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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000458/FIN-2017001-02 | 2017Q1 | River Bend | Failure to Properly Pre - Plan and Perform Maintenance on the Control Building Chilled Water System | Green . The inspectors identified a non- cited violation of Technical Specification 5.4, Procedures, for the licensees failure to properly pre-plan and perform maintenance on safety -related components in accordance with documented instructions appropriate to the circumstances. Specifically, the licensee used work order instructions that did not contain sufficient detail for the reassembly of SWP -PVY32C, a safety -related valve in the control building ventilation system . As a result, SWP -PVY32C developed a refrigerant leak, and on November 17, 2015 , the valve failed. This in turn caused the control building ventilation system to fail , and the high pressure core spray system was consequently declared inoperable. The licensee entered this condition into their corrective action program as Condition Report CR- RBS -2017- 02364. Corrective actions included incorporating the torque values into the model work order instructions for future maintenance and reassembly . The failure to properly pre-plan and perform maintenance on safety -related components in accordance with documented instructions was a performance deficiency. The performance deficiency was more than minor , and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the control building ventilation system failed, it impact ed the operability of the high pressure core spray system. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect the design or qualification of a mitigating structure, system, or component (and the structure, system, or component maintained its operability), it did not represent a loss of safety function, it did not represent an actual loss of function of at least a single train for greater than its technical specification outage time, and it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees Maintenance Rule program for greater than 24 hours. This finding has a cross- cutting aspect in the area of human performance, challenge the unknown, because individuals did not stop when faced with uncertain conditions. Specifically, workers proceeded with assembling the valve when the torque values or torqueing sequence were not specified (H.11) |
| 05000458/FIN-2017001-03 | 2017Q1 | River Bend | Failure to Enter Applicable Technical Specification Action Statements When Control Building Chillers Were O ut of Service | Green . The inspectors identified a non- cited violation of Technical Specifications 3.8.4, DC Sources - Operating, 3.8.7, Inverters Operating, and 3.8.9, Distribution Systems Operating, for the licensees failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. Specifically, electrical power systems required by the above limiting condition s for operation were inoperable due to the associated division of the control building chilled water system chillers being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that is needed for the associated electrical systems to perform their specified safet y functions. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by allowing configurations that did not conform to the single failure criterion. The lic ensee entered this issue into their corrective action program as Condition Report CR- RBS -2015 -02525 . Corrective actions included entering the appropriate limiting conditions for operation of affected safety -related systems when the non -safety related support system were non -functional. 4 The failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours wa s a performance deficiency . Specifically, electrical power systems required by the above limiting condition s for operation were inoperable due to the associated division of the control building chilled water system chillers being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that is needed for the associated electrical systems to perform their specified safety functions. The performance deficiency was more than minor, and therefore a finding, because it wa s associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that res pond to initiating events to prevent undesirable consequences. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by allowing configurations that did not conform to the single failure criterion. The inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to require a detailed risk evaluation because it represented a loss of system and/or function. A senior reactor analyst performed a det ailed risk evaluation for a previously identified performance deficiency associated with the licensees failure to account for a loss of all control building chilled water system cooling scenario, either quantitatively or qualitatively, which resulted in uncompensated impairment to all systems associated with the main control room (Agencywide Documents Access and Management System (ADAMS) Accession N o. ML16132A144). This previously performed detailed risk evaluation bounds the risk associated with the finding dispositioned in this write- up: the failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. Therefore, the finding was determined to be of very low safety significance (Green). No cross -cutting aspect was assigned as the performance deficiency is not indicative of current licensee performance |
| 05000458/FIN-2018001-01 | 2018Q1 | River Bend | Failure to Implement Procedure for Storage of Material in the Pools | The inspectors identified a non-cited violation of Technical Specification 5.4.1.a for the licensees failure to implement written procedures for activities referenced in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensee failed to implement radioactive material control Procedure ADM-0071, Fuel Pools Material Control, Revision 8, for the storage and movement of spent Tri-Nuke filters |
| 05000458/FIN-2018001-02 | 2018Q1 | River Bend | Installation of an Incorrectly Specified Relay Causes Plant Transient and Reactor Scram | The inspectors reviewed two examples of a self-revealed finding for the licensees installation of an incorrectly specified relay in 1) the control circuitry for the feedwater level control systemand 2) the turbine generator voltage regulator circuitry. In each instance, the incorrectly specifiedrelay failed in service, causing a plant transient and automatic reactor scram |
| 05000458/FIN-2018012-01 | 2018Q2 | River Bend | Failure to Conduct Adequate Transient Snap Shot Assessment Following Recirculation Pump Trip | The inspectors identified a finding for the licensees failure to adequately validate simulator response during a transient snap shot assessment following an unexpected trip of reactor recirculation pump A on December 19, 2012. |
| 05000458/FIN-2018012-02 | 2018Q2 | River Bend | Failure to Identify and Correct a Broken Feedwater Chemistry Probe | Two examples of a self-revealed non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified for the licensees failure to identify that a broken chemistry probe in the feedwater system had the potential to cause an adverse impact on plant safety, and promptly implement appropriate measures to address that condition. |
| 05000458/FIN-2018012-03 | 2018Q2 | River Bend | Failure to Establish Procedural Guidance for Determining Core Flow During Unanticipated Single Loop Operations | The inspectors reviewed a self-revealed,non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish appropriate instructions in the abnormal operating procedure for thermal hydraulic instabilities. Specifically, the procedural step for determining core flow when in single loop operations at low power did not provide appropriate instructions to operators. As a result, station personnel could not conclusively determine core flow and inserted a manual reactor scram. |
| 05000458/FIN-2018012-04 | 2018Q2 | River Bend | Failure to Submit a Licensee Event Report for a Manual Scram | The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to submit a required licensee event report (LER). Specifically, on February 1, 2018, after an unexpected trip of the recirculation pump B, the licensee initiated a manual scram of the reactor that was not part of a preplanned sequence and failed to submit an LER within 60 days. |
| 05000458/FIN-2018012-05 | 2018Q2 | River Bend | Failure to Develop an Adequate Operational Decision-Making Issue for Compensatory Measures Related to a Degraded Condition of the Feedwater System Sparger Nozzles | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to develop an adequate Operational Decision-Making Issue (ODMI) document per Procedure EN-OP-111, Operational Decision-Making Issue Process. Specifically, the licensee failed to develop an ODMI that provided adequate guidance to the operators for safely operating the plant with degraded feedwater sparger nozzles. |
| 05000458/FIN-2018012-06 | 2018Q2 | River Bend | Failure to Provide Adequate Procedures for Post-Scram Recovery | The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to establish, implement and maintain a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, Procedure OSP-0053, Emergency and Transient Response Support Procedure, Revision 22, which is required by Regulatory Guide 1.33, inappropriately directed operations personnel to establish feedwater flow to the reactor pressure vessel using the main feedwater regulating valve as part of the post-scram actions. This resulted in the main feedwater regulating valves being operated outside their design limits. This resulted in catastrophic failure of the main feedwater regulating valve variseals and subsequent damage to multiple fuel assemblies. |
| 05000458/FIN-2018012-07 | 2018Q2 | River Bend | Failure to Perform 10 CFR 50.59 Evaluation for Main Feedwater System Sparger Nozzle Damage | The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59 , Changes, Tests, and Experiments, for the licensees failure to provide a written safety evaluation for the determination that operation with compensatory measures for damaged feedwater sparger nozzles did not require a license amendment pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. Specifically, the licensee failed to recognize that compensatory measures prohibiting operation in single loop conditions required technical specification changes, and as such required prior NRC approval. |
