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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000390/FIN-2010201-01 | 2010Q3 | Green | NRC identified | Security | ||
| 05000390/FIN-2010201-02 | 2010Q3 | Green | NRC identified | Security | ||
| 05000390/FIN-2010201-03 | 2010Q3 | Green | NRC identified | Security | ||
| 05000390/FIN-2010201-04 | 2010Q3 | Green | NRC identified | Security | ||
| 05000390/FIN-2010402-01 | 2010Q2 | Green | NRC identified | Security | ||
| 05000390/FIN-2011002-01 | 2011Q1 | Green | H.5 | NRC identified | Failure to Establish Adequate Compensatory Measures for a Degraded Sprinkler system in Accordance with the Approved Fire Protection Plan | The inspectors identified a NCV of the Unit 1 Operating License Condition 2.F for the licensees failure to establish adequate compensatory measures for an impaired sprinkler system in accordance with the approved Fire Protection Plan (FPP). Specifically, a solid scaffold platform had been erected below the preaction sprinkler system protecting the 1B Charging Pump oil system without establishing compensatory measures, as required by the FPP. As a result, the capabilities of the sprinkler system protecting the 1B Charging Pump oil system were impaired from performing their designed function. The licensee entered this issue into the corrective action program as PER 332853. The finding was determined to be more than minor because it affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone, in that it impacted automatic fire suppression capability and affected the cornerstone objective of ensuring the availability of systems that respond to external events. This finding was evaluated using IMC 0609, Appendix F, Attachment 1, and was determined to be of very low safety significance because it represented a low degradation of the fixed fire suppression systems. The cause of the finding had a cross-cutting aspect in the area of human performance associated with the work control component. It was directly related to the licensees not planning and coordinating work activities consistent with nuclear safety to ensure that adequate compensatory actions were established for a degraded sprinkler system (H.3 (a)). |
| 05000390/FIN-2011002-02 | 2011Q1 | Green | NRC identified | Failure to Ensure Adequacy of AOI-13, Loss of Essential Raw Cooling Water, Revision 0038 | The inspectors identified a NCV of Technical Specification 5.7.1, Procedures, for the licensee failure to establish and implement an adequate Abnormal Operating Instruction (AOI) to address flooding in the Auxiliary Building from a pipe break in the Essential Raw Cooling Water System (ERCW). As a result, the inadequate procedure would have resulted in the increased flooding from an ERCW header leak. The licensee entered this issue into the corrective action program as PER 339112. The inspectors determined that the performance deficiency was more than minor, and therefore a finding, because it would have the potential to lead to a more significant safety concern if left uncorrected, in that, use of the inadequate procedure could have resulted in increased flooding from an ERCW pipe break. This finding was evaluated using the significance determination phase 1 screening criteria in accordance with IMC 0609, Attachment 4 and was determined to be of very low safety significance because the finding did not involve the total loss of any safety function, identified by the licensee through a PRA, IPEEE, or similar analysis, that contributes to the external event initiated core damage accident sequences. No cross-cutting aspect was assigned to this finding because it was not determined to be indicative of current licensee performance. | |
| 05000390/FIN-2011003-01 | 2011Q2 | Green | H.2 | NRC identified | Failure to Ensure The Operability of an Emergency Battery Lighting Unit in Accordance with the Approved Fire Protection Plan | The inspectors identified an NCV of the Unit 1 Operating License Condition 2.F for the licensees failure to maintain the operability of Appendix R emergency lighting in accordance with the approved Fire Protection Plan (FPP). Specifically, both lamps for an Appendix R emergency light in the Unit 2B 480 volt transformer room were not aimed in the direction required by design to accomplish the operator manual action to restore outside air ventilation to the room in the event of a fire, as required by the FPP. The licensee implemented compensatory measures and entered this issue into the corrective action program as Problem Evaluation Report (PER) 341645. The finding was determined to be more than minor because it affected the protection against external events attribute of the Mitigating Systems cornerstone, in that it affects the objective of ensuring reliability and capability of systems that respond to initiating events. This finding was evaluated using Inspection Manual Chapter (IMC) 0609, Appendix F, Attachment 1, and was determined to be of very low safety significance because it was not a major degradation of FSSD capability. The cause of the finding was directly related to the cross-cutting aspect of Effective Supervisory/Management Oversight in the Work Practices component of the area of Human Performance, in that the licensee did not ensure oversight of work activities that adversely affected the operability of Appendix R emergency lighting |
| 05000390/FIN-2011003-02 | 2011Q2 | Green | H.5 | NRC identified | Procedure AOI-30.2 C.36, Fire Safe Shutdown Room 737-A1A, Non-feasible Operator Manual Action | The inspectors identified an NCV of Technical Specification 5.7.1, Procedures, for the licensees failure to maintain a plant procedure to ensure that an operator manual action for fire safe shutdown (FSSD) could be feasibly performed under the current physical plant configuration. Specifically, post-fire safe shutdown procedure AOI-30.2 C.36, Revision 3, contained instructions for an operator manual action for FSSD that could not be feasibly performed following implementation of a plant design change. The licensee took immediate corrective action to install a temporary scaffold as a compensatory measure. The licensee entered this issue into the corrective action program as PER 356563 The finding was determined to be more than minor because it affected the protection against external events attribute of the Mitigating Systems cornerstone, in that it affects the objective of ensuring reliability and capability of systems that respond to initiating events. This finding was evaluated using IMC 0609, Appendix F, Attachment 1, and was determined to be of very low safety significance because the procedure step in question was not a time-critical step. The cause of the finding was directly related to the cross-cutting aspect of Work Activity Coordination in the Work Control component of the area of Human Performance, in that the licensee failed to appropriately coordinate work activities, consistent with nuclear safety, to ensure that changes to the physical plant configuration would not adversely affect the feasibility of operator manual actions |
| 05000390/FIN-2011003-03 | 2011Q2 | Green | H.5 | NRC identified | Failure to Perform a Transient Combustible Evaluation for Storage of Oil in a Safety Related Area in Accordance with the Approved Fire Protection Plan. | The inspectors identified an NCV of the Unit 1 Operating License Condition 2.F for the licensees failure to store transient combustible materials in a safetyrelated/ critical area of the auxiliary building in accordance with the approved FPP. Specifically, approximately 38 gallons of hydrocarbon oil was stored inside the entrance labyrinth of the 1B charging pump room without an approved transient combustible evaluation, as required by the FPP. As a result, this was an unapproved increase in fire loading due to an increase in the volume of the predominant combustible material in the area. The licensee took immediate corrective action to remove the drum of oil from the area. The licensee entered this issue into the corrective action program as PER 371383 and PER 380910. The finding was determined to be more than minor because it affected the protection against external events attribute of the Mitigating Systems cornerstone, in that it affects the objective of ensuring reliability and capability of systems that respond to initiating events. This finding was evaluated using IMC 0609, Appendix F, Attachment 1, and was determined to be of very low safety significance because it represents a low degradation of fire prevention and administrative controls. The cause of the finding was directly related to the cross-cutting aspect of Proper Work Planning in the Work Control component of the area of Human Performance, in that the licensee failed to appropriately plan work activities to minimize the risk associated with a large quantity of oil in a safety-related fire zone without compensatory measures. |
| 05000390/FIN-2011003-04 | 2011Q2 | Green | NRC identified | Failure to Translate Moderate Energy Line Break Study Output into a Plant Procedure | The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control for the licensees failure to correctly translate a design document into operating procedures. Specifically, from original Licensing until the beginning of this reporting period, the licensee failed to translate into procedures guidance that would ensure that the plant could be safely shut down following a non-isolable break in the piping connecting the refueling water storage tank (RWST) to the emergency core cooling system in the auxiliary building. The licensee entered this issue into the corrective action program as PER 341568 and corrective actions have been completed to address the issue. This finding was determined to be more than minor because it adversely affected the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, flood protection was degraded due to a lack of procedures to mitigate flooding from the RWST into the auxiliary building with accompanying damage to both trains of the containment spray motors. This finding was evaluated using the SDP Phase 1 screening criteria in accordance with IMC 0609, Attachment 4, and was determined to require a Phase 3 analysis. The phase 3 analysis was performed by regional SRAs and determined to be of very low safety significance. The cause of the finding extends back to original plant licensing. Therefore, it is not related to current performance and is not assigned a cross-cutting aspect. | |
| 05000390/FIN-2011003-05 | 2011Q2 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and was a violation of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Facility Operating License NPF-90 for Watts Bar Nuclear Plant Unit 1, Condition 2.F, requires that TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the FPR. Contrary to the above, on April 18, 2011, the licensee found Fire Door A027 latching mechanism taped in a manner that prevented the door from automatically latching. The tape had apparently been placed there by an unknown individual during the refueling due to the inability to open the door from the auxiliary building side to the pipe chase side. The tape was removed restoring the fire door function and the latch assembly was replaced the next day. This was identified in the licensees CAP as PER 356803. This finding was of very low safety significance in accordance with IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, because the fire barrier was only moderately degraded and there were no fixed or in-situ fire ignition sources that would subject the degraded fire barrier to direct flame impingement. | |
