly aligned which compensated for, rather than corrected, an underlying issue of improper alignment when tightening the alignment pin.
05000255/FIN-2015001-052015Q1PalisadesFailure to Evaluate the Adverse Effects of the Use of Non-Seismic Temporary JumpersA Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance was identified by the inspectors when licensee personnel failed to maintain a written safety evaluation that provided a basis that the use of temporary alligator clip jumpers to maintain emergency diesel generator (EDG) operability during certain maintenance activities did not require a license amendment. Specifically, the licensee did not address the adverse effects of the use of alligator jumpers on the design and qualification of the diesel generator (DG) circuit breaker used per Engineering Change 50310 and changes to procedure SPSE1, 2400 Volt and 4160 Volt Allis Chalmers and Siemens Vacuum Circuit Breaker Auxiliary Switch Adjustments, Revision 34. This issue was entered into the licensees CAP as CRPLP201404859, NRC Identified 50.59 Issue, dated October 7, 2014. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the change that was implemented adversely affected the seismic qualification of the electrical circuit that was relied upon to ensure safety bus 1C would be loaded by the 11 DG upon a loss of offsite power. The inspectors evaluated the underlying technical issue and determined the finding was of very low safety significance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the finding associated with this violation was determined to be of very low safety significance. This finding had a cross-cutting aspect in the Conservative Bias component of the Human Performance cross-cutting area. Specifically, the licensee did not use all available information and relevant guidance, such as Nuclear Energy Institute 9607, to demonstrate that the proposed activity was safe and did not require a license amendment prior to implementation.
05000255/FIN-2015001-032015Q1PalisadesTurbine-Driven AFW Pump Trip During Surveillance TestingOn November 14, 2014, during performance of surveillance procedure RO-127, the turbine-driven AFW pump overspeed trip mechanism actuated and tripped the pump, which resulted in entry into 72-hour TS LCO 3.7.5 Condition A. Personnel performing the test reported no abnormalities prior to the trip. The pump had been operated satisfactorily on November 13, 2014, during performance of quarterly surveillance procedure QO-21, Inservice Test ProcedureAuxiliary Feedwater Pumps, Revision 43. Following the November 14 pump trip, the licensee performed a sequence of testing in an attempt to replicate the overspeed trip to assist in cause identification, or to demonstrate operability if the trip could not be replicated. This sequence included test runs in accordance with procedure SOP12, Feedwater System, Revision 72; procedure RO145, Comprehensive Pump Test Procedure, Auxiliary Feedwater Pumps P8A, P8B and P8C, Revision 13; and a modified test run in accordance with procedure T186, Auxiliary Feedwater Turbine K8 Overspeed Trip Test and Governor Setting, Revision 18. Since all acceptance criteria were met, the licensee declared the turbine-driven AFW pump operable on November 16, 2014, and LCO 3.7.5 was exited. The licensee subsequently conducted an ACE that identified three possible causes: 1) excess condensate in the moisture removal system, 2) decreased margin between the as-found operating speed and the overspeed trip setpoint, and 3) inherent design conditions of the steam supply and steam control system affecting the speed of the turbine. The evaluation also identified discrepancies in the Maintenance Rule classification of the steam traps and the frequency of preventive maintenance activities that were conducted on the equipment. Corrective actions were generated to evaluate each of the possible causes during the next pump maintenance outage, which occurred from March 26, 2015. The licensee conducted procedure T186 to verify the overspeed trip mechanism would trip within its required band and that the governor was functioning properly to changes in steam supply and pump speed. Procedure RO145 was conducted to observe steam trap operation and any oscillations in the steam supply system. Finally, the overspeed trip setpoint was revised based on the vendors recommendation for the operating conditions of this particular pump. At the end of the inspection period, all of the data from these tests were under review by the licensee, with vendor support, to aid in determining the cause of the November 2014 pump trip. Pending inspector review of the licensees data, conclusions, and any revisions made to the ACE, this issue is unresolved.
05000255/FIN-2015001-072015Q1PalisadesInadequate Procedure Leads to Primary Coolant Pump Seal DegradationA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1(a) was self-revealed when the C primary coolant pump (PCP) seal degraded as a result of an inadequate maintenance procedure. Specifically, maintenance procedure PCSM54, N9000 Primary Coolant Pump Shaft Seal Assembly, did not identify critical steps in the assembly of the PCP seal and, as a result, the work activity was not adequately controlled. This issue was entered into the licensees Corrective Action Program (CAP) as CRPLP201403495, Planned Outage Required Due to Two Stages of the Primary Coolant Pump P-50C Seal Not Performing as Expected, dated June 21, 2014. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone 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. Specifically, the C PCP seal was not correctly assembled or installed during refueling outage (RFO) 1R23, which resulted in premature seal degradation. Based on a detailed risk evaluation performed by a Region III Senior Reactor Analyst (SRA) using SAPHIRE Version 8.20 and the Events and Conditions Assessment Feature of the Palisades Standardized Plant Analysis Risk (SPAR) model (Version 8.1.2), the inspectors determined the finding was of very low safety significance. This finding had a cross-cutting aspect in the Work Management component of the Human Performance cross-cutting area. Specifically, the licensee did not effectively screen the PCP seal assembly through the work management process to identify that it should have been classified as a critical maintenance activity. In addition, insufficient emphasis was placed on in-field vendor oversight during work execution.
05000255/FIN-2015001-062015Q1PalisadesFailure to Evaluate the Adverse Effects of the Use of Non-Seismic Temporary JumpersA Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance was identified by the inspectors when licensee personnel failed to maintain a written safety evaluation that provided a basis that the use of temporary alligator clip jumpers to maintain emergency diesel generator (EDG) operability during certain maintenance activities did not require a license amendment. Specifically, the licensee did not address the adverse effects of the use of alligator jumpers on the design and qualification of the diesel generator (DG) circuit breaker used per Engineering Change 50310 and changes to procedure SPSE1, 2400 Volt and 4160 Volt Allis Chalmers and Siemens Vacuum Circuit Breaker Auxiliary Switch Adjustments, Revision 34. This issue was entered into the licensees CAP as CRPLP201404859, NRC Identified 50.59 Issue, dated October 7, 2014. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the change that was implemented adversely affected the seismic qualification of the electrical circuit that was relied upon to ensure safety bus 1C would be loaded by the 11 DG upon a loss of offsite power. The inspectors evaluated the underlying technical issue and determined the finding was of very low safety significance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the finding associated with this violation was determined to be of very low safety significance. This finding had a cross-cutting aspect in the Conservative Bias component of the Human Performance cross-cutting area. Specifically, the licensee did not use all available information and relevant guidance, such as Nuclear Energy Institute 9607, to demonstrate that the proposed activity was safe and did not require a license amendment prior to implementation.
05000255/FIN-2015001-042015Q1PalisadesFailure to Verify the Adequacy of Credited High Energy Line Break BarriersA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified by the inspectors when the licensee credited fire doors for High Energy Line Break (HELB) protection without a supporting test or evaluation. Specifically, Procedure 4.02 credited fire doors with protection of safety-related equipment against a HELB when the primary HELB barrier was disabled without a test or evaluation to demonstrate the doors could withstand the HELB environment. This issue was entered into the licensees CAP as CRPLP201500371, NRC Concerns with Calculation EAPSACCWHELB0217, dated January 22, 2015. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not have an evaluation to demonstrate that barriers relied upon to protect mitigating systems from a HELB initiating event could perform the credited protection function. The inspectors answered No to the questions in Exhibit 2.A, Mitigating Systems Screening Questions, and as a result determined the issue was of very low safety significance. This finding was not associated with a cross-cutting aspect since the calculation in question was created in 2003 and therefore did not represent current performance.
05000255/FIN-2015001-022015Q1PalisadesInoperability of Safety Injection Tank Due to Long-Term LeakageA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors when licensee personnel failed to assure that leakage out of the B safety injection tank (SIT), a condition adverse to quality, was corrected in a timely manner. Specifically, although minor water leakage out of the B SIT had been occurring since at least 2010, the licensee had not corrected the leakage despite several plant outages that provided an opportunity to address the issue. This issue was entered into the licensee s CAP as CRPLP201404861, B SIT Declared Inoperable Due to Reaching Technical Specification Low Level Setpoint, dated October 7, 2014. The 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 affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the leakage out of the B SIT resulted in unexpected inoperability of the tank on October 7, 2014. The finding was determined to be of very low safety significance based on answering No to the screening questions in Exhibit 2.A, Mitigating Systems Screening Questions. This finding had a cross-cutting aspect in the Avoid Complacency component of the Human Performance cross-cutting area. Specifically, over time the licensee became confident that the long-term leakage out of the B SIT was minor and could be managed without an impact to equipment operability, which proved to be incorrect when the minor leakage resulted in B SIT inoperability on October 7, 2014.
05000255/FIN-2014005-012014Q4PalisadesFailure to Follow Procedure for Storage of Equipment in the Vicinity of Safety-Related EquipmentThe inspectors identified a finding of very low safety significance (Green) with an associated non-citied violation of Technical Specification (TS) 5.4.1, Procedures and Programs, for the failure to follow site procedures covering the storage of material in the vicinity of safety-related equipment. Specifically, on three occasions the inspectors identified ladders at ladder station 42 in the 590 elevation of the component cooling water room that were either in contact with safety-related equipment or were capable of toppling into safety-related equipment. For immediate corrective actions, licensee personnel properly stored the ladder after each issue was identified by the inspectors. This issue is documented in the licensees corrective action program (CAP) as Condition Report CR-PLP-2015-00126. The performance deficiency was determined to be more than minor based on Inspection Manual Chapter (IMC) 0612, Appendix E, Example 4.a, which determined that low-level procedural errors without a safety consequence are more than minor when they become a repetitive/routine occurrence. Specifically, unrestrained ladders could impact safetyrelated equipment during a design basis seismic event. The inspectors evaluated the significance of the finding in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings. In accordance with Table 2, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors answered No to the questions in Table 3 and continued the significance evaluation in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors answered No to the Mitigating Systems Screening Questions contained in Exhibit 2 and determined the finding was of very low safety significance (Green). This finding was associated with a cross-cutting aspect of Identification in the Problem Identification and Resolution cross-cutting area (P1).