| 05000482/FIN-2008005-02 | 2008Q4 | Wolf Creek | Residual Heat Removal Suction Piping Saturation Temperature and Pressure | An unresolved item was identified on October 3, 2008, when Wolf Creek issued Licensee Event Report 2008-008-00 which stated that, during reviews for its response to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, it was discovered that both trains of residual heat removal were inoperable in Modes 3 and 4. Wolf Creek places residual heat removal in service during outages while the temperature of the reactor coolant system and the residual heat removal system could be up to 350oF. The temperature of the residual heat removal suction piping could remain sufficiently high to cause the saturation pressure to be greater than the static head from the refueling water storage tank. This may cause the residual heat removal to refueling water storage tank suction check valve to be unable to open against such a pressure differential and cause a steam void to form. This condition could exist on one or both trains of residual heat removal and prevent them from injecting to the reactor upon demand. Wolf Creek stated that residual heat removal was not operable per Technical Specifications 3.5.2 and 3.5.3. This issue is considered unresolved pending additional NRC review of Wolf Creeks root cause evaluation, fluid flow analyses, any past evaluations, compensatory measures, licensee procedures, and corrective actions: Unresolved Item 05000482/2008005-02, Residual Heat Removal Suction Piping Saturation Temperature and Pressure |
| 05000482/FIN-2008010-01 | 2008Q4 | Wolf Creek | Post-Fire Safe Shutdown Procedure did Not Identify Diagnostic Instrumentation | The team identified an unresolved item concerning the availability of diagnostic instrumentation needed to respond to a loss of reactor coolant pump seal cooling during certain fire scenarios. Technical Specification 5.4.1.d states that written procedures shall be established, implemented, and maintained covering fire protection program implementation. One of the procedures covered by this requirement is Procedure OFN KC-016, Fire Response. Procedure OFN KC-016, Revision 19, identified that fire damage in the following four fire areas could isolate both reactor coolant pump seal injection and thermal barrier cooling: • Fire Area A-21 Control Room AC and Filtration Units (Room 1501) • Fire Area C-22 Upper Cable Spreading (Room 3801) • Fire Area C-30 South Vertical Cable Chase (Room 3617) Control Building Elevation 2047-6 to 2073-6 • Fire Area C-33 South Vertical Cable Chase (Room 3804) Control Building Elevation 2073-6 Reactor coolant pump seal injection and thermal barrier cooling are the two methods used to cool the reactor coolant pump seals. One method of seal cooling must be maintained during reactor coolant pump operation to prevent seal failure, which, in some cases, could lead to increased seal leakage beyond the capacity of the charging pump. Procedure OFN KC-016 requires operators to recognize when one or both seal cooling methods were lost and take specific mitigating actions. While checking the feasibility of manual actions, the team identified that neither Procedure OFN KC-016 nor any other fire protection program document identified the instrumentation needed to identify a loss of seal cooling. Since the procedure required operators to recognize the loss of cooling and take response actions and the procedure did not identify the instrumentation to be used, the team could not verify that it would remain free of fire damage for fires in these four fire areas. The team was unable to verify that manual actions used as compensatory measures for potential fire damage could be reliably performed (these compensatory measures were implemented in response to Apparent Violation 05000482/2005008-03, Failure to Ensure Redundant Safe Shutdown Systems Are Protected in Accordance with the Provisions of the Approved Fire Protection Program). If the operators failed to implement appropriate actions, seal failure and an uncontrolled loss of coolant could occur. Such a seal failure was not analyzed as part of the approved fire protection program. The team determined that this deficiency may be more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and could affect the availability, reliability, and capability of systems that respond to fire events to prevent undesirable consequences. Additional information is needed from the licensee in order to determine whether the instrumentation needed to promptly recognize and diagnose challenges to reactor coolant pump seal cooling was available and would be free of fire damage. The significance of this issue will be determined if this issue does involve a performance deficiency. The licensee implemented an hourly fire watch in the affected fire areas as a compensatory measure. The licensee entered this issue into their corrective action program as Condition Report 2008-005171. Enforcement. Technical Specification 5.4.1.d states that written procedures shall be established, implemented, and maintained covering fire protection program implementation. Report E-1F9910, Post-Fire Safe Shutdown Analysis, Revision 4, documents an area-by-area analysis of the post-fire safe shutdown capability. The post-fire safe shutdown analysis is part of the approved fire protection program as defined in License Condition 2.C.(5), Fire Protection. The post-fire safe shutdown analysis identified the potential for the loss of reactor coolant pump seal injection and thermal barrier cooling for fires in Fire Areas A-21, C-22, C-30, and C-33. Procedure OFN KC-016, Fire Response, Revision 19, identified the operator actions necessary to mitigate possible failures of equipment due to fire damage in order to implement the post-fire safe shutdown analysis. The team was concerned that the procedural guidance provided may not be adequate to allow operators to successfully perform the required post-fire safe shutdown actions required by the approved fire protection program. This issue could impact the ability to control reactor coolant system inventory and pressure to support the post-fire safe shutdown. Additional information is needed from the licensee in order to determine whether the instrumentation needed to promptly recognize and diagnose challenges to reactor coolant pump seal cooling was available and would be free of fire damage. This information is needed to determine whether a violation existed for this issue. Therefore, this issue is being treated as an unresolved item (URI): URI 05000482/2008010-01, Post-Fire Safe Shutdown Procedure Did Not Identify Diagnostic Instrumentation |
| 05000482/FIN-2008010-04 | 2008Q4 | Wolf Creek | Operator Actions May Create the Potential for Secondary Fires | The team identified an unresolved item concerning the potential that operator actions taken in response to fires in 14 fire areas may cause secondary fires and invalidate the safe shutdown analysis. Procedure OFN KC-016, Fire Response, Revision 19, specified operator actions to be taken in response to fires outside of the control room. This procedure provides the mitigating actions needed to maintain hot standby in the event of various failures and spurious actuations. The team identified the following 14 fire areas where the mitigating actions may cause secondary fires and invalidate the safe shutdown analysis: • Fire Area A-8 Auxiliary Building - 2000 Elevation, General Area • Fire Area A-11 Cable Chase (Room 1335) • Fire Area A-16 Auxiliary Building - 2026 Elevation, General Area • Fire Area A-17 South Electrical Penetration (Room 1409) • Fire Area A-18 North Electrical Penetration (Room 1410) • Fire Area A-27 Reactor Trip Switchgear (Room 1403) • Fire Area C-18 North Vertical Cable Chase (Room 3419) • Fire Area C-21 Lower Cable Spreading (Room 3501) • Fire Area C-22 Upper Cable Spreading (Room 3801) • Fire Area C-23 South Vertical Cable Chase (Room 3505) • Fire Area C-24 North Electrical Chase (Room 3504) • Fire Area C-30 South Vertical Cable Chase (Room 3617) • Fire Area C-33 South Vertical Cable Chase (Room 3804) • Fire Area RB Reactor Building (Containment) For these 14 fire areas, the procedure directs the operators to remove power to a power-operated relief valve if a fire causes the power-operated relief valve to spuriously open and operators are unable to close its associated block valve. Specifically, the procedure directs the operators to open circuit breakers on the associated 125 Vdc power supply. The inspectors noted that the failure of the block valve to close is considered fire damage and is not considered a spurious operation of the valve. The licensee specified this action in order to close the power-operated relief valve and preclude the potential for spurious opening due to inter-cable faults (i.e., cable-to-cable hot shorts). However, the team determined this action would also remove the control power used to operate 4160 Vac and 480 Vac circuit breakers. The removal of control power would prevent remote breaker operations and disable circuit breakers protective trips for the train affected by the fire. Removing control power to the circuit breaker results in a loss of its ability to automatically isolate faults before severe damage occurs. As a result, fire-induced faults (shorts to ground) in non-essential power cables of the affected 4160 Vac and 480 Vac supplies may not clear until after tripping an upstream feeder breaker to the supplies, which would remove power from equipment which was assumed by the safe shutdown analysis to be unaffected. This action would prevent breakers from automatically opening during an overload condition and has the potential to initiate secondary fires in plant locations outside of the initial fire area. The safe shutdown analysis assumed that a fire is present in only one fire area at any time. The team determined that the operator actions taken in response to fires in the listed fire areas had the potential to initiate secondary fires in other plant locations, which would invalidate the safe shutdown analysis and could impact the ability to achieve and maintain safe shutdown. The team was concerned that operator actions specified for responding to the fire-induced spurious opening of a pressurizer power-operated relief valve could remove electrical circuit protection and create the potential for secondary fires outside the initial fire area. Taking actions that could create secondary fires was potentially a performance deficiency. The team determined that this deficiency may be more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and could affect the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additional information is needed from the licensee to understand the potential for secondary fires and the possible locations of those fires based on the circuit design and remaining protection available. The licensee will provide the routing of power-operated relief valve cables in the 14 fire areas of concern and will identify the affected components where breaker coordination would be lost. The significance of this issue will be determined if this issue does involve a performance deficiency. As a compensatory measure, the licensee implemented an hourly fire watch in the affected fire areas, with the exception of the reactor building, which is not readily accessible during power operations. For the reactor building, the licensee is monitoring the containment temperature as a compensatory measure. The licensee entered this issue into their corrective action program as Condition Report 2008-005210. Enforcement. License Condition 2.C.