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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000263/FIN-2009007-07 | 2009Q4 | Green | NRC identified | Pipe Support Design Deficiencies | The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety-significance for the failure of two pipe supports to meet their design requirements. Specifically, the calculation for pipe support SR-526 failed to use the minimum yield strength in determination of the allowable bending stress of the pipe support baseplate as required in the American Institute of Steel Construction code. In addition, the calculation for pipe support PS-16 failed to use the design basis concrete compressive strength in determination of the anchor bolt allowable as required in the licensees design specification. This finding was entered into the licensees corrective action program and a preliminary analysis performed by the licensee concluded that the pipe supports were operable but nonconforming. The performance deficiency for pipe support SR-526 example was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of the safety-related residual heat removal and core spray pumps. This finding is of very low safety-significance (Green) because the design deficiency was confirmed not to result in loss of operability or functionality. The performance deficiency for pipe support PS-16 example was more than minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. This finding is of very low safety-significance (Green) because there was no actual barrier degradation. The inspectors did not identify a cross-cutting aspect associated with this finding because this was a legacy design issue and therefore was not reflective of current performance. | |
| 05000263/FIN-2009008-01 | 2009Q1 | Severity level III | H.8 | NRC identified | Failure to Notify the NRC of a Permanent Illness or Disability of a Licensed Operator | To Be Determined. On July 8, 2004, a licensed senior reactor operator (SRO) notified the stations medical staff that he began taking prescribed medication in June 2004 for a potentially disqualifying medical condition. The NRC was notified of the SROs potentially disqualifying medical condition on November 25, 2008. Title10 CFR 50.74(c), Notification of Change in Operator or Senior Operator Status, requires the licensee to notify the NRC within 30 days of the licensee being informed of a permanent change in a licensed operators medical condition. The time period between July 2004 and November 2008 exceeded the 30-day notification requirement. The licensee conducted a review of all licensed operator medical records to determine the extent of condition and initiated other compensatory measures to prevent recurrence of this condition. Because the issue affected the NRCs ability to perform its regulatory function it was evaluated using the traditional enforcement process. The finding was determined to be of low safety significance because the SRO was taking the medications as prescribed and had not made any operational errors during any emergency condition. The regulatory significance was important because your staff failed to notify the NRC of a permanent disability or illness of an SRO within 30 days. This was preliminarily determined to be an apparent violation of 10 CFR 50.74(c). The cause of the apparent violation is related to the cross-cutting element of human performance in the area of work practices. H.4(b)(Section 1R11 |
| 05000263/FIN-2009008-02 | 2009Q1 | Severity level III | H.8 | NRC identified | Failure to Provide Complete Information to the NRC which Impacted a Licensing Decision. (Section 1R11) | To Be Determined. Every six years an operators NRC operating license must be renewed. When the licensee submits the request for license renewal, the licensee must certify to the NRC that the operator is medically capable of performing license duties. This is done by completing an NRC Form 396, Certification of Medical Examination by Facility Licensee. When signed by senior station management, the NRC Form 396certifies that an operator is able to perform operator duties. The form contains several standard license conditions that restrict operator activities to ensure their ability to perform license duties. In this SROs case, the licensee certified to the NRC in a letter dated September 11, 2008, that the operator was capable of performing license duties with no restrictions except to wear corrective lenses. The licensee provided incomplete and inaccurate information on the accompanying NRC Form 396 in that the licensee failed to inform the NRC that the SRO was taking medication for a potentially disqualifying medical condition so the NRC could properly restrict the SROs operating license to have a Must Take Medication as Prescribed to Maintain Qualifications license restriction. Because the issue affected the NRC=s ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The finding was determined to be of low safety significance because the SRO had taken medications as prescribed and had not made errors during any emergency condition prior to the license being amended. However, the regulatory significance was important because the incomplete and inaccurate information was provided under a signed statement to the NRC and impacted a licensing decision for the SRO. This was preliminarily determined to be an apparent violation of 10 CFR 50.9, Completeness and Accuracy of Information. The cause of the apparent violation is related to the cross-cutting element of human performance in the area of work practices. H.4(b)(Section 1R11 |
