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05000461/FIN-2017004-032017Q4ClintonFailure to Identify the Extent of Condition for an Inadequate 10 CFR 50.59 EvaluationThe inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to follow a procedure that implemented the Quality Assurance Program requirements. Specifically, the licensee failed to follow procedure PIAA1251003, Corrective Action Program Evaluation Manual, and identify the extent of condition for a lack of proficiency in applying the licensing basis when performing 10 CFR 50.59 evaluations. The licensee documented this issue in their CAP as AR 04075581. The licensee planned an Updated Safety Analysis Report (USAR) Upgrade Project which reportedly would include a review of safety evaluations for USAR changes that dated back to 1986 and determined the scope of this project would be adequate to identify the extent of condition for this issue. The inspectors determined that this issue was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, because the extent of condition review was not adequate, there is a potential for other safety systems to have been adversely affected by a lack of proficiency in applying the licensing basis during safety related system changes. As a result, safety-related systems may not be able to perform intended safety functions as defined in the USAR. This issue would also adversely affect the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was screened against all cornerstones and determined to be of very low safety significance because the finding met each of the applicable screening questions to be characterized as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence, which stated individuals follow processes, procedures, and work instructions. Specifically, when the NRC violation was documented in the CAP previously it was not appropriately classified in accordance with PIAA120 and was also incorrectly closed to an unrelated evaluation. This contributed to the failure to appropriately perform an extent of condition. (H.8)
05000461/FIN-2017004-022017Q4ClintonFailure to Assess and Manage Risk Associated with the Performance of Control Rod Time TestingThe inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, the licensee failed to assess and manage the increase in risk associated with performing control rod scram time testing in Mode 1, prior to performing the activity. As corrective actions, the licensee assessed the increase in risk for performing control rod scram time testing at power and developed a risk mitigation plan that was used to complete the testing. This performance deficiency was determined to be more than minor because the finding was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to assess risk and develop a risk mitigation plan for control rod time testing at power contributed to an automatic reactor scram. Using IMC 0609, Attachment 4, Initial Characterization of Findings, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued May 5, 2005, and Appendix M, Significance Determination Process Using Qualitative Criteria dated April 12, 2002, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance. Specifically, the inspectors and the Region III Senior Reactor Analyst (SRA) determined that Appendix K was not directly applicable to this finding because the licensee performs qualitative evaluations of maintenance activities that have the potential to cause a transient rather than quantitative evaluations. The SRA used insights from Appendix K to support a qualitative SDP evaluation using the principles of Appendix M. The SRA determined that the maintenance activity could only result in an uncomplicated reactor transient event and that the increased risk of a transient compared to the baseline risk of the plant was of very low safety significance. The SRA considered the conditional core damage probability of an uncomplicated transient in this evaluation, which was less than 1E6, to conclude that the finding was Green. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. While performing the work order risk screening for completing the control rod scram time testing while the reactor was shut down, the screener identified that a new screening would be needed if the testing was performed at power. However, no holds were placed on the work order to ensure the risk screening was completed. (H.5)
05000461/FIN-2017004-012017Q4ClintonReactor Down Power due to Reactor Recirculation Pump Motor Lower Bearing Oil LeakA self-revealed finding of very low safety significance was identified for the licensees failure to comply with the requirements of station procedure MAAA716004, Compression Fittings Inspection, Installation, Remake and Repair, Revision 3. Specifically, the licensee failed to properly assemble a joint that was part of the reactor recirculation (RR) pump motor B oil level monitoring system that subsequently leaked requiring the plant operators to perform an unplanned power reduction to allow for identification and repairs of the leak. The licensee documented this issue in the corrective action program (CAP) as Action Request (AR) 04029024. As corrective actions, the licensee made repairs to the effected joint, inspected the remaining joints to ensure proper integrity, and filled the lower bearing reservoir. This issue was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and 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. Specifically, the leak on the reactor recirculation pump motor oil level monitoring system would have eventually resulted in the failure of the reactor recirculation pump causing a transient and upsetting plant stability. This finding was determined to be of very low safety significance because the event did not cause a reactor trip. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the licensee failed to ensure that all personnel installing the compression fittings received an installers brief prior to performing work on the reactor recirculation pump motor oil level monitoring system. (H.5)
05000331/FIN-2017004-012017Q4Duane ArnoldFailure to Maintain Reactor Water Level within Procedurally Required Level Band Results in Reactor Recirculation Pump RunbackThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to operations personnel failing to control reactor vessel water level in accordance with Integrated Plant Operating Procedure 2, Startup, Revision 160. Specifically, during a reactor startup, while at 55 percent reactor power with only one reactor feed pump running, the operating crew failed to maintain reactor water level within the procedurally required level band which resulted in a recirculation pump runback to 45 percent speed and an unplanned reactor power decrease from 55 to 43 percent. The licensee responded to the transient and verified that reactor power stabilized at 43 percent without complications, conducted a human performance review, and entered this issue into their corrective action program (CAP) as condition report (CR) 02233094.The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to control reactor water level within the procedurally specified water level band resulted in an unplanned recirculation pump runback and a decrease in reactor power from 55 to 43 percent. The finding was determined to be of very low safety significance because the finding did not cause a reactor trip. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of teamwork, where individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, a reactor operator dialed down the reactor water level control set point without notifying the control room supervisor, briefing the evolution, or obtaining a peer check. (H.4)
05000331/FIN-2017004-022017Q4Duane ArnoldFailure to Evaluate Site Fire and Explosion Hazards in Accordance with 10 CFR 72.212(b) (6)The inspectors identified a Severity Level IV NCV of 10 CFR 72.212(b)(6), Conditions of General License Issued under 72.210, for the failure of thelicensee as of June 9, 2003, to determine whether or not reactor site parameters were enveloped by the cask design bases as considered in the Updated Final Safety Analysis Report (UFSAR). Specifically, the licensee failed to evaluate site-specific fire and explosion hazards that were allowed to be near the dry cask storage systems under its Administrative Control Procedure (ACP) 1412.2, Control of Combustibles, Revision 48. The licensee documented this issue in its CAP as CR 02228514 and CR 02228558 and took timely corrective actions.The inspectors determined that the violation was of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. Example 4k is applicable to this issue in that the lack of evaluation showing that the quantity of combustible and flammable liquids stored near the dry cask storage system were bounded by the design basis in the UFSAR allowed for a credible unanalyzed fire and explosion scenario that could affect the important-to-safety dry cask storage system. The violation screened as a Severity Level IV NCV. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000461/FIN-2017004-042017Q4ClintonFailure to Perform an Evaluation in Accordance with 10 CFR 72.48 for Changes Made to the Time-to-Boil CalculationThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 72.48(d)(1), Changes, Tests, and Experiments, for the licensees failure to perform a written evaluation which provides the bases for the determination that changes do not require a Certificate of Compliance amendment pursuant to 10 CFR 72.48(c)(2). Specifically, the licensee accepted Engineering Change Order ECO501825R0 (1), ECO501848R1 (1), and ECO501848R1 (4) on June 20, 2016, to the time-to-boil calculation as described in the HI-STORM FW Final Safety Analysis Report and incorrectly screened out performing an evaluation of those changes in accordance with 10 CFR 72.48. The licensee documented this issue in its CAP as AR 02714091 and AR 04081583. The licensee is performing a 10 CFR 72.48 evaluation for Engineering Change Order ECO501825R0 (1) and ECO501848R1 (4) while planning to revise the acceptance of ECO501848R1(1). The inspectors determined that the violation was of more than minor significance as the inspectors could not reasonably conclude that the above changes did not require prior NRC approval. The violation screened as a Severity Level IV non-cited violation using example 6.1.d.2 of the NRC Enforcement Policy. No cross-cutting aspect was identified since cross-cutting aspects are not assigned to traditional enforcement violations.
05000282/FIN-2017003-042017Q3Prairie IslandLicensee-Identified ViolationPrairie Island Technical Specification 3.0.6 requires, in part, that an evaluation shall be performed in accordance with Technical Specification 5.5.13, Safety Function Determination Program, when a supported system LCO is not met solely due to a support system LCO not being met. Specifically, if a loss of safety function is determined to exist by the Safety Function Determination Program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.Contrary to this TS requirement, between August 18 and 22, 2017, control room operators did not evaluate Unit 2 A Component Cooling, Auxiliary Feedwater, and Cooling Water supported system LCOs while the 121 Safeguards Chilled Water support system LCO was not met. As a result, the appropriate Conditions and Required Actions were not entered during Unit 2 B Component Cooling and Auxiliary Feedwater supported system maintenance and testing activities for which a loss of safety function existed. Because the inspectors answered No to all questions under Exhibit 2.A of IMC 0609, Appendix A, The Significance Determination Process for Findings at-Power, the finding screened as very low safety significance (Green). Specifically, the finding did not represent (result in) an actual loss of function of two separate safety systems out-of-service for greater than their TS-allowed outage times. The above issues were documented in the licensees CAP as CAP 501000001929. Corrective actions included revisions to applicable station procedures for implementing TS 3.0.6 and the Safety Function Determination Program.