(5) states, in part, that the licensee shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPS Final Safety Analysis Report for the facility through Revision 17, the Wolf Creek Site Addendum through Revision 15, and as approved in the Safety Evaluation Report through Supplement 5. The Wolf Creek Updated Safety Analysis Report combined the SNUPPS Final Safety Analysis Report, Revision 17, and the Wolf Creek Site Addendum, Revision 15, into one document. Appendix 9.5B of the Updated Safety Analysis Report provides an area-by-area analysis of the power block that incorporated Drawing E-1F9905, Fire Hazards Analysis, Revision 2, by reference. Drawing E-1F9905 states that the overall intent is to demonstrate that a single plant fire will not negatively affect the post-fire safe shutdown capability and that if a circuit damaged by a fire is protected by an individual overcurrent protection device, that device is assumed to function to clear the fault. The team was concerned that operator actions specified for responding to the fire-induced spurious opening of a pressurizer power-operated relief valve could remove electrical circuit protection and create the potential for secondary fires outside the initial fire area. Specifically, removal of control power to 4160 Vac and 480 Vac circuit breakers prevents operation of the devices overcurrent protection function. Failure of circuit breakers to clear faults on power cables damaged by fire would create the potential for the overcurrent condition to start a secondary fire at another location. The plants post-fire safe shutdown capability has only been evaluated for damage due to a single fire. Additional information is needed from the licensee to understand whether there is a credible potential for secondary fires and the possible locations of those fires based on the circuit design and remaining protection available. The licensee will provide the routing of power-operated relief valve cables in the 14 fire areas of concern and will identify the affected components where breaker coordination would be lost. This information is needed to determine whether a violation existed for this issue. Therefore, this issue is being treated as an unresolved item: URI 05000482/2008010-04, Operator Actions May Create the Potential for Secondary Fires |
| 05000482/FIN-2009005-01 | 2009Q4 | Wolf Creek | Failure to Correct Discolored Boric Acid Deposits | The inspectors identified a cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to take action to stop leakage from the base of the refueling water storage tank or evaluate the leakage and wastage for acceptability. Specifically, the licensee did not take actions to prevent recurring discolored boric acid deposits for approximately 11 years. Failure to correct leakage from the refueling water storage tank base was the subject of a noncited violation in NRC Inspection Report 05000482/2007006. This issue was entered into the licensee\\\'s corrective action program as Condition Report 22866. The failure to implement corrective actions for the refueling water storage tank leakage was a performance deficiency. The inspectors determined this issue impacted the Mitigating Systems Cornerstone and was greater than minor because if left uncorrected, the failure to correct the presence of boric acid leakage could become a more significant safety concern in that continued wastage could impact tank operability. Using the Phase 1 worksheets in Inspection Manual Chapter 0609.04, Significance Determination Process, the finding was determined to have very low safety significance because it did not result in a system or component being inoperable and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors identified a crosscutting aspect in the area of human performance associated with resources. Specifically, Wolf Creek did not maintain long-term plant safety minimizing corrective maintenance deferrals and this long-standing equipment issue (H.2.c) (Section 1R05) |
| 05000482/FIN-2009005-02 | 2009Q4 | Wolf Creek | Control of Transient Ignition Sources | The inspectors identified a noncited violation of Technical Specification 5.4.1.a, for an inadequate Procedure AP-10-101, Control of Transient Ignition Sources. On October 21, 2009, the inspectors observed maintenance personnel performing weld preparation work on essential service water piping to containment cooler B using a flapper wheel. The inspectors observed that the ignition control barriers for the hot work were insufficient in that the sparks from the preparation work extended four to five feet from the job site and there was no fire watch posted. On December 4, 2003, a procedure revision inappropriately incorporated a change to the procedure where a fire watch did not have to be posted when using wire brushes, flapper wheels, polishing devices, or Rol-Lok type buffing pads mounted on power grinder motor drives or air tools. The maintenance supervisor stopped the work until a fire watch was posted. The licensee entered this into their corrective action system as Condition Report 20993. This finding is more than minor because it affected the Mitigating Systems Cornerstone attribute of Protection Against External Factors - Fires, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The lack of a posted fire watch could adversely affect the ability to achieve and maintain safe shutdown in the event of a severe fire in the affected area. Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, could not be used to effectively evaluate the finding and defense-in-depth strategies because the 2003 changes to the fire watch program affected multiple fire areas and conditions. Therefore, in accordance with Inspection Manual Chapter 0609, Appendix M, the safety significance was determined by regional management review who concluded that the finding was of very low safety significance (Green). This finding was reviewed for crosscutting aspects and none were identified. The original change occurred in 2003 and was not indicative of current performance (Section 1R05.2) |
| 05000482/FIN-2009005-03 | 2009Q4 | Wolf Creek | Failure to Identify Sources of Boron Leakage | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to identify sources of boron leakage and document them in a corrective action document. Specifically, prior to October 23, 2009, the licensee failed to accomplish the requirements of Procedure AP 16F-001, Boric Acid Corrosion Control Program, Revision 5, step 6.4.1, which states, in part, Sources of boron seepage/leakage shall be identified/verified and documented in the applicable corrective action document. During a boric acid walkdown, the inspectors identified 11 sources of boron leakage which had not been previously identified and documented by the licensee. The licensee entered this finding into their corrective action system as Condition Report 00021274. The finding was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because the issue would not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. The inspectors also determined that the finding had a crosscutting aspect in the area of problem identification and resolution, operating experience, where the licensee did not institutionalize operating experience through changes to station processes, procedures, equipment, and training programs (P.2.(b) |
| 05000482/FIN-2009005-04 | 2009Q4 | Wolf Creek | Failure to Incorporate Requirements of Regulatory Guide 1.182 into Daily Shutdown Risk Assessment | The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the failure to adequately perform shutdown risk assessments during Refueling Outage 17. Between October 10 and November 17, 2009, Wolf Creek did not appropriately consider electrical power, decay heat removal, and containment when assessing shutdown risk. This changed the outcome or color of the qualitative calculation on several occasions. The licensee entered this issue in their corrective action program as Condition Reports 22295 and 22296. The failure to meet shutdown risk assessment requirements in the daily shutdown risk assessment process is a performance deficiency. The inspectors determined this finding was associated with the Mitigating Systems Cornerstone and was more than minor because it involved incorrect risk assessment assumptions by omitting requirements specified in committed guidance without providing justification for that omission. Such errors of omission have the potential to change the outcome of the licensees maintenance risk assessment as described above. Per Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, licensees who only perform qualitative analyses of plant configuration risk due to maintenance activities, the significance of the deficiencies must be determined by an internal NRC management review using risk insights where possible in accordance with Inspection Manual Chapter 612, Power Reactor Inspection Reports. The NRC management review concluded that this finding was of Green safety significance because missing risk management actions did not result in loss of key shutdown risk functions. Additionally, the cause of the finding has a human performance crosscutting aspect in the area associated with the resources. Specifically, Wolf Creek did not ensure that Procedure APF 22B-001-02 was complete, accurate, and up-to-date (H.2(c)) (Section 1R13) |
| 05000482/FIN-2009005-05 | 2009Q4 | Wolf Creek | Mode Change Under Technical Specification 3.0.4.b Without Required Risk Management Actions | On November 18, 2009, the inspectors identified a noncited violation of Technical Specification 3.0.4.b for ascension from Mode 4 to Mode 3 without establishing required risk management actions. Wolf Creek used technical specification Limiting Condition for Operation 3.0.4.b to permit mode ascension after performance of a risk assessment and identification of risk management actions to maintain safety in the next mode. The turbine-driven auxiliary feedwater pump was inoperable per Technical Specification 3.7.5. As a risk management action, protected train signs would be placed on the doors to the motor-driven auxiliary feedwater Pump A and B room doors. A walkdown conducted by the inspector on the morning of November 18, 2009, found that the protected train signs on the motor-driven auxiliary feedwater pump rooms were not in place. Also, a maintenance crew was performing radiography in the motor-driven auxiliary feedwater pump Room B. The motor-driven auxiliary feedwater Pumps A and B were also made inoperable (at separate times) later on the morning of November 18, 2009. The licensee entered this issue in their corrective action program as Condition Report 21926. Mode ascension under Technical Specification LCO 3.0.4.b without establishing required risk management actions is a performance deficiency. The finding was more than minor because it was associated with the configuration control and alignment attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The configuration control issues not only included the work being completed on the turbine-driven auxiliary feedwater pump, but also included containment isolation valve testing and radiography that was performed on the motor-driven auxiliary feedwater pumps which was not included in the risk assessment. The inspector used Inspection Manual Chapter 0609.04, to determine that the finding was of very-low safety significance (Green) because it did not result in a loss of system safety function; did not exceed allowable technical specification outage time; and was not a seismic, flooding, or severe weather concern. Additionally, the cause of the finding has a human performance crosscutting aspect in the area associated with decision making. Specifically, Wolf Creek used a risk assessment form and an informal mode change form to communicate between departments the requirement for risk management actions. The two forms were in conflict and the personnel who implemented the risk management actions were not informed (H.1(c)) (Section 1R13) |
