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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000237/FIN-2017004-01 | 2017Q4 | Dresden | Failure to Follow Procedure,Results in Non-Functional Fire Door | The inspectors identified a finding of very-low safety significance and associated NCV of Technical Specification 5.4.1.c for the licensees failure to implement the established Fire Protection Program procedures which ensure Fire Barrier Integrity. Specifically, the licensee ran an electrical cable through the doorway of an automatically closing fire door. This was contrary to Procedure DFPP 417501, which requires in part that fire doors must not be blocked open by props or any other material in its closing path. The licensee took immediate actions to restore the fire door, by removing the obstruction and entered the issue into their Corrective Action Program (CAP). The inspectors determined that the performance deficiency was more-than-minor because it affected the Mitigating Systems cornerstone objective since the electrical cable could have prevented the fire door from performing its function. The finding was of very-low safety significance per Task 1.4.3A of IMC 0609, Appendix F. Specifically, the total combustible loading on both sides of the affected fire door was representative of a fire duration less than 1.5 hours. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, associated with the Training component, because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee believed the performance deficiency was caused by the one of the new temporary contractors brought onto the site to work in support of the D2R25 refueling outage. (H.9) |
| 05000254/FIN-2012010-01 | 2012Q4 | Quad Cities | Concerns with Meeting One Time Visual Inspections IAW ASME Section Xi Requirements | During the Phase I inspection of Commitment Items (IR 05000265/2012008; ML12129A226), the inspectors identified the licensee was not performing visual examinations in accordance with the ASME code. The inspectors identified this through direct observation, review of the work orders used to perform these examinations and interviews with the appropriate plant personnel. The commitment to perform visual examinations in accordance with the ASME Section XI code applies to the following programs: B.1.23 One Time Inspection Ventilation Systems, Compressed Gas Systems, SBLC Chemistry Program; B.1.24 Selective Leaching; B.2.8 Periodic Inspection of Plant Heating Steam; and B.2.9 Periodic Inspection of Components Subject to Moist Air Environments. The Commitment Item specifically states, The inspection will be performed in accordance with ASME Code requirements. Certified NDE examiners will conduct a VT-3 visual inspection (VT-1 for the Selective Leaching program, these inspections will consist of visual inspection consistent with ASME Section XI VT-1 visual inspection requirements.). For visual examinations performed in accordance with the code, ASME section XI IWA 2210 requires visual examination be performed in accordance with ASME Section V Article 9. This section requires a procedure be used when performing visual examinations that include instructions on how the visual inspection is to be performed as well as illumination and instruments to be used. ASME Section XI has specific requirements associated with the different levels of visual examinations (VT-1 or VT-3) including distance, illumination, and character card resolution requirements. Contrary to this the licensee did not perform the VT-1 or VT-3 examinations in accordance with written procedures which documented the requirements for lighting, distance and other key parameters as well as acceptance criteria. Failure to do so did not ensure an ASME qualified process is used and clear requirements are provided and met. The licensee submitted a Commitment Item change to clarify their position, indicating the intent was to have a qualified person performing the inspection, not to perform the examination in accordance with the requirements of the code. The licensee also indicated ASME requirements were verbally communicated to the examiners during their pre-job brief therefore, ensuring compliance with the code. In a letter dated May 18, 2012 (ML# 12173A423) the licensee requested a Commitment Item change. NRR staff subsequently issued a Request for Additional Information (ML 12291A831) and is awaiting the licensees response in order to make a decision regarding these Commitment Items. Therefore, this issue is considered unresolved pending resolution of the above issue. |
| 05000254/FIN-2012010-02 | 2012Q4 | Quad Cities | Questions Regarding Aging Management Inspections on the 16\\\' Diameter Discharge Pipes | As part of the review of licensee Commitment Item 31, the inspectors identified an unresolved item (URI) related to a 16 foot diameter discharge piping. Specifically, the licensee does not physically inspect the piping for aging effects so that the intended functions of this component will be maintained during the period of extended operation. As part of the review of Commitment Item 31 associated with the Water Control Structures inspection program, the inspectors reviewed the components that are within the scope of this program and how effects of aging are managed. NUREG 1796; Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2; dated 2004, Section 3.5.2.3.4 RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants described the scope of the program and how the effects of aging are managed. Page 3-426 of NUREG 1796 identified the 16 foot diameter discharge piping as an in-scope component. The 16 foot diameter discharge piping provides an Ultimate Heat Sink (UHS) function in that during a Lock and Dam No. 14 failure the discharge piping provides suction source for portable pumps to provide cooling water flow to the Residual Heat Removal Service Water (RHRSW) pumps and Diesel Generator Cooling Water (DGCW) Pumps. The licensee, however, does not physically inspect the discharge piping. The licensee provided aging management of the discharge piping by crediting the performance of a one-time inspection of a 96 inch diameter ice melt line. The ice melt line prevents freezing of the river water entering the plant intake and also performs a support function for the UHS. During this inspection, the licensee was unable to locate where the Office of Nuclear Reactor Regulation (NRR) had reviewed and approved this method of aging management of the discharge piping. In response to this concern, the licensee initiated Condition Report 01434957, (LR) Scoping Review Required for UHS Discharge Piping, dated November 2, 2012. Therefore, this issue is considered unresolved pending additional review of the information provided by the licensee and consultation with NRR to determine the appropriate aging management of the 16 foot diameter discharge piping. |
| 05000255/FIN-2009006-01 | 2009Q4 | Palisades | Agastat Time Delay Relays Design, Testing and Configuration Control Issues | A finding of very low safety-significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, was identified by the inspectors for the licensees failure to translate the design bases into design drawings, procedures and appropriate test instructions. Specifically, the design basis requirements for Agastat Time Delay Relays (TDR) settings, as well as vendor tolerances, were not accurately reflected in the design drawings, procedures and test instructions for numerous TDR calibrations. This issue was entered into the licensees corrective action program. The inspectors determined that the finding was more than minor because it was associated with the Mitigating System Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to ensure that safety-related TDRs would operate, within the design specified setpoints and allowed tolerances, could lead to the inability of safety-related systems and components to respond to design basis events (e.g., during load sequencing onto the EDG). The finding screened as being of very low safety-significance because the finding was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensees subsequent evaluation of the TDRs tolerances showed that available margin remained for satisfactory completion of the required safety function. This finding has an associated cross-cutting aspect in the area of problem identification and resolution because the licensee did not incorporate operating experience (OE)information, including internally generated lessons learned, to support plant safety. Specifically, even though the licensee was aware of the potential inadequacies of the Agastat TDR setpoints through internal OE, the licensee failed to adequately respond to the OE by implementing appropriate changes to station processes, procedures, equipment, and training program. P.2.(b) |
| 05000255/FIN-2009006-02 | 2009Q4 | Palisades | Failure to Translate the Design Basis for the CV-11 Control Room HVAC Chiller Into Specifications and Drawings | A finding of very low safety-significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to translate and incorporate design basis criteria that ensured the functionality of TDRs for the CR HVAC chillers into design drawings, procedures and work instructions for implementation. Specifically, even though the licensee reduced the replacement interval frequency for the chiller mounted TDRs due to high vibration levels to ensure functionality, and then initiated Work Orders (WOs) to perform this replacement, one WO was closed without replacing the TDRs as intended, and the second WO was not approved for implementation. This issue was entered into the licensees corrective action program. The inspectors determined that the finding was more than minor because this failure to establish measures to translate and incorporate design basis criteria to ensure the functionality of TDRs for the CR HVAC chillers could lead to the inability of the chillers to respond to design basis events. Specifically, the finding screened as of very low safety-significance (Green) because the finding did not represent loss of system safety function. This finding has an associated cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate problems such as that the resolution addresses causes and extent of condition, as necessary. This includes properly evaluating for operability conditions adverse to quality. P.1(c) |