05000282/FIN-2017003-032017Q3Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program shall be established to assure that all testing required to demonstrate that SSCs will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, the licensee established procedure 5 AWI 3.12.4, 26 Post-Maintenance Testing, Revision 24, as the program for selecting and documenting post maintenance tests (PMTs) and return to service tests to ensure that SSCs would perform their intended function when returned to service. Contrary to the above, on September 20, 2017, the licensee failed to assure that testing required the demonstrate that three safety injection system actuation relays would perform satisfactorily in service was identified and performed in accordance with written test procedures, which incorporated the requirements and acceptance limits contained in applicable design documents. The three safety injection system actuation relays had not been tested following replacement during planned maintenance. Specifically, while reviewing PMT activities performed on the D5 EDG on September 19, 2017, the licensee identified three safety injection system actuation relays that had not been tested following replacement during planned maintenance. As a result, the D5 EDG was declared inoperable at the time of discovery on September 20, 2017. In response, the licensee performed an in-depth review of all recent D5 EDG maintenance activities to ensure that all PMT requirements were met and performed SP 2150, D5 Diesel Generator Function Test, on September 21, 2017, to adequately test all three safety injection system actuation relays and an additional D5 EDG slow start test to fully demonstrate operability of D5. Because the inspectors answered No to all questions under Exhibit 2.A of IMC 0609, Appendix A, The Significance Determination Process for Findings at-Power, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CAP 501000002920. Corrective actions included performing an apparent cause evaluation, department clock reset, and planned changes to 5 AWI 3.12.4 to ensure all required PMT activities are performed satisfactorily prior to returning SSCs to service.
05000282/FIN-2017003-022017Q3Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.48(b)(2) requires, in part, that all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O. Appendix R, Section III.G.3 of 10 CFR Part 50, requires, in part, that alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section. In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration. Contrary to the above, on December 21, 2015, the licensee failed to provide an alternative or dedicated shutdown capability for 17 MOVs credited in the licensees Appendix R Safe Shutdown Analysis that did not satisfy the requirements of 10 CFR Part 50, Appendix R, Section G.2. Specifically the MOVs could have been rendered unavailable for manual operator action following a postulated fire in the control or relay rooms. These manual actions were required to achieve and maintain safe shut down in the event of a fire that resulted in functional loss and/or evacuation of the control and/or relay rooms. Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non compliances identified as a result of a licensees transition to the new risk informed, performance based fire protection approach included in 10 CFR 50.48(c), and for 25 certain existing non compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk informed, performance based approach is referred to as National Fire Protection Association (NFPA) 805, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. At the time of discovery, the licensee was in transition to NFPA 805 and therefore the licensee-identified violation was evaluated in accordance with the criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation: (1) the licensee identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. The finding also met additional criteria established in section 12.01.b of IMC 0305, Operating Assessment Program. In addition, in order for the NRC to consider granting enforcement discretion the violation must not be associated with a finding of high safety significance (i.e., Red). The licensee performed risk evaluation V.SPA.16.001, Revision 0, dated March 27, 2017, and determined that this issue was not associated with a finding of high safety significance. A Region III Senior Reactor Analyst (SRA) reviewed the evaluation and concluded that the result was reasonable and that the finding was less than Red and eligible for enforcement discretion. The dominant core damage sequence from the licensees evaluation was a fire in the Control Room or Cable Spreading Room which could cause spurious operation of several MOVs necessary for safe shutdown. The SRA used IMC 0609, Appendix F, Fire Protection Significance Determination Process, to review the results of the licensees evaluation. The SRA validated the licensees calculations through a series of walkdowns, reviews of the calculation and verification of the values used were consistent with NUREG-6850 and IMC 0609, Appendix F. The licensees results were approximately 1E6 deltaCDF and 2E8 deltaLERF for this finding and hence were significantly lower than the 1E4 deltaCDF threshold for a finding of high safety significance. In addition, the licensee entered this issue into their corrective action program as CAP 1506561. As a result, the inspectors concluded that the violation met all four criteria established by Section 9.1(a) and that the NRC was exercising enforcement discretion to not cite this violation in accordance with the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues.
05000282/FIN-2017003-012017Q3Prairie IslandFailure to Ensure Correct Operation of Meteorological TowerA finding of very-low safety significance, and an associated NCV of Technical Specification (TS) 5.4.1 was identified by the NRC inspectors for the failure to implement and maintain procedures to ensure adequate operation of a meteorological tower. The licensee entered this issue into their Corrective Action Program (CAP) as CAP 501000001091, dated July 27, 2017. The licensee had initiated efforts to assess and remove unnecessary vegetation growth. The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding impacted the Plant Facilities/Equipment and Instrumentation Attribute of the Public Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, existing meteorological tower procedures did not include the assessment and subsequent removal of trees that could impair the correct operation of sensors located at the 10 meter elevation of the tower. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008. The violation was of very-low safety significance (Green) because: it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or Title 10 of the Code of Federal Regulations (CFR), Part 20.1301(e) criteria. The inspectors concluded that the most significant contributing cause of the performance deficiency involved the Resolution cross cutting component in the area of problem identification and resolution because this issue was previously entered into the licensees CAP in 2015 and closed with no action taken. (P.3)
05000373/FIN-2016007-012016Q2LaSalleFailure to Monitor the Fouling Conditions of the CSCS Equipment Area CoolersThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to monitor the fouling conditions of the core standby cooling system (CSCS) equipment area coolers. Specifically, the licensee did not develop performance test procedures to assess the fouling conditions of the safety-related CSCS equipment area coolers and did not have acceptance criteria that delineate when to remove accumulations. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02665463 and established a standing order for operations to impose more restrictive service water temperature limits to reasonably assure the operability of the affected coolers until long term corrective actions were implemented to restore compliance. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee reviewed actual service water temperature values measured during the last 12 months, performed an operability evaluation, and concluded that the historical temperatures did not exceed the operability limits established by the operability evaluation. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance. Specifically, the test program for the CSCS equipment area coolers was developed in the decade of 1990s.
05000373/FIN-2016007-022016Q2LaSalleFailure to Ensure that Both Feed Supply Breakers for Swing DG Components Were Closed During Normal Plant OperationThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to have the capability to verify the supply breakers of both reactor units feeding the swing diesel generator (DG) components were closed during normal plant operation. Specifically, the circuit design and procedures for the swing DG room fan, fuel oil transfer pump, and fuel storage tank room exhaust fan did not ensure the detection of the condition where one of these feeder breakers was tripped in the open position during normal plant operation. The licensee captured this issue in their CAP as AR 02668759 and created a special log to monitor the associated breakers once per day. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of system and/or function, represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its Technical Specifications (TS) allowable outage time, and represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant for greater than 24 hours. Specifically, a historical review did not find an example where the swing DG was non-functional for a period greater than the applicable TS allowable outage time as a result of this finding during the last year. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the mean to detect an opened breaker associated with the affected loads was established more than 3 years ago.
05000373/FIN-2016007-032016Q2LaSalleInadequate Procedures for Containment Debris ManagementThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to establish procedures that were appropriate to manage containment debris consistent with the emergency core cooling system strainer debris loading design basis and supporting design information. Specifically, the procedures did not contain instructions for evaluating containment debris sources consistent with the associated analyses and other design documents. The licensee captured the team concerns in their CAP as AR 02663076 and AR 02656299. The immediate corrective actions included an operability evaluation that reasonably determined all of the affected emergency core cooling system strainers remained operable. 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability review and reasonably determined that only a portion of the unqualified coatings would be available for transport to the strainers and this quantity was bounded by the associated design basis analysis. In addition, this review reasonably determined that sufficient analytical margin existed to accommodate the quantities of the other debris types found during recent inspections. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated procedures were established more than 3 years ago.
05000373/FIN-2016007-042016Q2LaSalleAlternate Shutdown Procedures Failed to Ensure RCIC MOVs Supply Breakers Were ClosedThe team identified a finding of very-low safety significance (Green) and associated NCV of the LaSalle County Station Operating License for the failure to ensure that procedures were in effect to implement the alternate shutdown capability. Specifically, the abnormal operating procedures (AOPs) established to respond to a fire at the main control room did not include instructions for verifying that supply breakers for three reactor core isolation cooling motor-operated valves (MOVs) were closed to ensure they could be operated from the remote shutdown panel. Fire-induced failures could result in tripping MOV power supply breakers prior to tripping the MOV control power fuses. The licensee captured the team concerns in their CAP as AR 02668854 and established compensatory actions to reset the affected breakers, if required The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external events (fire), and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it was assigned a low degradation factor. Specifically, the procedural deficiencies could be compensated by operator experience/familiarity and the fact that the AOPs included steps to verify other breakers at the same locations were closed would likely prompt operators to close the remaining breakers. The team determined that this finding had a cross cutting aspect in the area of problem identification and resolution because the licensee failed to take effective corrective actions for a similar issue identified in 2014. Specifically, the resolution of this issue included actions to revise the affected AOPs to include verifying all the reactor core isolation cooling MOVs supplied breakers were closed. However, the licensee failed to include all of the MOVs in the revised AOPs. (P.3)
05000315/FIN-2016001-012016Q1CookIncorrect Auxiliary Feedwater Mission TimeThe inspectors identified a finding of very low safety significance and associated NCV of with Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that regulatory requirements and design bases were correctly translated into specifications and procedures, in that the licensee used an incorrect mission time for the turbine driven auxiliary feedwater (TDAFW) pump to determine operability. The licensee developed a procedure that permitted continued operability of the TDAFW pump without room ventilation provided room temperature remained below 104 F. The underlying engineering document assumed TDAFW pump mission time was 4 hours; however, this assumption was not supported by current license bases documents. This condition violates 10 CFR 50 Appendix B Criterion III, which requires licensees to establish measures to assure that applicable regulatory requirements and the design bases, as defined in 10 CFR 50.2 and as specified in the license application, for those systems structures and components to which the Appendix applies, are correctly translated onto specifications, drawings, procedures and instructions. The licensee has since restored the room coolers to an operable status, thus, no current safety concern exists. The licensee has entered the condition into the corrective action program (CAP). The licensees use of an incorrect mission time was a performance deficiency that warranted a significance review. Using IMC 0612 appendix B dated September 7, 2012, the inspectors determined that the finding was more than minor because it was associated with the Mitigating System cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events and adversely affected the attribute of design control. Specifically, the licensee applied an incorrect mission time when determining room temperatures to ensure TDAFW pump operability. Using IMC 0609 Appendix A, Exhibit 21, dated June 19, 2012, the inspectors answered no to Questions A. 1 thru 4. In particular, control room logs document about 6 hours with the TDAFW room ventilation not functioning; therefore the inspectors determined that the pump would not have been inoperable for longer than the 72 hour completion time in technical specifications. The inspectors also identified a cross cutting aspect of H.14, conservative bias, in the human performance area.