| 05000390/FIN-2011004-01 | 2011Q3 | NRC identified | Failure to Develop and Implement Corrective Actions for PMP Drainage Path Impact on Unit 1 | During review of corrective actions associated with PER 206105, the unauthorized placement of temporary structures supporting Unit 2 construction in the probable maximum precipitation (PMP) drainage path, the inspectors identified an unresolved item (URI). Drainage related to the PMP was impeded by temporary structures and additional items that had been erected in the drainage areas. It is unclear whether these items would increase the probability of water draining into the safetyrelated structures and damaging safety-related equipment. The inspectors noted that PER 206105 initiated on October 28, 2009, identified that Unit 2 temporary facilities had been placed inside the plant protected area surrounding Unit 1 and 2 without verifying impacts to the critical flood elevation. Some of the temporary structures supporting Unit 2 construction had been located in the PMP drainage path. The PMP event is an operating basis event wherein drainage is required to be directed away from plant safety-related and nonsafety-related equipment necessary for continued operation of the plant. Exceeding this elevation would impact the operability of safety-related equipment required for Unit 1 safe operation. In conjunction with this PER, a functional evaluation (FE) was prepared and approved on December 4, 2009, that identified several structures that would need to be moved or modified prior to March 1, 2010. The PER required a corrective action plan due date of February 14, 2010, but a corrective action plan was never developed and there was no FE to address the adverse condition impact on Unit 1 beyond February 28, 2010. In accordance with licensee procedure NPG-SPP-03.1.4, Corrective Action Program Screening and Oversight, the PER screening committee assigns the responsible organization for the PER. In this case the PER was assigned to an organization outside of the nuclear power group that did not have the ability to develop a corrective action plan. This assignment was outside the 10 CFR 50, Appendix B approved corrective action program process and there was no follow-up by the PER screening committee to ensure the corrective action plan development assignment was completed. No further action was taken until the inspectors identified the lack of corrective action to the PER screening committee on August 4, 2011. The licensee entered the issue into the corrective action program as PER 413818 and also initiated PER 417148 to address the continuing potential plant impact from the addition of more temporary structures since the initial problem was identified in 2009. A new drainage model subsequently determined that some of the current temporary structures could cause the PMP drainage to exceed the critical plant elevation impacting plant safety-related and nonsafety-related equipment necessary for continued operation. Pending additional information from the licensee involving potential for drainage to affect safety-related structures, systems, and components, this item is identified as URI 050000390/2011004-01, Failure to Develop and Implement Corrective Actions for PMP Drainage Path Impact on Unit 1. | ||
| 05000390/FIN-2011004-02 | 2011Q3 | Green | P.3 | NRC identified | Failure to Fully Implement Corrective Actions for a Motor Boat Necessary for Flood Mode Preparation | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to fully implement corrective actions to address a motor boat necessary for flood mode preparation in accordance with Abnormal Operating Instruction (AOI) 7.01, Maximum Probable Flood. As a result the inspectors found that the boat was not in serviceable condition, and there was no procedure to address preventive maintenance of the boat. The licensee entered the issue into the corrective action program as PER 417920 and developing a long-term maintenance strategy. The licensees failure to fully implement corrective actions to address a motor boat necessary for flood mode preparation in accordance with AOI 7.01, was a performance deficiency. The inspectors reviewed IMC 0612 and determined that the finding was more than minor because of the lack of an important piece of equipment (motor boat) necessary for coping with the probable maximum flood (PMF) impact on Unit 1. Using the Phase I screening worksheet of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green) because the licensee would have sufficient warning (27 hours) to obtain a replacement boat before it would be impacted by a PMF event. The cause of the finding had a crosscutting aspect in the area of Problem Identification and Resolution associated with the Corrective Action Program component. It was directly related to the licensee taking appropriate corrective actions to address a safety issue in a timely manner commensurate with its safety significance and complexity. Specifically, the licensee failed to fully implement corrective actions to address a motor boat necessary for flood mode preparation in accordance with AOI 7.01. (P.1(d)) (See Section 1R01). |
| 05000390/FIN-2011004-03 | 2011Q3 | Green | H.8 | NRC identified | Failure to Fully Implement Corrective Actions for the Unapproved Storage of Oil in a Safety Related Area | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to fully implement corrective actions to address the unapproved storage of a large quantity of oil in a safety-related area of the auxiliary building in accordance with the approved Fire Protection Plan (FPP). As a result, a drum containing approximately 38 gallons of new hydrocarbon oil was relocated, but not removed from a safety-related area of the auxiliary building, without addressing the FPP requirement for an approved transient combustible evaluation. The licensee entered the issue into the corrective action program as PER 380910 and PER 388926. The remaining oil was removed from the affected room or identified with an approved transient combustible evaluation. The licensees failure to fully implement corrective actions to address the unapproved storage of a large quantity of oil in a safety-related area of the auxiliary building in accordance with the approved FPP is a performance deficiency. The inspectors reviewed IMC 0612 and determined that the finding was more than minor because it affected the Protection Against External Factors attribute (i.e., fire) of the Mitigating Systems cornerstone, in that it affected the objective of ensuring availability, reliability, and capability of systems that respond to initiating events. Because the finding increased the fire loading due to an increase in the volume of the predominant combustible in the area, the inspectors completed a SDP Phase I analysis that indicated that the finding was not a major degradation of fire prevention or administrative controls. Using the Phase I screening worksheet of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green). The cause of the finding had a cross-cutting aspect in the area of human performance associated with the work practices component. It was directly related to the licensee defining and effectively communicating expectations regarding procedural compliance and personnel follow procedures. (H.4(b)) Specifically, the licensee failed to follow the control of transient combustibles procedure by allowing the unapproved storage of a large quantity of oil in a safety-related area of the auxiliary building. (See Section 1R05 |
| 05000390/FIN-2011004-04 | 2011Q3 | Green | P.3 | NRC identified | Inadequate Corrective Actions involving the Failure of a Shutdown Board to Transfer | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees repeat occurrence of a level A problem evaluation report (PER) 176604 written July 17, 2009. The licensees failure to ensure that all corrective actions for A level PER 176604 were complete is a performance deficiency. The inspectors reviewed IMC 0612 and determined that the finding was more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern; specifically, the subject safety-related breaker could have been installed in a more critical application or have been installed for a longer period of time, up to 18 months, in the alternate feeder application. Additionally, the finding was associated with the configuration control attribute of the Initiating Events 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 Phase I screening worksheet of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green) because it would not contribute to both a reactor trip and the likelihood that mitigation equipment or functions would not be available. The cause of the finding was directly related to the cross-cutting aspect for appropriate corrective actions to address safety issues in a timely manner commensurate with their safety significance and complexity in the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution, in that the licensee failed to take adequate corrective actions to prevent repetition of the fast transfer switch mal-adjustment. Specifically, effective corrective actions to preclude repetition were not implemented but signed as completed when the 1A shutdown board alternate feeder breaker was placed in service. (P.1(d)) (See Section 4OA2) |
| 05000390/FIN-2011004-05 | 2011Q3 | Green | H.2 | NRC identified | Failure to Adequately Control Non-Conforming or Degraded Equipment | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, for the licensees failure to ensure that 6.9Kv breaker 0-BKR-569-4605025-S, which had been identified as defective, was not labeled or otherwise segregated to prevent it from being installed into the plant. The licensees failure to ensure that defective 6.9Kv breaker 0-BKR-569-4605025-S was not installed into the plant is a performance deficiency. The inspectors reviewed IMC 0612 and determined that the finding was more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern; specifically, the subject safety-related breaker could have been installed in a more critical application or have been installed for a longer period of time, up to 18 months, in the Alternate Feeder application. Additionally, the finding was associated with the configuration control attribute of the Initiating Events 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 Phase I screening worksheet of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green) because it would not contribute to both a reactor trip and the likelihood that mitigation equipment or functions would not be available. The cause of the finding was directly related to the cross-cutting aspect that the licensee ensure supervisory and management oversight of work activities in the Work Practices component of the cross-cutting area of Human Performance, in that the licensee failed to ensure that a defective component was not installed into the plant. H.4(c) (Section 4OA2) |