| 05000255/FIN-2011009-01 | 2011Q3 | Palisades | Failure to Adequately Evaluate the Enclosure Installed Over the 1F/1G Buses | The inspectors identified a finding of very low safety significance involving the licensees failure to adequately evaluate the enclosure installed over the 1F/1G Buses to be in compliance with all applicable requirements. Specifically, the licensee did not ensure that the new enclosure would not affect start-up transformer 1-2 during a design basis wind event. There were no violations of NRC regulations identified. This finding was entered into the licensees corrective action program, which resulted in replacing inadequate eye-bolts. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of transient initiator (loss of offsite power) and affected the cornerstone objective to limit the likelihood of those events that upset plant stability. Specifically, there was reasonable doubt as to whether the enclosure could have withstood a design wind event, which would have increased the probability that severe weather could have affected the ability of startup transformer 1-2 to provide offsite power. The finding screened as very low safety significance (Green) because the transient initiator would not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding has a cross-cutting aspect in human performance because the licensee did not ensure reviews of safety significant decisions to verify the validity of the underlying assumptions or identify possible unintended consequences. Specifically, the licensees design reviews for the 1F/1G Bus enclosure modification did not address the potential impact on start-up transformer 1-2 if the enclosure failed during a design basis wind event. |
| 05000255/FIN-2011009-02 | 2011Q3 | Palisades | GL 2008-01 Design Reviews Did Not Adequately Assess the Potential to Accumulate Voids Within Piping Systems | The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to adequately review the design of emergency core cooling and containment spray systems with respect to the potential to accumulate voids. Specifically, the design reviews did not consider system interactions, evaluate the acceptability of locations believed to be inaccessible for periodic monitoring, and ensure the validity of the assumption that some high point vents were periodically used to ensure that some locations were full of water when excluding them from periodic monitoring. This finding was entered into the licensees corrective action program. The performance deficiency was associated with Mitigating System Cornerstone and determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding screened as of very low safety significance (Green) because the finding involved a design or qualification deficiency that did not result in a loss of operability. Specifically, based on a historical review of recent maintenance activities, current process parameters, and, in some locations, ultrasonic examinations, the licensees operability evaluation concluded there were no adverse voids at these locations. This finding had a cross-cutting aspect in the area of human performance because the licensee did not ensure supervisory oversight of work activities associated with the Generic Letter 2008-01 design reviews such that nuclear safety is supported. Specifically, oversight did not ensure that the contractors design reviews considered plant specific information such as system interactions and at-power operations. |
| 05000255/FIN-2011009-03 | 2011Q3 | Palisades | Procedures Were Not Appropriate To Address Gas Accumulation Issues | The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to establish appropriate procedures for managing gas accumulation issues. Specifically, three examples were identified as follows: (1) Procedure ESSO-10 did not ensure that identified voids would be successfully removed by flushing; (2) Procedure SOP-3 did not specify a maximum flowrate which analyzed net positive suction head and potential air entrainment due to vortexing during reduced inventory operations when in shutdown cooling; and (3) Procedure SOP-3 did not contain instructions to vent the steam that could form at the low pressure safety injection discharge piping following a shutdown loss of cooling accident prior to system initiation. This finding was entered into the licensees corrective action program. The performance deficiency was associated with the Initiating Events and Mitigating System Cornerstones, and determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding screened as of very low safety significance (Green) because: (1) Procedure ESSO-10 was a deficiency confirmed not to result in loss of operability in that a review of recent periodic gas monitoring results determined that the affected locations were full of water; (2) Procedure SOP-3 associated with reduced inventory operations did not meet any of the criteria that required a Phase II or III analysis in that it did not rise to the level that there was an increase in the likelihood of a loss of shutdown cooling; and (3) Procedure SOP-3 associated with the steam void formation did not require a quantitative assessment because it met each item for the core heat removal, inventory control, power availability, containment control, and reactivity guidelines. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate relevant external operating experience. Specifically, the licensees evaluation of gas related issues in response to Generic Letter 2008-01 was deficient in that, the licensee did not identify two potential gas sources, vortexing during reduced inventory and flashing following a shutdown loss of coolant accident, and did not address the minimum flowrate required to remove gas in piping when flushing. |
| 05000255/FIN-2011009-04 | 2011Q3 | Palisades | Void Size Acceptance Criteria is Non-Conservative | The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to develop conservative void size acceptance criteria. Specifically, the void size acceptance criteria was based on an incorrect safety injection and refueling water base tank elevation and a 10 percent degradation of the design rated flowrates of the pumps. When the correct base tank elevation and lower allowable pump flowrates were considered, the void acceptance criteria were non-conservative. This finding was entered into the licensees corrective action program. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of equipment performance 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 the finding was a design or qualification deficiency confirmed not to result in loss of operability. Specifically, a review of recent periodic gas monitoring results determined that no voids were present at the suction side of the affected pumps. This finding had a cross-cutting aspect in the area of human performance because the licensee did not ensure supervisory oversight of work activities associated with actions related to Generic Letter 2008-01 such that nuclear safety is supported. Specifically, oversight did not ensure that the contractors development of void acceptance criteria relied on limiting design values. |
| 05000255/FIN-2011014-09 | 2011Q4 | Palisades | Potential Loss of Preferred AC Sources in Harsh Environment | On September 25, 2011, a fault occurred on Panel D11-2, which resulted in reactor and turbine trip, and de-energiziation of Bus D-10. Breaker 72-37, which supplied DC power to Inverter D-06, was found tripped. According to the manufacturer, the inverters were capable of reverse-feeding DC short circuits for short durations and this could have caused Breaker 72-37 to trip. This was possible because the inverter had four 7700 microFarad parallel capacitors on the DC side of the inverter. During a DC short circuit, the capacitors would rapidly discharge and feed the fault. Breaker 72-37 had a rating of 100 Amps for the thermal setting and 700 Amps for the magnetic setting. According to the manufacturer an approximation for an inverter DC fault current contribution was about 1100 Amps per capacitor; therefore, this was approximately a total of 4400 Amps for Inverter D-06. This exceeded the magnetic rating of the breaker and explained why the breaker tripped during the fault condition. The PCP motor DC oil lift Pumps P-81A and P-81C were nonsafety-related loads, which received power from Bus D-10 via safety-related Breakers 72-13 and 72-14, respectively. The PCP motor DC oil lift Pumps P-81B and P-81D were also nonsafety-related loads that received power from D-20 via safety-related Breakers 72-23 and 72-24, respectively. The cabling for these loads was not environmentally qualified and was routed through containment, which could be susceptible to failure due to a harsh environment. The inspectors were concerned that if all four nonsafety-related cables for these pumps faulted due to a harsh environment during a design basis event, this could result in the loss of all preferred AC power busses due to the internal capacitors contributing to the fault as seen by each DC bus. However, without further analysis of the design and licensing basis, the inspectors could not determine if a postulated harsh environment affecting all four cables during a design basis event was a credible event. Therefore, the inspectors initial conclusion, based on the available information was that this event may not be credible; however, further analysis was required. In addition, all four PCP motor DC oil lift pump breakers were opened as one of the compensatory measures for the operability of the 125-Volt DC system. Therefore, this is not a current safety concern. Title 10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, Section b(2), requires nonsafety-related electric equipment to be environmentally qualified if the failure of the nonsafety-related electric equipment under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs (b)(1)(i) (A) through (C) of paragraph (b)(1) of this section by the safety-related equipment. The inspectors were concerned that the cables associated with the PCP motor DC oil lift Pumps P-81A, P-81B, P-81C, and P-81D were not evaluated for the effect on the safety-related equipment specifically the safety-related inverters and their associated preferred AC sources. The licensee entered this issue into their CAP as CR-PLP-2011-6210. This issue is a URI pending the licensee evaluation, and the inspectors review of the licensees design and licensing basis, and evaluation to determine if a performance deficiency existed (URI 05000255/2011014-09; Potential Loss of Preferred AC Sources in Harsh Environment). |
| 05000255/FIN-2013003-01 | 2013Q2 | Palisades | Inadequate Control of Welding at the F-East Nozzle Reinforcement Plate | The inspectors identified a finding of very low safety significance and an associated non-citied violation of 10 CFR 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to perform adequate pre-weld cleaning and control the welding process in a manner that ensured proper weld fusion of the F-East nozzle reinforcement plate weld joint within the safety injection refueling water storage tank (SIRWT). Consequently, this weld failed in service causing leakage from the SIRWT. The licensee subsequently replaced the floor of the SIRWT and included instructions in the floor replacement work order that required pre-weld cleaning with acetone or other approved solvents. The licensee entered the issue in their corrective action program (CAP) as CR- PLP-2013-03185. The finding was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the inspectors answered yes to the More-than-Minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern ? Absent NRC identification, the failure to adequately clean aluminum prior to welding and adequately control the repair welding techniques may have been repeated during future repairs to the SIRWT and resulted in lack of fusion type weld defects/cracks returned to service. Unstable cracks could propagate and create failure of the SIRWT pressure boundary resulting in loss of inventory and increase the risk for insufficient core cooling for post Loss-of-Coolant Accident (LOCA) conditions. Therefore, this finding adversely affected the mitigating systems cornerstone attribute of equipment performance (reliability). The inspectors determined this finding was of very low safety significance (Green) based on answering no to the questions in Part A of Exhibit 2, Mitigating Systems Screening Questions, in IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the small amount of leakage from the SIRWT weld leak did not result in loss of a mitigating system function. Therefore, this finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance for the resources component because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety was supported. |