| 05000482/FIN-2009005-06 | 2009Q4 | Wolf Creek | Failure to Follow Corrective Action Procedure | On October 15, 2009, the inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to follow Procedure AP 28A-100, Condition Reports. Wolf Creek failed to initiate a condition report for evaluation of corrosion on containment cooler A piping. After inspector challenging, Wolf Creek initiated condition reports, performed nondestructive testing, replaced corroded studs, and evaluated the cause of the corrosion. The inspectors determined that the failure to follow AP 28A-100, Appendix C, was a performance deficiency. This issue was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the issue screened to Green because there was not a loss of operability and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. A crosscutting aspect was identified in the problem identification and resolution area of the corrective action program. Specifically, Wolf Creek failed to implement a corrective action program with a low threshold for identifying issues (P.1.a) (Section 1R13) |
| 05000482/FIN-2009005-07 | 2009Q4 | Wolf Creek | Failure to Follow Procedures Results in Draining of Emergency Core Cooling System Pump Oil | On November 23, 2009, a self-revealing violation of Technical Specification 5.4.1.a was identified when a technician failed to follow procedure and emptied 45 gallons of oil from centrifugal charging Pump A rendering the pump inoperable. The technician was supposed to remove the temperature indicator for calibration but instead removed the thermowell which breached the lube oil subsystem of centrifugal charging Pump A. An unplanned entry into Technical Specification 3.5.2, Condition A, was made for approximately 10 hours. The licensee entered this issue in their corrective action program as Condition Report 21993. The failure to follow station procedures and correctly remove the detector was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and 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 significance of this finding using Phase 1 of Inspection Manual Chapter 0609.04, and determined that the finding was of very low safety significance (Green) because the pump was inoperable for less than 24 hours. Also, the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors identified a human performance crosscutting in the area of work practices because self-checking and communication with the supervisor failed to prevent the event (H.4.a) (Section 1R13) |
| 05000482/FIN-2009005-08 | 2009Q4 | Wolf Creek | Inadequate Operability Evaluation of Essential Service Water Pumps | On November 5, 2009, inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to perform an adequate operability evaluation required by procedure. The inspectors identified that Operability Evaluation EF 09-010, Revisions 0 and 1, did not demonstrate that the essential service water pumps could withstand a safe shutdown earthquake. Revision 2 of the operability evaluation included calculations to demonstrate acceptable stresses and included pump impeller clearances. This issue is captured in the corrective action program as condition reports 22798 and 21572. The failure to perform an adequate operability evaluation per Procedures AP 28-001 and AP 26C 004 was a performance deficiency. The inspectors determined that this finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, this issue relates to the availability and reliability examples of the equipment performance attribute because a latent common mode failure mechanism was not correctly evaluated. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, and determined that the finding was of very low safety significance (Green) because the issue was not a design or qualification deficiency confirmed to result in loss of operability or functionality, did not represent a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, an actual loss of safety function of a nontechnical specification risk-significant equipment train, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of the finding has a problem identification and resolution crosscutting aspect associated with the corrective action program because Wolf Creek failed to thoroughly evaluate the failure mechanism such that the resolutions address the causes and extent of conditions, as necessary (P.1.c) (Section 1R15) |
| 05000482/FIN-2009005-09 | 2009Q4 | Wolf Creek | Positive Reactivitiy Addition Prohibited by Technical Specifications while in Mode 2 | The inspectors identified a noncited violation of Technical Specification 3.3.1, Condition I, for making positive reactivity addition prohibited by technical specifications in Mode 2 because one source range nuclear instrument channel was inoperable. Following a reactor transient, one of the source range nuclear instrument channels experienced an unanticipated increased count rate and was declared inoperable. Wolf Creek restored the channel in an operability evaluation which cited the cause as a problem in a component which was later determined not to exist in the installed configuration; however, the improperly restored equipment had already been used for to support plant startup on August 22, 2009. Wolf Creek replaced the detector during Refueling Outage 17. This issue was entered into the correction action program as Condition Report 20208. Reactivity addition with source range channel Nuclear Instrument-31 inoperable is a performance deficiency. The finding was more than minor because it was associated with the configuration control (reactivity control) attribute of the Barrier Integrity Cornerstone, and it affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609.04, and determined that the finding screened to Green because the finding only affected the fuel barrier. Additionally, the cause of the finding has a human performance crosscutting aspect in the area associated with the decision making. Specifically, Wolf Creek did not use conservative assumptions in decision making and adopt requirements to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action, when performing an operability evaluation for the source range Nuclear Instrument 31 detector prior to restarting from a forced outage (H.1(b)) (Section 1R15) |
| 05000482/FIN-2009005-10 | 2009Q4 | Wolf Creek | Failure to Obtain Vendor Data Necessary for Plant Modification | On December 16, 2009, inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving failure to obtain vendor design data for a modification. In August 2009, a component cooling water modification was made to the reactor coolant pump thermal barrier heat exchangers flow rates as a corrective action to VIO 05000482/2009002 07 (EA-09-110). A flow rate above the previous design value was justified by an internal memo of a vendor opinion from a telephone conversation in 1992. The inspectors found this to be contrary to Procedure AP 05-005, for obtaining data from vendors. The notice of violation will remain open until full compliance has been restored. Wolf Creek consulted with Westinghouse, confirmed the acceptability of the increased flow rate, and requested a formal calculation. This issue is captured in Condition Report 22824. The inspectors determined that this finding was more than minor because this issue aligned with Inspection Manual Chapter 0612, Appendix E, example 2.f, in that the modification relied on verbal statements to raise the allowable flow through the heat exchanger. This is a significant deficiency in the modification package. The inspectors determined this finding was associated with the design control attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609.04 and determined that the finding was of very low safety significance because assuming worst case degradation, the finding would not result in exceeding the technical specification limit for identified reactor coolant system leakage and would not have likely affected other mitigation systems resulting in a total loss of their safety function because seal injection was available. This finding has a crosscutting aspect in the area of human performance associated with work practices in that management was unsuccessful in communicating expectations on procedure use and adherence in engineering (H.4.b) (Section 1R18) |
| 05000482/FIN-2009005-11 | 2009Q4 | Wolf Creek | Failure to Correct Vessel Head Vent Path | The inspectors identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, due to an inadequate vent path for the reactor vessel head. The inadequate vent path resulted in the formation of voids in the reactor vessel head during Refueling Outage 17. Failure to ensure an adequate vent path in the reactor vessel head was the subject of a noncited violation in NRC Inspection Report 05000482/2008004. During and after Refueling Outage 16, Wolf Creek initiated a root cause evaluation and corrective actions to prevent occurrence. When one of the possible root causes was disproven in Refueling Outage 17, no additional action was taken to determine the cause of the vessel head vent blockage. However, the licensee could not exclude blockage in the piping. This issue was entered into the corrective action program and the licensee plans to conduct a more thorough inspection of the piping during the next refueling outage. This issue is being tracked by the licensee as Condition Report 22501. The inspectors determined that the failure to provide adequate vessel head vent path to prevent gas accumulation in the reactor vessel during depressurized plant operations was a performance deficiency. The inspectors determined that this finding, which was associated with the Initiating Events Cornerstone, was more than minor because if left uncorrected, it would have become a more significant-safety concern. Specifically, without an adequate vent path the reactor vessel does not have an effective means of relieving noncondensable gases to prevent a loss of reactor coolant system inventory. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix G, Attachment 1, and determined it be of very low safety significance based upon the demonstrated availability of mitigating systems and the flooded reactor cavity inventory. The inspectors determined the cause of the finding had a problem identification and resolution aspect in the corrective action program. Specifically, Wolf Creeks corrective actions were not successful to address the vent path blockage in a timely manner (P.1(d)) (Section 1R20) |