05000315/FIN-2016001-022016Q1CookLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) with an associated NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, for the failure to ensure appropriate quality standards were specified and included in design documents associated with the Unit 1 and Unit 2 ESW strainer backwash valves. Specifically, this resulted in the use of non-dedicated parts in the backwash valves. The backwash function of the ESW strainers was originally classified as non-safety-related. However, in 2007, the backwash function became safety-related. When this change occurred, the Safety Classification Determination (SCD), which documented the safety classification of the various parts of the valves, was not updated accordingly. During a maintenance period on the ESW system in 2015, some licensee personnel questioned the adequacy of the SCD. The licensee later determined that non-dedicated replacement parts had been used in some of the strainer backwash valves since 2007. The issue was more than minor because per IMC 0612 Appendix B, it adversely affected the Mitigating Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The issue screened as Green based on the guidance in IMC 0609 Appendix A, Exhibit 2. Specifically, the finding was associated with the design or qualification of a mitigating SSC where the operability was maintained.
05000461/FIN-2015004-022015Q4ClintonFailure to Perform Activities Affecting Quality in Accordance with Prescribed ProceduresThe inspectors identified a finding of very low safety significance for the failure to ensure that activities were accomplished in accordance with prescribed procedures as required by station procedure HU-AA-104-101 Procedure Use and Adherence. Specifically, the inspectors identified two examples where the licensee failed to adhere to prescribed station procedures when performing activities in the plant. The licensee placed both issues in their corrective action program as Action Request (AR) 02600726 and addressed the nonconformances created by the failure to follow the procedures. The licensee planned to perform an apparent cause evaluation to determine why there was an adverse trend related to procedure adherence. The inspectors determined that the failure to perform activities in accordance with prescribed procedures as required by station procedure HU-AA-104-101, Procedure Use and Adherence, was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, by not performing activities in accordance with a procedure the licensee could manipulate equipment, challenge the operators, and cause unexpected transients. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not cause a reactor trip or the loss of mitigation equipment, and it did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which stated, Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with unknown conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to station equipment.
05000456/FIN-2015004-032015Q4BraidwoodFailure to Establish Adequate Feedwater Pump Operational Guidance During a Normal Plant ShutdownA finding of very low safety significance and an associated NCV of Technical Specification 5.4.1, Procedures, was self-revealed on October 5, 2015, due to the licensees failure to establish and maintain adequate guidance for operating the Unit 1 and Unit 2 motor driven main feedwater pump (MDFWP) during plant shutdown conditions. Specifically, on October 4, 2015, during a Unit 2 plant shutdown, the Unit 2 MDFWP was placed in service at low forward feedwater flow conditions and was manually tripped when the pumps main journal bearing temperature exceeded the procedural limit. Subsequent review, determined that the procedural limit was too low as previously recognized by historic station specific operating experience. This issue was entered into the licensees corrective action program (CAP) as Issue Report (IR) 2565486. The inspectors determined that the performance deficiency was more than minor because the issue was associated with the Procedural Quality attribute of the Initiating Event cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency contributed to a loss of main feedwater event that upset plant stability and challenged the critical safety function of removing decay heat via the steam generators in Mode 3. For Unit 1, the increased potential for a loss of main feedwater event existed under similar conditions. The inspectors determined that the finding was of very low safety significance based upon a detailed risk evaluation. The inspectors concluded that this finding did not have a cross-cutting aspect because the performance deficiency was greater than 3 years old and, therefore, not indicative of recent performance.
05000457/FIN-2015004-012015Q4BraidwoodLoss of Shutdown Cooling Train During Refueling Cavity Fill and Associated Reduced Inventory OperationsOn October 8, 2015, the inspectors identified an Unresolved Item (URI) regarding the failure of valve 2RH606, which is the 2A RHR heat exchanger flow control valve. The valves failure to open caused a loss of one train of shutdown cooling, and an unplanned Orange risk configuration with Unit 2 in Mode 6, and the reactor refueling cavity level less than 23 feet above the vessel flange. At the closure of the inspection period, the licensees investigation on the cause of the failure was ongoing. Resolution of this issue will be based on the inspectors review of the licensees completed investigation. A function of the RHR system in Mode 6 is to remove decay heat and sensible heat from the reactor coolant system (RCS). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the component cooling water system. The coolant is then returned to the RCS via the RCS cold legs. On October 8, 2015, valve 2RH606 became mechanically bound while in the process of filling the Unit 2 reactor refueling cavity to greater than 23 feet. This was identified when the operators attempted to open the valve from the control room. The failure of the valve to open caused Unit 2 shutdown risk to change from a planned Yellow configuration to unplanned Orange condition. Additionally, the licensee entered Limiting Condition for Operation 3.9.6, Residual Heat Removal and Coolant Recirculation-Low Water Level, Condition A, for one train of RHR cooling inoperable. This action required the licensee to initiate actions immediately to either restore the affected RHR loop to operable status or to initiate actions to establish greater than or equal to 23 feet of water above the reactor vessel flange. The licensee accomplished this action by raising water level in the cavity to greater than 23 feet. Troubleshooting of the failed valve revealed that a shaft key sheared, which prevented the valve from opening. The valve had been previously manipulated during the outage without an issue. The malfunctioning part was sent offsite for failure analysis. The valve was repaired. At the conclusion of the inspection, an apparent cause investigation was in process. This URI will remain open until the investigation is complete and the inspectors review the report to determine whether a performance deficiency exists.
05000457/FIN-2015004-022015Q4BraidwoodFailure of Startup Feedwater Pump to Start During Plant ShutdownThe inspectors identified an URI based upon the startup feedwater pumps (SUFWPs) failure to start during a plant shutdown. In addition to being used in plant startups and shutdowns, the SUFWP is also credited in the licensees emergency operating procedure as a means to add water to the steam generators for decay heat removal if the safety-related auxiliary feedwater systems failed to function properly during an event. On October 4, 2015, operations attempted to start the Unit 2 SUFWP at low power in Mode 1 during plant shutdown activities for a refueling outage. Upon start, the SUFWP automatically tripped. The licensee completed an apparent cause evaluation to determine the reason why the pump did not start and run. At the end of the inspection period, the inspectors were awaiting additional information to complete their review to determine if this issue of concern constituted a performance deficiency. This URI will remain open pending this review.
05000461/FIN-2015004-012015Q4ClintonFailure to Update the Final Safety Analysis Report (FSAR) Hydrogen Water Chemistry SystemThe inspectors identified a Severity Level IV Violation of 10 CFR 50.71(e), Periodic Update of the FSAR (Final Safety Analysis Report), for the licensees failure to update the FSAR after installing a hydrogen water chemistry system into the plant to reduce rates of intergranular stress corrosion cracking in recirculation system piping and reactor vessel internals. Specifically, the licensee did not update Section 5.4.15, Hydrogen Water Chemistry System, of the FSAR to include a design basis and description of the process and the system used to periodically injection noble metals. The licensee entered this issue into the corrective action program as AR 02594259 and planned to revise the FSAR to include a design basis and description of the process and the system used to periodically injection noble metals. The inspectors determined that the failure to update the FSAR in accordance with 10 CFR 50.71(e), Periodic Update of the FSAR, with the design basis and description of the process and the system used to periodically injection noble metals was a performance deficiency warranting a significance evaluation. The inspectors reviewed this issue in accordance with NRC IMC 0612 and the NRC Enforcement Manual. Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because cross cutting aspects are not assigned to traditional enforcement only violations.
05000461/FIN-2015004-032015Q4ClintonFailure to Follow Station Procedures for Plant ActivitiesThe inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of Title 10 Code of Federal Regulations (CFR), Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings. Specifically, the inspectors identified two examples where the licensee failed to perform activities affecting quality in accordance with prescribed procedures. The licensee entered this issue into their corrective action program as AR 02600726 and planned to perform an apparent cause evaluation to address the trend. Separate action requests were also written and immediate corrective actions were taken for each identified example to address the nonconformances created by the failure to follow procedures. The inspectors determined that the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings as required by 10 CFR 50, Appendix B, Criterion V, was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities affecting quality in accordance with a procedure the licensee could manipulate equipment and challenge the operators by causing unexpected transients or impact safety-related equipment. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, issued May 9, 2014, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the finding did not represent a loss of system safety function, it did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, it did not represent an actual loss of safety function of one or more non-TS trains of equipment during shutdown for equipment designated as risk significant for greater than 24 hours, and it did not degrade a functional auto-isolation of residual heat removal on low reactor vessel level. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which states, Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with uncertain conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to safety-related equipment and activities.