| 05000255/FIN-2013003-02 | 2013Q2 | Palisades | Failure to Follow Corrective Action Process for Service Water Leaks | A finding of very low safety significance with an associated non-citied violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to adhere to the requirements of the sites corrective action process. Specifically, the station failed to complete corrective actions to address cavitation-induced erosion of service water system components, which resulted in additional through-wall leaks and other adverse conditions in that safety-related system. Since 1993, this phenomenon caused several through wall leaks and the failure of a valve, which isolated normal service water flow to a component cooling water heat exchanger. Corrective actions to replace valves susceptible to this type of erosion were not implemented, and actions to utilize more effective non-destructive examination (NDE) techniques to assess piping or development of pre-emptive repair/replacement strategies were not performed, resulting in further leaks from the service water system. The current corrective action process procedure, EN-LI-102, states that corrective actions are determined, implemented, and adequate to resolve conditions. The licensee entered the issue in their corrective action program (CAP) as CR- PLP-2013-05813. The issue was determined to be greater than minor in accordance with IMC 0609 Appendix B, Issue Screening, issue date September 7, 2012, because it adversely affected the equipment performance attribute of the mitigating systems cornerstone whose objective is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a through wall leak can challenge the integrity of the piping and system function. The inspectors concluded the finding was of very low safety significance (Green) utilizing IMC 0609, Significance Determination Process, issue date June 2, 2011. Specifically, in Attachment 4, issue date June 19, 2012, utilizing Exhibit 2 of Appendix A, all questions in Section A were answered no since the leaks did not result in a loss of safety function. The finding had an associated cross-cutting aspect in the area of problem identification and resolution for the operating experience component. Specifically, the licensee did not implement and institutionalize operating experience through changes to station processes and procedures. |
| 05000255/FIN-2018003-01 | 2018Q3 | Palisades | Wire Not Landed on Safety Injection Initiation Relay Circuit | The inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018 |
| 05000263/FIN-2012007-01 | 2012Q2 | Monticello | Inadequate Station Battery Capacity Test Procedure | The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawing, for the licensees failure to ensure the bases for sizing of the 250 Vdc safety-related batteries was incorporated into the battery capacity test procedure. Specifically, the licensee did not incorporate the commitment to replace the 250 Vdc batteries when battery capacity drops more than 10 percent of rated capacity from its capacity on the previous test. The licensee verified current operability and entered this issue into their corrective action process as Action Requests 01333346 and 01334083. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality, and affected the cornerstone objective to ensure the availability, reliability, and capability of 250 Vdc batteries that are essential for the proper functioning of systems that respond to initiating events to prevent undesirable consequences. The finding screened as having very low safety significance because it did not represent an actual loss of safety function. The inspectors determined there was no cross-cutting aspect associated with this finding because it was not reflective of licensees current performance due to the age of the issue. |
| 05000263/FIN-2012007-02 | 2012Q2 | Monticello | Failure to Analyze Voltage Requirements for Operability of Non-Motor Loads and 120 Vac Instrument Panels | The inspectors identified a finding of very low safety significance(Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the operability of safety-related 120 Vac instrument bus loads and 480 Vac non-motor loads under degraded voltage conditions. The inspectors determined several loads and panels did not have the minimum required voltage specified in station procedures, USAR or the manufacturers specifications. The licensee entered this issue into their corrective action program as Action Requests 01332429, 01334571, and 01334562. The licensee performed testing and analyses, and implemented operating restrictions to obtain reasonable assurance of operability. The finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether 480 Vac non-motor loads and equipment supplied by 120 Vac instrument buses had adequate voltage to operate during degraded voltage conditions. The finding was considered to be of very low safety significance (Green) since this was a design deficiency confirmed not to have resulted in a loss of operability or functionality because of licensees compensatory actions. The inspectors determined the finding had a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to perform a thorough extent of condition review and an assessment of reasonable assurance of operability when similar issues were identified in the 2009 NRC CDBI and a self-assessment performed in 2011. |
| 05000263/FIN-2012007-03 | 2012Q2 | Monticello | Failure to Maintain the Degraded Voltage Function Time Delay Design | The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III for the licensees failure to translate the actual time delay of the degraded voltage relay scheme under all circumstances into the station procedures and Technical Specifications. Specifically, a modification which introduced a five second time delay to the degraded voltage scheme resulted in inconsistencies in Technical Specification Table 3.3.8.1-1 and functionality of the degraded voltage relay scheme when the safety buses are aligned to Transformer 1AR. The licensee entered this issue into their corrective action program as Action Request 01334146, and removed Transformer 1AR from service to match the design with the Technical Specifications. The finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether the degraded voltage scheme would perform as required by Technical Specifications during design basis conditions. The finding was considered to be of very low safety significance (Green) since the total degraded voltage protection scheme time delay of 15 seconds was commensurate with the current accident analysis in the Updated Safety Analysis Report (USAR). The inspectors determined there was no cross-cutting aspect associated with this finding because it was not reflective of licensees current performance due to the age of the issue. |
| 05000263/FIN-2012007-04 | 2012Q2 | Monticello | Failure to Analyze Effect of Degraded Voltage on proper operation of Thermal Overload Relays | The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly analyze thermal overload relays (TOLs) for Motor Operated Valves (MOVs) and continuous duty motors under degraded voltage conditions. The licensee entered this issue into their corrective action program as Action Requests 01332373, 01332567, and 01334042 and initiated modifications to ensure TOLs would perform as required. The finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether safety-related MOVs and continuous duty motors would continue to operate without tripping during degraded voltage conditions. The finding was considered to be of very low safety significance (Green) since this was a design deficiency confirmed not to have resulted in a loss of operability or functionality because of licensees compensatory actions. The inspectors determined there was no cross-cutting aspect associated with this finding because it was not reflective of licensees current performance due to the age of the issue. |
| 05000263/FIN-2012007-05 | 2012Q2 | Monticello | Inadequate Procedures for Alignment of 120 Vac Instrument Buses | The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure single failure criterion is not violated by the procedure for simultaneously aligning both divisions of 120 Vac uninterruptible instrument power to their alternate, non-battery backed power sources. The licensee entered this issue into their corrective action program as Action Request 01334510 and implemented restrictions to prevent simultaneous alignment of both Divisions 1 and 2 instrument buses to their alternate sources, pending resolution. The finding was more than minor because it affected the Mitigating Systems Cornerstone attribute of Procedure Quality, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the uninterruptible instrument power system and the systems supported by it would not be able to perform their required functions during events such as a loss of power or station blackout. The finding was considered to be of very low safety significance (Green) because it did not represent an actual loss of safety function since the licensee had not placed the equipment in this configuration. The inspectors determined there was no cross-cutting aspect associated with this finding because it was not reflective of licensees current performance due to the age of the issue. |