| 05000482/FIN-2009005-12 | 2009Q4 | Wolf Creek | Unevaluated Scaffold Against Component Cooling Water Piping | The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure AP 14A-003, Scaffold Construction and Use, when scaffolding was erected against operable safety-related equipment. On October 15, 2009, the inspectors walked down containment and identified scaffolding in contact with component cooling water piping. The tag on the scaffold explicitly stated that it was not seismically qualified. At the time, both steam generators were inoperable and both trains of residual heat removal were required to be operable. The inspectors reviewed the bases for Technical Specification 3.4.7, RCS Loops - Mode 5, Loops Filled, which required an operable heat sink path from residual heat removal to component cooling water to essential service water. This issue was entered into the corrective action program as Condition Report 22464. The construction of an unqualified scaffold against operable component cooling water piping was a performance deficiency. The inspectors determined that this finding was more than minor because it is associated with the equipment performance attribute for the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, this issue relates to the availability and reliability examples of the equipment performance attribute because a latent failure mechanism was not evaluated. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs. The inspectors determined that Checklist 3 was applicable because the unit was in cold shutdown with the refueling cavity level less than 23 feet. Using Appendix G, Attachment 1, Checklist 3, Phase 2 analysis was not needed and the finding was of very low safety significance (Green) because the licensee was able to demonstrate that the seismically unqualified scaffolding would not have resulted in a loss of safety function. The inspectors determined the cause of the finding had a human performance aspect in the area of resources. Specifically, Procedure AP 14A-003 was inadequate because it had conflicting guidance that allowed seismically unqualified scaffolds in Modes 5 and 6 (H.2.c) (Section 1R20) |
| 05000482/FIN-2009005-13 | 2009Q4 | Wolf Creek | Failure to Maintain Administrative Control of Keys to Locked High Radiation Areas | The inspector identified a noncited violation of Technical Specification 5.7.2.a.1 for failure to maintain administrative control of door and gate keys to high radiation areas with dose rates greater than 1 rem per hour but less than 500 rads per hour (referred to as locked high radiation areas). Specifically, as of October 21, 2009, the licensee did not have administrative controls over a single master key to locked high radiation areas. This issue was entered into the licensees corrective action program as Condition Report 20973. Failure to maintain administrative control of the master key to locked high radiation areas was a performance deficiency. This finding is greater than minor because if left uncorrected the finding has the potential to lead to a more significant safety concern in that an individual could receive unanticipated radiation dose by gaining access a locked high radiation area without the proper controls and briefing. This finding was evaluated using the occupational radiation safety significance determination process and determined to be of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning or work control issue, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, the violation has a crosscutting aspect in the area of human performance associated with the work practices component because the lack of peer and self-checking resulted in inadequate control of keys to locked high radiation areas (H.4(a)) (Section 2OS1) |
| 05000482/FIN-2009005-14 | 2009Q4 | Wolf Creek | Failure to Identify Inoperable P-6 Interlock and Intermediate Range Detector | On December 30, 2009, the inspectors identified a noncited violation of Technical Specification Table 3.3.1-1, Function 18.a, when Wolf Creek restarted on May 18, 2005. During a reactor shutdown on October 7, 2006, intermediate range neutron detector Nuclear Instrument-36 did not decrease below 6E -11 amps and energize source range detector Nuclear Instrument-32. The detector was not replaced until Refueling Outage 16 in March 2008. The licensee entered this issue in their corrective action program as Condition Report 22450 The inspectors determined that the failure to ensure that the P-6 interlock was operable per the technical specification as defined in the bases was a performance deficiency. The finding was more than minor because it was associated with the configuration control (reactivity control) attribute of the Barrier Integrity Cornerstone, and it affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609.04, and determined that the finding screened to Green because the P-6 interlock only affected the fuel barrier (Section 4OA2). This finding was not assigned a crosscutting aspect because the cause was not representative of current performance |
| 05000482/FIN-2009005-15 | 2009Q4 | Wolf Creek | Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function | The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73 in which the licensee failed to submit a licensee event report within 60 days following discovery of events or conditions meeting the reportability criteria. On December 31, 2009, the inspectors identified a licensee event report that was no timely. Licensee Event Report 2009-009-00 was not issued within 60 days for a condition prohibited by technical specifications, and the event report did not identify that the disabling of both trains of the P-4 interlock on August 22, 2009 was also reportable per 10 CFR 50.73(a)(2)(v). The P-4 interlock was required by Technical Specification 3.3.2, function 8.a, and is discussed in USAR, Section 7.3.8, NSSS Engineered Safety Feature Actuation System. Wolf Creek licensee event report 2009-009 was correct in that the interlock is not credited in accident analysis. However, NUREG 1022, Section 3.2.6, specifies that inoperable systems required by the technical specifications be reported, even if there are other diverse operable means of accomplishing the safety function. The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was affected. Specifically, the NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done, the regulatory function is impacted. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The finding was reviewed by NRC management, and because the violation was determined to be of very low safety significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation consistent with the NRC Enforcement Policy. This finding was determined to have a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the licensee failed to appropriately and thoroughly evaluate for reportability aspects all factors and time frames associated with the inoperability of the engineered safety features actuation system (P.1(c)) (Section 4OA3) |
| 05000482/FIN-2009005-16 | 2009Q4 | Wolf Creek | Operator Actions disable Curcuit Breaker Coordination and Could Initiate Secondary Fires | The inspectors identified a noncited violation of License Condition 2.C.(5), Fire Protection, for the failure to implement and maintain the approved fire protection program. Specifically, the licensee prescribed mitigating actions in response to certain fire scenarios that would result in a loss of circuit breaker coordination and could initiate secondary fires in plant locations outside of the initial fire area. The licensee entered this issue into their corrective action program as Condition Report 2008-005210. This finding was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The risk significance of this finding was determined using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The finding was determined to be of very low safety significance using a Phase 2 evaluation. This finding was not assigned a crosscutting aspect because the cause was not representative of current performance (Section 4OA5.2) |
| 05000482/FIN-2009005-17 | 2009Q4 | Wolf Creek | Licensee-Identified Violation | On October 22, 2009, at 12:06 p.m., the Wolf Creek control room received trouble annunciators for emergency diesel generator A. Emergency diesel generator B was out of service for planned maintenance. 10 CFR 50.47(b)(4) requires that a standard emergency classification action level scheme be used by the licensee. Wolf Creek EAL 6, Loss of Electrical Power/Assessment Capability, requires, in part, that when both emergency diesel generators are out of service for greater than 15 minutes, a Notice of Unusual Event be declared. Contrary to the above, on October 22, 2009, Wolf Creek did not declare a Notice of Unusual Event until 5 hours after both emergency diesel generators were out of service. This issue is of very low safety significance (Green) because it is associated with failure to report a Notification of Unusual Event. Wolf Creek initiated Condition Report 21058 regarding the late declaration |
| 05000482/FIN-2009005-18 | 2009Q4 | Wolf Creek | Licensee-Identified Violation | On July 3, 2008, Wolf Creek submitted Licensee Event Report LER 2008006 which described missed VT-2 weld inspections when modifying train B containment spray recirculation line in Refueling Outage 16, requiring the train to be declared inoperable. This issue has been entered in to the corrective action program as Condition Report 2008-2197. Technical Specification 3.0.4, states, in part, that when a limiting condition of operation is not met, that mode changes shall only be made: when actions to be entered permit continued operation for an unlimited period of time, after a risk assessment, or when an allowance is stated in the specification. Technical Specification Limiting Condition of Operation 3.6.6 requires, in part, two operable trains of containment spray in Modes 1 through 4. Contrary to the above, on May 8, 2008, Wolf Creek entered Mode 4 with only one operable containment spray system. This issue is of very low safety significance (Green) because there was no loss of function of the containment spray system |