05000255/FIN-2014004-012014Q3PalisadesInadequate Procedure for Protection against High WindsThe inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification (TS) 5.4.1 when licensee personnel failed to maintain and implement an adequate procedure covering Acts of Nature. Specifically, the licensees interpretation of Abnormal Operating Procedure (AOP)38 entry conditions resulted in a decision not to enter the procedure despite available information indicating the presence of high wind conditions in the vicinity of the plant. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PLP-2014-04155, NRC Questioned Entry into AOP-38, dated August 20, 2014. Planned corrective actions include a procedure revision to clarify the procedure entry conditions. The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone 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. Specifically, the preparatory actions prescribed by AOP-38 were directly related to the Initiating Events Cornerstone objective and inconsistent application of those actions in advance of high wind conditions increased the likelihood of debris-induced initiating events. In accordance with IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, Section B, Transient Initiators, because the finding did not result in a reactor trip or the loss of mitigating equipment, it was determined to be of very low safety significance. This finding was associated with a cross-cutting aspect of Training in the Human Performance cross-cutting area. Specifically, the licensees interpretation of procedure AOP-38 entry conditions was a result of the training provided to operators.
05000255/FIN-2014004-022014Q3PalisadesSpent Fuel Pool Region II Criticality AnalysisThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when licensee personnel failed to follow procedure EN-OP-104, Operability Determination Process. Specifically, Operability Evaluation CR-PLP-2013-04775 failed to include adequate technical information to support the basis for the reasonable expectation of operability, as required by Step 5.5.c of EN-OP-104. On March 25, 2014, the licensee entered the NRC questions into the CAP as Assignments 6 and 7 of CR-PLP-2013-04775, Issues Identified with Region II of SFP Criticality Analysis, with an initial due date of April 8, 2014. Both Assignments 6 and 7 were ultimately closed in late April to a new Assignment 9, which was created to complete a revised Operability Evaluation. The licensee determined that contracted technical support was necessary to adequately evaluate the NRC concerns. At the end of the inspection period, the contracted evaluation effort was ongoing. Planned corrective actions included documenting the conclusions of the ongoing evaluation in a revised Operability Evaluation for CR-PLP-2013-04775. The inspectors determined that the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity Cornerstone and adversely impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the Spent Fuel Pool (SFP) criticality analysis relied on certain physical conditions to maintain the effective neutron multiplication factor below 1.0, but actual physical conditions were not completely bounded by the existing criticality analysis. Because the inspectors answered No to all of the SFP questions in IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the finding was determined to be of very low safety significance. This finding was associated with a cross-cutting aspect of Operating Experience in the Problem Identification and Resolution cross-cutting area. Specifically, the licensee failed to collect and implement relevant external operating experience.
05000255/FIN-2014004-042014Q3PalisadesLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Title 10 CFR 50.55a(f)(4) requires, in part, that pumps and valves classified as ASME Code Class 1, 2, or 3 must meet the IST requirements set forth in the ASME OM Code and addenda, to the extent practical within the limitations of design, geometry, and materials of construction of the components. Contrary to that requirement, the licensee failed to test left train AFW flow control valves CV-0727 and CV-0749 in accordance with the ASME OM code requirements or the NRC-approved Valve Relief Request (VRR-18), for a period of approximately 6 years. In June 2013, the licensee identified that the alternate method for testing the AFW flow control valves, approved in 2007 by VRR-18 for the current 10-year IST interval, was not being performed. The approved alternate method for testing the regulating capability of these valves was to validate that the A/B AFW pump flow rates were within TS limits during the quarterly pump surveillance test. However, the quarterly surveillance test methodology was revised after the VRR was submitted to no longer provide AFW flow to the steam generators during testing, instead flowing water in a recirculation loop, and therefore not testing the regulating capability of the flow control valves. The licensee identified the issue during a focused self-assessment of the IST program and entered the issue into their CAP as CR-PLP-2013-2522, Alternate Testing Not Being Performed as Approved by Valve Relief Request for CV-0727 and CV-0749, on June 6, 2013. The valves were stroke time tested, in accordance with the ASME OM Code requirements, in November 2013, and every quarter thereafter, using the quarterly technical specification surveillance valve test procedure. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely impacted 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 did not perform Code-required timed valve strokes for several years, which adversely affected the ability to verify that the valves would operate as required. The inspectors evaluated the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, and 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? Therefore, the issue was determined to be of very low safety significance.
05000255/FIN-2014004-032014Q3PalisadesFailure to Evaluate the Adverse Effects of the Use of Non-Seismic Temporary JumpersOn September 10, 2014, the inspectors observed a preventive maintenance activity on 2.4kV breaker 152-106, which supplied power from Startup Transformer 1-2 to the 1C safety bus. During the activity, the inspectors noted that the licensee planned to install a temporary jumper to maintain operability of the 1-1 EDG. There was a b-contact on the breaker 152-106 auxiliary switch that was part of the 1-1 EDG auto-start circuit, which allowed the 1-1 EDG to automatically close on the 1C safety bus if both offsite power feeder breakers were open. During the preventive maintenance activity, the auxiliary switch was manipulated and re-positioned several times, which prevented the 1-1 EDG from automatically closing onto the 1C safety bus, rendering the 1-1 EDG inoperable. The licensee previously evaluated and approved a temporary modification, documented in EC 50310, for the use of temporary jumpers to maintain EDG auto-start circuit continuity, and therefore EDG operability, during preventive maintenance activities that manipulated the auxiliary switch for breaker 152-106 as well as six other breakers associated with the 1C and 1D safety buses. The evaluation acknowledged that the jumpers were being installed in seismically-qualified equipment and the jumpers should be installed using safety-related wire and ring tongue terminals. However, the evaluation also stated that, due to the design of each breaker, breakers 152-105 and 152-106 required the use of alligator clip jumpers, which was allowed by procedure EN-DC-136, Temporary Modifications, Revision 10, provided the alligator clip jumpers were not left unattended. The inspectors reviewed the process applicability determination (PAD) for the temporary modification documented in EC 50310. The licensees PAD concluded that, while some aspects of the activity were covered under maintenance risk regulations in 10 CFR 50.65, the use of temporary jumpers to maintain EDG operability was covered under 10 CFR 50.59. The 10 CFR 50.59 screening was documented in the PAD and the licensee determined that there were no adverse effects from the change. Section VI.B of the PAD referred to Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Section 4.2, for guidance in screening issues through the 10 CFR 50.59 criteria. The inspectors reviewed NEI 96-07, Section 4.2, and identified items that appeared to not be addressed by the licensee in the PAD. Section 4.2.1 contained a list of questions that illustrate the range of effects that may stem from a proposed activity. Two of the questions were as follows: Does the activity decrease the reliability of an SSC design function, including either functions whose failure would initiate a transient/accident or functions that are relied upon for mitigation? Does the activity degrade the seismic or environmental qualification of the SSC? The PAD did not address either of these questions despite the knowledge that alligator clip jumpers were not seismically qualified, required constant attention per procedure for that reason, and were being installed in seismically qualified equipment. The inspectors discussed the issue with 10 CFR 50.59 subject matter experts in the Region III Office, and collectively discussed the issue with the licensee. At the end of this inspection period, the licensee entered the concerns into their CAP as CR-PLP-2014-04859, NRC Identified 50.59 Issue, dated October 7, 2014. The inspectors were awaiting the licensees corrective action plan and evaluation of the temporary modification through the 10 CFR 50.59 criteria to determine whether a license amendment would have been necessary. This issue is an Unresolved Item pending review of the additional information.
05000255/FIN-2014003-012014Q2PalisadesWritten NRC Biennial Written Examinations Did Not Meet Qualitative StandardsThe inspectors identified a finding of very low safety significance associated with 10 CFR 55.59, Requalification, based on a determination that greater than 20 percent of the biennial requalification written exam questions administered to licensed operators during weeks three and five of the 2012 examination cycle were flawed. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PNP-2014-02521, Written Exam Quality, dated April 10, 2014. The inspectors determined that the finding was more than minor because it was associated with the Human 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 finding adversely affected the quality and level of difficulty of biennial written exams, which potentially impacted Palisades ability to appropriately evaluate licensed operators. The risk importance of this issue was evaluated using IMC 0609, Appendix l, Licensed Operator Requalification Significance Determination Process (SDP). The inspectors considered the number of written exam questions that did not meet the qualitative standard for written exam questions. The qualitative standards used by the inspectors are defined in NUREG-1021, Revision 9, ES-602, Attachment 1, Guidelines for Developing Open-Reference Examinations, and Appendix B, Written Examination Guidelines. Because more than 30 percent of the questions reviewed did not satisfy the guidance, Block 4 of Appendix I applied. Based on the screening criteria, the finding was characterized by the SDP as having very low safety significance (Green) because greater than 20 percent, but less than 40 percent, of the reviewed written exam questions were flawed. A review of the cross-cutting aspects was performed and no associated cross-cutting aspect was identified.
05000255/FIN-2014003-022014Q2PalisadesExam Security IssuesThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 55.49, Integrity of Examinations and Tests, which stated, Applicants, licensees, and facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Specifically, Palisades placed personnel in the simulator operating booth that were not identified in the security agreement, placed the scenario turnover sheet for a second scenario in the simulator during the first scenario, and left a job performance measure turnover sheet in the simulator after the applicant left the simulator and brought the next applicant into the simulator. This issue was entered into the licensees CAP as CR-PLP-2014-02533, Issues Were Identified During the Annual Exam Administered on April 10, 2014, dated April 10, 2014. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to become a more significant safety concern. Specifically, the failure to properly control operational examination material in a manner in which applicants were not prematurely exposed to the material provided opportunities to compromise the examination. The finding was screened as one of very low safety significance (Green) in accordance with IMC 0609, Appendix I, Licensed Operator Requalification SDP. This finding was associated with the cross-cutting aspect of Procedure Adherence in the Human Performance area.
05000255/FIN-2014003-042014Q2PalisadesFailure to Evaluate Long-Term Scaffolds in Accordance with ProceduresThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when licensee personnel failed to adequately implement procedure EN-MA-133, Control of Scaffolding. Specifically, multiple examples were identified of scaffolds installed in the plant for greater than 90 days that had not undergone process applicability determinations, were not appropriately documented in the scaffold control log, and/or did not contain proper tags. The licensee documented the issue in their CAP as CR-PLP-2014-2646, Two Scaffolds Near Safety-Related Equipment Not Being Controlled as Long-Term, dated April 17, 2014; conducted an extent-of-condition review of the entire scaffold log and identified additional discrepancies; completed the required process applicability determinations; and re-inspected scaffolds that had been categorized as long-term. The inspectors determined that the performance deficiency was more than minor because it was similar to Example 4.a) of IMC 0612, Appendix E, Examples of Minor Issues. This example described an engineering evaluation that was not performed for scaffolding erected near safety-related equipment and stated that it would be a more than minor issue if the licensee routinely failed to perform the engineering evaluations. For this specific finding, there were multiple examples of process applicability determinations not being performed within the procedurally required timeframe. The finding was determined to be of very low safety significance (Green) because it did not affect the operability/functionality of structures, systems and components (SSCs) and all required safety functions were maintained. This finding was associated with the cross-cutting aspect of Teamwork in the Human Performance area. Specifically, licensee and supplemental individuals and work groups did not sufficiently communicate and coordinate work activities associated with maintaining the scaffold control log or documentation related to scaffolding installed in the plant. The workers also did not understand how to account for time during refueling and forced outages when determining the long-term status of scaffolds, which could have been resolved with input from other work groups.
