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Significance:
Jun 21, 2008
Identified By: NRC
Item Type: FIN Finding
Failure to Follow Work Control Program Procedures
A self-revealing Green finding was identified associated with the licensee’s failure to follow Administrative Procedure 0.40, “Work Control Program,” requirements that would have ensured the 4160 V bus breaker fuse maintenance would not trip reactor recirculation pump “B”. Implementation of inadequate maintenance instructions resulted in an unexpected trip of the reactor recirculation pump B and an unplanned reduction in reactor power. Specifically, the licensee failed to perform a thorough review of the electrical drawings required to fully understand the consequences of pulling fuses for maintenance in the 4160 V breaker cubicle. The licensee entered this issue into their corrective action program as CR-CNS-2008-4400.
This finding was more than minor since it affected the Reactor Safety Initiating Events cornerstone attribute of procedure quality and resulted in a plant transient. It was considered to be of very low safety significance since it did not contribute to both the likelihood of a plant trip and the likelihood that mitigation equipment or functions will not be available. The cause of this finding is related to the human performance crosscutting component of work practices because personnel did effectively use self-checking techniques while determining the plant impact of the proposed NN fuse removal (H.4(a)).
Inspection Report# : 2008003 (pdf)
Significance: Dec 31, 2007
Identified By: Self-Revealing
Item Type: NCV NonCited Violation
Inadequate Tagout Procedure Results in Inadvertent Stroke of Motor Operated Valve
A self-revealing noncited violation of Technical Specification 5.4.1.a, “Written Procedures,” was identified involving an inadequate procedure for controlling work on energized circuits. Specifically, inadequate procedural guidance in Administrative Procedure 0.9, “Tagout,” allowed power to be restored to the control logic for residual heat removal injection valve RHR-MOV-27A while personnel were performing maintenance on the valve. This condition created a personnel hazard and resulted in the inadvertent opening of injection valve RHR-MOV-25A due to interlock logic with valve RHR-MOV-27A being satisfied. This issue was entered into the licensee's corrective action program as Condition Report CR-CNS-2007-06844.
The finding is more than minor because it affects the equipment performance attribute of the initiating events cornerstone, and affects 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. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because the performance deficiency did not result in a condition that could have resulted in exceeding the Technical Specification limit for any RCS leakage or could have likely affected other mitigating systems causing a total loss of safety function. The cause of this finding is related to the human performance cross cutting component of work control in that the licensee did not appropriately coordinate work activities by incorporating guidance to consider the impact of changes to the work scope on other maintenance that was in progress (H.3(b)).
Inspection Report# : 2007005 (pdf)
Significance: SL-IV Jun 21, 2008
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Notify the NRC of the Inability to Meet ASME Code Requirements
The inspectors identified a noncited violation of 10 CFR 50.55a(g)(5)(iii) for the licensee’s failure to notify the NRC of the inability to meet the requirements of the American Society of Mechanical Engineers Code for Class 1 and 2 welds performed during Refueling Outage 22 in February 2005. Specifically, on April 21, 2008, the inspectors identified that welds associated with design changes to the reactor feedwater and high pressure core injection systems performed during Refueling Outage 22 did not meet the 90 percent total area coverage, required by ASME Section XI, during the performance of the pre service inspection ultrasonic testing. The licensee failed to notify the NRC of the inability to meet the ASME Code requirements within 12 months from the end of the third 10 year Inservice Inspection interval as required by 10 CFR 50.55a(g)(5)(iv).