| 05000482/FIN-2010006-05 | 2010Q3 | Wolf Creek | Failure to Perform Adequate Evaluation for Significant Conditions | The inspectors identified a cited violation 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to perform an adequate evaluation to determine the cause of loss of offsite power induced water hammers and internal corrosion in the essential service water system and did not take corrective actions to preclude repetition of additional water hammer events and system leaks. Specifically, the licensee performed an apparent cause evaluation instead of a root cause evaluation as required, and the licensees evaluation did not consider metallurgical evaluations that were performed outside the corrective action program. The inspectors found that the licensee had not corrected a previous NCV 05000482/2009007-03, Failure to Correctly Screen ESW Piping Leaks for Significance, which resulted in the licensee failing to perform a root cause evaluation. Because the licensee failed to restore compliance within a reasonable time after NCV 05000482/2009007-03 was identified, this violation is being cited in a Notice of Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. The licensees corrective action to this cited violation was to initiate Condition Reports 27212, 26466, and 27075, to evaluate and correct the identified conditions, to start a root cause evaluation and, separately, to evaluate the licensees failure to properly respond to NCV 05000482/2009007-03. The issue was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions (P.1(c)) |
| 05000482/FIN-2011003-01 | 2011Q2 | Wolf Creek | No Procedure for Debris in Transformed and Tank Yards Prior to Severe Weather | The inspectors identified a noncited violation of Technical Specification 5.4.1.a, Administrative Procedures, for having no procedure to address onsite debris impacting plant equipment during severe weather. The inspectors walked down external areas of the plant on June 1 and June 9, 2011, prior to the onset of predicted severe thunderstorms and tornadoes. The inspectors found loose debris each time and brought it to the attention of the licensee who secured the materials. The inspectors walked down the transformer yard and tank yard during a thunderstorm on June 16 and found loose debris such as plywood, trash, wood planks, and fiberglass planks. The inspectors brought this to the attention of Wolf Creek and the materials were removed or secured. Wolf Creek initiated several condition reports but they only addressed immediate cleanup. Wolf Creek procedures had no steps for securing potential wind-driven projectiles prior to severe weather. After June 16, Wolf Creek wrote Condition Report 40573 which started a weekly maintenance activity to remove loose materials and added procedure steps to have operations walk down external areas prior to severe weather. This finding was more than minor because it impacted the protection against external factors attribute of the Initiating Events Cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, and determined that it was of very low safety significance (Green) for June 16, 2011, because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment would be unavailable since the reactor was shutdown. Inspectors used Manual Chapter 0609 Appendix G, Checklist 4 for the other occurrences because Wolf Creek was in Modes 4 or 5. The finding again screened to Green because it did not increase the likelihood of a loss of inventory, did not cause the loss of reactor coolant system instrumentation, did not degrade the ability of the licensee to terminate a leak path or add inventory when needed, or degrade the ability to recover residual heat removal if it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution, specifically the corrective action program attribute because licensees short-term corrective actions failed to ensure debris was secured or removed prior to severe weather |
| 05000482/FIN-2011003-02 | 2011Q2 | Wolf Creek | Failure to Properly Establish Clearance Order Boundary Isolation Resulting in Loss of Component Cooling Water Inventory | The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1a, Administrative Procedures, for a loss of component cooling water train B inventory caused by inadequate clearance order verification. Valve HBV110 was stuck in position and was partially open. When the clearance order was implemented, the operators concluded the valve was already closed. Subsequently, the valve created a leakage path which exceeded the surge tank makeup flow capacity and required manual isolation by the control room operators to protect safety-related components. Wolf Creek has taken corrective actions to include communication of expected as-found equipment positions in pre-job briefings and the clearance order template. This issue is captured in the corrective action program as Condition Reports 34505 and 40219. Failure to properly establish clearance order boundary isolation was a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance and human performance attributes of the Mitigating Systems Cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the finding was determined to be of very low safety significance because the finding did not result in the loss of operability or functionality of the component cooling water train or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors found that the finding had a cross-cutting aspect of work practices in the area of human performance associated with the communication of human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities |
| 05000482/FIN-2011003-03 | 2011Q2 | Wolf Creek | Failure to Assure Fillet Weld Met Size Requirements on Train B Charging Header Vent Line | The inspectors documented a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. Specifically, in October 2009, welders failed to ensure the fillet weld between the train B charging header and the half coupling used to attach two vent valves met the specified weld requirements. This weld failed in January 2011, rendering the train B charging system inoperable. The licensees extent of condition review identified 12 vent line welds which did not meet ASME code weld size requirements and/or procedural requirements for 2:1 weld taper configuration. Additionally, quality assurance inspectors failed to identify that the 2:1 taper weld requirements specified by procedure, and ASME minimum weld size requirements, were not met in multiple vent line welds. The weld was repaired and built up to the correct 2:1 aspect ratio. This issue was entered into the licensees corrective action program as Condition Reports 32648, 33686, 33689, and 36438. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The inspectors performed a Phase 1 screening in accordance with Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the issue did not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding had a cross-cutting aspect in the area of human performance, resources, because the licensee failed to ensure that personnel, specifically welders and quality assurance inspectors, were adequately trained in the procedural requirements and methods for measuring weld dimensions to assure nuclear safety. |
| 05000482/FIN-2011003-04 | 2011Q2 | Wolf Creek | Failure to Assure Separation of Stainless Steel and Carbon Steel Grinding and Cutting Equipment | The inspectors identified a noncited violation of 10 CFR Part 50 involving the failure of the licensee to ensure that weld preparation was protected from deleterious contamination in that drawers (located in the hot tool room) containing files, grinding wheels, flapper wheels, and cutting wheels, used for the purpose of weld preparation, contained a mixture of both stainless steel tools and carbon steel tools. The failure to separate tools used for stainless steel weld preparation from tools used for carbon steel preparation could result in the contamination of stainless steel welds by carbon steel and affect the material integrity and corrosion resistance. The licensee immediately removed the tools and replaced them with new tools stored separately for use on specific types of metal. This issue was entered into the licensees corrective action program as Condition Report 36444. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations, and if left uncorrected the finding would become a more significant safety concern. The inspectors performed a Phase 1 screening in accordance with Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the issue did not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding had a cross-cutting aspect in the area of human performance, resources, because the licensee did not provide complete, accurate, and up-to-date procedures for the preparation of stainless steel and carbon steel welds. |
| 05000482/FIN-2011003-05 | 2011Q2 | Wolf Creek | Failure to Assure Configuration Control of Safety-Related Systems | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure of the licensee to review the suitability of installing brass fittings and leaving test fittings on pressure, differential pressure, and flow transmitter equalizing block valve drain ports instead of the design specified stainless steel manifold plugs. During a boric acid walkdown, the inspectors identified that drain ports on the equalizing block of two separate reactor coolant system flow transmitters had brass fittings installed instead of the design specified stainless steel fittings. In response to inspector concerns about the brass fittings, the licensee subsequently discovered that a design configuration nonconformance existed by leaving the test fittings on the drain port during plant operation. Licensee Drawing J-17D22 specifies that manifold plugs be installed in the drain ports during plant operation. The licensee immediately replaced the brass caps with stainless steel fittings. This issue was entered into the licensees corrective action program as Condition Report 36439. The finding was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The inspectors performed a Phase 1 screening in accordance with Inspection Manual 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the issue would not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance, resources, because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns. |