| 05000263/FIN-2009201-01 | 2009Q4 | Green | P.2 | NRC identified | Security | |
| 05000263/FIN-2009404-01 | 2009Q4 | Green | Licensee-identified | Security | ||
| 05000263/FIN-2010002-01 | 2010Q1 | NRC identified | Reactor Building Crane Design and Licensing Basis Issues | The inspectors reviewed the following licensing documents for the reactor building crane: NRC letter to Northern States Power (NSP), Safety Evaluation by the Office of Nuclear Reactor Regulation (NRR) Supporting Approval of Crane Modification and Use of 70 Ton Spent Fuel Shipping Cask IF-300, dated May 19, 1977; NSP letter to NRC, Response to Request for Additional Information, dated February 28, 1977; and USAR, Section 10.2, page 4 of 24, and Section 12.2 page 28 of 49, and page 29 of 49, Revision 23 and Revision 25.The NRC letter to NSP, Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Approval of Crane Modification and Use of 70 Ton Spent Fuel Shipping Cask IF-300, dated May 19, 1977, established the reactor building overhead crane capacity as a maximum of 85 tons and the crane seismic analysis did not analyze for a maximum 85 ton lifted load concurrent with a seismic event based on extremely low probability of both events occurring simultaneously. The licensee subsequently changed the reactor building crane capacity from 85 tons in USAR, Section 10.2, page 4 of 24,and Section 12.2, page 28 of 49, and page 29 of 49, Revision 23 to a crane capacity of105 tons in USAR Section 10.2, page 4 of 24 and Section 12.2, page 28 of 49 and page 29 of 49, Revision 25.The inspectors noted that the licensee did not perform a written 10 CFR 50.59evaluation to assess the following: 1) whether the change of increasing design loads on the crane and the crane support structure required a license amendment and, 2) probabilistic analysis with consideration for a new maximum crane lifted load of105 tons that evaluates whether or not a lifted load must be considered during a seismic event for the design of the reactor building crane and crane support structure. The inspectors reviewed Calculation Nos. CA 76 138, Structural Requalification for New 85 Ton Crane, Revision 0; CA-05-103, Reactor Building Superstructure Seismic Response Analysis with 105 Ton Crane, Revision 0A; and CA-05-107,Structural Seismic Qualification Reactor Building Crane Upgrade for ISFSI, Revision 0B. The inspectors were concerned that the reactor building crane and reactor building crane support structure had been evaluated using friction in a linear elastic analysis to reduce seismic load effects applied to the reactor building crane and crane support structure. The licensee used much smaller seismic loads limited by the friction force and this resulted in a significant load reduction for qualification of the reactor building crane and reactor building crane support structure. In addition, the non-linear effects of friction have not been addressed in the aforementioned calculations. The licensee was unable to provide evidence that the NRR staff had approved friction in a linear elastic analysis as a method of evaluation for this application. The use of friction to reduce seismic load effects on the reactor building crane and reactor building crane support structure was not discussed in the USAR. The inspectors reviewed Calculation No. CA-05-101, Evaluation of Reactor Steel Superstructure for 105 Ton Reactor Building Crane, Revision 3A. The inspectors were concerned with the following: 1) The minimum yield strength for the American Society of Testing and Materials (ASTM) A1 trolley and the bridge rail has not been established in accordance with the American Institute of Steel Construction (AISC) code and,2) The restraint mechanism in the longitudinal direction of the trolley rail and bridge rail connection was based on the use of friction resistance between the bottom of the rail and the supporting beam to resist the sliding of the rail during a design and licensing basis event. In response to the concern, the licensee initiated corrective action program documents CAP 01214808, RB Crane Seismic Calc may not be Consistent W/License Basis, dated January 22, 2010 and CAP 01222530, Crane Heavy Lift Inspection URI:10 CFR 50.59 for Crane Upgrade, dated March 13, 2010. Near the end of the inspection period, the licensee provided the inspectors additional information relevant to the design basis and licensing basis of the reactor building crane and reactor building crane support structure which will require additional review. Therefore, this issue is considered an unresolved item (URI 05000263/2010002-01, Reactor Building Crane Design and Licensing Basis Issues) pending additional inspector review to determine design and licensing basis requirements | ||