| 05000263/FIN-2012007-06 | 2012Q2 | Monticello | Failure to Analyze Effect of System and Transient Harmonics on Proper Operation of Degraded Voltage Relays | The Monticello degraded voltage protection scheme features three ITE Type 27N relays for each 4.16kV safety bus, arranged in a two out of three tripping scheme. BBC Instruction Bulletin 7.4.1.7-7 states, the relay employs a peak voltage detector, and harmonic distortion on the AC waveform can have a noticeable effect on the relay operating point and the measuring instruments used to calibrate the relay. The bulletin also notes that the relay is available with an internal harmonic filter for applications where waveform distortion is a factor. The inspectors noted Calculation 92-220, Instrument Set-Point Calculation 4.16KV Degraded Voltage, identified the relays as a model not equipped with harmonic filters, but did not address the basis for excluding harmonic distortion as a factor affecting relay accuracy. The inspectors were concerned that persistent harmonics on the 4.16kV system could cause the relays to fail to actuate at the set point specified in Technical Specifications, and that transient harmonics, caused by switching operations, could cause the relays to spuriously reset during the time delay that occurs during an actual degraded voltage condition concurrent with a design basis accident. Persistent harmonics can be produced by factors external to the nuclear site or by internal phenomena. A typical internal source of harmonics at nuclear power plants is defects in rotating equipment. Persistent harmonics could cause dropout set point shift, and mask an actual degraded voltage condition. Transient harmonics could cause the relays to spuriously reset during an actual degraded voltage event, thereby delaying the protective function beyond the nominal 10 seconds (9.2 seconds Allowable Value) stipulated in Technical Specifications Table 3.3.8.1-1 and the 15 seconds assumed in the accident analysis. The relay is susceptible to this type of mal-operation because it features an instantaneous voltage sensor that could reset in less than two cycles in the presence of harmonics, thereby reinitiating the relays internal timer. Operating Experience available to the inspectors from another nuclear station indicated the transient voltages that occur during the operation of medium voltage circuit breakers could cause the relays to spuriously reset during an actual degraded voltage condition. The inspectors were specifically concerned that if a degraded voltage condition occurred concurrently with an accident, the automatic switching operations that occur on the 4.16kV electrical system following the onset of an accident could cause the relay to spuriously reset, thereby delaying the transfer to a reliable source of power beyond the required time. In response to the inspectors concerns, the licensee provided information regarding condition monitoring of large motors consisting of periodic measurement and analysis of motor bearing vibration from which various defects that may produce harmonics could be identified. However, the inspectors noted, there was no guidance in design documents that linked the presence of harmonics produced by motors during testing to mal-operation of the degraded voltage scheme. Although the motor vibration data taken by the licensee may be used to predict the presence of harmonics to some extent, the reason or the intent of the tests was not to evaluate or monitor harmonics. Also, the test data was not interpreted to evaluate and document the presence of harmonics. The inspectors further noted during normal bus voltage conditions, i.e., when voltage is above the degraded voltage relay reset set point, harmonics would shift system peak voltage away from the degraded voltage relay operating set point rather than closer to it, and so the presence of harmful harmonics would not self-reveal by spurious actuations. Also, there was no instrumentation or testing that demonstrated transient harmonic would not be presented during switching operations. The licensee has entered this item into their corrective action program as AR 01331618. This issue is unresolved pending consultation with NRC Headquarters for clarification of equipment qualification requirements of degraded voltage relays to withstand the effects of harmonics. |
| 05000263/FIN-2014403-01 | 2014Q3 | Monticello | Security | |
| 05000263/FIN-2014403-02 | 2014Q3 | Monticello | Security | |
| 05000263/FIN-2014403-03 | 2014Q3 | Monticello | Licensee-Identified Violation | |
| 05000266/FIN-2009005-01 | 2009Q4 | Point Beach | Failure to Meet GL 89-13 Program for Mussel Control | The inspectors identified a finding of very low safety significance for the failure to meet a commitment made in the Generic Letter (GL) 89-13 program. Specifically, the program states that biocide treatments at Point Beach are performed at least annually and are directly applied to the service water system for mussel control and eradication to prevent fouling of safety-related heat exchangers. However, the 2008 biocide treatment for mussel control was deferred until 2009. After the treatment in 2009, greater than expected tube blockage and reduced flow to safety-related heat exchangers due to mussels was identified. In response, the licensee adjusted flow through the affected heat exchangers and opened and cleaned the heat exchangers to remove mussels that caused the tube blockage. The licensee took corrective actions to ensure that future annual biocide treatments would be conducted annually. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The finding was determined to be of very low safety significance because the issue did not result in the actual loss of a safety function. This finding did not involve a violation of NRC regulatory requirements. The inspectors determined this performance deficiency was not indicative of current performance; therefore, no cross-cutting aspect was identified. (Section 1R12.1 |
| 05000266/FIN-2009005-02 | 2009Q4 | Point Beach | Failure to Ensure Adequate Control of Foreign Material in Safet-Related Systems | A self-revealed finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure to ensure adequate control of foreign material in accordance with the requirements of procedure NP 8.4.10, Exclusion of Foreign Material from Plant Components and Systems. Specifically, on October 17, 2009, foreign material was discovered inside the 2SI-897B valve after the valve failed to properly stroke during the performance of procedure IT-215, SI Valves - Cold Shutdown. The licensee took prompt corrective actions to repair the valve and perform an extent-of-condition review. Additionally, upon entering the issue into its corrective action program, the licensee performed a causal evaluation to determine any additional corrective actions. The finding was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of human performance and adversely affected the associated cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, due to the interference caused by the foreign material inside the 2SI-897B valve, the valve would not have been able to perform its safety function to close during the initiation of the post-LOCA (loss of coolant accident) sump-recirculation phase of safety injection. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, dated January 10, 2008. The finding was determined to be of very low safety significance because the issue did not represent a degradation of the radiological barrier function provided for the control room, the auxiliary building, or the spent fuel pool; represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, containment isolation system (logic and instrumentation)), and heat removal components; or involve an actual reduction in function of hydrogen ignitors in the reactor containment. No cross-cutting aspect was identified because the foreign material was determined to have been introduced into the system in the past and was not considered indicative of current performance. (Section 1R15.1 |
| 05000266/FIN-2009005-03 | 2009Q4 | Point Beach | Failure to Update Safe Load Path Manual to Include Safety-Related Cable Locations | The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to update the Safe Load Path Manual for the Unit 2 turbine building (SLP-3) as part of the mid-1990\'s modification that added the G-03 and G-04 emergency diesel generators. Specifically, it was identified that SLP-3 allowed unrestricted load lifts over the Unit 2 turbine building truck bay area based upon a 1980\'s evaluation, and was not updated to reflect a modification that added safety-related cables for emergency diesel generators under the Unit 2 truck bay. Due to the close proximity of the A train cables to the B train cables, a loss of both trains of emergency alternating current (AC) power could result if the underground cables were disabled by a dropped load of sufficient magnitude. The licensee addressed the immediate concern by installing temporary steel plates over the affected area of the truck bay to provide adequate protection for upcoming heavy load lifts. Additionally, the licensee revised SLP-3 to require additional risk mitigation measures be taken prior to heavy load lifts in that area. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the associated 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, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The finding was determined to be of very low safety significance because the issue did not result in the actual loss of a safety function. This finding had a cross-cutting aspect in the area of problem identification and resolution, corrective action program component, because the staff did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance. Specifically, in 2008, when questions were raised by licensee staff regarding the adequacy of SLP-3, the SLP was not revised (P.1(d)). (Section 1R18.1 |
| 05000266/FIN-2009005-04 | 2009Q4 | Point Beach | Potential Failure to Adequately Evaluate Seismic II/I Concerns for Units 1 and 2 B Containment Sump Strainers | The inspectors identified an unresolved item (URI) regarding the B containment sump strainers for Units 1 and 2. Specifically, the inspectors questioned 24 Enclosure whether the ventilation ducts located above containment sump strainers were adequately evaluated with respect to seismic II/I considerations. On October 27, 2009, the inspectors performed a walkdown of the containment sump strainers of Unit 2 and noted a ventilation duct located above the B containment sump strainer. The inspectors were concerned that during a seismic event the structure could collapse and affect the strainers ability to fulfill its accident mitigating function. Specifically, if the ventilation duct and its support structure collapsed, the structural integrity of the sump strainer could be compromised or the failed duct and support could block the strainers. The sump strainers are relied upon to simultaneously maintain an adequate post-loss-of-coolant-accident suction source while preventing debris from entering the emergency core cooling system. The licensee\'s immediate documentation search on the seismic evaluation of the ventilation duct was unsuccessful. The licensee initiated AR 01159937. The licensee also determined that the same condition existed in Unit 1 and performed a prompt operability determination for the Unit 1 B strainer. The licensee later determined that the installation modification documentation for Unit 1, Engineering Change (EC) 1602, indicated that the modification did not require analysis of non-seismic components located over or adjacent to seismic components because there was no evidence of a potential seismic II/I concern at the time the modification was completed. Specifically, a seismic interaction walkdown was required in