05000456/FIN-2015004-042015Q4BraidwoodFailure to Establish a Written Procedure for a Loss of Feedwater Event in Mode 3A finding of very low safety significance and an associated NCV of Technical Specification 5.4.1, Procedures, was self-revealed on October 5, 2015, due to the licensees failure to establish a written procedure for combating emergencies and other significant events, as required by Regulatory Guide 1.33, Quality Assurance Program Requirements. Specifically, upon a loss of feedwater in Mode 3 (Hot Standby), which is an expected design and licensing basis event, the licensee did not have a written procedure as established by the Regulatory Guide. This issue was entered into the licensees CAP as IRs 2566239 and 2565513. The inspectors determined the finding to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the Mitigating Systems cornerstone Procedural Quality attribute, and adversely impacted the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the absence of a procedure(s) complicated the operator response to the loss of feedwater event in Mode 3. The inspectors determined the finding to be of very low safety significance in accordance with IMC 0609, Appendix A, The SDP for Findings at Power, dated September 7, 2012, Exhibit 2, since the inspectors answered "No" to the Mitigating Systems questions under Section A, Mitigating Systems, Structures, and Components and Functionality. The inspectors did not identify a cross-cutting aspect associated with this finding, because it was confirmed not to be reflective of current performance due to the age of the performance deficiency.
05000456/FIN-2015003-012015Q3BraidwoodLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are appropriately translated into specifications, drawings, procedures, and instructions. Contrary to the above, as of October 10, 2014, the licensee failed to translate the design basis essential service cooling pond berm height into procedures and instructions. Specifically, procedure BwVSR 3.7.9.3, "Braidwood Cooling Lake Hydrographic Survey," did not ensure that the height of the essential service cooling pond berm was being verified. This issue was entered into the licensees CAP as IR 2400960; The UHS EL. At Top of the East Slope Found Less than 590 ft, dated October 24, 2014, and the procedure was corrected. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue 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). Specifically, following a seismic event that drains the non-essential main Braidwood cooling pond, the essential cooling pond (i.e., UHS) would have a decrease in available inventory at the start of a design basis event. This could reduce the available net positive suction head for the service water pumps that take suction from the UHS, as well as potentially resulting in the UHS design temperature of 100 degrees Fahrenheit being exceeded. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined that the finding affected the design of the UHS, but did not result in a loss of operability, and therefore screened the finding as having very low safety significance (Green).
05000456/FIN-2015003-022015Q3BraidwoodLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions. Procedures, and Drawings, requires, in part, that activities affecting quality shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, from April 23, 2015, to June 24, 2015, the licensee failed to translate specific acceptance criteria into procedures and instructions. Specifically, when the licensee modified the DG fuel oil system in a manner that reduced the DG fuel oil system train separation from two isolation points to one isolation point, the licensee failed to establish quantifiable acceptance criteria in the post-maintenance test, and failed to establish performance monitoring with acceptance criteria, as specified in the design change. This issue was entered into the CAP as IR 2519208 with immediate corrective actions of re-establishing the dual isolation point. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue 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 determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined that the finding was of very low safety significance (Green), because the issue did not prevent the 2B DG from operating for its specified probable risk assessment mission time of 24 hours.
05000282/FIN-2015003-012015Q3Prairie IslandFailure to Determine Compensatory MeasuresA finding of very low safety significance with two examples and an associated non-cited violation of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14. Specifically, on two occasions, the licensee failed to determine compensatory measures following the identification of a Updated Safety Analysis Report (USAR) non-conforming condition associated with the Units 1 and 2 residual heat removal (RHR) recirculation sump valves on August 31, 2015, and for a degraded condition of the Unit 1 B RHR recirculation sump valves on September 14, 2015. The licensee entered the issues into the Corrective Action Program (CAP) as CAPs 01491302 and 01491900. The inspectors determined that the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14, to properly determine compensatory measures for operable but degraded and operable but non-conforming conditions was a performance deficiency. The performance deficiency, with two examples, was determined to be more than minor and a finding in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed on two occasions to properly determine compensatory measures to maintain or enhance operability of Technical Specification (TS) Systems, Structures, and Components (SSCs) that were not fully qualified until final corrective actions were taken. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that because the finding was a qualification deficiency and did not impact operability of the TS SSCs, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor for the performance deficiency was associated with the cross-cutting aspect of Consistent Process in the Human Performance cross-cutting area, involving individuals using a consistent, systematic approach to make decisions. Specifically, the licensee did not apply a consistent, systematic approach for determining the appropriateness of compensatory measures while making operability decisions for the degraded and non-conforming conditions associated with the RHR recirculation sump valves.
05000282/FIN-2015003-032015Q3Prairie IslandLicensee-Identified ViolationTitle 10, CFR Part 50.72(b)(3)(xiii) states, in part, a licensee shall report (notify the NRC as soon as practical and in all cases within 8 hours of the occurrence) any event that results in a major loss of emergency assessment capability. Contrary to this requirement, over the past 3 years, the licensee identified six instances (on August 14, 2012; November 16, 2012; November 18, 2012; November 21, 2012; December 5, 2012; and January 16, 2013) of a failure to report the major loss of emergency assessment capability when the Seismic Monitoring Panel was non-functional for unplanned events. The licensee also identified three instances (on December 14, 2012; September 3, 2014; and September 30, 2014) of a failure to report the major loss of emergency assessment capability when the Seismic Monitoring Panel was non-functional for planned events for greater than 24 hours. The system degradation adversely impacted the sites ability to make an ALERT and a Notice of Unusual Event Emergency Action Level assessment in accordance with PINGP-1575, Emergency Action Level Matrix, and F3-2.1, Emergency Action Level Technical Bases. The licensee entered the issue into the corrective action program as CAP 01472229, OE Review of NRC Event Reports Related to Seismic Monitors, CAP 01472731, Missed Reportability for Seismic Monitor Out of Service, and CAP 01486147, Potential Licensee ID Violation from EP Inspection. The licensee completed the required report to the NRC on April 2, 2015 (Event Number 50948, Seismic Monitor Not Available for Emergency Plan Assessment). 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 Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, example d.9 in the NRC Enforcement Policy. Example d.9 specifically states, A licensee fails to make a report requirement by 10 CFR 50.72 or 10 CFR 50.73. Because the issues were entered into the licensees corrective action program as CAPs 01472229, 01472731, and 01486147, the violation is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000282/FIN-2015003-022015Q3Prairie IslandImproper Operability DeterminationA finding of very low safety significance and an associated non-cited violation of Title 10, CFR, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14. Specifically, on August 9, 2015, following the discovery of a non-functional D6 building ventilation system and declaration of inoperability of Buses 26, 221, 222, and the D6 DG, the licensee improperly declared the affected TS SSCs operable and fully qualified without restoring functionality of the ventilation TS support system or implementing appropriate compensatory measures per the requirements of FP-OP-OL-01. The licensee entered the issue into the Corrective Action Program as CAP 01490027. The inspectors determined that the licensees failure to accomplish the requirements of procedure FP-OP-OL-01, Operability/Functionality Determination, Revision 14 was a performance deficiency. The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee improperly declared the TS SSCs operable and fully qualified without restoring functionality of a TS support system or implementing appropriate compensatory measures. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. The inspectors answered No to all questions within Table 3, SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Per Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered No to all questions under Section A, therefore the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor for the performance deficiency was associated with the cross-cutting aspect of Challenge the Unknown in the Human Performance cross-cutting area, involving individuals stopping when faced with uncertain conditions and evaluating and managing risk prior to proceeding. Specifically, the licensee did not properly evaluate and manage uncertain conditions associated with the ventilation system and impacts on TS SSC operability prior to proceeding with declaration of full qualification.
05000440/FIN-2014005-042014Q4PerryFailure to Follow Procedures During Dry Cask OperationsThe inspectors identified a Severity Level IV NCV of very low safety significance of 10 CFR Part 72.150, Instructions, Procedures, and Drawings, for the licensees failure to follow procedures important to safety during dry cask operations. The licensee entered each example identified into its corrective action program as Condition Reports 201411637 and 201414279. The violation was determined to be more than minor in that both examples identified deficiencies in the performance of dry cask operations important to safety. In this determination, the inspectors considered example 4.a in IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, and concluded that, while the errors did not result in any actual safety concern, there were multiple examples of procedural non-compliance. Additionally, if left uncorrected, a failed weld could lead to a release of radioactive materials to the environment and a malfunction of the Fuel Handling Building crane could lead to a more significant safety concern such as a load drop. The significance of the violation was found to be similar to SLIV example 6.5.d.3, of the NRCs Enforcement Policy, in that the licensee failed to adequately implement Quality Assurance processes or procedures. The issue was not found to be similar to any examples of higher significance; as such, the violation screened as a SLIV violation. Since traditional enforcement was used to disposition the violation, a cross-cutting aspect is not applicable.
05000440/FIN-2014005-012014Q4PerryUnevaluated Preconditioning of Emergency Service Water Motor-Operated Valves and Check Valves Prior to Conducting As-Found Inservice Surveillance TestingThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees unevaluated preconditioning, on October 15, 2014, of emergency service water (ESW) pump discharge motor-operated valves and check valves prior to performing as-found inservice testing (IST). This finding was entered into the licensees corrective action program for resolution as Condition Report 2014-15759. The unevaluated preconditioning was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, unevaluated preconditioning of valves could mask their actual as-found conditions and result in an inability to verify their operability, as well as make it difficult to determine whether the valves would perform their intended safety function during an event. The inspectors determined that the finding was of very low safety significance because the finding was confirmed not to result in a loss of operability or functionality of the ESW system. The finding has a cross-cutting aspect in the area of human performance associated with the work management component because the licensee did not implement a process of planning, controlling, and executing work activities to prevent preconditioning of valves prior to testing.