05000255/FIN-2014003-082014Q2PalisadesLicensee-Identified ViolationTechnical Specification 5.7.2, High Radiation Areas with Dose Rates Greater than 1.0 Rem/Hour at 30 Centimeters from the Radiation Source or from 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, requires, in part, that each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment. Contrary to the above, on March 12, 2014, radwaste operators found that the south east steam generator bio-wall cage door, a locked high radiation area, was open and not locked. The licensee documented this issue as CR-PLP-2014-02083, Radwaste Operators Found a Locked High Radiation Area Gate Left Open, dated March 13, 2014. The finding was determined to be of very low safety significance (Green) because it was not an ALARA planning issue; there was no overexposure, nor substantial potential for an overexposure; and the licensees ability to assess dose was not compromised.
05000255/FIN-2014003-032014Q2PalisadesFailure to Notify the NRC Within 30 Days of Discovering Changes in Medical ConditionsA Severity Level IV non-cited violation of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status, was identified by the inspectors during a review of licensed operator medical records. Specifically, Palisades did not notify the NRC within 30 days of discovering a change in medical condition for a licensed operator. Subsequently, the licensee submitted the required notification for the operator on April 11, 2014, and entered the issue into their CAP as CR-PLP-2014-02518, NRC Informed the Palisades Training Department that an NRC Form 396 was Not Submitted, dated April 10, 2014. The inspectors determined that Traditional Enforcement applied because a failure to make a required report impacted the regulatory process. Specifically, the licensee had not notified the NRC within 30 days of learning of a change in medical condition for a licensed operator for which a license condition was required. Based on Example 6.9.d.1 of the NRCs Enforcement Policy, the inspectors determined that the issue represented a Severity Level IV violation. No associated Reactor Oversight Process finding was identified, thus there was no associated cross-cutting aspect.
05000255/FIN-2014002-012014Q1PalisadesInadequate Installation of Steam Generator Nozzle DamsA finding of very low safety significance and an associated non-citied violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was selfrevealed when licensee personnel failed to have an adequate procedure and work order (WO) to install steam generator nozzle dams. The licensee entered this issue in their Corrective Action Program (CAP) as Condition Report (CR) PLP-2014-00770, Improper Routing of Nozzle Dam Air Supply. As part of their corrective actions, the licensee planned to revise the nozzle dam installation procedure and the WO. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding was associated with the Procedure Quality attribute of the Initiating Events cornerstone and adversely impacted 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, and was similar to the more than minor criteria in Example 5.a of IMC 0612, Appendix E, Examples of Minor Issues. As it related to this finding, the intended design of the nozzle dam air supply system was not correctly translated into the installation procedure or the work instructions. Further, the nozzle dam air system was not properly tested prior to being placed into service. Since the plant was shutdown in Mode 6, the inspectors assessed the risk significance of the event in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. A Phase 2 risk evaluation was required that determined the total event risk was 3.6E-8 and was therefore of very low safety significance (Green). This finding had an associated crosscutting aspect in the Change Management (H.3) component of the Human Performance cross-cutting area. In particular, issues during the previous refueling outage led the steam generator project management team to review the configuration of the nozzle dam air system. Through this review, the licensee identified that changes to the alignment of air to the nozzle dams was required. However, due to turnover within the project management group and inadequate communications and documentation, the licensee failed to appropriately evaluate and implement those changes.
05000255/FIN-2014002-022014Q1PalisadesFailure to Complete Volumetric Examinations for DM Butt Welds in Branch ConnectionsThe inspectors identified a finding of very low safety significance and an associated non-citied violation of 10 CFR 50.55a(g)(6)(ii)(F)(3) when licensee personnel failed to complete required baseline volumetric examinations for nine dissimilar metal (DM) butt welds in the Primary Coolant System (PCS) that were fabricated from Inconel Alloy 82/182 weld metal and were susceptible to primary water stress corrosion cracking (PWSCC). The licensee entered this issue into their CAP as CR-PLP-2014-01742, NRC Question on Whether Hot and Cold Leg Branch Connection Welds are In Scope of ASME (American Society of Mechanical Engineers) Code Case (CC) N-770-1. As part of their corrective actions, the licensee submitted a request for relief to the NRC to allow substitution of a visual and dye penetrant surface examination of these welds as an alternative to volumetric examinations. The NRC granted verbal relief on March 13, 2014, which stated the licensee could implement the proposed alternative to 10 CFR 50.55a(g)(6)(ii)(F), which included a commitment to perform enhanced leakage monitoring during the current operating cycle and perform the required volumetric examinations during the next refueling outage. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding was associated with the Equipment Performance (Reliability) attribute of the Initiating Events cornerstone and adversely impacted 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 inspectors also determined that if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure to complete volumetric examinations on the nine DM butt welded PCS branch connections fabricated with Alloy 82/182 weld metal could have allowed PWSCC susceptible material to remain in service, which could propagate and result in a Loss-of-Coolant-Accident (LOCA). The inspectors performed a Phase I Significance Determination Process screening using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions. The inspectors answered the Phase I SDP LOCA Initiators Questions A1 and A2 No because undetected cracks, if present, were not yet through-wall and did not challenge the structural integrity of the welds. Therefore, this finding screened as having very low safety significance (Green). This finding had an associated cross-cutting aspect in the Evaluation (P.2) component of the Problem Identification and Resolution cross-cutting area because the licensee did not ensure that the resolution of the issue appropriately addressed causes and the extent of condition. Specifically, when determining the applicability of CC N-770-1, the licensee failed to thoroughly evaluate the scope of welds susceptible to PWSCC that required volumetric examination commensurate with the safety significance of this issue.
05000255/FIN-2014002-042014Q1PalisadesIntroduction of Foreign Material Into the SW SystemA finding of very low safety significance and an associated non-citied violation of Technical Specification (TS) 5.4.1, Procedures, was identified by the inspectors when licensee personnel failed to follow procedure EN-MA-118, Foreign Material Exclusion (FME), during work on the safety-related critical service water (SW) system during refueling outage (RFO) 1R23. Specifically, Sections 5.2(1) and 5.2(6) of EN-MA-118 stated that planners and procedure writers should evaluate FME considerations for work activities and include job-specific FME controls in work instructions and procedures. Additionally, EN-MA-188 stated that during the planning stage, the planner should designate the FME Zone type, risk level, pathways to FME sensitive equipment, and work practice restrictions, as applicable, in all work packages. However, adequate controls were not established and documented when the decision was made to use an inflatable bladder inside the SW system when work was being performed on the system. As a result, on two separate occasions during RFO 1R23, bladders were inadvertently entrained into the return header of the SW system by the relative vacuum created by system flow. The licensee entered this issue into their CAP as CR-PLP-2014-00715, Vacuum was So Great that Bladder was Ripped Off Lanyard and Lost in Piping, and CR-PLP-2014-01176, FME Bladder Lost During Work Near CV-0823. As part of their corrective actions, the licensee successfully completed a comprehensive SW system test, which validated acceptable system parameters. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. In accordance with Checklist 3, PWR (Pressurized Water Reactor) Cold Shutdown and Refueling Operation RCS (Reactor Coolant System) Open and Refueling Cavity Level < 23' Or RCS Closed and No Inventory in Pressurizer Time to Boiling < 2 hours, following the loss of the first bladder, and Checklist 4, PWR Refueling Operation: RCS Level > 23' Or PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, following the loss of the second bladder of Attachment 1, Phase 1 Operational Checklists for both PWRs and BWRs (Boiling Water Reactors), of IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors determined that mitigation capabilities were not adversely impacted. Additionally, utilizing Table 1, Losses of Control, of IMC 0609, Appendix G, the inspectors determined there was no loss of control. As a result, the finding screened as having very low safety significance (Green). This finding had an associated cross-cutting aspect in the Work Management (H.5) component of the Human Performance cross-cutting area because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. In particular, the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities.
05000255/FIN-2014002-052014Q1PalisadesFailure to Follow Procedures During Reactor Vessel Head LiftA finding of very low safety significance and an associated non-citied violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self revealed when licensee personnel failed to follow maintenance procedure RFL-R-16, Reactor Vessel Closure Head Installation. Specifically, during the reactor vessel head lift on March 5, 2014, to support reinstallation onto the vessel flange, workers failed to identify an interference with the reactor head lift structure, causing the head to impact a jack screw on the structure and increasing the total load weight to approximately 283,000 pounds, which was greater than the procedural maximum polar crane load rating of 270,000 pounds. The licensee entered this issue into their CAP as CR-PLP-2014-01903, Reactor Head Flange Contacted Jacking Screw While Raising it Off the Head Stand. As part of their corrective actions, the licensee conducted a Level 1 Human Performance Evaluation, generated a site-wide Human Performance error communication, and performed work crew stand downs to discuss crane and rigging expectations. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding was associated with the Human Performance attribute of the Barrier Integrity cornerstone and adversely impacted the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Additionally, the inspectors determined that the performance deficiency could reasonably be viewed as a precursor to a significant event and that if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the operability of the containment polar crane was required to be evaluated and the reactor vessel head was required to be inspected after the event occurred to verify no significant damage was caused and the maximum design limit of the crane could have been exceeded if the evolution was not stopped when it was, which increased the risk of dropping the head during the lift. The finding was screened in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Phase 1 Operational Checklists for both PWRs and BWRs. The finding was determined to be of very low safety significance (Green) based on not requiring a quantitative assessment after reviewing the five shutdown safety functional areas in Checklist 3, PWR Cold Shutdown and Refueling Operation RCS Open and Refueling Cavity Level < 23' Or RCS Closed and No Inventory in Pressurizer Time to Boiling <2 hours. This finding had an associated cross-cutting aspect in the Challenge the Unknown (H.11) component of the Human Performance cross-cutting area. Specifically, human performance investigations identified that workers exhibited a lack of rigor when performing interference verifications prior to and during the reactor head lift, and an inadequate stop when unsure mentality when assessing the situation before continuing with the head lift. In addition, the workers and supervisors for this task did not understand that the load cell increase exceeded the procedural maximum value and did not inform decision-makers outside of the immediate work area to validate it was safe to proceed with the evolution.