the installation work plan prior to the installation of the strainers. The walkdown was completed by two civil engineers who were Seismic Qualification Users Group (SQUG) qualified. The licensee determined, through discussions with the engineers who performed the walkdown, that the ventilation ducts were reviewed. Based on these facts, the licensee concluded that: (1) the ventilation ducts were seismically evaluated; (2) the evaluation determined that there are no seismic II/I concerns; and (3) that this is a documentation issue. The same conclusions applied to Unit 2. However, the inspectors were concerned with the use of SQUG methodology to evaluate the seismic II/I interactions with respect to the duct ventilation and the strainer. Specifically, the inspectors questioned whether this methodology could be applied to ventilation ducts because this type of structure did not appear in the equipment classes of the implementing procedure for SQUG. As a result of the inspectors\' questions, the licensee performed a prompt operability determination, in accordance with EN-AA-203-1001 that determined the Unit 1 B sump strainer was operable. The basis for this conclusion was documented in EC 14790. This EC performed a structural analysis that concluded that the ventilation duct support structure would be able to support loads induced by a seismic event. Again, this evaluation applied to Unit 2. In addition, the inspectors noted that the FSAR, Appendix A5.6, stated that Modified, new, or replacement equipment classified as Seismic Class I may be seismically designed and verified (after installation) for seismic adequacy using seismic experience data in accordance with a methodology developed by the SQUG. It was not clear whether this statement applied for all new modifications or to the replacement of previously SQUG-qualified equipment with similar equipment. The inspectors were also concerned with the level of documentation maintained by the licensee for the walkdowns performed using the SQUG methodology. Specifically, the inspectors noted that the documentation did not provide the necessary details to permit independent auditing of the inferences or conclusions. This issue is unresolved pending further NRC review of the licensing basis for the use of SQUG methodology and determination of further NRC actions to resolve the issues (URI 05000266/2009005-04; 05000301/2009005-04) |
| 05000266/FIN-2009005-05 | 2009Q4 | Point Beach | Momentary Loss of Unit 2 Reactor Vessel Level Indication in the Control Room | A self-revealed finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, \"Instructions, Procedures, and Drawings,\" was identified for performing an Instrumentation and Control (I&C) procedure that was inappropriate to the circumstances, and resulted in the momentary loss of all available channels of reactor vessel level indication in the control room. As part of the immediate corrective actions, the licensee suspended the performance of the procedure and sent an operator into containment to verify reactor vessel level via the local standpipe level indicator and to ensure level indication was reestablished. Additionally, the licensee applied a work planning logic-tie to this activity to ensure the reactor was de-fueled prior to performing this calibration and was currently evaluating the need for revisions to the procedure. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality 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 (i.e., core damage). The inspectors assessed the significance of the finding in accordance with IMC 0609, Appendix G, \"Shutdown Operations Significance Determination Process,\" and determined that this issue required a Phase 2 analysis since the finding increased the likelihood of a loss of reactor coolant system inventory. The inspectors and a senior reactor analyst determined through the analysis that this issue is best characterized as a finding of very low safety significance. This finding had a cross-cutting aspect in the area of human performance, work control component, in that the licensee did not appropriately coordinate work activities for the existing plant conditions to ensure the operational impact on reactor vessel level indication while at a water level above reduced inventory was fully understood (H.3(b)). (Section 1R20.1 |
| 05000266/FIN-2009005-06 | 2009Q4 | Point Beach | Failure to Maintain Proper Control of Radioactive Material Within the Radiologically Controlled Area | A self-revealed finding of very low safety significance and associated Non-Cited Violation of 10 CFR 20.1101(b) was identified for the failure to adequately control radioactive material to prevent its migration outside the radiologically controlled area (RCA), as required by licensee procedures. On May 21, 2009, a contract worker performing inspections of the main electrical transformers located outside the RCA picked-up a wadded-ball of debris (unmarked tape) and placed it in his front pants pocket. The debris was later found to be radioactively contaminated when the worker alarmed the protected area exit radiation monitors a few hours later as he attempted to leave the site. The tape was likely used to cover contaminated hoses that were previously used within the Point Beach RCA, but had escaped the licensee\'s control and migrated (blew) into the transformer area outdoors where it was found by the worker. The licensee\'s storage of radioactive material in an outdoor satellite RCA and/or the licensee\'s radioactive material control practices during refueling outages when the containment building equipment hatch was open to the environment led to the escape of the material outside the RCA. The contractor\'s assigned work duties should not have involved exposure to radioactive material; consequently, the worker was unnecessarily exposed to radiation from the contaminated tape. A dose evaluation completed by the licensee\'s consultant determined that the effective dose equivalent to the worker\'s thigh from exposure to the contaminated ball of tape was approximately one mrem. The licensee\'s corrective action called for expanded radiation protection oversight during movement of material in outdoor areas. Procedures were revised to include a post-outage walkdown of outdoor areas near the RCA yard. Additionally, the licensee planned to construct an enclosure so that storage/transfer of contaminated materials could be performed indoors. The finding was more than minor because it impacted the program and process 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 radiation, in that, unnecessary radiation exposure was received by an individual from inadequately controlled radioactive material. The finding was determined to be of very low safety significance because: (1) it involved a radioactive material control problem that was contrary to NRC requirements and the licensee\'s procedure; and (2) the dose impact to a member of the public (the contract worker) within the licensee\'s restricted area was less than 5 millirem total effective dose equivalent. The cause of the radioactive material control problem involved a cross-cutting component in the human performance area for inadequate work control, in that, job site conditions including environmental conditions (high winds, night time work, etc.) impacted human performance and consequently, radiological safety, during movement of material/equipment in outdoor areas (H.3.(a)). (Section 4OA5.1 |
| 05000266/FIN-2010002-01 | 2010Q1 | Point Beach | Untimely Corrective Actions to Address Longstanding Issue of Submerged Cables | A finding of very low safety significance and associated Non-Cited Violation of inspectors for the licensees failure to implement timely corrective actions to address the Specifically, this issue was first identified in 1997, with numerous condition reports condition adverse to quality. The licensee entered this issue into its corrective action program. Corrective actions completed include increased monitoring and pumping of manholes; proposed actions include design changes to support automatic monitoring and/or water removal from the manholes. The finding was more than minor because it was associated with the Initiating Events objective of limiting the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown as well as power operations. submerged cable issue in a timely manner; if left uncorrected, would lead to other cable failures as a result of the continued cable very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that cross-cutting aspect in the area of human performance, resources, because the licensee did not appropriately maintain long-term plant safety by maintenance of design margins, minimization of longstanding equipment issues, minimizing preventive maintenance deferrals, and ensuring maintenance and engineering backlogs were managed low enough to support safety (H.2(a)). (Section 1R06) |
| 05000266/FIN-2010002-02 | 2010Q1 | Point Beach | Inappropriate Application of A Dedicated Operator During A System Venting Surveillance | A finding of very low safety significance and associated Non-Cited Violation of 10 CFR 50.65(a)(4) was identified by the inspectors for the failure to properly assess risk that resulted from risk-significant maintenance being performed on the residual heat removal, safety injection, and containment spray systems. Specifically, the licensee inappropriately applied criteria for the use of a dedicated operator to meet availability requirements. As part of its corrective actions, the licensee stopped work that required the use of a dedicated operator pending further evaluation. The issue was more than minor because the licensees risk assessment for January 12, 2010, failed to consider multiple systems unavailable during maintenance. Specifically, the failure to account for the unavailability of the residual heat removal, safety injection, and containment spray systems, resulted in an inadequate daily risk assessment and could affect the unavailability time of this system in related performance and maintenance rule indicators. The inspectors evaluated the finding using the Significance Determination Process in accordance with IMC 0609, Significance Determination process, Attachment K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined the issue screened as having very low safety significance, because the incremental conditional core damage probability was less than 1E-6 due to the test condition lasting only four hours. This finding had a cross-cutting aspect in human performance, decision-making, because the licensee did not have a process or use a systematic approach regarding facets of a dedicated operator (H.1(a)). |