05000440/FIN-2014005-022014Q4PerryInadequate Procedure for Performing an Acceptable Technical Specification Required Channel CheckThe inspectors identified a finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1.a., Procedures, was identified for the licensees failure to establish and maintain a correct surveillance inspection procedure for redundant reactivity control system (RRCS) channel checks. The licensee entered the issue into the corrective action program as Condition Report 201417635 and took immediate actions for a missed surveillance in accordance with TS. The inspectors determined that the failure to establish and maintain a correct surveillance procedure required by TS 5.4.1.a. was a performance deficiency and resulted in the licensees failure to perform a channel check that meets the TS definition of a channel check. The performance deficiency was determined to be more than minor, and thus a finding, because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the channel check surveillance procedure did not compare the channel indication and status to other indications or status derived from available independent instrument channels measuring the same parameter. The inspectors determined that the finding was of very low safety significance because the finding (1) did not affect a reactor protection system trip signal and the function of other redundant trips or diverse methods of reactor shutdown, (2) did not involve control manipulations that unintentionally added positive reactivity, and (3) did not result in a mismanagement of reactivity by operators. No cross-cutting aspect is assigned as this performance deficiency first occurred in 1986 and is not indicative of current licensee performance.
05000440/FIN-2014005-032014Q4PerryFailure to Follow Licensee Procedure to Properly Screen and Evaluate Temporary Changes to Plant Facilities / Structures, Systems, or ComponentsThe inspectors identified a finding of very low safety significance and associated NCV of Technical Specification 5.4.1.a, Procedures, for the licensees failure to implement the requirements of Nuclear Operating Business Practice (NOBP)LP4003A, FENOC 10 CFR 50.59 User Guidelines. This finding was entered into the licensees corrective action program for resolution as Condition Report 201500284. The inspectors determined that the failure to complete a Regulatory Applicability Determination (RAD) specified in NOBPLP4003A was a performance deficiency. The performance deficiency was more than minor, and thus a finding, because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to have very low safety significance because the finding: (1) was not a design or qualification issue confirmed not to result in a loss of operability or functionality; (2) did not represent an actual loss of safety function and/or system; (3) did not result in the loss of one or more trains of TS equipment; and (4) does not represent the loss of a non-TS train of equipment. The finding has a cross-cutting aspect in the area of human performance associated with the change management component, in that leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000440/FIN-2014005-052014Q4PerryLicensee-Identified ViolationTitle 10 CFR, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states in part that, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances. The requirements of Criterion V apply to Criterion XVI of Appendix B, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above requirements, on October 27, 2014, the licensee completed a surveillance to time the operation of main steam line drain valves and could not determine the stroke time for one of the valves; 1N22F420C had a failed position indication element resulting in the inability to time the valve. The valve was declared inoperable and TSs were entered as appropriate. However, no corrective action document was initiated, contrary to procedure NOPLP2001, Corrective Action Program, which requires that, Condition Reports will be initiated upon discovery of any degraded condition that affects a safety structure, system, or component or any USAR described system, structures, or component. The main steam drain system is described in the USAR and the CAP is intended to meet the requirement of Criterion V to provide a quality procedure to support the requirements of Criterion XVI. The performance deficiency of failing to write a corrective action document was discovered when the valve timing failed again during a performance of the test on December 15, 2014, and the operators noted that no repairs had been made since the conduct of the test in October 2014, during which the valve stroke timing test initially failed. The performance deficiency was documented in the CAP as CR 201418329. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, if left uncorrected, the performance deficiency would become a more significant safety concern. The finding was evaluated using IMC 0609, Significance Determination Process, Attachment 0609.4, Initial Characterization of Findings, dated June 19, 2012. Exhibit 2 of Appendix A, the Mitigating Systems Screening Questions, Section A.1, Mitigating SSCs and Functionality, was checked as Yes because the finding is a deficiency affecting design or qualification. As a result, the finding screens as very low safety significance.
05000456/FIN-2014004-022014Q3BraidwoodMultiple Failures to Follow Operability Evaluation Process Following Discovery of a Non-Conforming Condition in the Ultimate Heat SinkThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when licensee personnel failed to follow procedure OPAA108115, Operability Determinations. Specifically, licensee personnel failed to adhere to numerous Operability Determination Process standards after identifying a non-conforming condition that had the potential to impact the operability of the Ultimate Heat Sink (UHS). This issue was entered into the licensees CAP as IR 1674557, Question on UHS License Amendment Request Impact on Pumps, and IR 1675291, Unanalyzed Condition Identified During IR 1674557 Response. Corrective actions included correcting the non-conforming condition by revising the abnormal operating procedures to be aligned with the current licensing basis (CLB). The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue 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). Specifically, based on the analysis of record, at the time of discovery there was reasonable doubt that the UHS could meet its mission time of 30 days. The inspectors determined that the finding was of very low safety significance in accordance with IMC 0609, Appendix A, Exhibit 2, since it was determined to not represent a confirmed loss of operability. The inspectors concluded that this finding had a cross-cutting aspect in the Conservative Bias component of the Human Performance cross-cutting area because the licensee failed to use conservative assumptions in their decision-making when evaluating the operability of the UHS. Specifically, operations did not request a documented evaluation to support understanding why the UHS was operable and to verify that their assumptions regarding operator actions were feasible (H.14).
05000456/FIN-2014003-012014Q2BraidwoodIssues That Could Adversely Affect the UHSUnresolved Item: Issues That Could Adversely Affect the Ultimate Heat Sink Introduction: The inspectors identified four potential issues of concern after the licensee discovered that station procedures to address a failure of the Braidwood cooling lake dike did not include steps to secure nonsafety-related pumps, although the UFSAR stated and design calculations assumed that all circulating water pumps and nonsafety-related service water pumps would be secured.15. Description: Issue 1: TS 3.7.9, Ultimate Heat Sink, Limiting Condition for Operation Applicability After Identifying that a Non-Conforming Condition Could Challenge and/or Exceed the Associated Ultimate Heat Sink 30 Day Mission Time. The Braidwood cooling lake dike allows the ultimate heat sink (UHS) level to be maintained greater than the TS minimum level of 590. A failure of this nonsafety-related dike would cause a loss of level in the UHS to the 590 TS minimum level. During the inspection period, the licensee discovered that station procedures to address a failure of the Braidwood cooling lake dike did not include steps to secure nonsafety-related pumps, including circulating water pumps and service water pumps, that take a suction from the UHS and discharge to a location outside the UHS. As a result, and because the UFSAR stated and design calculations assumed that all nonsafety-related pumps, including circulating water pumps and service water pumps, would be secured to conserve UHS inventory following a dike failure, a non-conforming condition was identified. The licensee concluded that this non-conforming condition did not render the UHS inoperable as discussed in IR 1675291, Unanalyzed Condition Identified During IR 1674557, and IR 1676076, Discrepancy in the UFSAR Ultimate Heat Sink Description (Section 2.4.11.6), based upon the following: The issue was process-related and only concerned future planned actions for increasing the maximum UHS temperature; All TS 3.7.9, Ultimate Heat Sink, surveillance requirements were met; The Braidwood cooling lake did not actually reach the minimum TS level of 590; A cooling lake dike failure did not actually occur; and A statement in the UFSAR concerning the ability of the UHS to handle an assumed loss-of-coolant-accident coincident with a design basis seismic event that the licensee believed was erroneous. Specifically UFSAR Section 2.4.11.6, Ultimate Heat Sink Design Requirements included the following statement: ...The essential service water cooling pond (ESCP) is an excavated area located within the cooling pond designed to provide a sufficient volume to permit plant operation for a minimum 30-day period without requiring makeup water in accordance with Regulatory Guide 1.27. The ESCP has been reviewed to determine its ability to handle the total heat dissipation requirement of the station assuming a loss of coolant accident (LOCA) coincident with a loss of offsite power on one unit and a concurrent orderly shutdown and cooldown from maximum power to cold shutdown of the other unit using normal shutdown operating procedures, a single active failure, a coincident design basis seismic event... The inspectors noted that IR 1674557, Question on Ultimate Heat Sink License Amendment Request Impact on Pumps, documented that the licensee had preliminarily determined that operation of a single nonsafety-related service water pump at full flow would deplete the UHS in about 3.6 days and, as a result, the UHS would not be able to satisfy the 30-day post-accident volume requirements required by the plants design basis. The licensee concluded that even though procedural guidance did not 16 explicitly direct that nonsafety-related pumps be secured following a design basis accident, operators would recognize the problem and take actions to ensure that the UHS would still be able to perform its safety function and meet all design basis requirements. At the end of the inspection period, the licensee planned to more formally document the bases for UHS TS operability consistent with the definition of operability in the site-specific TSs and the licensees Operability Determination procedure. The licensee subsequently corrected this non-conforming condition by revising procedures to secure nonsafety-related pumps upon reaching a low lake level condition consistent with plant design calculations. Therefore, the inspectors did not have a current operability concern. Issue 1 will remain open pending the completion of the inspectors review of the licensees past operability determination. Issue 2: Timeliness of Actions to Inform the Shift Manager and/or Unit Supervisor of an Issue that May Affect Ultimate Heat Sink Operability On June 25, 2014, the inspectors reviewed IR 1674557, which documented that AOP BwOA ENV3, Braidwood Cooling Lake Low Level, did not direct nonsafety-related pumps that take a suction from the UHS and discharge outside of the UHS to be secured following a dike failure. In particular, although the Operability section of IR 1674557 was left blank, the Reviewed section concluded the following: There were no equipment deficiencies identified. This is a process