| 05000390/FIN-2011005-01 | 2011Q4 | Green | P.3 | NRC identified | Failure to develop and implement corrective actions for PMP Drainage Path Impact on Unit 1 | The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion XVI, Corrective Action, for the licensees failure to develop and implement corrective actions to address the unauthorized placement of temporary structures supporting Unit 2 construction in the probable maximum precipitation (PMP) drainage path. Problem Evaluation Report (PER) 206105, initiated by the licensee on October 28, 2009, identified that Unit 2 temporary structures had been placed inside the plant protected area surrounding Unit 1 and 2 without verifying impacts to the PMP critical flood elevation of 729 feet. The PER required a corrective action plan (CAP) due date of February 14, 2010. The condition as it then existed was bounded by a functional evaluation which expired February 28, 2010. The inspectors determined that the corrective actions had not been implemented and that the original adverse condition had worsened due to the addition of other temporary structures. Based on this observation, the licensee reentered the issue into the corrective action program as PER 413818 and also initiated PER 417148 to address the continuing plant impact from the addition of more temporary structures. The licensees failure to develop and implement corrective actions to preclude the unauthorized placement of temporary structures in the PMP drainage path was a performance deficiency. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding was more than minor because if left uncorrected, would have the potential to lead to even more temporary structures being added to the PMP drainage path thus exceeding the critical flood level of 729 feet, at which point safety-related structures, systems, and components would be adversely impacted. Additionally, the finding was associated with the protection against external factors attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the frequency of those events that upset plant stability and challenge critical safety functions, during shutdown as well as power operations. The inspectors determined that the finding was of very low safety significance (Green) because, at the point the licensee gained control of adding temporary structures, the modeled PMP basin water level was two tenths of a foot below the elevation leading to flooding of safety-related structures, systems, and components. The cause of the finding had a cross-cutting aspect in the area of problem identification and resolution associated with the CAP component because the licensee failed to develop and implement corrective actions to address the unauthorized placement of temporary structures supporting Unit 2 construction in the PMP drainage path. |
| 05000390/FIN-2011005-02 | 2011Q4 | Green | H.5 | NRC identified | Failure to store transient combustible materials in a safety-related/critical area of the Auxiliary Building in accordance with the approved Fire Protection Plan | An NCV of the Unit 1 Operating License Condition 2.F was identified for the licensees failure to store transient combustible materials in a safety-related/critical area of the auxiliary building in accordance with the approved Fire Protection Plan (FPP). Specifically, an excessive amount of combustible trash and laundry was stored on the auxiliary building refueling floor. The stored combustible material was approximately two and a half times the allowable limit, and the amount in excess of that limit was stored without an approved transient combustible evaluation, as required by the FPP. As a result, this was an unapproved increase in fire loading due to an increase in the volume of the combustible material in the area. The licensee took immediate corrective action to issue a transient combustible evaluation and then remove the excess combustibles from the area. The licensee entered this issue into the CAP as PER 432883 and PER 455545. The finding was determined to be more than minor because it affected the protection against external events attribute of the Mitigating Systems cornerstone, in that, it affected the objective of ensuring reliability and capability of systems that respond to initiating events. This finding was evaluated using IMC 0609, Appendix F, Attachment 1, and was determined to be of very low safety significance because it represented a low degradation of fire prevention and administrative controls. The cause of the finding was directly related to the cross-cutting aspect of proper work planning in the work control component of the area of Human Performance, in that, the licensee failed to appropriately plan work activities to minimize the risk associated with an excessive amount of combustible trash and laundry in a safety related fire zone without compensatory measures. |
| 05000390/FIN-2011005-03 | 2011Q4 | NRC identified | Failure to comply with Technical Specification Requirement 3.8.4.14 for Vital Batteries III and IV | On February 13, 2011, Vital Battery IV was tested per 0-SI-236-54, 125 VDC Vital Battery IV 60 month Performance Test and 125 VDC Vital Battery Charger IV Test, as required by Technical Specification 3.8.4, surveillance requirement 3.8.4.14 which requires, in part, that the battery capacity be Y 80% of the manufacturer s rating. The battery tested at 82.5% capacity, down from 108.75% at its last test performance. No additional actions were taken at that time. In June 2011, the licensee noted that cell # 51, was at 2.12 vdc which was out of the Category A and B limits of the associated surveillance requirement. Following cell replacement, on June 27, 0-SI-236-44, 125 VDC Vital Battery IV 18 month Service Test and Vital Battery Charger IV Test was attempted and aborted due to multiple cell voltages dropping below procedural limits. The licensee decided to replace 8 cells and re-perform the test. The test passed with no margin to support a two unit load profile requirement which was the test acceptance criteria. The licensee ordered new battery cells for both Vital Batteries IV and III. The two unit load profile had been in use on previous tests. In that Vital Battery III was the same age as Vital Battery IV, in November 2011, the inspectors inquired as to the reason for postponing the Vital Battery III performance test which had been due since November of 2010. The licensee responded that they had ordered new cells for the battery and that the cells would be replaced prior to the end of the surveillance grace period, which would end in February of 2012 and that they had confidence in Vital Battery III due to the performance of 0-SI-236-43, 125 VDC Vital Battery III 18 month Service Test and Vital Battery Charger III, in February of 2011, which showed no degradation at that time. At that point, the inspectors questioned the operability of Vital Battery III. Based on test history and battery age, batteries I, II, and V were not considered at risk for current operability. To address the current operability questions, the licensee decided to perform 0-SI-236- 53, 125 VDC Vital Battery III 60 month Performance Test and 125 VDC Vital Battery Charger III Test. Inspectors observed this test on November 21, 2011, which failed to achieve the required 80% capacity. This battery was subsequently left out of service until new cells arrived and replaced the old cells. The licensee modified the service test load profile for each battery to include only Unit 1 loads and Unit 2 loads that were connected during this period in question. This modified test was successfully performed on Vital Battery III and Vital Battery IV with the 8 new cells replaced in June. Current plans are to re-install the original 8 cells in Vital Battery IV and perform the modified service test to prove its past capability of meeting established design base loads for the time frame between February and June of 2011. The licensee plans to perform this test in January 2012. Until then, spare Vital Battery V will remain in service carrying Vital Battery IV loads. Pending this additional testing, this item is identified as URI 050000390/2011005-03, Operability concerns for Vital Batteries III and IV | ||
| 05000390/FIN-2011005-04 | 2011Q4 | Green | H.7 | NRC identified | Inadequate Procedures for Identifying Accumulated Gas in ECCS Systems | The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish adequate procedures to identify accumulated gas in emergency core cooling systems (ECCS). Specifically, the operations surveillance test procedures, 1-SI-63-10.1-A, ECCS Discharge Pipes Venting Train A Inside Containment, Rev 1 and 1-SI-63-10.2-A, ECCS Discharge Pipes Venting Train A Outside Containment, Rev 1, could allow accumulated gases inside ECCS to be vented without being quantified and evaluated for potential adverse impacts on system operability. The licensee entered this in their corrective action program as PER 478095. The inspectors determined that the licensees failure to establish adequate procedures to identify accumulated gas in ECCS was a performance deficiency (PD). The PD was more than minor because if left uncorrected, the PD had the potential to lead to a more significant safety concern. Specifically, if left uncorrected the potential existed for an unacceptable void that could affect ECCS operability to remain undetected. The inspectors screened this finding in accordance with Inspection Manual Chapter (IMC) 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) since it did not represent the loss of any system safety function and it did not screen as potentially risk significant due to seismic, flooding or severe weather. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance, because the licensee did not have accurate procedures to measure gas vented from ECCS systems in order to determine if gas voids existed which could impact ECCS system operability |
| 05000390/FIN-2011008-01 | 2011Q1 | Green | NRC identified | Failure to Submit a Licensee Event Report for a Condition Prohibited by Technical Specifications Associated with the Essential Raw Cooling Water System | The inspectors identified a Severity Level IV, non-cited violation (NCV) of 10 CFR 50.73(a)(2)(i)(B) for the licensees failure to submit a Licensee Event Report (LER) within 60 days for a condition which was prohibited by Technical Specifications (TS). On June 22, 2009 the licensee failed to recognize that they exceeded the limiting condition for operation (LCO) action time of TS 3.0.3 when both trains of the Essential Raw Cooling Water (ERCW) system were made inoperable when a cross connect valve was inadvertently opened and remained in that position for more than nine hours. Subsequent to this, the condition was discovered, TS 3.0.3 was entered, the valve was shut, and TS 3.0.3 was exited. The licensees initial reportability evaluation concluded that the event was not reportable because the ERCW system would have been able to perform its safety function. However, with the 2A and 2B header cross connected, the system did not meet TS 3.7.8 requirements and was inoperable. This evaluation failed to identify that operating for nine hours with the system inoperable exceeded the TS 3.0.3 LCO action time and thus placed the unit in a condition prohibited by TS, which is a reportable event. The licensee entered this issue into their Corrective Action Program as Problem Evaluation Report 314950. The inspectors determined that this issue was subject to Traditional Enforcement because it had the potential to impact the NRCs ability to perform its regulatory function and was not suitable for evaluation using the significance determination process. This Violation matched example nine in Section 6.9.d of the NRC Enforcement Policy and was, therefore, determined to be more than minor and a SL-IV Violation. Cross cutting aspects are not assigned to violations being dispositioned through the traditional enforcement process. | |
| 05000390/FIN-2011403-01 | 2011Q2 | Green | H.8 | NRC identified | Security | |