| 05000266/FIN-2010002-03 | 2010Q1 | Point Beach | Failure to Follow Temporary Modification Procedure | A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensee\'s failure to follow the temporary modifications procedure FP-E-MOD-03, Revision 6. Specifically, the Applicability section of this procedure was not properly applied to the temporary condensate storage tank (CST) modification such that the system was not appropriately characterized as a temporary modification. As a result, the licensee failed to adequately document an evaluation of the potential impacts to operating equipment. As of the conclusion of the inspection, the licensee had entered this issue into its corrective action program. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee inappropriately applied the exemption criteria of the temporary modification procedure to the fill point connected to the newly classified vent of the permanent CST and failed to assess the impact of the temporary CST system on plant design. The finding screened as having very low safety significance (Green) because the finding was not a design or qualification deficiency resulting in a loss of functionality, did not represent a loss of system safety function or loss of a single train for greater than its allowed technical specification time, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events. This finding had a cross-cutting aspect in the area of human performance, decision-making, because the licensee did not appropriately use conservative assumptions in decision-making and verify the validity of underlying assumptions for the temporary CST modification (H.1(b)). (Section 1R18) |
| 05000266/FIN-2010002-04 | 2010Q1 | Point Beach | Potential Failure to Adequately Assess Risk During CST Modifications | The inspectors identified an Unresolved Item (URI) regarding the risk-management aspects of the use of the temporary CST system while the permanent CST was out-of-service for various activities. As described in the previous section, in January 2010, the licensee used a temporary CST system that was installed as an alternate water supply for the permanent CSTs to accommodate several plant modifications for several issues including coating refurbishment, license renewal inspection, and installation of new motor-driven auxiliary feedwater system mechanical piping connections. During this evolution, the licensee credited the out-of-service CST as available from a risk perspective through the use of the temporary CST modification. The inspectors reviewed the risk management aspects of this issue and found that for the maintenance activities on the permanent CST, procedures did not exist to support this risk mitigation activity. Additionally, the design of the system did not appear to meet the established guidance for risk mitigation. Therefore, the risk mitigation management aspects are considered unresolved pending future evaluation of these activities (URI 05000266/2010002-04; 05000301/2010002-04; Potential Failure to Adequately Assess Risk During CST Modifications) |
| 05000266/FIN-2010002-05 | 2010Q1 | Point Beach | Inadequate Communications, Incomplete ALARA Job Planning and Ineffective Implementation of Radiological Work Controls | The inspectors identified a finding of very low safety significance for inadequate ALARA job planning and ineffective implementation of radiological work controls. This issue adversely impacted the licensees ability to minimize dose for the containment sump fibrous insulation removal project during the Unit 2 refueling outage (U2R30). Specifically, radiological controls were not effectively implemented to reduce ambient radiation levels and minimize in-field work hours for craft personnel. This resulted in an actual dose outcome that was not consistent with the planned, intended dose for work associated with the fibrous insulation removal project. Corrective actions were implemented to address the organizational communication deficiencies that led to the incomplete ALARA job planning and ineffective implementation of radiological work controls for the project. The finding was more than minor because it impacted the Occupational Radiation Safety Cornerstone objective for ensuring adequate protection of worker health and safety from exposure to radiation in the attribute of program and process for ALARA planning, in that, incomplete ALARA job planning and radiological work control deficiencies contributed to an actual increase in worker doses in excess of 5 person-rem and exceeded the licensees initial intended dose estimates by more than 50 percent. The finding did not involve: an overexposure; a substantial potential for an overexposure; or an impaired ability to assess dose. While the finding involved ALARA planning and controls, the 3-year rolling average dose for the Point Beach Nuclear Plant was less than the significance determination process threshold of 135-person-rem for pressurized water reactors at the time the performance deficiency occurred. Therefore, the inspectors determined that this is a finding of very low safety significance. The finding had a cross-cutting aspect in the area of human performance in decision-making, in that, the licensee did not communicate decisions and the basis for decisions to personnel who have a need to know the information in order to perform work safely in a timely manner (H.1(c)). |
| 05000266/FIN-2010002-06 | 2010Q1 | Point Beach | Failure to Establish Required Fire Watches | A finding of very low safety significance and associated Non-Cited Violation of Technical Specification 5.4.1.h for Units 1 and 2 was identified by the inspectors for the licensees failure to establish appropriate fire watches required as compensatory measures to address identified fire protection impairments. Specifically, on three occasions, the licensee failed to issue, and properly implement, fire watch surveillances as required by procedure OM 3.27. The licensee had entered all instances into its corrective action program. The finding was more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external factors (fire) and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to implement fire watches required as compensatory measures degraded the defense-in-depth elements of the fire protection program that is necessary to ensure safe shutdown in the event of a fire. The issue was of very low safety significance based on the low degradation rating for the finding. The finding had a cross-cutting aspect in the area of human performance, resources, because the licensees preliminary apparent cause evaluation attributed the underlying cause of these events to less than adequate procedures, or procedures that did not adequately link to each other, and pre-job briefing materials that did not address fire protection considerations (H.2(c)). |
| 05000266/FIN-2010002-07 | 2010Q1 | Point Beach | Failure to Evaluate Seismic Piping Interactions | A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, \"Design Control,\" was identified by the inspectors for the licensees failure to evaluate seismic piping interactions. Specifically, for a plant configuration where the stem of a spent fuel pool cooling system valve contacted an adjacent service water pipe, the licensee\'s evaluation to demonstrate that the existing spent fuel pool cooling system piping and valves met the design basis acceptance criteria of United States of America Standard (USAS) B31.1-1967 used a method of analysis that did not evaluate the dynamic effect of impact forces as specified by the design basis piping code. The licensee entered this issue into its corrective action program. The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, compliance with the seismic Category I design basis requirements of USAS B31.1-1967 was to ensure valve SF-2, the valve connection between two sections of spent fuel pool cooling system piping, would function as required during a seismic Category I design basis event. The finding screened as having very low safety significance (Green) because it was a design deficiency of the structural integrity of the spent fuel pool cooling piping system that: did not result in loss of cooling to the spent fuel pool; did not result from fuel handling errors that caused damage to fuel clad integrity or a dropped assembly; and did not result in loss of spent fuel pool inventory greater than 10 percent of spent fuel pool volume. The finding had no cross-cutting aspect because it was a legacy design issue, not reflective of current performance. |
| 05000266/FIN-2013003-01 | 2013Q2 | Point Beach | Failure to Control Materials Classified as High Winds/Tornado Hazards | The inspectors identified a finding of very low safety significance for the licensees failure to maintain control over the proper storage and placement of materials that were classified as high winds/tornado hazards, in accordance with procedure NP 1.9.6, Plant Cleanliness and Storage. Specifically, the inspectors identified that the licensee failed to perform weekly high wind missile hazards inspections since April 17, 2013. As a result, unsecured wooden pallets, wooden planks, metal rods and a metallic desk were discovered by the inspectors near Units 1 and 2 transformer areas. The issue was entered into the licensees corrective action program (CAP) for resolution as action request AR01882921. The licensee took immediate corrective action to remove and/or properly store the material after the tornado warning on June 17, 2013. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, the unsecured items would have the potential to lead to a more significant safety concern during high wind and tornado events. The inspectors determined the finding to be of very low safety significance because the inspectors answered No to each question listed in IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions . The inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not provide supervisory or management oversight of work activities such that nuclear safety was supported. Specifically, the licensee failed to provide appropriate oversight of work activities such that, when the program owner of the weekly high wind inspection changed, the requirement to perform weekly high winds tornado hazard walkdowns was not understood (H.4(c)). |
| 05000266/FIN-2013003-02 | 2013Q2 | Point Beach | Failure to Follow Operability Evaluation Process Following Water Leakage into the Control Room | The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V for the licensees failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, following water leakage into the control room, the licensees immediate operability determination failed to evaluate the effect the leakage had on the control room envelope operability. Additionally, the licensee did not address the functionality of the degraded flood barrier and its impact on operability. This issue was entered into the CAP as AR01877185. Corrective actions for this issue included performing a test of the control room envelope to demonstrate that appropriate positive pressure could be maintained with the known degraded barrier, and repair of the degraded flood barrier following performance of a functionality assessment. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Protection Against External Factors attribute of the Initiating Event Cornerstone, and adversely affected the Cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors determined the finding to be of very low safety significance in accordance with IMC 0609, Appendix A, Exhibit 1, because they answered No to the questions under Transient Initiators and External Event Initiators. The inspectors concluded that this finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate this problem such that the resolution addressed the cause and evaluated the condition for operability (P.1(c)). |