issue regarding future planned actionsthere are no TS/Technical Requirements Manual/Offsite Dose Calculation Manual/GOCAR (General Operation Corrective Action Requirement) actions applicable; reportability criterion affected; or any SSC (structure, system and component) availability or functionality concerns raised by this issue. The inspectors determined that although the context of IR 1674557 suggested that this issue only impacted future planned actions that, in fact, the issue could affect the current operability of the UHS. Therefore, the inspectors promptly discussed this issue with the Operations Shift Manager who was not aware of any operability concerns associated with the issue or station actions to address the issue. Later that shift, the Shift Manager determined that the issue was reportable under 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition. At the end of the inspection period, it was not clear if the station had adhered to OPAA108115, Operability Determinations to inform the Shift Manager and/or Unit Supervisor of this issue in a timely manner. Issue 2 will remain open pending the licensees completion of a timeline of events and an inspector review of the station standards and implementation of those standards for this issue. Issue 3: Implementation of Operations Standing Order Upon Reaching a Low Lake Level Condition Without Performing a 10 CFR 50.59 and/or Generic Letter 8610 Review17. Upon discovery of the non-conforming and unanalyzed condition of the UHS, the licensee implemented an operations standing order that directed the nonsafety-related service water system, fire protection water system, and circulating water system to be secured following a cooling lake dike failure and low lake level of 590. This operations standing order augmented AOP BwOA ENV3, which did not direct any of these actions. In developing the subject standing order, the licensee did not perform a 10 CFR 50.59 evaluation and/or an associated review in accordance with Generic Letter 8610, Implementation of Fire Protection Requirements. At the end of the inspection period it was not clear if the licensees standing order process, or any other process, permitted this type of change without performing a 10 CFR 50.59 and/or associated Generic Letter 8610 evaluation. Additionally, it was not clear if the licensees temporary change was adequate (i.e. tripping both units, securing all circulating water and non-essential service water system pumps, and securing all running Fire Protection pumps just prior to reaching a low lake level of 590). Issue 3 will remain open pending the licensees completion of a timeline of events and additional inspector review. Issue 4: Safety Category II Structure, Systems and Component Interaction with the Ultimate Heat Sink The turbine building and a number of systems and components within the turbine building are designated as Safety Category II SSCs. The licensee defined Safety Category II SSCs as SSCs that were not designed to Safety Category I Standards. Specifically, Braidwood UFSAR Section 3.2.1.2 defined Safety Category II as follows: Those SSCs which are not designated as Safety Category I are designated as Safety Category II. This category has no public health or safety implication. Safety Category II structures, systems, and components are not specifically designed to remain functional in the event of the safe shutdown earthquake or other design-basis events (including tornado, probable maximum flood, operating basis earthquake, missile impact, or an accident internal to the plant). A reasonable margin of safety is, however, considered in the design as dictated by local requirements. Many Safety Category II items in Category I buildings are supported with seismically designed supports. These items and their supports are not Safety Category I or Seismic Category I as defined by Regulatory Guide 1.29. Structures and major components not listed in Table 3.2-1 as Safety Category I are Safety Category II. Safety Category II systems or portions of systems and components do not follow the requirements of Appendix B to 10 CFR 50. The quality assurance standards for these systems and components follow normal industrial standards and any other requirements deemed necessary by the Licensee. The licensee determined that a circulating water system line break and/or main condenser expansion joint rupture was not credible based on a review of postulated safe shutdown earthquake loads, and therefore a failure of this system following a design basis event such as a safe shutdown earthquake was not within the current licensing basis. The inspectors identified that a failure of the Safety Category II circulating water system could impact safety. For example the Braidwood cooling lake dike was also a Safety Category II structure. A failure of the cooling lake dike and establishment of the UHS18 level of 590 followed by a circulating water line break/expansion joint failure in the turbine building would result in a condition not currently evaluated (i.e., less useable UHS volume due to the displacement of a fraction of the UHS volume into the turbine building). At the end of the inspection period it was not clear how a Safety Category II SSC such as the circulating water system could be credited in a manner to not fail during a safe shutdown earthquake or other associated design basis event since, by definition, Safety Category II SSCs are not specifically designed to remain functional during these events. Additionally, the inspectors planned to review the Safety Category II Lake Screen House structure design to ensure that it could not adversely affect the intake in a manner that would prevent the UHS from performing its intended safety function. Issue 4 will remain open pending NRC review to ensure that the licensee is in compliance with their current licensing basis. (URI 05000456/201400301; 05000457/201400301, Issues That Could Adversely Affect the UHS)
05000237/FIN-2014003-012014Q2DresdenFailure to Take Appropriate Corrective Action When a Maintenance Rule Performance Goal for the Standby Coolant System was Not MetThe inspectors identified a finding of very low safety significance and non-cited violation of 10 CFR 50.65(a)(1), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to take corrective actions by performing an (a)(1) determination when the standby coolant supply system preventative maintenance (a)(2) demonstration was failed. Specifically, in November 2013, the standby coolant supply system exceeded its maintenance rule performance criteria when it experienced an additional maintenance preventable functional failure. The licensee failed to appropriately account for this failure in their Maintenance Rule Program and, as a result, the site failed to perform appropriate corrective action, by failing to perform an (a)(1) determination in accordance with Procedures ERAA310, Implementation of the Maintenance Rule, and ERAA-3101005, Maintenance RuleDispositioning Between (a)(1) and (a)(2), Revision 6. Corrective actions taken by the licensee to address this issue included performing a maintenance rule (a)(1) determination and placing the system into (a)(1) status. The issue was entered into the licensees corrective action program as issue report (IR) 1644740, NRC Questions D2R23 Performance of DOS 390001, and IR 1650033, MRule A1 Determination Needed for Missed MRFF Z391. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstones attribute of Equipment Performance and affected the cornerstones objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to identify a functional failure during a periodic (a)(2) demonstration purposed to provide reasonable assurance that the structures, systems, and components (SSCs), the standby coolant injection valve MO 23902, was capable of performing its intended function as specified in licensee emergency operating procedure DEOP 050003, Alternate Water Injection Systems, Revision 22. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems cornerstone. The inspectors answered Yes to the question Does the finding represent a loss of system and/or function and determined that a Detailed Risk Evaluation was required. The Senior Reactor Analysts (SRAs) evaluated the finding using the Dresden Standardized Plant Analysis Risk (SPAR) model version 8.18 and Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.0.9.0 software. The exposure time for the unavailability of the Standby Coolant Supply Valve 23902 was assumed to be the maximum value of one year. The result was a delta core damage frequency (CDF) of 6.6E8/yr. The dominant sequence was a medium loss of coolant accident initiating event with a failure of suppression pool cooling, a failure of power conversion system recovery, and a failure of late injection. Based on the Detailed Risk Evaluation, the SRAs determined that the finding was of very low safety significance (Green). This finding had a crosscutting aspect in Human Performance, Procedure Adherence, because the licensee failed to appropriately document the failure of a standby coolant supply valve in accordance with periodic test procedure DOS 390001, Standby Coolant Supply Functional Test. (H.8)
05000237/FIN-2014003-022014Q2DresdenLicensee-Identified ViolationTS 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. Station Procedure RPAA503, unconditional release survey method requires, in part, that materials have no detectable radioactivity for unconditional release from the site. Contrary to the above, on June 9, 2014, a contracted sewage truck transporting contaminated sewage from Unit 1 ejector pit to the licensees sewage treatment plant was unconditionally released after the truck was emptied. Specifically, the sewage truck was unconditionally released to the contractors facility without the proper authorization from RP Management. On the following day, the truck was returned to the licensees facility for survey and decontamination by the RP staff. The empty sewage truck contained traced amount of radioactivity of Co60 and tritium above minimal detectable activities. The licensee investigation determined that the empty sewage truck did not leak or cause contamination during transit on the public road. This event was entered into the licensees CAP as CR 01673475. The Radiation Protection Department immediately stopped work. Future transport of sewage between the licensed facility and the licensees sewage treatment plant will be escorted by radiation protection personnel to ensure that drivers follow licensee direction. The significance of the finding was determined by using Inspection Manual Chapter 0609, Appendix D, "Public Radiation Safety SDP." The issue is 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.
05000315/FIN-2014002-012014Q1CookDegraded Latch Prevents Closure of Fire DoorThe inspectors identified a finding of very low safety significance (Green) and associated non-citied violation of License Condition 2.C.4 for Unit 1, for the licensees failure to ensure that a fire door would be closed at the time of a fire. Specifically, fire door 1-DR-AUX387 was found with a degraded latch that prevented the door from closing. Donald C. Cook is required to comply with the National Fire Protection Association (NFPA) 80, 1970 which requires a closing device to ensure fire doors close and latch at the time of a fire. Contrary to this requirement, fire door 1-DR-AUX-387 would not close and latch because the latching mechanism for the inactive leaf had failed in a manner preventing the door from closing. As immediate corrective action, the licensee started hourly fire watches on the door and performed an interim repair to restore the door to a functional status. The licensee has entered the condition into the corrective action program as AR 2014-0802. The inspectors determined the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events (Fire) and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to maintain door 387 such that it could perform its required function as a 3 hour fire barrier. Using IMC 0609, Appendix F, the inspectors concluded that the finding was of very low safety significance (Green) because the fire loading was below the screening criteria of 120,000 btu/ft2. The inspectors concluded the finding included a cross-cutting aspect of H.5, Work Planning, in the area of human performance because the licensee did incorporate risk insights.