| 05000390/FIN-2012002-01 | 2012Q1 | Green | H.5 | NRC identified | Procedure AOI-30.2 C.36, Fire Safe Shutdown Room 737-A1A, NON-FEASIBLE Operator Manual Action | The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion XVI, Corrective Action, for the licensees failure to ensure that an operator manual action for fire safe shutdown (FSSD) could be feasibly performed under the current physical plant configuration. Specifically, post-fire safe shutdown procedure Abnormal Operating Instruction (AOI)-30.2 C.36, Fire Safe Shutdown Room 737-A1A, Revision 3, contained instructions for an operator manual action for FSSD that could not be feasibly performed following implementation of a plant design change. A temporary scaffold which was previously installed as a corrective action compensatory measure was removed without authorization. The licensee entered this issue into the corrective action program as Problem Evaluation Report (PER) 485043. The finding was determined to be more than minor because it affected the protection against external events attribute of the Mitigating Systems cornerstone, in that it affects the objective of ensuring reliability and capability of systems that respond to initiating events. This finding was evaluated using, IMC 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 1, and was determined to be of very low safety significance because the procedure step in question was not a time-critical step. The cause of the finding was directly related to the cross-cutting aspect of work activity coordination in the work control component of the area of human performance, in that the licensee failed to appropriately coordinate work activities, consistent with nuclear safety, to ensure that changes to the physical plant configuration would not adversely affect the feasibility of operator manual actions |
| 05000390/FIN-2012002-02 | 2012Q1 | Green | H.12 | Self-revealing | Failur to Comply with Technical Specification 3.4.12 by Allowing a Safety Injection Pump to Inject Into the RCS in Mode 5 | A Green, self-revealing NCV of Technical Specification (TS) 3.4.12 was identified for failure to ensure that no safety injection pump was capable of injecting into the reactor coolant system while in Mode 5. The finding was determined to be greater than minor because it was associated with the human 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. This finding was evaluated using the significance determination Phase 1 screening criteria in accordance with Inspection Manual Chapter (IMC) 0609 Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and was determined to require review in accordance with IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. This finding was determined to have a cross-cutting aspect in the area of human performance associated with the work practices component. The licensee failed to adequately implement human error prevention techniques, such as self and peer checking, to ensure that the work activity was being performed on the correct component. |
| 05000390/FIN-2012002-03 | 2012Q1 | Green | H.14 | NRC identified | Failure to Comply with Technical Specification 3.8.4, 3.8.5 and 3.0.3 by Failing to Recognize Vital Batteries Iii and Iv Degradation | A Green, NRC-identified NCV of TS 3.8.4, DC Sources Operating, was identified. The licensees failure maintain TS operability by accurately identifying that vital battery III was approaching end-of-life was a performance deficiency. It is more than minor because, if left uncorrected, it could lead to a more serious safety concern, that of loss of functionality. Additionally, the finding 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 (i.e., core damage). Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green), because subsequent functional testing by the licensee, witnessed by the inspectors, showed that vital batteries III and IV would meet all design basis analysis requirements. This finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision-making component. The licensee failed to use conservative assumptions in decision making and to adopt a requirement 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. |
| 05000390/FIN-2012002-04 | 2012Q1 | Green | NRC identified | Failure to Demonstrate Corrective Actions for the Auxiliary Charging Pumps | The inspectors identified a lack of documentation to verify proper implementation of testing necessary to verify the corrective action requirements of the Augmented Inservice Test (AIST) Program. During the review of corrective actions related to previous December 2009 NCV 05000390/2009005-01, Failure to Implement Analysis for Failed Auxiliary Charging Pumps, the inspectors determined that no documentation of acceptable testing could be found which verified the functionality of the auxiliary charging pumps (ACPs) 1A and 1B until March 23, 2012. During that testing, which was observed by the inspectors, only the 1B ACP met its acceptance criteria,1A ACP failed. The flood mode boration makeup system relies on the capability of these pumps to support Technical Requirement 3.7.2, Flood Mode Protection Plan. Pending additional information from the licensee which can verify the adequacy of the AIST program corrective action for the auxiliary charging pumps, this item is identified as unresolved item (URI) 050000390/2012002-04, Failure to Demonstrate Corrective Actions for the Auxiliary Charging Pumps. | |
| 05000390/FIN-2012003-01 | 2012Q2 | NRC identified | Failure to Demonstrate Corrective Actions for Check Valve 0-CKV-040-0604 | The inspectors identified a unresolved item (URI) related to 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for failure to take timely corrective action on CKV-040-0604, 0-PUMP-040-003B discharge check valve. During the semi-annual internal flooding inspection, inspectors were verifying the licensees capability to insure that the intake pumping station (IPS) sump pumps were capable of removing assumed leakage into the ERCW strainer rooms, A train and B train. This system relies on the capability of two 25 gpm sump pumps in each strainer room to remove leakage from ERCW strainers and fire protection strainers. Additionally, they are credited in the external flooding study to be operable to remove any leakage through the exterior of the building. The pumps discharge through individual check valves into a common pipe which dumps into the sluice trough external to the room. These piping outlets are submerged during the probable maximum flood. Therefore, the individual check valves, in addition to preventing recirculation between the pumps, also prevent flood waters from entering the room. These ERCW strainers are required for operation of the ERCW system A train and the fire protection system A train. Inspectors reviewed the maintenance history of the check valves to verify that they were being maintained in a manner that would fulfill their operability requirements. These check valves are in the Augmented Inservice Testing (AIST) Program under a two-year test frequency. Licensee records indicate that on March 8, 2008, 3A sump pump exhibited low flow below the allowable limits of TI-50.021, Intake Pumping Station Strainer Room A Sump Pump A Performance Test, and exhibited bubbling from the opposite sump pump suction 3B, indicating back leakage past check valve CKV-040-0604, 0-PUMP-040-003B discharge check valve. This resulted in problem evaluation report (PER) 139387 and work order (WO) 08-812124. This PER was closed to previously existing PER 128435 dated August 4, 2007, which also was for flow-related issues. WO 08-812124 has not been located and check valve CKV-040-0604 was not entered into the work control process at that time. PER 128435 was an all-encompassing PER for both A train sump pumps 3A and 3B. The next performance of TI-50.021 on March 1, 2009, the 3A sump pump yielded zero flow and the operator noted bubbles coming from the opposite sump pump, which again implicated the opposite train check valve 0-CKV-040-0604 as leaking backward past the seat sufficiently to prevent the 3A sump pump from removing water from the sump it shares with the 3B sump pump. WO 09-812234 was written for this check valve specifically. This WO was performed and its associated post maintenance test was signed off as satisfactory on April 15, 2010. TI 50-021 was performed the next day, April 16, 2010, at which time the 3A sump pump failed on low flow. However, there appeared to be no back leakage of CKV-040-0604. PER 225913 was written as a result of the test failure which closed to WO 110952174 for check valve 0-CKV-040-0606, 0- PUMP-040-003A discharge check valve for apparently being partially stuck shut. On October 19, 2011, in a situation of high demand due to fire pump strainer leakage, the licensee determined that check valve CKV-040-0604 was stuck open sufficiently to render sump pump 3A incapable of lowering level due to back leakage through check valve CKV-040-0604. As a result, service request (SR) 448624 was initiated and resulted in WO 112833360 which is scheduled to work December 11, 2012. On June 18, 2012, TI-50.021 was performed and failed. As such, no satisfactory testing has been shown to verify the functionality of CKV-040-0604 since April 16, 2010. Pending additional information from the licensee which can verify the adequacy of the corrective action for check valve 0-CKV-040-0606 | ||
| 05000390/FIN-2012003-02 | 2012Q2 | NRC identified | Failure to Demonstrate Corrective Actions for IPS Sump Pumps 3A and 3B | The inspectors identified a URI related to 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for failure to take adequate corrective actions to ensure the reliability of A train sump pumps 3A and 3B. During the semi-annual internal flooding inspection, inspectors were verifying the licensees capability to insure that the IPS sump pumps were capable of removing assumed leakage into the ERCW strainer rooms, A train and B train. This system relies on the capability of two 25 gpm sump pumps in each strainer room to remove leakage from ERCW strainers and fire protection strainers. Additionally, they are credited in the external flooding study to be operable to remove any leakage through the exterior of the building. These ERCW strainers are required for operation of the ERCW system A train and the fire protection system A train. Inspectors reviewed the maintenance history of the sump pumps to verify that they were being maintained in a manner that would fulfill their operability requirements. These pumps are in the Augmented Inservice Testing (AIST) Program under a two-year test frequency. Licensee records indicate that on November 4, 2007, the 3A sump pump was replaced. Each subsequent test following this time frame has exhibited continual decreasing flow below the allowable limits of TI-50.021, Intake Pumping Station Strainer Room A Sump Pump A Performance Test, up until the present. On November 2, 2007, 3B sump pump was replaced with the power leads reversed leading to reverse rotation and low flow. This was not corrected until January 24, 2008. The sump pump tested satisfactorily until January 16, 2011, when the scheduled test per TI-50.022, Intake Pumping Station Strainer Room A Sump Pump B Performance Test, was aborted due to a failed breaker disconnect switch which had been in the work planning system since 2009. On October 19, 2011, in a situation of high demand due to fire pump strainer leakage, the licensee determined that sump pump 3B would not start in local manual control. Additionally, the 3A sump pump was pumping, but all flow was being pumped backward through 3B sump pump. This resulted in an inability of the sumps pumps to remove water from the room. Operator actions were required in this remote structure to stop the level of water rise. Pending additional information from the licensee which can verify the adequacy of the corrective action for IPS sump pump 3B start failure | ||