| 05000263/FIN-2010002-02 | 2010Q1 | Green | H.13 | NRC identified | SRV Low Low Set Surveillance Procedure Implementation | The inspectors identified a finding of very low safety significance and NCV of Technical Specification 5.4.1 for the licensee failing to appropriately implement an applicable procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, when unexpected local alarms were received during the performance of the safety relief valve (SRV) low low set system quarterly test, Instrument and Control (I&C) personnel elected to attempt to clear the alarms prior to notifying operations and without fully understanding which alarms were present. The surveillance procedure provided no guidance on how to clear the unexpected module trip alarms and relay energized lights. The licensee entered this issue into their corrective action program. The inspectors determined that the performance deficiency affected the cross-cutting area of Human Performance, having decision-making components, and involving aspects associated with using conservative assumptions indecision making. H.1(a)The inspectors determined that the performance deficiency was more than minor and a finding because it was associated with the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations, using the Phase 1 Worksheet for the Barrier Integrity Cornerstone. Since the inspectors answered no to all four questions in the Containment Barrier column of the Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance |
| 05000263/FIN-2010002-03 | 2010Q1 | Green | P.2 | NRC identified | Inadequate Corrective Actions for Unexpected SRV Low Low Set Trips Encountered During Surveillance Testing | The inspectors identified a finding of very low safety significance and NCV of10 CFR 50, Appendix B, Criterion XVI, for the licensees failure to adequately evaluate and take corrective actions for a condition adverse to quality. Specifically, the licensee failed to appropriately evaluate the implications of the unexpected trips of high/low pressure switches, PSHL-4065A and PSHL-4066A, during the January 28, 2009, performance of the SRV low low set system quarterly tests and implement appropriate corrective actions. The failure to adequately evaluate the unexpected trips and correct the condition adverse to quality directly contributed to a repeat occurrence and subsequent unplanned Technical Specification Action entry during the January 27, 2010, performance of the same surveillance test. The licensee entered the issue into their corrective action program. The inspectors determined that the performance deficiency affected the cross-cutting aspect in the area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with the licensee thoroughly evaluating problems such that the resolutions address causes and extent of conditions, as necessary. P.1(c) The inspectors determined that the performance deficiency was more than minor and a finding because it was associated with the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations, using the Phase 1 Worksheet for the Barrier Integrity Cornerstone. Since the inspectors answered no to all four questions in the Containment Barrier column of the Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance |
| 05000263/FIN-2010002-04 | 2010Q1 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified that during the Monticello 2008 ISFSI loading campaign a contractor performing non-destructive examinations (NDE) on two dry shielded canisters (DSCs) failed to follow a step of TriVis Procedure QP-9.202,Revision 3, Color Contrast Liquid Penetrant (PT) Examination Using the Solvent-Removable Method, which required the base metal temperature of the surface for the NDE to be below 250 degrees F. Specifically, the base metal temperature for DSC 002 was 270 degrees F and 253 degrees F for DSC 005 when the NDE was performed. The licensee immediately evaluated the situation and held a stand down to emphasize the importance of procedural adherence. Since the loaded and welded canisters were already stored in the Horizontal Storage Modules on the ISFSI pad at the time of discovery of the discrepancies, the licensee performed qualification PT examinations using two comparator blocks at a procedural non-standard temperature of 325 degrees F. The NDE products adequately identified flaws on the comparator blocks even beyond the qualified procedure range for the base metal temperature. Title 10 CFR 72.150, Instructions, Procedures, and Drawings, requires, in part, that the licensee prescribe activities affecting quality by documented instructions, procedures, or drawings of a type appropriate to the circumstances and requires that these instructions, procedures, and drawings be followed. The violation was addressed by traditional enforcement since 10 CFR Part 72 is not risk based and is not covered under the Reactor Oversight Process. The inspectors reviewed the examples in the Enforcement Policy, Supplement I and determined that the failure to follow the procedure was a violation that had more than minor safety or environmental significance but did not rise to a Severity Level I, II, or III violation due to the above mentioned results of the comparator blocks and the fact that the test port plug (weld no. 5) did not serve a structural or pressure retaining function. The inspectors determined that the violation had more than minor safety significance because failure to follow procedures and keep the base metal temperature below the temperatures at which the PT examination materials function properly could have lead to errors in identifying potential flaws in more critical welds that serve as pressure boundaries. The licensee entered these issues into its corrective action program as AR 01156986 and AR 01155771 | |