| 05000266/FIN-2013003-03 | 2013Q2 | Point Beach | Lack of Acceptance Criteria for Containment Visual Examinations | The inspectors identified a non-cited violation of 10 CFR 50.55a(g)(4), for failure to define acceptance criteria for containment visual examinations. Consequently, active containment liner degradation (pitting) was identified and the liner returned to service without defined criteria for accepting this condition. The licensee entered this issue into the CAP as action requests AR01858862 and AR01861158, and developed visual examination acceptance criteria to restore compliance with this NRC regulation. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening dated September 7, 2012, because it adversely affected the Barrier Integrity Cornerstone attribute of maintaining the functional integrity of containment. The inspectors also answered Yes to the more-than-minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, the lack of acceptance criteria in site procedures for containment visual examinations would become a more significant safety concern in that active liner degradation may not be properly evaluated and/or promptly corrected, resulting in a containment liner breach. In accordance with Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness, of IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, the inspectors checked the box under the Barrier Integrity Cornerstone because the corrosion induced pitting degraded the containment barrier. The inspectors determined this finding was of very low safety significance based on answering No to the Exhibit 3, Barrier Integrity Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, issued on June 19, 2012. Specifically, the inspectors answered No to the screening question associated with an actual open pathway (e.g., breach) in the containment and No to the question associated with reduction in function of hydrogen igniters in containment. The inspectors determined that the primary cause of the failure to define containment visual examination acceptance criteria was related to the cross-cutting component of human performance, decision making, because licensee staff did not apply a systematic process, when faced with unexpected plant conditions, to ensure safety was maintained. Specifically, a systematic process for developing acceptance criteria was not applied for the containment visual examinations (H.1(a)). |
| 05000266/FIN-2013003-04 | 2013Q2 | Point Beach | Incorrect Equipment Selected for Ultrasonic Examination | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for failure to select an appropriately contoured ultrasonic examination search unit wedge in accordance with procedure NDE-173, PDI Generic Procedure for the Ultrasonic Examination of Austenitic Piping Welds. Consequently, three elbow-to-pipe socket welds on the chemical and volume control system (CVCS) line were examined with the incorrectly contoured search unit and this examination would not provide a demonstrated level of accuracy necessary to reliably detect and size thermal fatigue cracks. The licensee entered this condition into the CAP as AR 01860155. To restore compliance with NRC regulations, the licensee considered the option of repeating these weld examinations using a qualified ultrasonic examination technique or the option to seek NRC approval to deviate from the American Society of Mechanical Engineers (ASME) Code Section XI requirements for ultrasonic examination. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because the inspectors answered Yes to the more-than-minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically , the examination of three chemical and volume control system welds was presumed adequate and absent NRC intervention, would have been returned to service for an indefinite period of service, which would have placed the piping at increased risk for undetected thermal fatigue cracking, leakage, or component failure. In accordance with Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness, of IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, the inspectors checked the box under the Initiating Events Cornerstone because leakage at this chemical and volume control system letdown line could result in a primary system loss of coolant accident. The inspectors determined this finding was of very low safety significance based on answering No to the questions in Part A of Exhibit 1, Initiating Events Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, issued on June 19, 2012. The inspectors answered these questions No because of the small diameter (2-inch) of the line and because the affected pipe welds were subjected to a VT-2 visual and penetrant testing (PT) examination that did not identify rejectable defects. The primary cause of the failure to select ultrasonic equipment (search unit contour) in accordance with procedure NDE-173 was related to the cross-cutting component of human performance, work practices, because the licensees management staff did not adequately set up clear expectations for procedure control and adherence for this activity. Specifically, insufficient direction was provided to vendor staff for simultaneous use of two procedures, NDE-178 and NDE-173, with different equipment requirements and restrictions (H.4(b)). |
| 05000266/FIN-2013003-05 | 2013Q2 | Point Beach | Failure to Implement Risk Management Actions During Bus D-40 Outage | A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(4) occurred on April 29, 2013, as a result of the licensees failure to properly manage and assess risk during a scheduled maintenance outage for emergency diesel generator G-04. Specifically, not all ongoing maintenance activities had been taken into account in the risk assessment for the in-progress maintenance activities and an unplanned entry into yellow risk occurred when they isolated bus D-40. The licensee entered this issue into the CAP as action request AR01870208. Corrective actions for this issue included restoring bus D-40 to service and initiating an evaluation of the issue through the condition reporting process. The inspectors determined the finding to be more than minor because it was similar to Example 7.e of IMC 0612, Appendix E, Example of Minor Issues, dated August 11, 2009, and because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone. The finding also affected the Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The inspectors determined that the finding was a mitigating systems contributor; evaluated the risk deficit for each instance; and found that the issue screened as having very low safety significance. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work control, because the licensee failed to appropriately plan and coordinate work activities. Specifically, when the licensee attempted to remove bus D-40 isolation work from the work schedule, the work package was not updated to reflect the change; and there was a failure to communicate and/or coordinate the changes in the work scope to the appropriate groups (H.3(b)). |
| 05000266/FIN-2013003-06 | 2013Q2 | Point Beach | Failure to Account for Plant-Specific Maintenance History in the Development of Preventive Maintenance Frequency | The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V for the licensees failure to follow procedure FP-PE- 90-01, Preventive Maintenance Program. Specifically, in 2009, when setting the preventive maintenance frequency for containment isolation valve 1MS-02083, t he licensee determined that a 15-year frequency was appropriate instead of the recommended 10 years. The licensees justification was based on internal maintenance history showing good performance. However, the inspectors review revealed that the maintenance history for this category of valves did not support this determination. The valve subsequently failed during surveillance on March 21, 2013, after 13 years of service. The licensee entered this issue into the CAP as AR01858451; corrective actions included replacing the valve and an action to review the preventive maintenance frequencies of critical solenoid-operated valves. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Barrier Performance attribute of the Barrier Integrity Cornerstone, and adversely affected the Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated this finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Checklist 3, and determined that the finding was of very low safety significance because the inspectors determined that a quantitative assessment was not required. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding did not reflect current performance due to the age of the performance deficiency. |
| 05000266/FIN-2013003-07 | 2013Q2 | Point Beach | Licensee-Identified Violation | Title 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part and the planning standards of 10 CFR 50.47(b). The Point Beach Nuclear Plant Emergency Plan Implementing Procedure EPIP 1.2, Emergency Classification, Section 5.1 states in part, Review the basis for the selected Emergency Action Level (EAL) to determine/confirm that the EAL applies, and, if an event meets the threshold of the EAL, then classify the emergency. Contrary to the above, on April 25, 2012, the licensee failed to follow its Emergency Plan during an actual emergency and that resulted in a failure to properly implement EALs. Specifically, inaccurate communications resulted in the over classification of an alert emergency classification level based on (EAL) HA3.1. Using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated February 24, 2012, Section 4.0 Actual Event Implementation Issue (Failure to Implement), and the inspectors determined that the violation was not greater than very low safety significance (Green) because no public official protective actions were implemented as a result of this event over classification. The issue is documented in the licensees Corrective Action Program as AR 01759720. |