05000255/FIN-2014002-062014Q1PalisadesFailure to Maintain Radiation Exposure ALARA on CRDM 24 RepairsA finding of very low safety significance was self-revealed when workers received unplanned and unintended occupational radiation dose during a maintenance outage conducted in August 2012 due to deficiencies in the licensees Radiological Work Planning and Work Execution Program. Specifically, the licensee failed to properly incorporate As-Low-As-Reasonably-Achievable (ALARA) strategies and insights while planning and executing Control Rod Drive Mechanism (CRDM) 24 housing work. The licensee entered this issue into their CAP as CR-PLP-2014-05812, UT (Ultrasonic Testing) Exams of the Additional CRDM Stalk Housings Has Exceeded the Dose Estimate for the RWP (Radiation Work Permit). Corrective actions were implemented to address the outage planning and work execution issues. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely impacted the cornerstone objective of ensuring the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the finding was similar to the more than minor criteria in Example 6.i of IMC 0612, Appendix E, Examples of Minor Issues. The inspectors screened this finding in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that the finding did not involve: (1) a radiological overexposure; (2) a substantial potential for an overexposure; or (3) a compromised ability to assess dose. The inspectors also determined that the finding involved ALARA planning and work controls and that the licensees 3-year rolling collective dose average was above 135 person-Rem at the time the performance deficiency occurred. However, because the work activity was a single occurrence that involved an actual dose outcome that was within the licensees control of less than 25 person-Rem, this finding was determined to be of very low safety significance (Green). This finding had an associated cross-cutting aspect in the Work Management (H.5) component of Human Performance cross-cutting area because the licensee did not plan work activities that appropriately incorporated radiological safety.
05000456/FIN-2014002-012014Q1BraidwoodInadequate AOP Entry Criteria for Intake Frazil Icing ConditionsThe inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, when licensee personnel failed to specify adequate entry conditions in Abnormal Operating Procedure (AOP) 0BwOA ENV-1, Adverse Weather Conditions, utilized to monitor and mitigate a frazil ice event at the lake screen house (LSH). Specifically, the licensee had established a frazil ice entry condition without adequately considering the plant data available to control room operators and without accounting for instrument accuracy and uncertainty. The licensee entered this issue into their Corrective Action Program (CAP) as Issue Report (IR) 1613056, NRC Identified Ice Forming at the LSH CW (Circulating Water) Trash Bars, and IR 1617385, NRC Questions Regarding Frazil Ice. Corrective actions included revising the frazil ice entry conditions based upon essential service water temperature with margin to account for instrument uncertainty and essential service water heat input. The inspectors determined that the performance deficiency was more than minor, because it was associated with the Procedure Quality 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 failure to establish and maintain adequate entry conditions into 0BwOA ENV-1 could result in additional time for ice to accumulate on plant components before mitigating actions would be initiated. Any delay in mitigating this type of event could increase the likelihood of a loss or partial loss of essential service water event or other transient. A detailed risk evaluation was performed by an NRC Regional Senior Risk Analyst (SRA) and the finding was determined to be of very low safety significance (Green). This finding did not have an associated cross-cutting aspect because the inspectors determined that the most significant error occurred when the entry criteria was established in November 2010, and therefore was not indicative of current performance.
05000456/FIN-2014002-022014Q1BraidwoodFailure to Ensure Mitigating System Availability and Reliability During Weather Conditions that Could Promote Frazil Ice at the LSHThe inspectors identified a finding of very low safety significance when licensee personnel failed to ensure that the LSH trash rake would be capable of clearing ice buildup on the trash rake bars. Specifically, the licensee failed to ensure that the trash rake system was functional prior to the onset of weather conditions that could promote frazil ice production and after a repair following a trash rake failure during those conditions. The licensee entered this issue into their CAP as IR 1613767, LSH Trash Rake Will Not Traverse on Rails. The licensee corrected this issue by utilizing a vendor to re-furbish and repair the trash rake. Additionally, the licensee revised their procedures to include additional methods to clear ice from the trash bars. The inspectors determined that the performance deficiency was more than minor because it 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 failure to have any mitigating systems available during weather conditions that could promote frazil icing of the lake intake increased the likelihood of a plant transient including a loss of essential service water event. A detailed risk evaluation was performed by an NRC Regional SRA and the finding was determined to be of very low safety significance (Green). This finding had a cross-cutting aspect in the Restoration component of the Problem Identification and Resolution cross-cutting area because the organization failed to take effective corrective action to address a non-functioning LSH trash rake in a timely manner commensurate with safety after restoring the equipment to Operations for use during weather conditions that could promote frazil icing conditions (P.3).
05000255/FIN-2014002-032014Q1PalisadesSpent Fuel Pool Region II Criticality AnalysisThe inspectors identified an Unresolved Item (URI) regarding assumptions used in the criticality analysis for Region II of the licensees spent fuel pool. Specifically, several assumptions in the applicable criticality analysis, which supported compliance with TSs and NRC regulations for criticality, did not appear to bound the characteristics of some fuel assemblies stored in Region II of the spent fuel pool. Description: On November 5, 2013, the licensee initiated CR-PLP-2013-04775, Issues Identified with Region II of Spent Fuel Pool Critically Analysis, which documented that the spent fuel pool criticality analysis was not updated following a power uprate that had been implemented in 2004. This was identified during the licensees review of industry operating experience documenting a similar issue at a different power plant. The licensee identified the following concerns with the criticality analysis for Region II of the spent fuel pool: 1) the assumed fuel temperature depletion parameter did not appear to bound the actual temperature for Batch A fuel, 2) the assumed cycle boron concentration did not appear to bound the actual cycle boron concentration after Cycle 20, and 3) the assumption that all Promethium-149 has decayed to Samarium-149 prior to placement of fuel into the spent fuel pool did not appear to be directly translated into site procedures. These concerns ultimately focused on whether fuel had achieved adequate burnup prior to placement in Region II of the spent fuel pool. The criticality analysis stated Batches A, B, and C fuel from Cycle 1 would not qualify for storage in Region II of the spent fuel pool due to extremely low burnup. However, Batch A fuel had been stored in Region II since a spent fuel pool re-rack project in 1987. Most of the Batch A fuel was relocated to dry storage in 1994 and 1995, but nine Batch A fuel assemblies currently remain stored in Region II. As a result of the assumptions that appeared to not bound actual conditions, Operations requested an Operability Evaluation to further evaluate the issue. Operability Evaluation CR-PLP-2013-04775 was assigned on November 5, 2013, and completed on December 5, 2013. The inspectors reviewed the Operability Evaluation along with staff from the Spent Fuel Team in the Office of Nuclear Reactor Regulation (NRR), Division of Safety Systems (DSS). On March 20, 2014, the NRC discussed the following questions regarding the Operability Evaluation with the licensee: The licensee compared the post-uprate hot leg temperature to the reactor core temperature in the analysis of record to justify that the analysis was bounding. However, the core temperature was hotter than the hot leg temperature, thus the Operability Evaluation did not appear to demonstrate that the existing core temperature was bounded by the core temperature in the analysis of record. Technical Specification Table 3.7.16-1 did not appear to ensure compliance with 10 CFR 50.68, which addressed spent fuel criticality, or TSs 4.3.1.3.a or 4.3.1.3.b, both of which addressed design assumptions in the Region II fuel storage racks. The methodology used in the development of the analysis of record contained non-conservatisms that appear to be mitigated by design margins that were already credited elsewhere in the Operability Evaluation. On March 25, the NRC questions were entered into the licensees CAP as assignments 6 and 7 of CR-PLP-2013-04775 with due dates of April 8. At the conclusion of the inspection period, the NRC staff was waiting to review the responses to the questions provided on March 20. This is an URI pending NRC review of the requested additional information.
05000456/FIN-2013005-052013Q4BraidwoodLicensee-Identified ViolationTechnical Specification 5.4.1 requires that written procedures shall be established, implemented, and maintained for procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Section 7.e.2 includes implementation of the Radiation Survey Program. Procedure RP-AA-503, Unconditional Release Survey Method, required, in part, that materials have no detectable radioactivity for unconditional release from the site. Contrary to the above, on September 17, 2013, Radioactive Shipping Specialist discovered that nine sample bottles containing radioactivity above minimal detectable activity for Co-58, Co-60 and Cs-137 were unconditionally released and shipped by the licensees warehouse staff to a licensed facility without proper authorization from RP Management. All sample bottles were accounted for and secured at the licensed facility on September 16, 2013. The licensee investigation determined that the shipment of the sample bottles did not leak or cause contamination during the shipment. This event was entered into the licensees CAP as IR 1560642, Radioactive Oil Samples Shipped From Site Without RP Shipper Review, dated September 18, 2013. The RP department immediately stopped work and retrained the RP staff. The significance of the finding was determined by using IMC 0609, Appendix D, "Public Radiation Safety Significance Determination Process." The issue was determined to be of very low safety significance (Green) because it involved radioactive material control, was not a finding involving transportation, and did not result in public exposure greater than 0.005 rem.
05000456/FIN-2013005-042013Q4BraidwoodLicensee-Identified ViolationBraidwood TS 3.7.8, Essential Service Water System, states, in part, that two unit-specific SX trains shall be operable. Technical Specification 3.8.1, Alternating Current (AC) Sources Operating, states, in part, that two EDGs capable of supplying the onsite Class 1E AC electrical power distribution shall be operable. If one unit specific SX train is inoperable, Condition A of TS 3.7.8 requires the licensee to enter the applicable conditions of TS 3.8.1 for the EDG made inoperable by the inoperable SX train. Contrary to the above, on August 1, 2013, the licensee failed to properly determine that operability of the 1A SX and 1A EDG could not be supported and subsequently failed to enter TS 3.7.8 and 3.8.1, as required. Specifically, the licensee initially applied ASME Code Case 513-3 after identifying a pinhole leak on an elbow fitting on the SX return isolation line from the 1A EDG and concluded that the SX train was operable. However, upon re-evaluation, the licensee identified that operability in accordance with ASME Code Case 513-3 was limited to straight pipe and not elbow fittings. On August 2, 2013, the licensee declared the 1A SX train and 1A EDG inoperable and entered TS 3.7.8 Condition A; TS 3.8.1 Condition B; and Technical Requirements Manual 3.0.c. Weld repairs were completed and operability was supported on August 3, 2013. The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance and Configuration Control attributes 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 inspectors screened the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, answered No to all of the Mitigating System Screening questions, and determined the finding was one of very low safety significance (Green). This issue was entered into the licensees CAP as IR 1542372, Essential Service Water Piping Leak - 1SX27DA, dated August 1, 2013. As part of the licensees corrective actions, Operability Determination procedure OP-AA-108-115 was planned to be revised to provide clearer guidance regarding the application of ASME Code Case 513-3.