| 05000390/FIN-2012003-03 | 2012Q2 | Green | NRC identified | Failure to Maintain Steam Generator Blowdown Isolation Valves in the Environmental Qualification Program | The team identified a Green non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control , for the licensees failure to maintain steam generator blowdown (SGBD) isolation valves 1-FCV-1-181, 182, 183, and 184 in the environmental qualification (EQ) program. Removing the valves from the EQ database resulted in internal components (lower bottom gasket and reed switch) in the SGBD valves exceeding their qualified life and replacement intervals as stated in the licensees existing EQ and revised EQ calculations. The licensee entered this issue into their corrective action program as problem evaluation report (PER) 495239 and service request (SR) 562298, and performed additional analyses and evaluations to provide reasonable assurance of operability of components. The team determined that the failure to maintain SGBD isolation valves 1-FCV-1- 181, 182, 183, and 184 in the EQ program, which resulted in two subcomponents in these valves exceeding their qualified life and replacement interval, is a performance deficiency. In addition, the licensee failed to perform an adequate functional evaluation to confirm operability of these valves after the NRC identified that the reed switch was not included in the original functional evaluation. The revised EQ calculation performed by the licensee to address the lower bottom gasket indicated the reed switch had exceeded its qualified life of 13.5 years; however, this was not addressed in the licensees functional evaluation until identified by the NRC. This performance deficiency was more than minor because it affected the Mitigating System Cornerstone attribute of design control to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. In addition, this performance deficiency also closely parallels Inspection Manual Chapter 0612, Appendix E, example 3.j because the error resulted in a condition where there was a reasonable doubt of the operability of safety related components as a result of the revised EQ calculation. The team screened this finding in accordance with NRC IMC 0609, Initial Screening and Characterization of Findings, Attachment 4, Phase 1, and determined the finding was of very low safety significance (Green). The team determined that no crosscutting aspect was applicable because this finding was not indicative of current licensee performance. | |
| 05000390/FIN-2012003-04 | 2012Q2 | Green | H.2 | NRC identified | Failure to Identify Degraded Auxiliary Charging Pump and Initiate Corrective Actions | The inspectors identified a Green NCV of 10CFR50 Appendix B Criterion XVI for failure to identify that the 1A auxiliary charging pump (ACP) was degraded based on previous questionable testing results. The inspectors determined that no acceptable testing had been performed which verified the functionality of 1A and 1B ACP until March 23, 2012. During subsequent testing, only the 1B ACP met its acceptance criteria. This system relies on the capability of these pumps to support Technical Requirement 3.7.2, Flood Mode Protection Plan. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 1A ACP was unable to perform its function in the event of severe external flooding for > 1 year. The inspectors performed a Phase 1 evaluation per Inspection Manual Chapter 0609, Attachment 4 and determined that the finding was potentially risk significant due the degradation of equipment specifically designed to mitigate external events (e.g., flooding mitigation). Consequently a Phase 3 analysis was performed by a Senior Reactor Analyst. The analyst determined that the risk significance of the issue was very low (i.e., ACDF < 1.0E-7). The dominant sequence was a significant flooding event which would require the licensee to implement their Flood Mode Mitigation strategy, with the subsequent failure of a single train of ACP pumps for the system. The finding directly involved the cross-cutting area of human performance under the supervisory and management oversight of work activities component, in that, the failures of the ACPs were left unresolved for an extended period of time over a number of failed tests. |
| 05000390/FIN-2012004-01 | 2012Q3 | NRC identified | Lack of Ability to Execute Flood Mode Configuration within the Time Frame Required by the Technical Requirements Manual, Section 3.7.2, Flood Protection Plan | Inspectors identified an unresolved item related to Technical Requirements Manual 3.7.2, Flood Protection Plan. This requirement specifies communications between the licensee and the TVA River Operations organization and time frames for these communications. Based on these communications, which are broken up into Stage I and Stage II, the licensee is required take actions. Stage I activities are essentially preparatory in nature for the plant site to receive flooding levels above plant grade. These include shutting down the reactor and commencing cool-down to 350 degrees and movement of equipment. These Stage I activities are to be complete within 10 hours of the determination that Stage I should be implemented. Based on communications with River Operations, the licensee remains in Stage I until River Operations determines that flood levels may reach plant grade level. At this point, Stage II is entered where significant plant system realignments occur including connecting the essential raw cooling water (ERCW) system to the component cooling system (CCS) system, ERCW to the raw cooling water (RCW) system, the Fire Protection system to the auxiliary feedwater (AFW) system and in some plant conditions, spent fuel pool (SFP) cooling to the residual heat removal (RHR) system. These connections are made with spool-pieces that are staged at various locations throughout the plant. Stage II activities are to be completed within 17 hours. Inspectors observed the licensee simulating installation of the spool-pieces utilized to implement AOI-7.01. Based on the observation of the tools, procedures and manpower requirements, the inspectors questioned if the licensee would be successful at reconfiguring the plant within the 17-hour window, and therefore the total time of Stage I plus Stage II activities may exceed the assumed 27 hours. Initial efforts at integrating these observed maintenance procedures within the master AOI-7.1 Maximum Probable Flood, yielded a time of approximately 39 hours. With input from the field and improved resource loading, sequencing of the support procedures over a three-day effort, the time was reduced to 32 hours and 37 minutes. With additional focus, based on previous field demonstration of one particular supporting procedure, MI-17.021, Installation of Spool-pieces Between ERCW and Component Cooling Systems, the time was reduced to 27 hours and 34 minutes. This was accomplished by assigning two maintenance teams working in parallel on the two largest, heaviest spool-pieces. A further reduction in the time requirements of AOI-7.10, Flood Mode Electrical Systems Alignment, by working parallel teams on the four shutdown boards versus in series yielding a time reduction to 25 hours and 57 minutes. Excluded from these times was the manpower that would have been required to build a temporary wall protecting the thermal barrier booster pumps (TBBP), which is approximately 4 people for 10 hours. This TBBP modification, although currently installed as of August 2012, would have been required to be built as part of the licensees flood mitigation strategy. Additionally, the two largest spool-pieces were relocated to make installation faster. The time saved by repositioning these spool-pieces would also be added to the above total times to achieve a realistic estimate of the time needed for the all required flood mode mitigation measures to be implemented. The Watts Bar Final Safety Analysis Report (FSAR), Section 2.4.14.4.3 says the following: The steps needed to prepare the plant for flood mode operation can be accomplished within 24 hours of notification that a flood above plant grade is expected. An additional 3 hours are available for contingency margin. Based on NRC observation of the aggregate activities and time frames needed to implement adequate flood protection measures as described in the FSAR, the licensee may not be able to demonstrate an acceptable capability in the required time. The ability of the licensee to perform these activities in the time allotted by the Technical Requirements Manual may not have been assured given the number of days of refinement required by the licensee to reduce the time to 25 hours and 57 minutes. Compounding the issue is the fact that not all of the activities that would have been required, specifically building the TBBP temporary barrier and relocating two of the spool-pieces, were included in the estimated time. Currently, based on the licensees refinements in their flood protection implementation plan involving more effective resource allocation and job planning, installation of the TBBP temporary barrier, and relocation of key spool-pieces, the licensee has calculated their ability to adequately prepare for a flood event. NRC evaluation of the licensees ability to implement a successful flood protection plan prior to the refinement in the flood mode protection implementation plan requires further response from the licensee as to how they could have previously met this requirement. Pending additional information from the licensee which can verify the timeliness of the licensees ability to reconfigure the plant for flood mode operation URI 050000390/2012004-01, Lack of Ability to Execute Flood Mode Configuration within the Time Frame required by the Technical Requirements Manual, Section 3.7.2, Flood Protection Plan, was identified. | ||
| 05000390/FIN-2012004-02 | 2012Q3 | Green | H.2 | NRC identified | Failure to Follow Scaffold Procedure Threatens ERCW Pump Operability | The inspectors identified a NCV of Technical Specification 5.7.1, Procedures, for the licensees failure to properly implement Maintenance Procedure MMTP-102, Erection of Scaffolds/Temporary Work Platforms and Ladders, Revision 7. Specifically, a temporary scaffold erected in close proximity to an essential raw cooling water (ERCW) pump was not adequately restrained to prevent interaction with the pump motor during a seismic event. The licensee entered the issue into the corrective action program as Problem Evaluation Report (PER) 588895, removed the subject scaffold, and implemented corrective actions to inspect all scaffolding in Seismic Category I areas for similar conditions. The licensees failure to erect the scaffold in accordance with procedures in the vicinity of safety-related equipment was a performance deficiency. The inspectors reviewed IMC 0612 and determined that the finding was more than minor because, if left uncorrected, scaffold interaction with the pump motor during a seismic event could render the pump inoperable. The finding was associated with the Mitigating Systems Cornerstone. Using the Phase I screening worksheet of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green) because no actual loss of safety function occurred and the finding did not screen as potentially risk significant due to external events. The cause of the finding had a cross-cutting aspect in the area of effective supervisory/management oversight in the Work Practices component. It was directly related to the licensee not ensuring adequate supervisory and management oversight of work activities, including contractors that erected the scaffold and licensee engineering personnel that reviewed and approved the deficient scaffold installation that could adversely affect nuclear safety. |