| 05000266/FIN-2013003-08 | 2013Q2 | Point Beach | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and an associated NCV of Limiting Condition for Operability (LCO) 3.0.4. Licensee LCO 3.0.4 states in part, When an LCO is not met, entry into a MODE or other specified condition in the applicability shall only be made: When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Contrary to this, on April 14, 2013, the Unit 1 entered Mode 4 with the Spray Additive System inoperable and LCO 3.6.7 did not permit continued operation in Mode 4 with the system inoperable. Specifically, the outage senior reactor operator elected to walkdown all existing tags prior to Uni1 transitioning from Mode 4 to Mode 3. The senior reactor operator identified a tag was hung on 1SI-831A, Spray Additive Tank Outlet Valve, and was unsure about the purpose of the tag because the system was required to be operable in Mode 4. Investigations showed that the tag should have been removed prior to entry into Mode 4, but was left hung due to a sequencing error between the Mode change checklist and the tagging process. The licensee immediately entered the LCO 3.6.7, Spray Additive System, restored the system alignment and exited the LCO; this evolution lasted approximately 16 minutes. This issue was entered into the licensees CAP as AR01865777 and immediate corrective actions to restore the system to operable were taken, and a root cause evaluation was assigned. The inspectors determined the performance deficiency to be more than minor in accordance with IMC 0612, Appendix B, because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone. The finding also affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the Spray Additive Tank outlet valve closed it rendered the Spray Additive System inoperable in Mode 4. The inspectors and the R-III Senior Reactor Analyst reviewed IMC 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process, and determined that a Quantitative Assessment was not required; close spaces the issue was of very low safety significance. |
| 05000266/FIN-2015008-01 | 2015Q1 | Point Beach | Failure to Promptly Correct Conditions Adverse to Quality Regarding Electrical Power Cable Sizing and Protection | The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to implement timely corrective actions to address the longstanding issue of electrical power cables that have not been verified to be sized or protected in accordance with their design bases, as described in PBNPs Final Safety Analysis Report Section 8.0.1. Specifically, the licensee failed to correct known deficiencies regarding: (1) power cables with operating currents in excess of their current-carrying capacities; (2) power cables that are not protected against overload in accordance with the National Electrical Code; and (3) power cables for which their current-carrying capacities are undetermined. Although various corrective action documents have been initiated since these issues first came to light in the 1990 to 1991 time period, the licensee has not taken appropriate actions to correct the conditions adverse to quality to this date. The licensee entered this finding into their Corrective Action Program as Condition Report (CR) 02035020 and CR 02035680, with recommended actions to perform ampacity analysis for applicable cables, verify cables are protected against overload in accordance with the National Electrical Code, verify cable ampacities are higher than their respective load currents, and perform an evaluation to determine why this issue has not been resolved and address the safety culture aspect. The inspectors determined the licensees failure to promptly correct the conditions adverse to quality regarding electrical power cables was a performance deficiency warranting a significance determination. The performance deficiency was determined to be more than minor, and a finding in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding in accordance with IMC 0609.04, Phase 1, Initial Screening and Characterization of Findings. The finding screened as having very-low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function on the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The inspectors identified a crosscutting aspect associated with this finding in the area of Human Performance, associated with the Design Margin component, because the licensee failed to ensure equipment is operated within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process. |
| 05000282/FIN-2011012-01 | 2011Q4 | Prairie Island | Flammable Gas Bottles Installed and/or Stored in the Auxiliary Building | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to check the adequacy of design for flammable gas bottles installed in areas located within the auxiliary building and their impact on safety-related cables and equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed locations could impact nearby safety-related structures, systems, or components. The licensee entered this issue into their corrective action program to review the placement of the flammable gas bottles. The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstones attribute of Protection against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance. |
| 05000282/FIN-2011012-02 | 2011Q4 | Prairie Island | Failure to Correct a Condition Adverse to Quality | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee s failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to submit a license amendment request (LAR) to correct the non-conservative Technical Specification (TS) surveillance requirements in Section 3.8.1 for the emergency diesel generators (EDGs) allowable steady state frequency. The issue was originally identified and entered into the licensee s corrective action program on September 8, 2006. During this inspection, the licensee entered the finding into their corrective action program to evaluate how to resolve the issue. The inspectors determined that the finding was more than minor because the finding was associated with the Mitigating Systems cornerstone s attribute of Equipment Performance and affected the cornerstone s objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee could not be assured that the design requirements for the EDGs system loads would operate within the appropriate design specifications if the EDGs were allowed to operate within the non-conservative TS allowable steady state frequency of Y 58.8 Hertz (Hz) and U 61.2 Hz. As a result, the licensee established an administrative limit to limit operation of the EDGs to a frequency between 59.5 Hz and 60.5 Hz. The finding was of very low safety significance because it did not result in a loss of operability. The finding had a cross-cutting aspect in the area of human performance, decision-making because the licensee repeatedly delayed submitting the license amendment until a resolution was developed by an industry working group. |
| 05000282/FIN-2014008-01 | 2014Q2 | Prairie Island | No Compensatory Measure were established for Lack of Fuses Coordination associated with Safe Shutdown Power Supplies | The inspectors identified a finding of very low safety significance and associated NCV of the Prairie Island Nuclear Generating Plant Facility Operating License Condition 2.C.(4) for the licensees failure to implement the requirements as specified in the Fire Protection Program (FPP) for impaired safe shutdown equipment. Specifically, the licensee failed to establish appropriate compensatory measures when they identified lack of coordination between DC panel fuses and upstream panels supply fuse under fault conditions for several safe shutdown power supplies. The licensee replaced all miss-coordinated fuses and entered the issue into their Corrective Action Program. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to fire events prevent undesirable consequences (i.e., core damage). Specifically, the failure to establish compensatory measures for lack of fuse coordination degraded the defense and depth element of the Fire Protection Program. The finding represented a low degradation and therefore the inspectors determined that the finding screened as having very low safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence for the licensees failure to follow instructions as specified in Procedure FP-E-CAL-01 Calculations. |
| 05000282/FIN-2014008-02 | 2014Q2 | Prairie Island | Licensee-Identified Violation | The licensee identified a Severity Level IV violation of 10 CFR 50.59, (Changes, Tests, and Experiments, for the failure to demonstrate in a written evaluation that prior NRC-approval was not required for changes made to an accident analysis. Specifically, the licensee incorrectly concluded in written Evaluation 1102, Waste Gas Tank Rupture Dose Analysis, Revision 0 that higher activity levels and dose rates at the Exclusion Area Boundary and Low Population Zone associated with extended plant life due to license extension did not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR. The performance deficiency was determined to be more than minor because it was associated with the Radiation Safety cornerstone attribute of program and process and 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. The inspectors determined the violation was of Severity Level IV because the associated finding was of very low safety significance (Green) as there was no actual radioactive material release. The licensee entered this issue into their Corrective Action Program as AR 1417573 and AR 1427150 and intended to submit a license amendment request for review by the NRC. |
| 05000298/FIN-2013002-01 | 2013Q1 | Cooper | Failure to Maintain Design Control of the High Pressure Coolant Injection System | The inspectors reviewed a self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to maintain design control of high pressure coolant injection relief valve HPCI-RV-12RV. The licensee entered this issue into their corrective action program as Condition Reports CR-CNS-2013-00474 and CR-CNS-2013-00507. The failure to maintain design control of high pressure coolant injection system relief valve HPCI-RV-12RV was a performance deficiency. This performance deficiency was more than minor and therefore, a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone. Specifically, the licensee failed to adequately analyze the effects of the change in flow rate of the replacement relief valve, thereby affecting the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety- significance in accordance with the licensees maintenance rule program. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000298/FIN-2013002-02 | 2013Q1 | Cooper | Failure to Maintain Design Control of the Internal Flooding Analysis | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to assure that the applicable design basis requirements, associated with the stations internal flooding analysis in response to a medium energy line break, were correctly translated into the plant design. Specifically, the licensee used incorrect assumptions for a time critical operator action, and this resulted in a nonconservative analysis for a moderate energy line break in the 903 feet control building corridor. The licensee entered this deficiency into their corrective action program for resolution as Condition Reports CR-CNS-2013-00579, CR-CNS-2013-00619, and CR-CNS-2013-01553. The failure to maintain design control with respect to the internal flooding analysis was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone. Specifically,the licensees failure to use correct assumptions for time-critical operator actions resulted in a nonconservative analysis for a moderate energy line break in the 903-foot control building corridor, thereby affecting the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes. |