05000456/FIN-2013005-032013Q4BraidwoodFailure to Submit Report Required by 10 cfr 50.72(b)(3)(xiii)The inspectors identified a Severity Level IV NCV of 10 CFR 50.72(b)(3)(xiii) when licensee personnel failed to submit a report required by 10 CFR 50.72 for a loss of emergency assessment capability when an unplanned degradation was identified associated with the Technical Support Center (TSC) ventilation filtered make-up train. Specifically, the discharge damper for the TSC ventilation filtered make-up fan was found unexpectedly closed, which adversely impacted the ability to supply filtered air to the TSC absent implementation of compensatory actions. Corrective actions included making the required Event Report on January 14, 2014. The licensee entered this issue into their CAP as IR 1598598, Wording Differences Between NUREG-1022 and Reportability Manual, and IR 1608133, ENS (Event Notification System) Call Made Due to TSC Ventilation Impact in October 2013. The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the generic communications input that 10 CFR 50.72 reports serve. Since the issue impacted the regulatory process, it was dispositioned through the traditional enforcement process. The inspectors determined that this issue was a Severity Level IV violation based upon Example 6.d.9 in the NRC Enforcement Policy. Example 6.d.9 specifically stated, The licensee fails to make a report requirement by 10 CFR 50.72 or 10 CFR 50.73. Because a more-than-minor Reactor Oversight Process finding was not identified, there was no cross-cutting aspect associated with this violation. (Section 4OA2.2b)
05000456/FIN-2013005-022013Q4BraidwoodFailure to Follow Procedure and Technical Specification Associated with Control for High and Locked High Radiation AreasThe inspectors identified a self-revealed finding of very low safety significance and an associated NCV of Technical Specification 5.7.1 when licensee personnel failed to adequately monitor and provide positive control over activities within a high radiation area that was greater than 100 millirem per hour (mrem/hr) but less than or equal to 1000 mrem/hr from a radiation source which was created during the cycling of valve 1RH8701B inside the missile barrier in containment. A slug of material dislodged from the valve and was transported to a location that resulted in localized elevated dose rates where an individual was performing work. As an immediate corrective action, the licensee instituted appropriate radiation protection controls and initiated an Apparent Cause Evaluation (ACE) to review the event in more detail. The licensee entered this issue into their CAP as IR 1559430, ED (Electronic Dosimeter) Dose Rate Alarm Received. The performance deficiency was more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, not evaluating the radiological impact of the slug of radioactive material being transported to an area where a worker was performing work caused the worker to receive unnecessary and unplanned exposure to radiation that if left uncorrected could lead to a more significant safety concern in that a worker could receive a much higher dose under different circumstances. The inspectors determined that the finding was of very low safety significance (Green) using IMC 0609, Appendix C. This finding had a crosscutting aspect in the Work Practices component of the Human Performance crosscutting area because licensee personnel failed to validate and communicate th changing dose rates of the work area after Operations personnel performed work that affected the dose rates in the work area (H.4(a)). (Section 2RS1.6b).
05000456/FIN-2013005-012013Q4BraidwoodFailure to Maintain Accurate Operator LogsThe inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, when licensee personnel failed to provide complete and accurate operator logs of record. Specifically, operator log entries of record on May 9, 2013, did not accurately document entry into and exit from Limiting Condition for Operation (LCO) 3.0.3. Initial corrective actions included additional late log entries and issuance of Operations Standing Order 13-10, Corrections to Electronic Log Entries, which provided interim guidance to operators regarding how to make revisions to electronic log entries. The Operations Director also initiated discussions with the fleet Operations Director peer group to determine how to incorporate guidance on revising electronic logs into procedure OP-AA-111-101, Operating Narrative Logs and Records. The licensee entered this issue into their Corrective Action Program (CAP) as Issue Report (IR) 1519660, Lack of Details in Log Entries. In consultation with regional enforcement staff, the inspectors determined that the issue was more than minor because operator logs of record are material documents to the NRC, in that inspection activities are planned and conducted based, in part, on the review of operator logs and the presumption of their accuracy. In determining the significance of the violation, the inspectors referenced the examples of violations in Section 6.9, Inaccurate and Incomplete Information or Failure to a Make a Required Report, of the NRC Enforcement Policy. Because the issue was determined to be more than minor, but did not meet the threshold of the examples of Severity Level I, II, or III violations, the inspectors determined this issue was a Severity Level IV violation. Because a more-than-minor Reactor Oversight Process finding was not identified, there was no cross-cutting aspect associated with this violation. (Section 1R22.2.b)
05000456/FIN-2013004-012013Q3BraidwoodFailure to Perform a Required 10 CFR 50.59 EvaluationThe inspectors identified a finding of very low safety significance and an associated Severity Level IV NCV of 10 CFR 50.59, Changes, Tests, and Experiments, when licensee personnel failed to perform and maintain a written evaluation to demonstrate that a procedure change did not require a license amendment. Specifically, the licensee implemented a change to procedures 1/2BwOA SEC-4, Loss of Instrument Air, Revision 3, that revised the actions to address a loss of component cooling water (CC) to the reactor coolant pump (RCP) thermal barrier heat exchange such that a complete loss of seal cooling could occur, which would result in damage to the RCP seals and a subsequent loss of coolant accident (LOCA). As part of the licensee corrective actions, procedures 1/2 BwOA SEC-4 were revised to address the issue. A revised 10 CFR 50.59 evaluation was also developed and approved. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it could be reasonably viewed as a precursor to a significant event. 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, Phase 1 - Initial Screening and Characterization of Findings, Table 2, for the Initiating Events cornerstone. The inspectors then answered No to all of the screening questions in Table 3. The finding was further evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1. The inspectors answered No to all of the questions contained therein. Therefore, the inspectors concluded the finding was of very low safety significance (Green). Because the associated finding was determined to be of very low safety significance in accordance with the SDP, the traditional enforcement aspect of this issue was determined to be at the Severity Level IV level. The inspectors did not identify a cross-cutting aspect associated with this finding since it was not indicative of current performance.
05000456/FIN-2013004-022013Q3BraidwoodFailure to Perform a Required 10 CFR 50.59 EvaluationThe inspectors identified a finding of very low safety significance and an associated Severity Level IV NCV of 10 CFR 50.59, Changes, Tests, and Experiments, when licensee personnel failed to perform and maintain a written evaluation to demonstrate that a procedure change did not require a license amendment. Specifically, the licensee implemented a change to procedures 1/2BwOA SEC-4, Loss of Instrument Air, Revision 3, that revised the actions to address a loss of component cooling water (CC) to the reactor coolant pump (RCP) thermal barrier heat exchange such that a complete loss of seal cooling could occur, which would result in damage to the RCP seals and a subsequent loss of coolant accident (LOCA). As part of the licensee corrective actions, procedures 1/2 BwOA SEC-4 were revised to address the issue. A revised 10 CFR 50.59 evaluation was also developed and approved. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it could be reasonably viewed as a precursor to a significant event. 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, Phase 1 - Initial Screening and Characterization of Findings, Table 2, for the Initiating Events cornerstone. The inspectors then answered No to all of the screening questions in Table 3. The finding was further evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1. The inspectors answered No to all of the questions contained therein. Therefore, the inspectors concluded the finding was of very low safety significance (Green). Because the associated finding was determined to be of very low safety significance in accordance with the SDP, the traditional enforcement aspect of this issue was determined to be at the Severity Level IV level. The inspectors did not identify a cross-cutting aspect associated with this finding since it was not indicative of current performance.
05000456/FIN-2013004-032013Q3BraidwoodLicensee-Identified ViolationBraidwood License Condition 2.E requires, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the UFSAR, as supplemented and amended. Section 2.3.5.2 of the approved fire protection report describes the Division 22 engineered safety feature (ESF) switchgear room and refers to the description contained in Section 2.3.5.1. Section 2.3.5.1 describes the Division 12 ESF switchgear room and states that the floor is a 5-inch clear cover of structural reinforced concrete with a 3-inch concrete topping over 3-inch fluted steel decking formwork. The floor is supported by structural steel beams protected with a fire resistant covering and carries a 3-hour fire rating. Contrary to the above, on July 28, 2013, a licensee individual performing a routine firewatch activity in the Division 22 ESF switchgear room identified a 25-inch long by 6.5-inch wide section of the poured concrete floor missing and a small nickel-sized hole in the metal floor plate that opened into the 2B EDG room. The un-poured portion of the floor and the hole were previously hidden from view and were recently revealed during installation of an unrelated plant modification. The inspectors screened the issue in accordance with IMC 0612, Appendix B, Issue Screening, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined the finding was of very low safety significance (Green). This issue was entered into the licensees CAP as IR 1540434, GOCAR Un-poured Hole in Floor 426 L-30.
05000456/FIN-2013004-042013Q3BraidwoodLicensee-Identified ViolationBraidwood TS 5.4.1.a requires that the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, shall be established, implemented, and maintained. Section 1.l of Regulatory Guide 1.33 recommends procedures covering the plant fire protection program. Contrary to the above, on July 30, 2013, the licensee identified that hourly firewatches in the 2A EDG room required by procedure BwAP 1110-1A1, GOCAR Required Compensatory Measures Action Response Fire Detection Instrumentation, Revision 8, had not been completed between 10:24 p.m. on July 29, 2013 and 3:30 p.m. on July 30, 2013, a span of approximately 17 hours. The inspectors screened the issue in accordance with IMC 0612, Appendix B, and IMC 0609, Appendix A, and determined the finding was of very low safety significance (Green). This issue was entered into the licensees CAP as IR 1541347, Missed Firewatches for 2A EDG Room.