The failure to notify the NRC constituted a performance deficiency of 10CFR 50.55a(g)(5)(iii). In accordance with Manual Chapter 0612, Appendix B, Section 2, this finding has the potential to impact the NRC’s ability to perform its regulatory function since the licensee did not notify the NRC within the designated time period, and as a result, impeded the NRC’s ability to evaluate and decide on the potential ASME code relief in a timely manner. This finding is greater than minor because it is associated with the mitigating system cornerstone, in that the licensee failed to ensure the reliability of safety-related equipment due to the failure to meet ASME Code requirements for the Class 1 and 2 system pressure boundary welds since February 2005. In accordance with Supplement 1 of the Enforcement Policy, the violation was characterized as Severity Level IV because it involved a failure to meet regulatory requirements that have more than minor safety significance. This finding is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy, due to the NRC review and acceptance of the licensee’s fourth cycle Risk Informed - Inservice Inspection program which no longer requires these welds to be periodically inspected in accordance with ASME Code, Section XI.
Inspection Report# : 2008003 (pdf)
Significance: Jun 21, 2008
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Properly Manage Elevated Risk
The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) with three examples regarding the licensee’s failure to manage the increase in risk that resulted from maintenance activities. Specifically, the licensee did not post protected equipment signs on risk sensitive equipment during periods of elevated risk as required by Administrative Procedure 0-PROTECT-EQP. This issue was entered into the licensee’s corrective action program as Condition Report CR-CNS-2008-03555.
The finding is more than minor because it is related to the licensee’s failure to implement prescribed significant compensatory measures. The inspectors reviewed Manual Chapter 0612, Appendix E, example 7.g and determined that it was not applicable to this situation due to the fact that the licensee does not maintain a shutdown probabilistic risk analysis model, and as such an incremental core damage probability cannot be estimated for the plant conditions that existed at the time of the performance deficiency. For the same reason, the inspectors determined that Manual Chapter 0609 Appendix K, “Maintenance Risk Assessment and Risk Management Significance Determination Process,” could not be used to determine the risk significance the finding. Using the qualitative review process of Manual Chapter 0609 Appendix M, “Significance Determination Process Using Qualitative Criteria,” the inspectors determined that this issue was of very low safety significance because it did not result in any additional loss of defense in depth systems. The cause of this finding is related to the human performance cross cutting component of work practices because licensee personnel did not follow the requirements of Procedure 0-PROTECT-EQP (H.4(b)).
Inspection Report# : 2008003 (pdf)
Significance: Jun 21, 2008
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Construct Drywell Shielding According to Design Documents
The inspectors identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion V, “Instructions, Procedures, and Drawings,” regarding the licensee’s failure to build a permanent drywell shielding system in accordance with the approved design documents. During a pre-startup inspection of the drywell, inspectors discovered numerous assembly errors and unevaluated piping interactions with safety-related piping in the as-left configuration. This issue was entered into the licensee’s corrective action program as Condition Report CR-CNS-2008-05208.
The finding is more than minor because it was associated with the design control attribute of the mitigating systems cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Manual Chapter 0609 Appendix G, “Shutdown Operations Significance Determination Process,” flowchart, the inspectors determined that the finding is of very low safety significance because it did not result in the loss of any mitigation capability identified in the Manual Chapter 0609 Appendix G, Attachment 1 worksheet and that no qualitative risk assessment is required. The cause of this finding is related to the human performance cross cutting component of work practices because licensee personnel provided inadequate management oversight of contractors erecting permanent scaffolding in the drywell (H.4(c)).
Inspection Report# : 2008003 (pdf)
Significance: Mar 22, 2008
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Follow Scaffold Inspection Procedures
The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a regarding the licensee’s failure to follow the requirements of Maintenance Procedure 7.0.7, “Scaffolding Construction and Control.” Specifically, licensee personnel failed to inspect all existing scaffolds and failed to identify multiple scaffolding interactions with safety-related equipment during a required annual scaffold inspection on January 21, 2008. This issue was entered Into the licensee’s corrective action program as Condition Report CR CNS 2008-01576.