| 05000263/FIN-2010003-01 | 2010Q2 | Green | H.5 | NRC identified | Failure to Comply with Turbine Floor Heavy Lift Procedure | A finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensee, on two occasions during the lift and transfer of the General Electric Zinc Injection Passivation (GEZIP) skid, failing to adhere to the load height restrictions documented in Procedure 8117, Turbine Maintenance Procedure Heavy Load Movement over Safe Shutdown Equipment on the Turbine Floor, a procedure affecting quality. This resulted in the licensee not evaluating and managing the risk associated with moving a heavy load above and in close proximity to the Division I emergency service water piping. The licensee immediately placed a restriction on moving heavy loads on the turbine floor until the appropriate corrective actions can be implemented. The inspectors determined that the performance deficiency affected the cross-cutting area of Human Performance, having work control components, and involving aspects associated appropriately planning work activities by incorporating risk insights. (H.3(a) The inspectors determined that the failure to adequately evaluate two deviations from the acceptable heavy load path for the transport and placement of the new GEZIP skid was a performance deficiency, because it was the result of the failure to meet a requirement and the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Appendix B, and determined that the issue was more than minor because it could reasonably be viewed as a precursor to a significant event. Specifically, the licensee failed to manage the risk of moving a heavy load above and in close proximity to the Division I emergency service water piping. |
| 05000263/FIN-2010003-02 | 2010Q2 | Green | H.4 H.5 | NRC identified | Unacceptable Preconditioning of 250 Vdc Battery Chargers | The inspectors identified a finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1 for the licensees failure to appropriately implement an applicable procedure recommended in Regulatory Guide 1.33, Appendix A, Revision 2, February 1978. Specifically, the licensee approved TS surveillance activities to commence for the 250 Vdc battery chargers in 2008 without ensuring that the equipment was tested in the as-found condition. Due to improper sequencing of preventive maintenance activities for the battery chargers, and subsequent inadequate review of the maintenance and testing order, the 250 Vdc battery chargers were unacceptably preconditioned prior to performing testing to satisfy the 24 month TS Surveillance Requirement 3.8.4.2. These issues were identified by the inspectors prior to the 2010 performance of the same surveillance tests. The licensee took immediate corrective actions and entered the issues into their corrective action program. The inspectors determined that the performance deficiency affected the cross-cutting area of Human Performance, having work control components, and involving aspects associated with appropriately coordinating work activities by incorporating actions to address the impact of the work on different job activities. (H.3(b) The inspectors determined that the issue was a performance deficiency because it was the result of the failure to meet a requirement, and the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. The inspectors determined that the performance deficiency was more than minor and a finding because, if left uncorrected, it would have had the potential to lead to a more significant safety concern. The inspectors applied IMC 0609, Attachment 4, \"Phase 1 Initial Screening and Characterization of Findings\" to this finding. Under Column 2 of the Table 4a worksheet, the inspectors answered \"Yes\" to Question 1 because the finding did not result in loss of operability or functionality. Therefore, the finding was considered to be of very low safety significance. |
| 05000263/FIN-2010004-01 | 2010Q3 | Green | H.5 | Self-revealing | Inadequate Electrical Isolation during Demolition Activity | A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to adequately implement the requirements of their fleet tagging procedure, a procedure affecting quality, during the demolition of the A train of the combustion gas control system (CGCS). This failure directly led to workers being unprotected from existing 24 Vdc, and potentially 120 Vac, during the removal of cables C259-SV40008A/1 and C259-SV4009A/1. In addition, cutting of the energized cables resulted in the loss of position indication for three primary containment isolation valves which are required by Technical Specifications. The licensee promptly took actions to restore the affected containment isolation valves to an operable status and entered this event into their corrective action program for further evaluation. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having work control components, and involving aspects associated with appropriately coordinating work activities by incorporating job site conditions which may impact human performance and plant systems and components. (H.3(a)) The inspectors determined that the licensees failure to adequately implement their work order planning and tagging processes to protect workers and equipment from existing electrical hazards during the demolition of the A train of the CGCS system was a performance deficiency because it was the result of the failure to meet a requirement; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. Since the finding directly resulted in the loss of position indication for three containment isolation valves which are required by Technical Specifications, the inspectors evaluated the finding under the Containment Barrier Cornerstone. Utilizing Column 4 of the Table 4a worksheet, the inspectors answered Yes to question 1. Since the finding only resulted in the degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment (SBGT) system, the finding was screened to be of very low safety significance. |