05000456/FIN-2013003-042013Q2BraidwoodImplementation of Lake Chemistry Management ProgramThe inspectors identified an URI associated with the licensees implementation of station procedural standards to notify Senior Site Management and Operations at the first sign of a lake softening event, and to implement AOP BwOA-ENV-7, Adverse Cooling Lake Conditions, when pre-determined calcium precipitation rate limits were exceeded on three occasions from March 2012 through April 2013. The licensees root cause analysis performed following a 2004 Braidwood Lake Precipitation Event (IR 199206, Lake Chemistry Trend Calcium Carbonate Issue, Assignment 13) identified that the Lake Chemistry Plan had not been formalized into operational procedures and, as a result, guidelines for administrative controls, actions limits and levels, and contingency actions had not been established for managing lake chemistry. As one of the corrective actions to address this issue, the licensee developed and implemented AOP BwOA-ENV-7, Adverse Cooling Lake Conditions, to address any future adverse lake precipitation event (IR 199206, Assignment #35). On November 10, 2004, BwOA-ENV-7, Adverse Cooling Lake Conditions, was approved, placed within the station procedures, and was required to be followed in accordance with station standards. This AOP stated that prompt actions may be required to minimize any adverse effects on plant operation. Procedure BwOA-ENV-7 required that several actions be performed to minimize the impact of a significant lake precipitation event. For example, the procedure directed numerous actions to determine whether there had been an adverse impact on plant systems. These actions included the observation of traveling screen operation, monitoring of safety-related and nonsafety-related service water system strainer performance, trending of main condenser pressure, and the monitoring of component cooling system heat exchanger performance, fire protection jockey pump performance, and reactor containment fan cooler service water flow. Upon the identification of any adverse impact, the procedure directed notification of the Braidwood Station Duty Team to ensure appropriate actions would be taken commensurate with safety. Additionally, immediately following entry, BwOA-ENV-7 required that the Emergency Director evaluate Emergency Plan conditions. The procedure also required that the licensee minimize SX and auxiliary feedwater pump, main control room chiller, and EDG operation to preclude chemical or biological fouling. Following issuance, BwOA-ENV-7 had been revised numerous times to modify the thresholds and standards for informing Senior Site Management and Operations of lake precipitation events and to specify the standards upon which Operations would be notified to implement the procedure. For the period of January 2012 through May 2013, CY-BR-120-412, Lake Chemistry Data Sheet, Revision 7 was in effect and required the following: At the first sign of a precipitation event or natural softening, NOTIFY Senior Site Management and Operations (Reference Section 3.5). COMPARE Calcium Hardness and Total Alkalinity trends to determine behavior of these parameters during period of softening and non-softening. (Reference Section 4.6.5) - REVIEW CW Makeup and blowdown flow history, as well as recent weather precipitation. - If lake softening rate exceeds 15 ppm (parts per million) Calcium Hardness in a 2-3 day period, NOTIFY Operations to enter BwOA-ENV-7. The inspectors reviewed Braidwood Lake chemistry data from January 2012 through April 2013. The inspectors identified that the licensee appeared to have not followed the standards discussed above for three of the five potential lake softening events during this period. Specifically, the inspectors identified that Senior Site Management and Operations notification and entry into procedure BwOA-ENV-7, Adverse Cooling Lake Conditions, was delayed for up to several days after the licensee had performed lake water sampling, had analyzed the sample, and had documented the results. The following specific examples were identified: 2012 First Lake Softening Event (BwOA-ENV-7 Entered on March 5, 2012 3 Days After Entry Conditions were Present Date Calcium Delta Between Prior Day Sample 2/29/2012 257 3/2/2012 231 (26) - 2012 Third Lake Softening Event (BwOA-ENV-7 Entered on April 15, 2012 2 Days After Entry Conditions were Present) Date Calcium Delta Between Prior Day Sample 4/11/2012 194 4/13/2012 167 (27) - 2013 Second Lake Softening Event (BwOA-ENV-7 Entered on April 4, 2012 1 Day After Entry Conditions were Present) Date Calcium Delta Between Prior Day Sample 4/1/2013 209 4/3/2013 191 (18) The inspectors determined through interviews with licensee personnel and through the review of Operations logs that the licensee had not notified Senior Management and Operations at the first signs of the listed lake softening events or had implemented BwOA-ENV-7 earlier than was documented in the Operations logs. As a result of not implementing BwOA-ENV-7, Adverse Cooling Lake Conditions, when required, the licensee did not appear to perform the actions required by the AOP in a time frame commensurate with station standards. Therefore, the licensee failed to meet the standards that they had originally developed and modified over the years to minimize the possible adverse effects of lake precipitation events. The inspectors discussed this issue of concern with licensee staff, management, and senior management who disagreed with the inspectors assessment. The main points of the disagreement were on the meaning of the term at the first sign and on the acceptability of allowing a sample to be taken and analyzed on one day but not reviewed by a supervisor through the Lake Chemistry Control Program until chemistry staff were available, potentially several days later. The inspectors inferred from the term at the first sign that actions were required to be performed without an undue delay and that these actions were not dependent upon readily available chemistry staff. In the past two lake precipitation events, plant equipment was adversely impacted relatively soon after the onset of the event. The inspectors recognized that the elevated differential calcium concentration samples identified during this inspection did not actually result in a lake precipitation event. As of the end of the inspection period, the licensee planned to determine the impact of a 2-3 day delay in implementing BwOA-ENV-7 on the ability to mitigate a lake softening event. Pending a review of this information, this issue is considered a URI. (URI 05000456/2013003-04; 05000457/2013003-04, Implementation of Lake Chemistry Management Program)
05000440/FIN-2013009-012013Q2PerryFailure to Implement the Operational and Radiological Controls Necessary to Prevent Plant Manipulations from Adversely Impacting Dose Rates or Airborne Radioactivity LevelsThe inspectors identified a finding of very low safety significance and associated non-cited violation of Technical Specification (TS) 5.4, Procedures. Specifically, TS 5.4 Procedures , Step 5.4.1 states, in part, that the licensee shall establish, implement, and maintain applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A. Section 7 of Appendix A of RG 1.33 specifies radiation protection procedures for control of radioactivity for limiting personnel exposures. Licensee procedure NOP-OP-4107, Radiation Work Permit, requires that radiological controls identify critical steps or critical instructions for positive radiological control of the work to ensure no change on unexpected change in radiological conditions, and prevent unplanned exposure. Contrary to this, on six occasions during the spring 2013 refueling outage, the licensee failed to implement operational and radiological controls necessary to prevent plant manipulations from adversely impacting ambient radiological dose rates or airborne radioactivity levels in the plant when workers were in the areas. The licensee documented this issue in its corrective action program as condition report 2013-09891. As an immediate corrective action, the licensee instituted the appropriate operational and radiological controls to ensure personnel safety. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening and determined that the issue was more than minor because, if left uncorrected, the performance deficiency could have led to a more significant safety concern. Specifically, not implementing the operational and radiological controls necessary to prevent plant manipulations from adversely impacting ambient radiological conditions in the plant could result in unnecessary and unplanned radiation exposures. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupation Radiation Safety Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance, work-control, because the licensee did not appropriately plan work activities when developing the work packages and authorizing the work.
05000440/FIN-2013009-022013Q2PerryFailure to Lock or Continuously Guard Doors to Prevent Unauthorized Entry to an LHRAA finding of very low safety significance and an associated non-cited violation of Technical Specification 5.7, High Radiation Area, was self-revealed when the access point to the locked high radiation area of the auxiliary steam tunnel on the 620-elevation of the turbine building was left unattended on May 1, 2013, for about 8 minutes. This issue was entered into the licensees corrective action program as condition report 2013-06892. As immediate corrective actions, access to the area was guarded and appropriate controls were instituted. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues, and determined that the issue was more than minor because it was similar to Example 6(g). The inspectors also determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported.
05000440/FIN-2013009-032013Q2PerryFailure to Post and Barricade a HRA in the Under-Condenser Area Turbine Building Cubicles 13 and 14The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification 5.7. High Radiation Area, when the inspectors identified an unposted, unbarricaded high radiation area under the condenser in turbine building cubicles 13 and 14 that was accessible to personnel by scaffold. This issue was entered into the licensees corrective action program as condition report 2013-06139. As an immediate corrective action, the scaffold was removed and appropriate controls were instituted. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues, and determined that the issue was more than minor because it was similar to Example 6(g). The inspectors also determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. This finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee did not thoroughly evaluate and address this issue when initially identified by the NRC in 2011 or during the licensees extent of condition evaluations.
05000440/FIN-2013009-042013Q2PerryFailure to Follow Procedural Requirements for RWCU System Fill and VentA finding of very low safety significance and associated non-cited violation of Technical Specification 5.4, Procedures, was self-revealed when the licensee failed to adhere to procedural requirements during the filling and venting of the reactor water cleanup (RWCU) system. Specifically, on April 26, 2013, valves 1G33-F008A and F556A were left in the open position, contrary to the requirements of step 7.16.9 of procedure SOI-G33, revision 36, and resulted in the RWCU system being aligned to the condensate transfer and storage system. This valve misposition event also resulted in the TS 3.6.1.3 inoperability of the containment isolation valve 1P11F0545. Upon discovery of the condition, the licensee promptly corrected the error and the entered the condition into its corrective action program as condition report 2013-07483, and performed an apparent cause evaluation. The inspectors reviewed Inspection Manual Chapter (MC) 0612, Appendix B, Issue Screening, and determined that the issue was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance, work practices, for the licensees failure to successfully incorporate human error prevention techniques, such as self and peer checks.
05000440/FIN-2013009-052013Q2PerryFailure to Implement a Procedure Appropriate to the Circumstances Leads to Reactor Overfeed EventA finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to perform a procedure that was appropriate to the circumstances. Specifically, on May 12, 2013, work instruction PTI-N27-P0012, Revision 5, was performed when the condition of the plant, i.e., the specific configuration of the feedwater system and the relatively low reactor pressure, was incapable of supporting the test and resulted in a reactor overfill event. The issue was entered into the corrective action program as condition report 2013-07473. The licensee performed an apparent cause evaluation to identify the most likely causal factors, citing the inadequacy of the procedure and the lack of proper planning as contributing causes. The inspectors reviewed Inspection Manual Chapter (MC) 0612, Appendix B, Issue Screening, and determined that the issue was more than minor because it was associated with the Initiating Events Cornerstone attribute of procedure quality 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 inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance, work control, for the licensees failure to plan work activities such that they could be performed while the plant was in an appropriate operational condition. Specifically, the licensee rescheduled the activity without performing an adequate impact review of the different plant conditions on the activity.