The finding is more than minor because if left uncorrected the failure to perform annual scaffold inspections could become a more significant safety concern. Specifically, annual inspections failed to inspect all existing scaffolds and failed to identify multiple scaffolding interactions with safety-related equipment. Using the Manual Chapter 0609, “Significance Determination Process,” Phase 1 Worksheet, the finding is determined to have a very low safety significance because it did not result in the loss of function of a Technical Specification required system for greater than its allowed outage time. The cause of this finding is related to the human performance crosscutting component of work practices because maintenance personnel did not follow the requirements of Maintenance Procedure 7.0.7 (H.4(b))
Inspection Report# : 2008002 (pdf)
Significance:
Mar 22, 2008
Identified By: NRC
Item Type: VIO Violation
Failure to Establish Adequate Procedures for Maintenance of Emergency DG Electrical Connections
Two examples of a self-revealing apparent violation of Technical Specification 5.4.1.a were identified regarding the licensee’s failure to establish procedural controls for maintenance of electrical connections on essential equipment. In the first example, the licensee failed to include amphenol connections within the scope of existing periodic electrical connection inspections to identify loosening connections. In the second example, the licensee failed to incorporate internal operating experience into work control procedures to ensure that diesel generator-mounted amphenol connections were solidly attached following maintenance. These failures to establish adequate procedural controls led to the trip of Diesel Generator 2 during testing on January 15, 2008. This issue was entered into the licensee’s corrective action program as Condition Report CR CNS 2008 00304.
The finding affected the mitigating systems cornerstone and is more than minor because it is associated with the cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The Phase 1 worksheets in Inspection Manual Chapter 0609, "Significance Determination Process," were used to conclude that a Phase 2 analysis was required because the finding represents an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time (7 days). A Phase 2 risk analysis was conducted using the guidance of Manual Chapter 0609, Appendix A, “Determining the Significance of Reactor Inspection Findings for At-Power Situations.” Entering the site-specific pre-solved table with an assumed exposure time of greater than 30 days yielded a result of red, or very high significance. A Phase 3 analysis conducted by a risk analyst preliminarily determined the finding to be of white, or low to moderate significance. The cause of the finding is related to the corrective action component of the crosscutting area of problem identification and resolution in that the licensee failed to take appropriate corrective actions for a 2007 NRC inspection finding which identified inadequate maintenance procedures for checking the tightness of diesel generator electrical connections (P.1(d))
Inspection Report# : 2008002 (pdf)
Significance: Mar 18, 2008
Identified By: NRC
Item Type: VIO Violation
Failure to ensure that some steps contained in emergency procedures would work as written
White. A violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for failure to ensure that some steps contained in emergency procedures at Cooper Nuclear Station would work as written. Inspectors identified that steps in Emergency Procedures 5.4POST-FIRE, "Post-Fire Operational Information," and 5.4FIRE-S/D. "Fire Induced Shutdown From Outside Control Room," intended to reposition motor-operated valves locally, would not have worked as written because the steps were not appropriate for the configuration of the motor-starter circuits. This condition existed between 2004 and June, 2007. Appendix B to 10 CFR 50, Criterion V, was not met because these quality-related procedures would not work to allow operators to bring the plant to a safe shutdown condition in the event of certain fires. This finding had a cross-cutting aspect in Problem Identification and Resolution, under the Corrective Action attribute, because the licensee did not thoroughly evaluate the 2004 NRC violation to address causes and extent of condition (P.1.c - Evaluations).
This finding is of greater than minor safety significance because it impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences. This finding affected both the procedure quality and protection against external factors (fires) attributes of this cornerstone objective. This finding was determined to have a White safety significance during a Phase 3 evaluation. The scenarios of concern involve larger fires in specific areas of the plant which trigger operators to implement fire response procedures to place the plant in a safe shutdown condition. Since some of those actions could not be completed using procedures as written, this would challenge the operators' ability to establish adequate core cooling.
Inspection Report# : 2008008 (pdf)
Significance: Dec 31, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Inaccurate Assumptions in Control Room Flooding Calculation
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, “Design Control,” involving the failure to verify the adequacy of input assumptions to a design basis calculation. Specifically, a design basis control room flooding analysis assumed operators could terminate a turbine equipment cooling system pipe leak in the control room within 30 minutes when it is not possible to do so. This issue was entered into the licensee's corrective action program as Condition Report CR-CNS-2007-07708.