| 05000390/FIN-2012004-03 | 2012Q3 | Green | P.3 | NRC identified | Inadequate Corrective Actions for the C ERCW Pump Breaker | The inspectors identified a NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct an identified deficiency in the C-A ERCW pump breaker on July 25, 2012. This uncorrected deficiency led to the inability of the breaker to trip and is a performance deficiency. The inspectors reviewed IMC 0612 and determined that the finding was more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern; specifically the failure of the C-A ERCW pump to load shed on a loss of offsite power. Additionally, the finding was associated with the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using the Phase I screening worksheet of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green) because the associated shutdown board is a Unit 2 board and is lightly loaded. Additionally, the failure of the C-A ERCW pump breaker to trip and thus be immediately loaded onto 2A emergency diesel generator is within the transient capability of the emergency diesel generator. The cause of the finding was directly related to the cross-cutting aspect for appropriate corrective actions to address safety issues in a timely manner commensurate with their safety significance and complexity in the corrective action program component of the crosscutting area of Problem Identification and Resolution, in that the licensee failed to take adequate corrective actions to repair the C-A ERCW breaker when the initial deficiency was discovered on July 25, 2012. |
| 05000390/FIN-2012005-01 | 2012Q4 | NRC identified | Engineering Justification for Design of Control Building Watertight Hatches | The inspectors reviewed Final Safety Analysis Report (FSAR), Section 3.8.4.1.1, which requires the two equipment hatches at elevation 708.0 in the control building to be watertight. Design Criteria WB-DC-20-21, Miscellaneous Steel Components for Category I Structures, Revision 13, covers the requirements for the equipment and, according to TVA Drawing 48N1306, the hatches are designed to be watertight. Drawing 48N1306 notes: 9. Seals and gaskets by Heavy Equipment Group, and 13. Gaskets shall be affixed to cover plates with waterproof cement or equal to assure that they remain attached to covers when covers removed. There is no information on the drawing regarding the material specifications of the gaskets, only the thickness: 1/8 gasket. Design Criteria Document WB-DC-40-60, Special Hatches and Manways, Revision 6, Section 3.12.2.2, states that the hatches must withstand a pressure of 1.3 psi from topside (water to El. 711.0 due to a turbine building flood resulting from a rupture in the CCW system). Service Request (SR) 427917 (initiated September 5, 2011) reported Water leaking down into the EBR Chiller Room Water leaking through the equipment hatch (WBN-2-EQH-271-0008) seals and dripping down into the EBR Chiller room and pooling in front of the door. Repair/replace leaking hatch seals and the seals around the coffer dam. The source of water appears to be from rain water entering the TB around the Steam & Feed line penetrations in the NE corner elevation 729 Unit-2 side. Cover/close/or seal these penetrations to prevent rain water from entering TB. This SR was closed to WO 112678945 and the WO was cancelled on May 9, 2012. It is unlikely that the hatches are capable of being watertight at a pressure of 1.3 psi from topside (water to El. 711.0) if they leak rain water from the 708.0 El. floor. Calculation WCG-1-1591, Seismic Evaluation of Watertight Equipment Hatches in Control Building at El. 708.0, provides some evaluation of structural elements of the hatches. This calculation does not provide any evaluation of the shear stress in the connecting screws due to hatch deflection/elastic deformation nor does the calculation evaluate the design adequacy of the mounting frame attached to the concrete opening under a live load (22,500 lbs.). Pending additional information from the licensee which can verify that there was adequate engineering justification for the design of floor hatches (WBN-1-EQH-271-0008 and WBN-2-EQH-271-0008) in the 708.0 El. of the control/turbine building, this item is identified as unresolved item (URI) 050000390/2012005-01, Engineering Justification for Design of Control Building Watertight Hatches. | ||
| 05000390/FIN-2012005-02 | 2012Q4 | Green | H.2 | NRC identified | Failure to Adequately Develop and Implement Ice Condenser Ice Basket Repairs | A NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, for the licensees failure to adequately develop and implement ice condenser ice basket repairs in accordance with approved engineering and maintenance documents. Specifically, the inspectors observed that repairs to six damaged ice condenser ice baskets, previously signed off as complete in the work order (WO) by the installers and following Quality Control inspection and acceptance were not in accordance with the design and maintenance WO documents. The licensee initiated Problem Evaluation Reports (PERs) 623040 and 626983 to address the inspector-identified deficiencies. The licensees failure to adequately develop and implement ice condenser ice basket repairs in accordance with approved engineering and maintenance documents was a performance deficiency. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding was more than minor because the deficiencies were not identified by the licensee and would have remained unidentified at least for the duration of the upcoming fuel cycle. Without the specified repairs being properly implemented on the damaged ice baskets, there was no reasonable assurance they were capable of performing their design function, and there was also potential for damage to adjacent ice baskets obstructing open flow paths, in the event the ice condenser was required to perform its design function. Using the Initial Characterization of Findings guidance of IMC 0609, the inspectors determined that the finding was of very low safety significance (Green) because no actual loss of safety function occurred. The cause of the finding had a cross-cutting aspect in the area of effective supervisory/management oversight in the Work Practices component. It was directly related to the licensee not ensuring adequate supervisory and management oversight of work activities, including the licensee engineering personnel that prepared and reviewed the ECP, the contractors that performed the repair work and the Quality Control personnel that performed the repair inspection and acceptance. |
| 05000390/FIN-2012005-03 | 2012Q4 | NRC identified | Engineering Justification for Modifications to NON-CONFORMING Ice Baskets | The inspectors reviewed outage WO 113393057 which specified the installation of new hardware on a total of six ice baskets that had been damaged, apparently due to ice condenser maintenance. The ice condenser is located within the primary containment and was designed, tested, qualified and fabricated by Westinghouse, the nuclear steam supply system original equipment manufacturer (OEM). The ice condenser contains a total of 1944 vertically supported, perforated 14 gauge sheet metal, ice baskets (12.1 inches in diameter and 48 feet tall) each weighing a maximum of 2200 pounds (ice column + basket). Each of the six damaged baskets (non-conforming components) had suffered plastic deformation (local compressive buckling) of support ligaments in the vicinity of the bottom three feet of the baskets. Instead of replacing the damaged portion of the baskets, as permitted by FSAR Section 6.7.4, licensee engineering had designed hardware to add to the damaged portions of the baskets. EDC E-50607, Revision A, specified the installation of vertical supports mechanically attached on the outside of the damaged area and EQV 60275, Revision A, specified the use of wire rope (steel cable) laced through the damaged area. According to the referenced calculation, WCG-1-1912, Qualification of the Optional Lower Ice Basket Support, the vertical supports were intended for compressive loading and the wire rope was intended for tensile loading. Per FSAR Table 6.7-2, during a deadweight load or deadweight and seismic loads the vertical load on the ice baskets is in compression. When subjected to a design basis accident (DBA) load in combination with a deadweight, or deadweight and earthquake load, the vertical load on all the ice baskets is in tension and the compressed ice basket would tend to elongate. Review of FSAR Section 6.7.4.3, Design Evaluation, Loading Conditions, part 2., Blowdown Loads, subpart E. Horizontal Ice Basket Forces, states that the tangential and radial forces acting on the ice baskets due to cross flow are assumed to act on the bottom, three feet of ice basket (one-half of the span between the top of the lower support structure and the attachment of the ice baskets to the first lattice frame). The inspectors did not find that the licensee-developed design changes adequately considered these dynamic tangential and radial loads on the damaged ice baskets. Their modifications only addressed either tensile or compressive forces on the ice baskets. Also the addition of the hardware appeared to be more appropriately governed by the requirements of a temporary alteration control form (TACF) per procedure NPGSPP- 09.5, Temporary Alterations, since information contained in the WO indicated the damaged baskets would have to be replaced during the next refueling outage. In addition, the 10CFR50.59 screening processes employed for the addition of hardware did not adequately consider the key elements that would be addressed for a TACF, since hardware was being added to safety-related components. The modifications may not be adequate for the damaged ice baskets to withstand all static and dynamic loads they were originally designed, tested, and qualified to be subjected to. Although there appears to have been some verbal contact between the licensee and the OEM engineering organization regarding the damaged ice baskets, there was no formal OEM review and acceptance of the licensee modifications as an acceptable alternative to ice basket replacement or repair per FSAR Section 6.7.4. Pending additional information from the licensee which can verify that there was adequate engineering justification for the use of an EDC and an EQV for hardware modifications to ice baskets, this item is identified as unresolved item (URI) 050000390/2012005-03, Engineering Justification for Modifications to Non-Conforming Ice Baskets. | ||
| 05000390/FIN-2012005-04 | 2012Q4 | Green | H.13 | Self-revealing | Late State Notification of Unusual Event | A self-revealing non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.54q(2) for failure to follow the approved emergency plan. Specifically, on August 10, 2012, state officials were not notified within 15 minutes of the declaration of an Unusual Event. State notification is a risk-significant planning standard requirement required by 10 CFR 50.47(b)(5), 10 CFR 50 Appendix E, Section IV.D.3 and Section 5.2.1, of the licensees Radiological Emergency Plan. The issue was greater than minor because it was associated with the Emergency Planning cornerstone attribute of Emergency Response Organization performance during an actual event. The finding affected the cornerstone objective in that timely notification is critical to ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors reviewed this finding using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Attachment 1, Failure to Implement (Actual Event) Significance Logic. The finding was determined to be of very low safety significance because it was a failure to implement during an Unusual Event. The finding had a cross-cutting aspect in the area of Human Performance, Decision-Making, because the unit supervisor, in the absence of the shift manager, did not effectively fulfill his responsibility to direct or perform required state communications within the required 15 minute time period as required by the Radiological Emergency Plan. |