05000440/FIN-2013009-062013Q2PerryLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation. Technical Specification 5.7.2 states, in part, that areas accessible to personnel with radiation levels such that a major portion of the whole body could receive in 1 hour a dose greater than or equal to 1000 millirem shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the shift supervisor on duty or the radiation protection supervisor. Contrary to this, on April 4, 2012, the licensee inappropriately down-posted the reactor water clean-up backwash receiving tank room from a locked high radiation area to a high radiation area. This issue was documented in the licensees CAP in CR 2012-18277. Immediate corrective actions included restoring the required locked high radiation area posting and instituting the appropriate associated access controls. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised.
05000456/FIN-2013003-032013Q2BraidwoodImplications of Control Room Ventilation Monthly SurveillanceThe inspectors identified an Unresolved Item (URI) regarding the use of TS Limiting Condition for Operation (LCO) 3.7.10 during the monthly control room ventilation system surveillance. Specifically, the inspectors questioned whether a step in procedure 0BwOSR 3.7.10.1-1, Control Room Ventilation Filtration Surveillance (Train A), to realign the VC suction source, and which appeared to defeat an automatic engineered safety feature (ESF) realignment, impacted the filtration system (Condition A) or control room envelope (CRE) boundary (Condition B) of the LCO. At 4:05 p.m. on May 8, 2013, the licensee commenced a routine monthly surveillance of the A VC filtration train using procedure 0BwOSR 3.7.10.1-1, Control Room Ventilation Filtration Surveillance (Train A). During performance of the surveillance, at 7:09 p.m., the licensee noted that B VC train damper 0VC08Y was unexpectedly open when it should have been closed. Approximately 25 minutes later, the damper repositioned closed. Operators were dispatched to inspect the damper and heard an abnormal grinding noise coming from the hydramotor. Consultation with the system engineer indicated that the grinding noise was likely caused by a degraded bearing. As a result, the licensee declared the B train of VC inoperable and entered LCO 3.7.10, Condition A, One VC Filtration System Train Inoperable for Reasons Other Than Condition B. Condition B stated, One or More VC Filtration System Trains Inoperable Due to Inoperable CRE Boundary in Mode 1, 2, 3, or 4. The licensee elected to continue with the routine surveillance on the A VC train. Step F5.1 of procedure 0BwOSR 3.7.10.1-1 directed Operations to enter LCO 3.7.10, Condition A, for the A VC train while the makeup filter selector switch was repositioned from auto to outside air then turbine building and back to auto as part of a contact check. The licensee entered LCO 3.7.10, Condition A, for the A VC train at 4:33 a.m. on May 9, 2013, and exited that Condition at 4:35 a.m. For those 2 minutes, both Units also entered LCO 3.0.3, since the A and B VC trains were simultaneously inoperable due to LCO 3.7.10, Condition A. During plant status activities on the morning of May 9, 2013, the inspectors noted discussions among senior plant personnel about whether LCO 3.7.10, Condition B (not Condition A) was actually the correct Condition to be entered while performing Step F5.1 of procedure 0BwOSR 3.7.10.1-1. The inspectors reviewed the TSs and discussed the system design with the VC system engineer. The VC system is designed such that when the makeup air suction is from outside air, the system would automatically realign the source air to the turbine building upon an air intake high radiation signal or a safety injection signal. When the makeup filter selector switch is not in the auto position, this automatic realignment will not occur, and manual actions would be required for the system to perform its ESF function. Additionally, the inspectors reviewed the licensees Control Room Habitability Program (CRHP), which included the following definitions: CONTROL ROOM ENVELOPE (CRE) BOUNDARY: A combination of walls, floor, roof, ducting, doors, penetrations, and equipment that physically form the CRE. CONTROL ROOM HABITABILITY SYSTEMS (CRHS): The plant systems that help ensure CRE habitability. This includes the Control Room emergency ventilation/filtration system and the Control Room HVAC systems. The CRE boundary is considered as an integral part of the CRHS, since it is critical to maintaining CRE habitability. The inspectors view was that the automatic realignment feature of the A VC train, which was blocked at the time the switch was not in auto, did not constitute part of the CRE boundary as defined in the CRHP. In addition, manual actions were required for the safety-related system to perform its ESF design function. As a result, the inspectors communicated to licensee management their view that Condition A was the correct Technical Specification Action Statement (TSAS) to be entered when performing the surveillance. Following this discussion, the licensee continued to believe that Condition B was the correct TSAS to enter when performing this surveillance. The inspectors also communicated their concerns that main control room logs, as officially recorded, did not completely and accurately capture the events that occurred on the night shift from May 8 to May 9, 2013. During plant status activities on May 9, the inspectors reviewed the main control room operating logs at approximately 6:30 a.m., and noted the log entries for entering LCO 3.7.10, Condition A, for the 0A VC train, and LCO 3.0.3, at 4:33 a.m. and exiting those LCOs at 4:35 a.m. However, later that morning when the logs were reviewed again, the inspectors noted those log entries had been revised. The log entries were annotated with, Late Entry 1030 5/9/13, and referenced entry into LCO 3.7.10, Condition B, and made no mention of LCO 3.0.3. There was no indication that anything had been revised or that LCO 3.0.3 had been entered. As a result of the inspectors concerns, the licensee generated IR 1519660, Lack of Detail in Log Entries, on May 30, 2013. Additionally, an Operations Noteworthy Event briefing sheet was created on June 12, 2013, and discussed with all Operating crews. The Noteworthy Event briefing sheet included the statement, Initially, LCO 3.0.3 was entered, but was retracted on days. LCO 3.7.10, Condition B, was determined to be the correct LCO entry. On July 8, 2013, the licensee again performed the monthly VC surveillance. Upon review of the main control room logs, the inspectors noted that LCO 3.7.10, Condition A, had been entered from 11:14 a.m. to 11:33 a.m. while alternating the suction source between outside and turbine building air. When questioned why the Noteworthy Event briefing sheet instructed Operating crews to enter Condition B and yet the crews entered Condition A, the licensee stated they were waiting for a more comprehensive review of the issue before revising the surveillance procedure. At the end of the inspection period, the inspectors were in the process of discussing the issue with NRC staff in the Office of NRR, reviewing the licensees determination of LCO applicability, and reviewing control room ventilation system design documentation. Pending additional information from the NRR staff, a complete understanding of the licensees position, and a more detailed understanding of the VC system design, this issue is considered a URI. (URI 05000456/2013003-03; 05000457/2013003-03, Implications of Control Room Ventilation Monthly Surveillance)
05000456/FIN-2013003-012013Q2BraidwoodFailure to Identify and Correct Degraded DOST Room Sump Pump Discharge Check ValvesThe inspectors identified a finding of very low safety significance when licensee personnel failed to identify degraded Diesel Oil Storage Tank (DOST) room sump discharge check valves in 2013 and after performing periodic testing in 2005. The licensee entered this issue into their Corrective Action Program (CAP) as Issue Report (IR) 1526652, IR Not Generated as Required 2005 OD Check Valve UT (Ultrasonic Testing) Results. Corrective actions included the repair of the degraded DOST room sump check valves. The inspectors determined that the failure to identify issues associated with degraded DOST room sump pump discharge check valves was a performance deficiency. The inspectors determined that the performance deficiency was 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). Since the finding resulted in the potential for a loss of the emergency power function during a turbine building flooding event, and based upon an actual DOST room sump check valve failure, a detailed risk evaluation was performed, which determined that the finding was of very low safety significance. This finding had a cross-cutting aspect in the Corrective Action Program component of the Problem Identification and Resolution (PI&R) cross-cutting area because the licensee failed to take appropriate corrective actions in a timely manner to address degraded DOST room sump check valves.
05000456/FIN-2013003-022013Q2BraidwoodFailure to Scope Nonsafety-Related Turbine Building to Auxiliary Building Sump Pump Discharge Check Valves into the Maintenance RuleThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(b)(2)(ii) when licensee personnel failed to scope four Unit 1 and Unit 2 Essential Service Water (SX) pump room sump pump discharge check valves and eight Unit 1 and Unit 2 DOST room sump pump discharge check valves into the Maintenance Rule as required. The licensee entered this issue into their CAP as IR 1498897, Review 1/2WF040A/B Valves for Inclusion Into MRule (Maintenance Rule), and planned to scope the components into the Maintenance Rule. The inspectors determined that the failure to scope the Unit 1 and Unit 2 SX pump room sump pump discharge check valves and Unit 1 and Unit 2 DOST room sump pump discharge check valves into the Maintenance Rule was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Since a degraded SX or DOST sump check valve would degrade one or more trains of a system that supported a risk-significant system or function, a detailed risk evaluation was performed that determined the finding was of very low safety significance. This finding had a cross-cutting aspect in the Decision-Making component of the Human Performance cross-cutting area because the licensee failed to use conservative assumptions readily available in the applicable guidance document to demonstrate that not scoping the components into the Maintenance Rule was in accordance with Maintenance Rule requirements and therefore maintained safety.
05000456/FIN-2013003-052013Q2BraidwoodInadequate Functionality Evaluations for a Degraded Unit 1 BAST BladderA finding of very low safety significance was self-revealed when licensee personnel performed inadequate functionality evaluations after previously identifying that the Unit 1 Boric Acid Storage Tank (BAST) bladder was degraded. The licensee entered this issue into their CAP as IR 1498696, Secured Boric Acid Tank Transfer Earlier Than Expected. Corrective actions included the replacement of the Unit 1 and Unit 2 BAST bladders. The inspectors determined that the failure to adequately evaluate Unit 1 BAST system functionality after identifying that the Unit 1 BAST bladder had substantially degraded was a performance deficiency. The inspectors determined the performance deficiency was more than minor because it 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). The inspectors screened the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors answered No to all of the Mitigating System Screening questions for Reactivity Control Systems, therefore the finding screened as having very low safety significance. This finding had a cross-cutting aspect in the Operating Experience component of the PI&R cross-cutting area because the licensee failed to implement and institutionalize Operating Experience that specifically discussed the potential adverse consequences that a degraded tank bladder could have on plant safety.