The finding is more than minor because it affects the design control attribute of the mitigating systems cornerstone, and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix M, “Significance Determination Process Using Qualitative Criteria,” a bounding quantitative analysis was performed resulting in the determination that the finding was of very low safety significance.
Inspection Report# : 2007005 (pdf)
Significance: Dec 31, 2007
Identified By: Self-Revealing
Item Type: NCV NonCited Violation
Inadequate PMT Results in Inoperable Emergency Diesel Generator
A self-revealing noncited violation of Technical Specification 5.4.1. a, “Written Procedures,” was identified because the licensee failed to establish an adequate postmaintenance test procedure to verify component performance following maintenance. Specifically, the licensee’s postmaintenance test instructions were inadequate to verify an essential shutoff function of the Diesel Generator 1 day tank float valve following replacement on August 28, 2007. This issue was entered into the licensee's corrective action program as Condition Report CR-CNS-2007-07594.
The finding is more than minor because it affects the procedure quality attribute of the mitigating systems cornerstone, and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Though the failure of the float valve did impact operability of the diesel generator it would not have prevented the diesel generator, from starting and loading in response to an accident. Using the Manual Chapter 0609, Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not represent a loss of safety system function. The cause of this finding is related to the human performance crosscutting component of resources in that the licensee’s postmaintenance test procedure was inadequate to verify the essential shutoff function of the float valve [H.2(c)].
Inspection Report# : 2007005 (pdf)
Significance: Dec 31, 2007
Identified By: Self-Revealing
Item Type: NCV NonCited Violation
Procedure Noncompliance Causes Reactor Equipment Cooling System Leakage
A self-revealing noncited violation of Technical Specification 5.4.1. a, “Written Procedures,” was identified for the failure of maintenance personnel to follow procedures. Specifically, maintenance personnel failed to follow site administrative procedures that require verification of component identification prior to starting work. This resulted in maintenance personnel inadvertently attempting to remove a relief valve associated with the reactor equipment cooling system instead of the fuel pool cooling system. This error was identified while maintenance personnel were removing the wrong relief valve and an unexpected leak occurred. This issue was entered into the licensee's corrective action program as Condition Report CR-CNS-2007-07519.
The finding is more than minor because it affects the configuration control attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 Screening Worksheet, the issue screened as having very low safety significance because the maintenance personnel immediately restored the system integrity on noting the system leakage so that this did not represent a loss of safety system function. The cause of this finding is related to the human performance crosscutting component of work practices because maintenance personnel failed to implement an expected human error prevention technique [H.4(a)] .
Inspection Report# : 2007005 (pdf)
Significance: Nov 28, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to correct battery surveillance requirements
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for failure to correct a nonconservative technical specification. The licensee determined on March 14, 2007 that Technical Specification Surveillance Requirements 3.8.4.2 and 3.8.4.5 were nonconservative, but did not initiate any corrective action to address the degraded condition. The licensee determined that these surveillance requirements were nonconservative with respect to safety related 125 Vdc battery intercell resistance measurements.
The failure to correct an inadequate technical specification surveillance requirement is a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of procedure quality and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable circumstances (i.e., core damage). Using the Manual Chapter 0609, ASignificance Determination Process,@ Phase 1 Worksheet, the finding is determined to have a very low safety significance because it did not result in the loss of a mitigating system safety function. This finding has a cross-cutting aspect in the corrective action program component of the cross-cutting area of problem identification and resolution because the licensee did not take appropriate corrective action to address a condition adverse to quality (P.1(d)) (Section 4OA2).