| 05000390/FIN-2012005-05 | 2012Q4 | Green | H.8 | Self-revealing | Failure to Follow Procedure Resulted in Failing to Remove Jumpers Inhibiting Proper Operation of the Steam Generator Blowdown System | A self-revealing NCV of Technical Specifications (TS) 5.7.1, Procedures, was identified for failing to adhere to OPDP-1, Conduct of Operations, Section 5.1, Procedure Adherence. The licensee failed to ensure a jumper was removed prior to placing the steam generator blow-down system into service per System Operating Instruction 90.01, Rev. 29, Liquid Process Radiation Monitors, step 5.5 (10). This was a performance deficiency and a finding. The finding was more than minor because, if left uncorrected, it could lead to a more significant safety issue, a radioactive release, and was associated with the Mitigating Systems Cornerstone attribute of equipment performance (reliability) and adversely affected the cornerstone objective. The finding was evaluated using the SDP Phase I and was determined to be a finding of very low safety significance because actual high contamination levels did not occur within the steam generators during the period that the jumper was installed. The licensee entered this issue into the corrective action program as PER 637279. The finding directly involved the cross-cutting area of Human Performance under the procedural compliance aspect of the work practices component; in that the procedural requirements of System Operating Instruction 90.01 were not met. |
| 05000390/FIN-2012005-06 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | Technical Requirements 3.4.2, Pressurizer Temperature Limits, required that the pressurizer heat-up rate be limited to U100a F in any 1-hour period. The TR action statement A.1 required that the limit be restored within 30 minutes. Contrary to the above, at or around 0834, on October 10, 2012, while Unit 1 was in Mode 5, the licensee determined that the pressurizer vapor space heat-up rate limit of U100a F in any 1-hour period had been exceeded. The heat-up rate was 103a F per hour. The heat-up rate was returned to within limits in less than the limiting condition of operation (LCO) action time of 30 minutes. The finding was screened in accordance with IMC 0609 Appendix G, Shut-down Operations SDP and was characterized to be of very low safety significance (Green) because the pressurizer water temperature did not exceed the TR heat-up rate, only the vapor temperature by a small (3a F per hour) amount and the engineering staff review using OEM documentation concluded that there were no adverse consequences. | |
| 05000390/FIN-2012005-07 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50 Appendix B, Criterion III, Design Control, states in part that measures shall be established to assure that regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee failed to control both ASME III code and non- ASME code materials during relocation of WBN-1-PDT-030-0042-G, WBN-1-PDT- 030-0045-D, and WBN-1-PDT-030-0043-F installed by Design Change Notice 58382 stages 8, 10, and 11. Portions of the installed material did not met ASME Section III Class 2, TVA Class B design requirements resulting in non-ASME code material being used in the fabrication of ASME code components. The finding was screened in accordance with IMC 0609 Appendix G, Shut-down Operations SDP and was characterized to be of very low safety significance (Green) because the nonconforming condition documentation concluded that there were no adverse functional consequences. | |
| 05000390/FIN-2012008-01 | 2012Q2 | Green | H.14 | NRC identified | Failure to Establish Test Procedures to Assure Satisfactory Acas Performance During Design Basis Accidents | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for failure to perform capacity (volumetric flow) testing on the safety-related auxiliary control air subsystem (ACAS). The licensee had documented that, for worst case environmental conditions, the air compressor capacity had little margin when compared to required air demand, even for single unit operation. This issue was entered into the licensees corrective action program as problem evaluation report 501941 for further evaluation of corrective actions. The team determined that the failure to perform capacity testing to ensure the ACAS would meet the required air demand in response to a design basis event was a performance deficiency. This performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to develop a test procedure that would reliably ensure that the ACAS would meet required air demand for its safety-related loads during design basis accidents. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. The inspectors determined that the finding had a cross- cutting aspect in the use of conservative assumptions in the decision-making component of the human performance area. Specifically, the licensee did not use conservative assumptions in making the decision to discontinue capacity testing of the ACAS system in 2002, and stated that if that decision had been made more recently (using available internal guidance and practices regarding the testing of safety-related systems), the resulting decision would have been the same |
| 05000390/FIN-2012008-02 | 2012Q2 | Green | P.6 | NRC identified | Failure to Adequately Test the AFW Discharge Check Valves | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish a test program that demonstrated the adequacy of the auxiliary feedwater (AFW) discharge check valves. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flow-rates to the steam generators. This issue was entered into the licensees corrective action program as problem evaluation report 499950. The licensee performed a functional evaluation and determined that the AFW system was operable based on the pumps not currently being degraded to the design limits, and the existence of additional conservatisms in the licensees design basis hydraulic analysis. The team determined that the licensees failure to establish a test program to ensure that back leakage through the AFW discharge check valves would not challenge the ability of the AFW system to provide design basis flow to the steam generators was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it could have the potential to lead to a more significant safety concern. Specifically, AFW check valve back leakage could challenge the systems ability to support removal of decay heat from the reactor, which would not be identified by the licensees test program. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green), utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1- Initial Screening and Characterization of Findings. Because the licensee performed a self-assessment in December 2011 that included missed opportunities to identify that check valve leakage could negatively impact the AFW system, this finding was assigned a cross-cutting aspect in the self- and independent assessments component of the problem identification and resolution area |
| 05000390/FIN-2012008-03 | 2012Q2 | Green | H.7 | NRC identified | Inadequate Acceptance Criteria in Maintenance and Surveillance Procedures (5 Examples) | The team identified five examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures. The five examples were entered into the licensees corrective action program. The inspectors determined that the failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures was a performance deficiency. The performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding is similar to Inspection Manual Chapter 0612, Appendix E (example 4.a), because the failure to ensure correct translation of acceptance criteria into procedures was not isolated. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. This finding had a cross-cutting aspect in the resources component of the human performance area, because the licensee had not ensured that complete, accurate, and up-to-date procedures were consistent with vendor and design specifications; and therefore, the procedures were not available and adequate to assure nuclear safety |
| 05000390/FIN-2012008-04 | 2012Q2 | NRC identified | Effect of System Harmonics on Degraded Voltage Relay Function | The team identified an issue of concern and an unresolved item related to the effect of electrical system harmonics on safety-related degraded voltage relays. Specifically, in 1993, the licensee identified that harmonic distortions adversely affected the 6.9 kilovolt (kV) bus overvoltage relays by causing them to alarm unnecessarily. The licensee entered this issue into their corrective action program and modified the overvoltage relays to minimize the effects. However, the licensee did not identify (or otherwise evaluate) the adverse effect that harmonics could have on the ability of the degraded voltage relays to perform their safety function, as required by limiting condition for operation 3.3.5 of the plants technical specifications. The Watts Bar degraded voltage protection scheme features three ABB type 27N relays for each 6.9 kV safety bus, arranged in a two out of three tripping scheme. The ABB instruction bulletin 7.4.1.7-7 contained in vendor manual WBN-VTD-AS04-0080 states that (1) the relay employs a peak voltage detector, and (2) harmonic distortion on the AC waveform can have a noticeable effect on the relay operating point and the measuring instruments used to calibrate the relay. The bulletin also notes that the relay is available with an internal harmonic filter for applications where waveform distortion is a factor. The team noted that calculation WBPE2119202001, 6.9kV Shutdown & Logic Boards Under-voltage Relay Requirements/Demonstrated Accuracy Calculation, identified the relay as a model not equipped with a harmonic filter, but did not address the basis for excluding harmonic distortion as a factor which affected relay accuracy. In response to the teams inquiries, the licensee provided PER 930397 that addressed spurious actuations of the ABB type 59H overvoltage relays which are similar in design to the ABB type 27N degraded voltage relays. Troubleshooting tests performed to identify the cause of the 59H spurious actuations revealed that high levels of 6.9 kV system harmonics from sources both external and internal to the station accompanied the spurious operations. The causal factor section of PER 930397 stated that the relays sometimes trip on harmonic distortion although the root mean square voltages are at acceptable levels. Corrective actions consisted of replacing the type 59H overvoltage relays with a model equipped with harmonic filters. The team further noted that the extent of condition section of PER 930397 did not identify or address whether the degraded voltage relays operating point could also be affected by the same harmonics implicated in the mal-operation of the overvoltage relays. The team was concerned that harmonics on the 6.9 kV system could cause the degraded voltage relays to fail to actuate at the set-point specified by technical specifications. Persistent harmonics can be produced by factors external to the nuclear site or by internal phenomena. A typical internal source of harmonics at nuclear power plants is