05000456/FIN-2013003-062013Q2BraidwoodInadequate Control of a Special Lifting DeviceThe 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 licensee personnel failed to adhere to design requirements specified for a special lifting device used to handle a transfer cask containing spent nuclear fuel in the vicinity of the spent fuel pool. The licensee entered this issue into their CAP as IR 1509204, Required NDE (Nondestructive Examination) Not Performed on Lift Yoke, and IR 1509602, Lift Yoke Stud Nuts Not Lock Wired. As part of their corrective actions, the licensee performed required tests and installed lock wire in accordance with design drawings prior to conducting additional lifts with the special lifting device. The inspectors determined that the failure to adhere to design drawings and American National Standards Institute (ANSI) requirements for annual testing of a special lifting device was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because it was associated with the Design Control attribute of the Barrier Integrity Cornerstone and adversely impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radioactive releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors answered No to all the screening questions in Appendix A, Exhibit 3, and therefore the finding screened as having very low safety significance. This finding had a cross-cutting aspect in the Resources component of the Human Performance cross-cutting area since the licensee failed to have complete, accurate, and up-to-date design documentation and procedures that ensured personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically the licensees procedures for annual testing of a special lifting device lacked specific guidance, and design changes were made that conflicted with design drawings.
05000456/FIN-2013003-072013Q2BraidwoodInadvertent Removal of the Design Basis Requirement to Commence a Cooldown Within 2 Hours Following the Establishment of Natural Circulation Conditions and Loss of Air to ContainmentThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when licensee personnel failed to maintain the procedural requirement to commence a reactor coolant system (RCS) cooldown within 2 hours following a design basis seismic event that included a reactor trip, failure of all nonsafety-related equipment, and limiting single active failure. The licensee entered this issue into their CAP as IR 1496506, NRC Identified PZR (Pressurizer) PORV (Power-Operated Relief Valve) Natural Circulation Cooldown Analysis. Corrective actions included development of a revised instruction in the Emergency Operating Procedures (EOPs). The inspectors determined that the failure to adequately revise an EOP was a performance deficiency. Specifically, the licensee removed a procedural requirement to commence an RCS natural circulation cooldown if instrument air was lost to containment, which inadvertently could adversely affect a safety-related PZR PORV function. The inspectors determined that the performance deficiency was more than minor because it was associated with the Procedural Quality 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 inspectors evaluated this finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined that this finding was of very low safety significance because the issue was determined to not be a confirmed loss of operability or functionality. This finding had a cross-cutting aspect in the Corrective Action Program component of the PI&R cross-cutting area because licensee personnel failed to thoroughly evaluate a problem and ensure that the resolution adequately addressed the cause and extent of condition, as necessary. Specifically, the licensee failed to adequately evaluate a prior NRC finding such that the corrective actions adequately addressed the problem.
05000456/FIN-2013003-082013Q2BraidwoodFailure to Account for PZR PORV Accumulator Leakage During Hot Standby and Subsequent Cooldown Period Following a Postulated EarthquakeThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when licensee personnel failed to account for PZR PORV accumulator air system leakage during the assumed 2 hour time spent in hot standby following a limiting seismic event. The licensee entered this issue into their CAP as IR 1481590, NRC Question Regarding Pressurizer PORV Accumulator Leakage. As part of their corrective actions, the licensee planned to revise procedures and seek clarification from the NRC concerning the licensing basis of the auxiliary spray system. The inspectors determined that the failure to ensure that the PZR PORVs could perform their credited safety function following a limiting seismic event was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because it was associated with the Design Control 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 inspectors evaluated this finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined that the finding was of very low safety significance because the issue was determined to not be a confirmed loss of operability or functionality. This finding had a cross-cutting aspect in the Corrective Action Program component of the PI&R cross-cutting area because the licensee failed to thoroughly evaluate a problem such that the resolution addressed causes and extent of condition, as necessary. Specifically, the licensee failed to adequately evaluate not accounting for PZR PORV air accumulator leakage in the natural circulation cooldown current licensing basis (CLB) due to the reliance on another system to provide the credited safety function.
05000456/FIN-2013002-062013Q1BraidwoodCurrent Licensing Basis Requirements for RCS Pressure Control Function During a Postulated Seismic Event in Reference to NRC RSB BTP 5-1The inspectors identified an URI regarding the licensees interpretation of their CLB requirements pertaining to the RCS Pressure Control Safety Function during a postulated seismic event and assumed 2 hour period in hot standby. Specifically, the inspectors identified three issues of concern that questioned the licensees ability to maintain RCS pressure control without the reliance of the primary safety valves and in a manner that could accomplish an RCS cooldown within a timeframe required by RSB BTP 5-1. Description: The licensees CLB utilized the standards in NRC BTP RSB 5-1, Design Requirements of the Residual Heat Removal System, Revision 2, dated July 1981, to meet aspects of 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 19 and GDC 34. In summary, the station was licensed to demonstrate the capability to reach a cold shutdown condition assuming a design basis earthquake resulting in a LOOP and the failure of all non-safety, non-seismically qualified equipment. Design functions necessary to maintain hot standby and cold shutdown conditions include inventory control, reactivity management, decay heat removal, and RCS pressure control. The three issues of concern discussed in this URI are related to the RCS pressure control function during the assumed 2 hour hot standby period. The licensees Analysis of Record (AOR) assumed the following: 1) the time spent in hot standby will be limited to 2 hours, 2) the safety-related PZR PORV and associated instrument air accumulators could maintain RCS pressure in hot standby without the reliance on the RCS code safety valves, and 3) every attempt would be made to open key CVCS valves needed for auxiliary spray in the case that the PZR PORVs were not available. Since instrument air was considered nonsafety-related, instrument air was assumed to be unavailable during this postulated seismic event. The licensees UFSAR stated, however, that every attempt would be made to either restore the instrument air compressors (in the case of a LOOP) or to utilize nitrogen bottles to open the necessary air valves to restore the nonsafety-related auxiliary spray system if the PZR PORVs were not available. Issue of Concern 1: Inadvertent Removal of the Design Basis Requirement to Commence a Cooldown within 2 Hours Following the Establishment of Natural Circulation Conditions and Loss of Instrument Air to Containment...... Issue of Concern 2: Failure to Account for Allowable PZR PORV Accumulator Air Leakage During 2 Hour Hot Standby Period. .....Issue of Concern 3: No Procedures for Crediting the Use Auxiliary Spray Utilizing Portable Nitrogen Bottles. ....Based d on the above, the inspectors questioned whether the licensee had appropriately addressed the issues both individually and collectively to the standards required by NRC regulations. At the conclusion of the inspection period, the inspectors were reviewing the licensees CLB. This URI will remain open pending additional review. (URI 05000456/2013002-06, 05000457/2013002-06, Current Licensing Basis Requirements for RCS Pressure Control Function During a Postulated Seismic Event in Reference to NRC RSB BTP 5-1)
05000456/FIN-2013002-052013Q1BraidwoodNonSafety-Related Turbine Building Waste Disposal System to Safety-Related Essential Service Water Pump Room Sump Design InteractionOn January 21, 2013, the licensee documented in IR 1465027, 1WF040A Not Seating Properly, that SX sump pump discharge check valves 1WF040A and/or 1WF040B might be leaking by based on data that indicated that when the TB sump pump(s) operated, the Unit 1 and Unit 2 A train SX pump room sump pump(s) would start shortly after. This condition suggested that the TB sump pump(s) were filling the Unit 1 and Unit 2 A train SX sump to a level that caused the SX sump pump(s) to start. The licensees prompt operability evaluation was documented in IR 1473152, Single Point Vulnerability for SX Pump Room Flooding, and concluded that the SX pumps were operable since the SX pump room sump pumps can pump water out of the SX pump room sumps and, therefore, prevent water from accumulating in the SX pump room. However, the inspectors noted that previous IRs indicated degraded performance of both A train SX pump room sump pumps (IR 1426946, 1WF06PB Does Not Develop Adequate Discharge Pressure, and IR 1464644, 1WF06PA and B Degraded Insufficient Urgency to Correct. ) On February 13, 2013, the licensee updated their operability review to credit isolating the TB from the SX pump rooms by closing nonsafety-related isolation valves 1WF055 and 2WF055 until the final operability evaluation was complete. On February 14, 2013, the licensee documented that alarm response procedure BwAR OPL02J-2-A6, TB Floor Drain Sump Level High High, was being revised to provide operator direction to align the SX pump room sump to the Radioactive Waste system in the event of TB flooding. Additionally, credit was given to the nonsafety-related SX pump room sump high level alarm to alert operators to an off-normal level condition. The licensee credited the SX pump room sump pumps to be able to pump against the head pressure from the flood water in the TB, though reference was not given to their degraded condition. Issue Report 1473152 referenced UFSAR 10.4.5, Circulating Water System, and identified that the worst case flood in the TB could theoretically reach 396 feet. The lowest elevation of the SX sump pumps was 322 feet. The IR stated that the discharge of the SX room sump pumps was given as 100 gpm at 106 feet which would prevent inflow from the TB. The IR also stated that the NRC Standard Review Plan (SRP) requirement to prevent flooding of a safety-related area was maintained. On March 18, 2013, WO 1497423 was performed and identified that the disc for 1WF040B (SX sump discharge check valve) was stuck in the mid-position. NRC SRP 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, BTP SPLB 3-1 B.3.b, stated, In analyzing the effects of postulated piping failures, the following assumptions should be made with regard to the operability of systems and components: (1) Offsite power should be assumed to be unavailable if a trip of the turbine-generator system or reactor protection system is a direct consequence of the postulated piping failure; (2) A single active component failure should be assumed in systems used to mitigate consequences of the postulated piping failure and to shut down the reactor, except as noted in Item B.3.b.(3) below. The single active component failure is assumed to occur in addition to the postulated piping failure and any direct consequences of the piping failure, such as unit trip and loss of off-site power (LOOP). Additionally, SRP 9.3.3, Equipment and Floor Drainage System, required that the equipment and floor drainage system be capable of preventing a backflow of water that might result from maximum flood levels to areas of the plant containing safety-related equipment. SRP 10.4.5, Circulating Water System, required compliance with General Design Criteria 4, Environmental and Dynamic Effects Design Bases, based on meeting the following: 1) Means should be provided to prevent or detect and control flooding of safety-related areas so that the intended safety function of a system or component will not be precluded due to leakage from the Circulating Water system; and 2) Malfunction or a failure of a component or piping of the Circulating Water system including an expansion joint should not have unacceptable adverse effects on the functional performance capabilities of safety-related systems or components. Based on the above, the inspectors questioned whether the failure of the 1WF040B check valve would result in water from a postulated TB flood to backflow into the common Unit 1 and Unit 2 A train SX pump room sumps resulting in the loss of the 1A and 2A SX Pumps. The inspectors were unable to determine during the inspection whether the licensees justification was acceptable and therefore this issue will be considered an URI pending further NRC review.
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