Inspection Report# : 2007006 (pdf)
Significance: Nov 02, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to consider vortexing and available net positive suction head impact on the emergency diesel fuel oil system
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for the failure to verify the adequacy of design for the emergency diesel generator fuel oil system. Specifically, the licensee did not complete the necessary vortexing and net positive suction head calculations on the emergency diesel generator fuel oil storage tank and associated transfer pumps, and the fuel oil day tanks and associated booster pumps. These calculations were required to establish that adequate design margins exist to demonstrate air entrainment or cavitation does not occur during the mission time for these pumps. This finding was entered into the corrective action program under Condition Reports CNS-2007-07421 (fuel oil storage tank) and CNS-2007-07585 (fuel oil day tank).
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of “Design Control.” It impacts the cornerstone objective to ensure the availability, reliability, and capability of the emergency diesel generator system to respond to initiating events and prevent undesirable consequences. Using Manual Chapter 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At- Power Situations," Phase 1 screening, this issue was determined to be of very low safety significance (Green) because it was determined that there was no loss of safety function. This finding has cross cutting aspects in the area of problem identification and resolution, with the Operating Experience attribute [P.2(b)]. The licensee failed to evaluate and apply various industry events associated with safety-related storage tanks vortexing into station design basis calculations.
Inspection Report# : 2007011 (pdf)
Significance: Nov 02, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Installation of essential electrical cable with inadequate fault current ratings and not in accordance with original design basis requirements
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for a design change, associated with the emergency diesel generator, that failed to be subjected to control measures commensurate with those applied to the original design. Specifically, a design change installed an emergency diesel generator feeder cable that could fail prior to protective device actuation on postulated asymmetrical short-circuit current values. This issue was entered into the corrective action program under Condition Report CNS-2007-07409.
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of “Design Control.” It impacts the cornerstone objective to ensure the availability, reliability, and capability of the emergency diesel generator system to respond to initiating events and prevent undesirable consequences. Using Manual Chapter 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At- Power Situations," this issue screened as having very low safety significance (Green) during a Phase 1 review because the condition did not represent a loss of system safety function.
Inspection Report# : 2007011 (pdf)
Significance: Nov 02, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to ensure design basis information remains consistent within affected design documents
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance, for the failure to correctly translate the emergency core cooling system design basis into instructions, procedures, and drawings. Specifically, the licensee failed to ensure design bases information was consistent within affected design documents. The licensee failed to identify that Calculation NEDC 91-078, "System Level Design Basis Review of High Pressure Coolant Injection (HPCI) System Program MOVs," and Design Calculation NEDC 98-001, "Vortex Limit for the Emergency Condensate Storage Tanks A & B," were documents that affected each other. This issue was documented in the licensee’s corrective action program as Condition Report CNS-2007-07459.
The finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of “Design Control.” It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At- Power Situations," this issue screened as having very low safety significance (Green) during a Phase 1 review because these deficiencies were determined not to result in loss of system safety function.
Inspection Report# : 2007011 (pdf)
Significance: Nov 02, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to comply with design control program requirements
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for the failure to correctly translate the design basis into specifications, drawings, procedures, and instructions. Specifically, the design criteria documents were defined as being controlled documents that provided the criteria, requirements, and bases for safety-related/important-to-safety portions of Cooper Nuclear Station. Procedure 3.32 and the related series procedures specified certain types of information to be included in the design criteria documents (i.e., logic diagrams or system templates containing system safety objectives, functional and design criteria requirements, components and parameters essential to the ability of the system to achieve its required safety functions; four different configuration matrices used to validate that current plant configuration is consistent with the design basis criteria; and various appendices, including an acceptance criteria appendix for each component, sub-system and system). The team noted during review of the design criteria documents that much of this required information was not being maintained. These issues were documented in the licensee’s corrective action program as Condition Reports CNS-2007-07461 and CNS 2007-07608.
This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of “Design Control.” It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding screened as very low safety significance (Green) because it was a design control deficiency confirmed not to have resulted in loss of safety function. A crosscutting aspect was identified involving the human performance component area for resources to ensure that design documentation is complete, accurate, and up-to-date (H.2(c)).