motor defects. The team was also concerned that transient harmonics could cause the relays to spuriously reset during an actual degraded voltage event, thereby delaying the protective function beyond the 10 seconds stipulated in technical specification limiting condition for operation 3.3.5. Specifically, the degraded voltage relays design features an instantaneous reset characteristic that could allow reset of the degraded voltage relay in less than two cycles in the presence of harmonics, thereby reinitiating the external 10 seconds timer. The reset function of the existing degraded voltage relays is identical to the tripping function of the overvoltage relays that actuated due to transient harmonics in 1993. In 1993, transient harmonics were measured at levels of greater than 10% total harmonic distortion during the troubleshooting for PER 930397 versus the 0.3% distortion deemed acceptable by the relay vendor. The transient harmonics documented in PER 930397 were attributed to events that included the trip of the nearby Sequoyah generating station, and to breaker operations at the Watts Bar station. The team noted that similar conditions could exist during an accident scenario when proper performance of the degraded voltage scheme time delay would be critical with respect to satisfying the response time assumptions in the accident analysis. In response to the teams concerns, the licensee provided information regarding condition monitoring of large motors that consisted of periodic measurement and analysis of motor bearing vibration from which various defects that may produce harmonics could be identified. The team noted, however, that there was no written guidance or acceptance criteria for these tests that would prompt engineering to investigate whether suspected motor defects could produce harmonics that would adversely affect the accuracy of degraded voltage relays. Specifically, there was no recognition in design or maintenance documents regarding the susceptibility of the degraded voltage relays to harmonic distortion, or the need to investigate suspected motor defects with respect to this susceptibility. The team further noted that during normal bus voltage conditions when voltage is above the degraded voltage relay reset set-point, harmonics would shift system peak voltage away from the degraded voltage relay operating set-point rather than closer to it, and so the presence of harmful harmonics would not self-reveal by spurious actuations. The overvoltage relays are now equipped with harmonic filters so they will also not reveal the presence of either transient or persistent harmonics. Based on the teams observations, the licensee has entered these concerns into their corrective action program as PER 515413 and PER 546072. The team determined that additional review of information recently received from the licensee regarding Watts Bars design and licensing bases was necessary to determine if the licensees performance constituted a violation of NRC regulatory requirements. Additionally, the team determined that additional consultation with the Office of Nuclear Reactor Regulation was warranted before reaching a final disposition of the unresolved item. This unresolved item is open pending (1) the review of additional information from the licensee regarding the design and licensing basis of the degraded voltage relays and (2) consultation with the Office of Nuclear Reactor Regulation: URI 05000390/2012008-04, Effect of System Harmonics on Degraded Voltage Relay Function. | ||
| 05000390/FIN-2012009-02 | 2013Q1 | Severity level III | NRC identified | Failure to Report Unanalyzed Condition Related to External Flooding | The inspectors identified an AV of 10 CFR 50.72(b)(3)(ii)(B), Immediate Notification Requirements for Operating Nuclear Reactors, for failure to report within eight hours an unanalyzed condition that significantly degraded plant safety. Specifically, the licensee failed to notify the NRC upon discovery that a postulated PMF would result in the overtopping of earthen dams not previously assumed in the plant design. The failure to report this unanalyzed condition resulted in the NRC not being made aware of a condition which would have resulted in additional NRC review. Specifically, the failure to notify the NRC within eight hours of discovery of an unanalyzed condition that significantly degraded plant safety and resulted in an unacceptable change to the facility or procedures. The inspectors determined an evaluation for cross-cutting aspect was not applicable because this is a traditional enforcement violation. | |
| 05000390/FIN-2012009-03 | 2013Q1 | Yellow | H.1 | NRC identified | Failure to Maintain an Adequate Abnormal Condition Procedure to Implement the Flood Mitigation Strategy | The inspectors identified an AV of Technical Specification 5.7.1, Procedures, for the licensees inability to demonstrate that the required Stage I and Stage II activities could be performed within 27 hours as required by AOI-7.1, Maximum Probable Flood. The licensees failure to adequately demonstrate the ability to realign plant systems into their flood mode configuration using AOI-7.1, Maximum Probable Flood, within the time frame required by TRM 3.7.2 and Watts Bar UFSAR Section 2.4, which could directly lead to the inability to remove decay heat from the reactor core resulting in core damage, was a performance deficiency. This performance deficiency was considered more than minor because it was associated with the Protection Against External Factors attribute of the Reactor Safety/ Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inability to realign plant systems into their flood mode configuration within the required time frame could directly lead to the inability to remove decay heat. The combination of the seismic and rainfall event frequencies and types of rainfall events which would lead to flooding above site grade and the inability to realign plant systems into their flood mode configuration within the 27-hour required time frame could directly lead to the inability to remove decay heat from the reactor core resulting in core damage which has an impact of substantial safety significance. The NRC concluded that the significance of the finding is preliminarily substantial safety significance (Yellow). The cause of the finding had a cross-cutting component of Resources in the area of Human Performance with an aspect of ensuring that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, inadequacies in those procedures, equipment, and personnel training necessary to realign plant systems within the required time frame to cope with all anticipated external flooding events. |
| 05000390/FIN-2012009-04 | 2013Q1 | Green | H.2 | NRC identified | Failure to Adequately Protect SAFETY-RELATED Equipment During Flood Mode Preparation | The inspectors identified an AV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licenseefs failure to adequately protect safety-related equipment during flood mode preparation. The licenseefs failure to adequately protect safety-related equipment during flood mode preparation as implemented by AOI-7.1, Maximum Probable Flood, was a performance deficiency. This performance deficiency was considered more than minor because it was associated with the Protection Against External Factors attribute of the Reactor Safety/Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the TACF was incapable of preventing water damage, during a PMF event, to both trains of important equipment, specifically the thermal barrier booster pumps (TBBPs), necessary for coping with the PMF impact on Unit 1. Without the TBBPs and with reactor coolant pump (RCP) seal injection lost, there is no engineering assurance that RCP seal damage would not occur, leading to an RCP seal loss of coolant accident (LOCA). The performance deficiency involved external events. Consequently a Phase 2 analysis could not be performed and therefore a Phase 3 analysis was conducted. The increase in core damage frequency (CDF) for this issue was estimated at 6.35~10-6; which has an impact of low to moderate safety significance. The NRC concluded that the significance of the finding is preliminarily of low to moderate safety significance (White). This finding has a cross-cutting aspect in the Work Practices component of the Human Performance area because it was directly related to the licensee not ensuring adequate supervisory and management oversight of engineering design work activities associated with a plant design change to protect the TBBPs during certain flood events. |
| 05000390/FIN-2012009-05 | 2013Q1 | Green | H.2 | NRC identified | Failure to Correct Conditions Adverse to Quality Related to IPS CKV-040-0604 and IPS 3A and 3B Sump Pumps | The inspectors identified two examples of an NCV of the 10 CFR 50 Criteria XVI, Corrective Action, for failure to correct conditions adverse to quality for the intake pumping station (IPS) CKV-040-0604, pump 3B, discharge check valve which resulted in it being non-functional for an extended period of time, and both the IPS 3A and 3B sump pumps, which resulted in the pumps remaining in a degraded condition for an extended period of time. The licensees failure to maintain these components in accordance with the requirements of the augmented in-service testing program and WB-DC-40-29, Flood Protection Provisions, were performance deficiencies. The performance deficiencies were determined to be more than minor because, if left uncorrected, would lead to a more significant safety concern. Specifically, internal flooding of the IPS mechanical equipment room housing the train A essential raw cooling water (ERCW) strainers could occur. The inspectors performed a Phase 1 evaluation per IMC 0609, Attachment 4, and determined that the finding was potentially risk significant because it involved the degradation of equipment specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors). Consequently a Phase 3 analysis was performed by a Senior Reactor Analyst. The analyst determined the finding was of very low safety significance, Green. These findings directly involved the cross-cutting area of Human Performance under the Work Practices component, in that, the licensee failed to provide adequate supervisory and management oversight to ensure corrective actions were taken to maintain the functionality of IPS equipment for extended periods of time. |
| 05000390/FIN-2012009-06 | 2013Q1 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, conditions adverse to quality related to the main control room chilled water circulating pumps A-A and B-B and shutdown board room chilled water circulating pumps A-A and B-B were not corrected which resulted in the chillers being inoperable and reportable. The inspectors performed a Phase 1 evaluation per IMC 0609, Attachment 4, and determined that the finding was potentially risk significant because it involved the degradation of equipment specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors). Consequently, a Phase 3 analysis was performed by a Senior Reactor Analyst. The analyst estimated the frequency of the PMP event, adjusted for the time of year when room cooling would be necessary and multiplied by a conservative value for ACCDP (0.1) representing the likelihood for core damage due to alternate shutdown from outside the control room. The analyst determined that the risk significance of the issue was very low (i.e., ACDF < 1.0E-6). Because this finding is of very low safety significance (Green) and has been entered into the corrective action program as PER 641937, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy. |