05000315/FIN-2014002-022014Q1CookFailure to Establish Procedures for Vacuum FillA finding and associated non-cited violation of technical specification (TS) 5.4.1, Procedures, self-revealed pertaining to establishing and maintaining procedures to ensure reliable indication of reactor vessel level during reduced RCS inventory and vacuum fill operations. Specifically, the licensee failed to include in procedures for vacuum fill methods to ensure the level detection system sensing lines were vacuum tight and to include provisions to normalize level indications. During the vacuum fill evolution for Unit 1, the licensee made 5 attempts to draw vacuum because of diverging level indications. The additional time spent in reduced inventory as well as the additional drain downs resulted in increased plant risk. As immediate corrective actions, the licensee corrected the leaking fitting, normalized level readings, and completed the vacuum fill evolution. The licensee has entered this issue into the corrective action program (CAP) as action request (AR) 2013-6907. The inspectors concluded the finding was more than minor because it adversely affected the Initiating Event cornerstone objective of limiting the likelihood of events that upset plant stability while shutdown. Specifically, the issue impacted the Procedure Quality attribute. Based on the screening criteria of IMC 0609, the inspectors and regional SRA concluded a phase 2 or 3 evaluation was needed. The Office of Nuclear Reactor Regulatory (NRR) performed a phase 3 assessment and estimated the conditional core damage probability at 5.9E-7. Therefore, the finding is of very low safety significance (Green). The finding included a cross-cutting aspect of H.9, Training, in the human performance area because the licensee lacked understanding of the precision level instruments.
05000316/FIN-2014002-032014Q1CookFailure to Establish Procedures for Vacuum FillA finding and associated non-citied violation of TS 5.4.1, Procedures, self-revealed pertaining to establishing and maintaining procedures to ensure reliable indication of reactor vessel level during reduced RCS inventory and vacuum fill operations. Specifically, the licensee failed to include in procedures for vacuum fill methods to ensure the level detection system sensing lines were vacuum tight. Although the licensee implemented some corrective actions prior to the scheduled vacuum fill evolution, the actions taken failed to prevent recurrence. During the vacuum fill evolution for Unit 2, the licensee made 2 attempts to draw vacuum because of diverging level indications. The additional time spent in reduced inventory as well as the additional drain down resulted in increased plant risk. As immediate corrective actions, the licensee corrected the leaking fitting, normalized level readings, and completed the vacuum fill evolution. The licensee has entered this issue into the CAP as AR 2013-18146. The inspectors concluded the finding was more than minor because it adversely affected the Initiating Event cornerstone objective of limiting the likelihood of events that upset plant stability while shutdown. Specifically, the issue impacted the Procedure Quality attribute. Based on the screening criteria of IMC 0609, the inspectors and regional SRA concluded a phase 2 or 3 evaluation was needed. Since the issue in Unit 2 was bounded by the phase 3 assessment performed for Unit 1, the inspectors and SRA concluded the finding was of very low safety significance, (Green). The finding included a cross-cutting aspect of P.3, Resolution, in the corrective action area because the licensee failed to implement corrective actions that prevented recurrence.
05000461/FIN-2013005-012013Q4ClintonFailure to Evaluate Failures of Individual Safe Shutdown Emergency Lighting Units for Maintenance Preventable Functional Failures and Repetitive Maintenance Preventable Functional FailuresThe inspectors identified an unresolved item (URI) regarding the licensees failure to evaluate failures of individual safe shutdown emergency lighting units for maintenance preventable functional failures and repetitive maintenance preventable functional failures in accordance with the licensees Maintenance Rule implementation procedures. During review of the licensees Maintenance Rule program, the inspectors identified that the licensee only used condition monitoring criteria to demonstrate effective maintenance of the plants safe shutdown lighting system. The licensee limited this system to fewer than 9 condition monitoring failures on 33 emergency lighting battery packs at 264 quarterly testing demands in a rolling 24 month period, which was equivalent to a 3 percent failure rate, to demonstrate effective maintenance of the system. The licensees definition of a condition monitoring failure for the battery packs is the failure of any emergency lighting battery pack to provide the required illumination. However, the licensee does not have any established reliability performance criteria to monitor for functional failures of this system. The licensees process for evaluating condition monitoring failures and reliability failures is different. For reliability failures, step 4.5.5 of ER-AA-310, Implementation of th Maintenance Rule, directs the licensee to evaluate whether a Maintenance Rule Functional Failure (MRFF) has occurred. If one has, the procedure directs the license to determine if the MRFF is a Maintenance Preventable Functional Failure (MPFF), an if so, determine if the MPFF is a Repetitive Maintenance Preventable Functional Failur (RMPFF). If a system has a RMPFF, the procedure directs the licensee to perform a evaluation to determine if maintenance is effective, and if not, the licensee is required t monitor the system under 10 CFR 50.65(a)(1). Conversely, condition monitoring failure are not evaluated for any of these criteria, and therefore do not get evaluated for RMPFFs which require (a)(1) evaluations. In NUMARC 93-01, Industry guideline for Monitoring the effectiveness of Maintenance at Nuclear Power Plants, Revision 4A, a MPFF is defined as, an unintended event or condition such that a SSC within the scope of the rule is not capable of performing its intended function and that should have been prevented by the performance of appropriate maintenance actions by the utility. The inspectors identified that there are areas of the plant that are required for safe shutdown activities that have only one or two safe shutdown emergency lighting units installed in the area. Under the licensees current Maintenance Rule program, each individual lighting unit failure would be considered a condition monitoring failure and would not be evaluated for MPFFs o RMPFFs. The inspectors need to determine whether a failure of one or more lights in a area such that the area no longer has functional installed safe shutdown lighting shoul be considered a Functional Failure subject to a MPFF review. The licensee informed the inspectors that the licensees condition monitoring criteria was developed based on a memo from the licensees corporate office (formerly ComEd) that was developed in 1999 in response to a NRC violation at Braidwood Station. The memo recommended that each station implement condition monitoring performance criteria because, the NRC pointed out that by measuring only reliability, Braidwood was allowing important functional failures to occur without requiring moving the system to (a)(1), and thus not repairing the attendant (a)(1) action plan. In response to this memo, the licensee and other Exelon (ComEd) stations implemented only Condition Monitoring for their emergency lighting. During the inspectors review, however, the inspectors identified that some other Exelon stations, including Braidwood, perform both reliability and condition monitoring. The inspectors need to determine whether it is appropriate for the licensee to perform condition monitoring without reliability monitoring on the Safe Shutdown Emergency Lighting system when step 6 of Attachment 1 to the licensees procedure ER-AA-310-1003, Maintenance Rule Performance Criteria Selection, states, Condition monitoring augments existing criteria ((Availability Performance Criteria), (Reliability Performance Criteria), Plant Level), where the existing monitoring may be insufficient. The issue is categorized as an URI pending licensees completion of a revised evaluation and the NRCs review of it (URI 05000461/20013005-01, Failure to Evaluate Failures of Individual Safe Shutdown Emergency Lighting Units for Maintenance Preventable Functional Failures and Repetitive Maintenance Preventable Functional Failures).
05000373/FIN-2013008-022013Q4LaSalleFailure to Ensure Battery Margin Maintained for Station BlackoutThe inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure that 5 percent battery margin would be maintained for station blackout (SBO). Specifically, the capacity value used for an acceptance criterion by the battery test procedure did not ensure that battery capacity was sufficient to maintain the required 5 percent remaining battery margin through the next surveillance test. This finding was entered into the licensees Corrective Action Program. The licensee planned to revise their battery test procedure to ensure the required 5 percent margin would be maintained. The finding was determined to be more than minor because, if left uncorrected, it would become a more significant safety concern. Specifically, the battery performance test procedure criteria would not ensure that the batteries retained sufficient margin to support SBO loads through the next scheduled surveillance test. The finding screened as of very low safety significance (Green) because it did not result in loss of operability or functionality of mitigating systems. Specifically, the most recent test results showed that the capacity of the battery was sufficient to supply the calculated load demands under SBO conditions at the time of this inspection. The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution, operating experience because the licensee did not properly evaluate relevant operating experience, i.e., NRC Information Notice 2013-05, Battery Expected Life and its Potential Impact on Surveillance Requirements.
05000373/FIN-2013008-012013Q4LaSallePump Test Instruments Were Not Maintained Within Required Accuracy LimitsThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to use instrumentation that met the data collection requirements of America Society of Mechanical Engineers Operation and Maintenance Code. Specifically, the licensee did not maintain the pressure instruments used during pump comprehensive in-service testing within the required Code accuracy limits. This finding was entered into the licensees Corrective Action Program to evaluate operability of the affected pumps and revise the calibration procedures of the affected instruments to reflect the Code accuracy requirements. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, inaccurate test instrumentation could reasonably result in an unrecognized degraded condition of safety equipment. In addition, recent test results required to be reanalyzed taking into account the actual as-left calibration data of the instruments to ensure the affected safety pumps remained operable. The finding screened as of very low safety significance (Green) because it did not result in loss of operability or functionality of mitigating systems. Specifically, the licensee reviewed recent as-found in-service test (IST) calibration data of the affected pumps, adjusted the as-found IST collected data using the actual calibration data, and reasonably determined the applicable IST acceptance criteria were met. In addition, the finding example associated with the spent fuel pool cooling did not result in actual adverse spent fuel pool conditions such as excessive temperatures, fuel clad damage, and inadequate water inventory. The inspectors did not identify a cross-cutting aspect associated with this finding because it did not reflect current performance due to the age of the performance deficiency.