Inspection Report# : 2007011 (pdf)
Significance: Nov 02, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Inadequate corrective actions associated with multiple workmanship issues on safety related valves
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," having very low safety significance for the failure to adequately evaluate the extent of equipment failures resulting from workmanship issues, and to determine the causes and corrective actions for this significant condition adverse to quality to prevent recurrence. During Refueling Outage 23, multiple examples of workmanship issues were identified that resulted in safety related valve failures discovered during post-maintenance testing. Subsequent to the implementation of corrective actions to address this issue, a directly related workmanship condition was identified involving Safety-Related Valve HPCI-MOV-MO16. This valve was returned to service, for approximately 10 months, before identifying that a nonconforming condition involving workmanship existed that required correction prior to returning the valve to service. The licensee entered this condition into their corrective action program as Condition Report CNS-2007-07609.
This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to evaluate the extent of condition for the valves, which were potentially affected, and to determine the causes for the multiple workmanship issues. Using Manual Chapter 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," this issue screened as having very low safety significance during a Phase 1 review because the valve workmanship issues were corrected prior to returning to service with the exception of one valve, which was determined to be functional in the nonconforming condition. The cause of this finding had crosscutting aspects associated with problem identification and resolution, related to the Corrective Action Program attribute [P.1.(c)], for thoroughly evaluating problems. The resolutions address causes and extent of conditions, as necessary.
Inspection Report# : 2007011 (pdf)
Significance: Nov 02, 2007
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to comply with the requirements of 10CFR50.71.e and to assure the updated safety analysis report has the latest information developed
The team identified a noncited Severity Level IV violation for the failure to comply with the requirements of 10 CFR 50.71(e). The correct value for the automatic depressurization system accumulator minimum pressure was not used to revise the Updated Safety Analysis Report. Specifically, the licensee’s technical specifications and Design Calculation NEDC 88-306 require a minimum of 88 psig to assure five actuations of the safety relief valves with the drywell at atmospheric conditions. The Updated Safety Analysis Report lists a minimum pressure of 68.6 psig for this function. The Updated Safety Analysis Report stated pressure of 68.6 psig was incorporated as part of the licensee’s Updated Safety Analysis Report rebase line project and became effective on March 10, 2000. The licensee was unable to provide a basis for the lower pressure stated in the Updated Safety Analysis Report.
This violation was subject to traditional enforcement because it had the potential to impact the regulatory process. This finding is considered more than minor because use of this lower pressure value could render the automatic depressurization feature incapable of performing its design function. In accordance with NRC Enforcement Policy, the NRC has concluded that this is a Severity Level IV violation. Because this violation was of very low safety significance, was not repetitive or willful, and it was entered into the licensee’s corrective action program as Condition Report CNS-2007-07468, this violation is being treated as an noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2007011 (pdf)
Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Significance: N/A Nov 28, 2007
Identified By: NRC
Item Type: FIN Finding
Assessment of the licensee's problem identification and resolution program
The team reviewed approximately 208 risk significant issues, apparent and root cause analyses, and other related documents, to assess the effectiveness of the licensee=s problem identification and resolution processes and systems. The team concluded that the licensee=s management systems were effective, although several examples (historical and current) of failure to implement appropriate and timely corrective actions existed, especially early in the assessment period. But, overall, corrective actions were appropriate to the circumstances. The licensee implemented an effective program for evaluating operational experience. However, three examples existed where ineffective use of operating experience contributed to issues. The licensee overall performed effective and critical self assessments.
The team concluded that the licensee maintained an overall safety-conscious work environment. An increasing trend in anonymous condition reports written was being addressed by the licensee to ensure that issues affecting the safety conscious work environment did not exist. In addition, the team (as well as a licensee self-assessment) received isolated comments that it was easier to quickly repair items upon identification, rather than entering the items into the corrective action program. Plant personnel interviewed generally considered the employee concerns program a viable option to pursue safety issues. However, the team received isolated comments that individuals lacked confidence in the ability of the employee concerns program to resolve issues.
Inspection Report# : 2007006 (pdf)
Last modified : November 26, 2008