| 05000482/FIN-2011003-06 | 2011Q2 | Wolf Creek | Inadequate Acceptance Criteria for Postmaintenance Testing of the Startup Feedwater Pump | The inspectors identified a finding involving the failure to follow the requirements of Procedure AP 16E-002, Post Maintenance Testing Development, for the startup feedwater pump. On November 4-6, 2010, Wolf Creek workers disassembled the startup feedwater pump for numerous preventive and corrective activities including removing the rotating element. On November 17, 2010, Wolf Creek conducted surveillance Procedure STN AE-007, Startup Main Feedwater Pump Operational Test, following reassembly. The only acceptance criteria listed in this procedure is that the motor-driven feedwater pump starts from the control room with no local operator action. The inspectors found this contrary to Procedure AP 16E-002, which requires acceptance criteria for a pump flow capacity test, vibration, bearing and lubrication temperatures, motor current, external leakage, and lubrication level be found satisfactory. This issue is captured in the corrective action program as Condition Report 39494. Wolf Creek issued a new work package to conduct a single-point pump capacity test and complete the required postmaintenance testing. Wolf Creek found, pending final review, that initial calculations show that the pump design is capable of enough flow to provide a heat sink in emergency operating procedures. Failure to follow Procedure AP 16E-002 for developing test criteria for plant equipment after the completion of maintenance activities is a performance deficiency. The finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the inspectors determined that the finding had very low safety significance (Green) because it did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, Wolf Creek created a testing procedure in response to a root cause evaluation, but did not consider acceptance criteria to ensure that the pump performs acceptably. |
| 05000482/FIN-2011003-07 | 2011Q2 | Wolf Creek | Failure to Correct Procedure for Opening Main Steam Isolation Valves | The inspectors identified a cited violation of Technical Specification 5.4.1.a, Administrative Procedures, involving Wolf Creeks failure to correct Procedure SYS AB-120 for main steam isolation valve operation. Specifically, between March 3, 2010, and March 19, 2011, Wolf Creek experienced repeat cases of safety-system actuations due to Procedure SYS AB-120 containing inadequate steps to establish conditions necessary to open a main steam isolation valve. Corrective actions were previously limited to steam header pressures below 300 psi. Wolf Creek commenced a root cause evaluation of the March 19, 2011, safety injection under Condition Report 34964. Due to Wolf Creeks failure to restore compliance from previous NCV 05000482/2010004-01 within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with the Enforcement Policy. Failure to correct deficiencies in Procedure SYS AB-120 for steam pressures above 300 psi was a performance deficiency. The inspectors determined that this finding was more than minor because it impacted the equipment performance attribute for the Initiating Events Cornerstone, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, this issue relates to the configuration control attribute for shut down equipment alignment. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609.04. Assuming worst case degradation, the finding resulted in exceeding the technical specification limit for reactor coolant system leakage due to the pressurizer power-operated relief valve cycling. Therefore, the inspectors screened the finding to a Phase 2 review by the senior reactor analyst. The senior reactor analyst used the Wolf Creek SPAR model and concluded that the incremental core damage probability was 3.7E-7 (Green). The inspectors found that the cause of the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program. Specifically, several evaluations failed to have an adequate extent of condition review and did not find that procedures were inadequate for opening a main steam isolation valve above 300 psi |
| 05000482/FIN-2011003-08 | 2011Q2 | Wolf Creek | Failure to Maintain Reactor Coolant System Pressure Below Relief Valve Setpoint | The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Administrative Procedures, for failure to follow procedural requirements to maintain reactor coolant system pressure below 350 psig. Control room operators increased charging flow at too great a rate with the reactor coolant system water-solid which caused the pressurizer power-operated relief valve to cycle three times over several minutes until adjustments to letdown could be made to reduce reactor coolant system pressure. Also, the letdown pressure controller was left in manual when automatic control would have lessened the pressure increase. Wolf Creek wrote Condition Report 35244 to correct the deficiency by changing several procedures for water-solid plant operations. The failure to maintain pressure below the power-operated relief valve setpoint was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone objective of configuration control to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of the finding was determined using Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, Checklist 2, and determined to be of very low safety significance (Green), because it did not cause the loss of mitigating capability of core heat removal, inventory control, power availability, containment control, or reactivity control. Additionally, the finding also did not cause any low temperature overpressure technical specifications to be exceeded. The inspectors found that the cause of the finding had a cross-cutting aspect in the area of human performance. Specifically, operators had to rely on skill of the craft when procedures should have supplied more instruction for manipulating charging and letdown with a water-solid plant. |
| 05000482/FIN-2011003-09 | 2011Q2 | Wolf Creek | Inadequate Fire Watch Defeats Halon Fire Suppression in Vital Switchgear Rooms During Fire | The inspectors reviewed a self-revealing noncited violation of License Condition 2.C.5 for failure to implement adequate fire watches which affected both trains of vital ac and dc switchgear. The inadequate fire watches occurred during an actual fire which negated the Halon system discharge because internal fire doors were not shut, as required, by the fire watch. The inspectors found problems with fire impairments and watches from 2008 that had not been corrected. Subsequent to the fire, Wolf Creek again briefed and trained its personnel on the requirements for fire watches. This issue is captured in the corrective action program as Condition Report 36719. Failure to implement adequate fire impairments such that the fire watches ensured the success of the Halon system was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone and its objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the protection against external factors attribute was impacted by the fire impairment. To determine significance, the inspectors used Inspection Manual Chapter 0609.04 to screen the finding to Inspection Manual Chapter 0609, Appendix F, because the fire protection defense-in-depth strategies involving automatic suppression, fire barriers, and administrative controls were degraded. The senior reactor analyst conducted a Phase 3 review of this finding and concluded that the incremental core damage frequency was 1.6E-8 per year, or very low safety significance (Green). The inspectors found that the cause of the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, corrective actions from ineffective fire watches in 2008 did not prevent recurrence of the inadequate fire watch on April 5, 2011. |
| 05000482/FIN-2011003-10 | 2011Q2 | Wolf Creek | Failure to Analyze for Vortexing in Containment Spray Additive Tank | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate the design basis into instructions, procedures, and drawings. The inspectors found that the licensee failed to assess whether vortexing occurred in the containment spray additive tank in the event of a design-basis accident. Wolf Creek entered this issue in the corrective action program as Condition Report 38715. Failure to implement design control measures to analyze whether containment spray piping remained full of water was a performance deficiency. This finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of the containment spray system to respond to initiating events and prevent undesirable consequences. Specifically, the inspectors had reasonable doubt on the capability of the containment spray system to properly inject because of vortexing in the containment spray additive tank. The inspectors performed the significance determination using Inspection Manual Chapter 0609.04. The finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Although the failure to have this calculation had existed since original construction, the inspectors determined this finding reflected current performance since the licensee was required to evaluate likelihood of tanks allowing gas intrusion into the emergency core cooling systems in response to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems. Consequently, this finding had problem identification and resolution cross-cutting aspects associated with the corrective action program in that the licensee did not thoroughly evaluate the potential for gas intrusion from all possible tanks |
| 05000482/FIN-2011003-11 | 2011Q2 | Wolf Creek | Licensee-Identified Violation | Title 10 CFR 50.54(hh)(2)(ii) states: Each licensee shall develop and implement guidance and strategies intended to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities under the circumstances associated with loss of large areas of the plant due to explosions or fire, to include strategies in the following area of operations to mitigate fuel damage. On April 13, 2011, while performing procedure reviews as part of industry-wide self-assessments in response to the core damage events at Fukushima Daiichi, Wolf Creek engineers identified two instances of mitigating strategy procedures which did not contain sufficient information to accomplish those strategies successfully. The first example was the ability to refill the refueling water storage tank, and the second example involved flashing the diesel generator field using alternate dc sources. These issues were documented in the licensees corrective action program as Condition Report 37374. The inspectors evaluated these findings under Inspection Manual Chapter 0609, Appendix L, and determined these findings to be of very low safety significance because the findings did not involve unrecoverable unavailability of multiple mitigating strategies such that spent fuel pool cooling, injection to the reactor vessel, or injection to steam generators cannot occur, or unrecoverable unavailability of on-site, self-powered, portable pumping capability, or substantial inability to perform command and control enhancements. |
| 05000482/FIN-2011003-12 | 2011Q2 | Wolf Creek | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in the applicable design documents. On May 13, 2011, Wolf Creek identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, test control for stroking residual heat removal containment sump valve 8811B prior to its as-found diagnostic test. Wolf Creek stroked the valve for a clearance order and as such, preconditioned the valve prior to its test. Plant computer data from this stroke, data from the diagnostic stroke, and valve disassembly showed no deficiencies. Using Inspection Manual Chapter 0609.04, the inspectors determined the finding to be of very low safety significance because it was confirmed not to result in the loss of operability or functionality. This issue is captured in Condition Report 37244. |