05000266/FIN-2013005-032013Q4Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which requires, in part that activities affecting quality shall be accomplished in accordance with instructions, procedures and drawings. Specifically, licensee procedure PI-AA-205, Condition Evaluation and Corrective Action, establishes that conditions adverse to quality be entered into the corrective action program. Exceeding license power limits was a nonconformance with the operating license and therefore a condition adverse to quality. This issue was entered into the CAP as AR 0190339. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, failure to document conditions adverse to quality could lead to a more significant event since over power conditions would not be evaluated. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. Since the finding involved the mismanagement of reactivity by operator(s) (e.g., reactor power exceeding the licensed power limit), the inspectors determined that IMC 0609, Appendix M, needed to be used. Since the highest power level achieved was within the analyzed limits in the FSAR the inspectors screened the finding as having very low safety significance.
05000266/FIN-2013005-022013Q4Point BeachFailure to Follow Maintenance and Test Equipment ProcedureThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow procedure NP 8.7.1, Measurement and Test Equipment (M&TE). Specifically, the inspectors identified multiple examples where the licensee did not document the withdrawal and use of M&TE in either the M&TE usage log or its electronic equivalent. This issue was entered into the licensees corrective action program (CAP) as action request (AR) 01925171. The finding was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, without accurate M&TE usage logs the licensee may not evaluate all past surveillances affected by failed M&TE, potentially resulting in a failed TS surveillance going undetected. The inspectors determined that the finding was associated with the Mitigating Systems Cornerstone, because not evaluating the prior use of inaccurate M&TE could permit equipment required to mitigate the consequences of the accident to not perform its design and licensing basis functions when called upon. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. The inspectors concluded that this finding has a cross-cutting aspect in the area of human performance, decision making, because the licensee failed to effectively communicate the station expectations related to changes in responsibilities for implementing NP 8.7.1.
05000266/FIN-2013005-012013Q4Point BeachFailure to Provide Adequate Work InstructionsA self-revealed finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, occurred when a surveillance procedure was performed with several steps marked not applicable which resulted in Unit 1 power rising over the license limit. Specifically, when the Unit 1 turbine-driven auxiliary feedwater pump was operated as part of a post-maintenance test, the discharge isolation valves remained open which resulted in a small unplanned positive reactivity change. This issue was entered into the licensees CAP as AR 01920721. The inspectors determined that this finding was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure of the control room operators to respond promptly could have led to the final reactor power being higher than during this issue. The inspectors determined that the finding was associated with the Initiating Events Cornerstone, specifically the configuration control attribute of operating equipment lineup. The inspectors determined that the finding could be evaluated using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Initiating Events Screening Questions. The finding was determined to be of very low safety significance (Green) because the inadequate work instructions did not result in a reactor trip. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, work control, planning, because a human performance error was made during the planning process in an effort to reduce the work load during the test, and due to a cognitive error, the post-maintenance test was made inadequate. Specifically, steps were marked non-applicable that would have placed the pump discharge valves in their required position for the next portion of the surveillance test.
05000315/FIN-2013004-032013Q3CookImproper Setting in Digital Control SystemA self-revealed finding of very low safety significance (Green) occurred because the licensee failed to adjust a key parameter, (KWINIT), in the turbine digital control system after replacing and calibrating the turbine control system linear variable differential transformers. Vendor documents for the generator recommended an initial load of 2 to 5 percent of full load when the turbine generator is synchronized to the grid. For Cook Unit 1, this equates to 22 to 54 megawatts. However, the licensee did not adjust the KWINIT parameter, which is used to determine control valve position, after the turbine control system linear variable differential transformers were replaced and subsequently calibrated using a tighter tolerance than previously used. Consequently, when the turbine generator was synchronized to the grid the turbine control valves opened more than on previous synchronizations, which resulted in picking up excessive load. As a result, reactor cooling system (RCS) temperature momentarily lowered below the minimum temperature for criticality. As an immediate corrective action, the licensee stabilized the plant by taking manual control of the turbine generator. The licensee has entered the condition into the corrective action program (CAP) as AR 2013-7472. Using IMC 0612 the inspectors concluded that this issue was more than minor because it is associated with the equipment performance attribute in the Initiating Events Cornerstone and it adversely impacted the cornerstone objective of limiting the likelihood of events that upset plant stability. Using IMC 0609, Appendix A, Exhibit 1, the inspectors concluded the finding was of very low safety significance (Green) because it did not cause both a reactor trip and a loss of mitigating equipment. The inspectors concluded the finding had an aspect in the Work Control component of the Human Performance cross-cutting area because the licensee did not coordinate work activities to address the impact of changes to work activities on plant performance.
05000315/FIN-2013004-042013Q3CookReactor Trip Due to Improper Control Valve SetpointA self-revealed finding of very low safety significance occurred because the licensee failed to program the automatic controller for the condensate heater condensate bypass control valve, 2-CRV-224, with the correct setpoint. Specifically, the automatic controller (2-RU-2) setpoint was not set at the required 240 psig because licensee personnel incorrectly interpreted information in SD-ENG-05400, System Description Condensate System. Consequently, an incorrect set point of 188 psig was incorporated in procedure 2-OHP-4021-001-006, Power Escalation, which was used to program the automatic controller. As a result, 2-CRV-224 did not open as designed to mitigate the lowering main feedwater pump suction pressure, which resulted in the west main feedwater pump tripping on low suction pressure and a subsequent manual reactor trip. For corrective actions, the licensee programmed the correct setpoint into the automatic controller; revised the associated procedures to ensure setpoint changes are accurately incorporated and reviewed prior to implementation; developed plans to communicate lessons learned to the site; and entered the condition into the CAP. Using IMC 0612 the inspectors determined that this issue was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, failure to set the 2-CRV-224 automatic controller to the design setpoint of 240 psig resulted in the subsequent loss of the west main feedwater pump during a feedwater heater level transient, which caused steam generator water levels to lower and required the operators to manually trip the reactor. The inspectors determined the finding was of very low safety significance (Green) using Exhibit 1 of IMC 0609, Appendix A, because the finding did not cause both a reactor trip and a loss of mitigating equipment. The inspectors concluded that this finding was associated with an aspect in the Resources component of the Human Performance cross-cutting area. Specifically, the procedure used to program the automatic controller for 2-CRV-224 was not accurate in that it did not contain the correct design setpoint.
05000315/FIN-2013004-022013Q3CookFaulted 4KV Qualified Off-site CircuitA finding of very low safety significance was self-revealed on April 24, 2013, because the licensee failed to comply with requirements contained in procedure PMI-7030, Corrective Action Program, prior to restoring power to the Unit 1 reserve auxiliary transformer CD-101. Specifically, following multiple trips of supply breaker 12 CD, the licensee failed to correct an issue, defined as a condition adverse to quality in their corrective action program, prior to restoring power to the transformer on April 21. This ultimately led to the supply breaker to the Unit 1 and 2 reserve auxiliary transformers opening due to a faulted cable. No violations of NRC requirements were identified for this issue since the degraded cable was on a non-safety related portion of the electrical system. The licensee entered the issue into the corrective action program as AR 2013-6194. The corrective actions for this issue included replacing the faulted cables and testing the unaffected cables. Using IMC 0612, the inspectors concluded that the issue was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and it adversely impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded insulation failed causing a loss of the qualified circuit; a condition which lessened the likelihood of its availability for some events. Using IMC 0609, Appendix A, Section 6, a detailed risk evaluation, assuming inoperability of four days, determined the delta Core Damage Frequency was less than 1E-6; therefore the finding screens as very low safety significance (Green). The inspectors concluded this finding was associated with an aspect in Operating Experience component of the Problem Identification and Resolution cross-cutting area because the licensee did not implement and institutionalize operating experience information from the Electric Power Research Institute (EPRI) and Institute of Electrical and Electronics Engineers (IEEE) into processes and procedures.
05000346/FIN-2013002-032013Q1Davis BesseLicensee-Identified ViolationThe requirements of the NRC Maintenance Rule, 10 CFR 50.65(a)(4) state, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to this requirement, on January 28, 2013, the licensees probabilistic risk assessment did not accurately reflect the increase in online probabilistic risk associated with startup transformer X01 being unavailable during planned maintenance. Specifically, while removing the 345 kV Lemoyne transmission line from service for planned maintenance, startup transformer X01 unavailability during switchyard manipulations was inadvertently omitted from the stations risk assessment. Re-performance of the risk assessment after the switchyard manipulations were already completed indicated that an elevated yellow risk category that required additional work controls had actually existed for approximately 16 minutes during the switchyard manipulations. Licensee personnel initiated CR 2013-01309 to document the issue. The objective of the Mitigating Systems Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A key attribute of this objective is equipment performance and availability. Using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the inspectors determined that the violation was of very low safety significance since the incremental core damage probability deficit calculated for the issue was less than 1E-6.
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