| 05000263/FIN-2010005-01 | 2010Q4 | Green | H.8 | NRC identified | Failure to Properly Store Loose Material in Close Proximity to Safety Related Equipment | A finding of very low safety significance was identified by the inspectors when the licensee failed to properly control loose material located above the sensing lines for the safety-related residual heat removal pump minimum flow switches. No violation of NRC requirements associated with this finding was identified. Once informed of the issue, the licensee took action to relocate the material to a proper storage location. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having Work Practices components, and involving aspects associated with the licensee defining and effectively communicating expectations regarding procedural compliance and personnel following procedures. (H.4(b)) The inspectors determined that the licensees failure to properly store loose material located in close proximity to safety-related equipment was a performance deficiency, because it was the result of the failure to meet a requirement; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it impacted the protection against external events attribute of the Mitigating System Cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 2 of the Table 4a worksheet to screen the finding. As a result of the inspectors answering No to all five questions, the finding was screened to be of very low safety significance. |
| 05000263/FIN-2010005-02 | 2010Q4 | Green | P.2 | NRC identified | Failure to Implement Corrective Actions to Address a Deficiency Associated with the Door Interlock on Airlock 413 | A finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors when the licensee failed to implement corrective actions for a condition adverse to quality. The condition adverse to quality was a deficiency associated with the door interlock on airlock 413 which contributed to a loss of secondary containment boundary event. Subsequent to the August 5, 2010, event, the licensee initiated administrative controls on all airlocks with a similar design to airlock 413 and are currently evaluating other means of addressing air lock integrity. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, having Corrective Action components, and involving aspects associated with thoroughly evaluating problems such that the resolution addresses the causes and extent of condition as necessary. (P.1(c)) The inspectors determined that the licensees failure to implement corrective actions for a condition adverse to quality was a performance deficiency because it was the result of the failure to meet a requirement; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it impacted the configuration control attribute of the Barrier Integrity Cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors applied Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. Since the finding resulted in a momentary loss of the secondary containment boundary, the inspectors evaluated the finding under the Containment Barrier Cornerstone. Utilizing Column 4 of the Table 4a worksheet, the inspectors answered Yes to Question 1. Since the finding only resulted in the degradation of the radiological barrier function provided for the control room; auxiliary building; spent fuel pool; or standby gas treatment system; the finding was screened to be of very low safety significance. |
| 05000263/FIN-2010005-03 | 2010Q4 | Green | Licensee-identified | Licensee-Identified Violation | Title 10 CFR 50.72(b)(3)(v) requires, in part, that the licensee shall notify the NRC as soon as practical, and in all cases within eight hours, of the occurrence of any condition that at the time of discovery could have prevented the fulfillment of the safety function of systems that are needed to: shut down the reactor and maintain it in a safe shutdown condition, and mitigate the consequences of an accident. Contrary to this requirement, the licensee failed to report to the NRC on November 5, 2010, the inoperability of all four average power range monitors (APRMs); a condition applicable to 10 CFR 50.72(b)(3)(v), within eight hours. Specifically, because the APRMs provide the reactor protection system with scram safety functions, the inoperability of all APRMs resulted in a condition that could have prevented the fulfillment of specific reactor protection system safety functions. Per NRC Enforcement Policy, Section 6.9.d.9, the failure to make a report required by 10 CFR 50.72 is categorized as a Severity Level IV violation. Because this violation was not repetitive or willful, and it was entered into the licensees corrective action program as CAP 01257379, this violation is being treated as a Severity Level IV NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. | |
| 05000263/FIN-2010006-01 | 2010Q1 | Severity level IV | H.9 | NRC identified | Failure to Perform 10 CFR 50.59 Evaluation For Isolation of Room Cooler Which Addressed Temperature Limitations | The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.59, Changes, Tests, and Experiments, Section (d) 1 for the licensees failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis which addressed room temperature limitations as to why the isolation of a high pressure coolant injection (HPCI) room cooler did not require prior NRC approval. The licensee entered this issue into their corrective action program and determined that no immediate corrective actions were necessary because administrative controls were in place to ensure that the HPCI room temperature would not exceed the calculated initial room temperature limitation. The inspectors determined that the finding was more than minor because they could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that the finding was of very low safety significance because the finding did not result in loss of operability or functionality. The finding affected the Mitigating Systems cornerstone attribute of Equipment Performance to ensure the availability and reliability of systems (HPCI) that respond to initiating events to prevent undesirable consequences. This finding has a cross-cutting aspect in the area of human performance within the resources component because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety in that training of personnel was not sufficient. |