| 05000482/FIN-2015001-01 | 2015Q1 | Wolf Creek | Question Related to Ultrasonic Examination of Reactor Vessel Flange Stud Hole Threads | The inspector identified an unresolved item pertaining to 10 CFR 50 Appendix B, Criterion IX, Control of Special Processes, associated with the licensees method of performing ultrasonic examination of the reactor vessel flange stud hole threads in accordance with applicable American Society of Mechanical Engineers (ASME) Code requirements. The inspector identified several issues of concern while observing the licensees ultrasonic examination of the reactor vessel flange stud hole threads. The inspector questioned whether the licensee would be able to detect any reportable indication within the ASME Code examination zone using the technique employed. The inspector identified that in 2003 the licensee had modified the method used to perform the examination scanning, but never verified that the new methodology was capable of detecting relevant indications within the examination zone. The new method placed the one inch diameter zero angle transducer on a radial arm at the end of an approximately 30 foot pole. The pole is aligned on the handle of the protective cap that covers the stud hole in the flange. The inspector reviewed examination Procedure UT-11, Ultrasonic Stud Hole Threads, Revision 13, and Examination of Reactor Vessel Flange made note of the following: - The inner edge of the transducer is at a nominal distance of 3.875 inches from the center of the stud hole. - The protective cap has a nominal diameter of 7.25 inches or a radius of 3.625 inches while the stud hole diameter is 6.822 +0, -.01 inches. - This places the examination zone of inspection starting at a radius of approximately 3.411 inches and extending to a radius of 4.411 inches. The configuration of the transducer on the pole and the alignment mechanism results in the inside edge of the transducer being placed approximately 0.465 inches from the edge of the stud hole, which is the start of the one inch examination area. Because the technique employs a "zero" angle transducer and the examination area is not directly beneath the transducer, there is a concern with instrument signal coverage. The inspector also identified several procedural compliance issues while reviewing the licensees implementation of UT-11. The inspector questioned the following statements in the procedure: - Procedure UT-11, Section 11.1.1, states in part, The examination volume is a one inch annular band around each stud hole, extending to one stud diameter into the flange. - Procedure UT-11, Section 11.2.2, states in part that, Straight beam examination of ligaments shall be performed. - Procedure UT-11, Section 12.1.1, states in part, All indications which are found that are orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 in, as measured radially from the root of the thread, shall be reported to the LMT Site Supervisor and recorded on the Ultrasonic Examination report form. There is a concern that the technique currently being utilized by the licensee may not provide adequate coverage of the required examination area and may not be capable of detecting indications orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 inch, as measured radially from the root of the thread, as required by the licensee's procedure and Section XI of the ASME Code. Additional analysis and simulations need to be completed to determine if the licensee is meeting ASME Code requirements. This issue is being tracked as URI 05000482/2015001-01, Questions Related to Ultrasonic Examination of Reactor Vessel Flange Stud Hole Threads. |
| 05000482/FIN-2015001-02 | 2015Q1 | Wolf Creek | Failure to Assess the Operability of Emergency Diesel Generator B during Emergent Work Activities | The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the failure to properly preplan maintenance such that it would not affect safety-related equipment in accordance with procedure AP 22C-008, On-Line Qualitative Risk Management, Revision 3. Specifically, during planning of emergent work activities on January 29, 2015, the licensee failed to recognize that when electrical cabinet doors containing safety-related under voltage and under frequency relays were opened to accomplish troubleshooting activities, the cabinet was not in a seismically qualified configuration. Thus the maintenance had the potential to impact the reliable operation of emergency diesel generator B during a seismic event. The licensee initiated Standing Order 37, Safety Related Cabinet Operability Requirements, Revision 0, to provide the requirements for assessing operability of opening safety-related electrical cabinet and panel doors out of their seismically qualified configuration during maintenance activities and entered this issue into their corrective action program for resolution as Condition Reports 91501 and 94605. The licensees failure to properly preplan maintenance such that it would not affect safety-related equipment during emergent work activities was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Specifically, the licensees failure to properly preplan maintenance resulted in emergency diesel generator B being placed in a condition that did not meet its seismic design requirements. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and 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 human performance associated with work management. Specifically, the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work (H.5). |
| 05000482/FIN-2015001-03 | 2015Q1 | Wolf Creek | Failure to Complete an Adequate Operability Evaluation for Declaring the Train A Control Room Air Conditioning Unit Operable | The inspectors identified non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to complete an adequate operability evaluation in accordance with procedure AP-28001,Opeability Evaluations, Revision 24 following the failure to meet a surveillance test acceptance criteria. Specifically, the licensee did not have an accurate technical basis for declaring the train A control room air condition unit operable when the minimum air flow rate was not met. The licensees operability evaluation, which declared the train A control room air condition unit operable, incorrectly applied instrument uncertainty and used a superseded minimum air flow value. When these inaccuracies were addressed, the licensee determined the train was inoperable. The licensee entered this issue into their corrective action program as Condition Report 92274. The licensees use of an inadequate technical basis for an operability evaluation of a non-conforming condition resulting in the train A control room air conditioning air condition unit being declared operable when it was actually inoperable was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and 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 human performance associated with conservative bias component because the licensee did not use a decision making-practice that emphasized prudent choices over those that are simply allowable. A proposed action was determined to be safe in order to proceed, rather than unsafe in order to stop (H.14). |
| 05000482/FIN-2015001-04 | 2015Q1 | Wolf Creek | Failure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust System Boundary Door | The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, associated with the licensees failure to follow the requirements of Station Procedure AP 10-104, Breach Authorization, Revision 32. Specifically, the licensees failure initiate a breach permit and station a boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was opened multiple times for transporting scaffolding from the turbine building to the auxiliary building. Opening this door without compensatory measures rendered the auxiliary building emergency exhaust system inoperable. The license entered this issue into their corrective action program for resolution as Condition Reports 92315 and 92630. The licensees failure to initiate a breach permit and implement required compensatory measures for when the auxiliary building emergency exhaust system boundary door 41015 was open was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier functionality of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door required entry of Technical Specification 3.7.13 Limited Condition of Operation Condition B. The longest period door 41015 was open was approximately one hour without the required compensatory measure. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because the finding only involved a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work (H.5). |
| 05000482/FIN-2015001-05 | 2015Q1 | Wolf Creek | Licensee-Identified Violation | Technical Specification Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, February 1978. Section 1.c of Regulatory Guide 1.33 requires procedures for equipment control (e.g. locking and tagging). Station Procedure AP 21E-001, Clearance Orders, Revision 37, requires that the shift manager, ensure that plant conditions can support establishing the clearance order boundaries, including activities such as removing equipment from service. Contrary to the above, on January 28, 2015, the licensee failed to ensure that plant conditions could support the clearance order boundaries during preparation and implementation of clearance orders. Specifically, the preparation and implementation of clearance order EJ-A-005 unintentionally rendered both trains of the residual heat removal system inoperable and necessitated an unplanned entry into Technical Specification 3.0.3 for 2 hours. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e. core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined a detail risk evaluation was required because this finding represented a loss of system and/or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst noted that the isolation of valve EJ HV8716A would only affect the reliability of hot leg injection for train B. Hot leg injection is a necessary function to ensure that there will not be unacceptably high concentrations of boric acid in the core region (resulting in precipitation of a solid phase) during the long-term cooling phase following a postulated large-break loss of coolant accident. Consequently, valve alignments affecting hot leg injection are only of concern during large-break loss of coolant accidents. Using the simplified plant analysis risk model, the analyst noted that the frequency of a large-break loss of coolant accident (LLOCA) was 2.5 x 10-6 /year. As stated above, the exposure period was two hours or 2.28 x 10-4 years. The analyst then calculated the upper bound risk impact of the performance deficiency to be 5.7 x 10-10. Therefore, this finding is of very low safety significance (Green). |