| 05000263/FIN-2010007-01 | 2010Q4 | NRC identified | HELB Analysis Potentially Non-Conservative | As part of the review of the ACE for an adverse trend in double disc gate valve (DDGV) local leak rate testing (LLRT) performance documented in CAP 1202466, the inspectors noted that the ACE had determined that the valves performance degradation did not prevent the valves from performing their safety function. The ACE only addressed the valves safety function of providing containment isolation. The inspectors questioned if the safety function of the high pressure coolant injection (HPCI), reactor core isolation cooling (RCIC) and reactor water cleanup (RWCU) steam supply valves to close after detection of a HELB should have been considered. The licensee responded that the ACE did not need to consider the effect of the valves increased leakage on the HELB analyses because any leakage would not impact the alternate shutdown path. The inspectors reviewed the assumptions and acceptance criteria of the HELB calculations for HPCI, RCIC and RWCU line breaks and identified potential inconsistencies between the calculations assumptions with Technical Specifications and UFSAR allowed values for valve closure times, incorporation of delay actuations, and isolation initiation signals. The licensee entered the NRC concerns with these potential inconsistencies into the CAP by initiating CAP 01252363 on October 1, 2010. The licensee stated that the calculations were appropriate and provided the inspectors with some original licensing documents for the HELB analyses; however, additional questions remained. This issue will be tracked as URI 05000263/201007-01 pending further NRC review of the licensee responses and the HELB analyses and determination of the original and current licensing bases. | ||
| 05000263/FIN-2010008-01 | 2009Q2 | NRC identified | Potential concern with the one-time inspection program related to butt welds | In the March 25, 2006, license renewal annual update (ML060800360), the licensee provided details of changes made to the original license renewal application. With respect to Class 1 Small-Bore piping, the licensee committed to the following: Upon further review subsequent to submitting the LRA to identify inspection locations, it was determined that all piping in this inspection group is of actual diameter 2 inches and less. In accordance with the plant piping specification, only socket weld connections are used in such applications and no butt welds. Therefore, inspections of this piping for the One-Time Inspection Program will consist of VT-2 examinations during pressure testing for system leaks upon return to service from outages and destructive examinations of any socket welds removed from service prior to the period of extended operation. In a letter dated May 14, 2010 (ML101370259), the licensee notified the NRC of the existence of a limited number of small-bore stainless-steel butt weld connections (less than 4 inches in diameter) which are contrary to the requirements of plant piping specifications. As a result, the licensee changed the one-time inspection commitment again and committed to perform augmented ISI volumetric examinations of ASME Class I stainless steel small-bore piping butt welds with a 2-inch nominal pipe size through less than 4-inch nominal pipe size in accordance with the ISI aging management program. The inspectors questioned whether the identification of stainless-steel butt welds constituted a newly identified component or whether the commitment change was appropriate. This is considered an unresolved item (URI 05000263/2010008-01) pending further discussions with the NRC program office. | ||
| 05000263/FIN-2010009-01 | 2010Q1 | Severity level IV | NRC identified | Failure of an NDE Technician to Follow an Ultrasonic Thickness Examination Procedure | A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified by the inspectors for a contract Non-Destructive Examination (NDE) technicians failure to follow a procedure during an Ultrasonic (UT) examination of the Reactor Core Isolation Cooling (RCIC) barometric condenser shell. Specifically, the technician failed to properly perform a calibration of the UT examination equipment. The underlying performance deficiency (PD) associated with this violation did not result in a finding due to the minor safety-significance of the PD and hence the PD was not evaluated for cross-cutting aspects. Specifically, the PD was similar to Example 4b of IMC 0612, Appendix E, Examples of Minor Issues, in that, it involved an insignificant procedural error, failure to calibrate UT equipment per procedure. The failure had minimal impact on the UT readings (within UT test equipment tolerances). However, due to the willfulness of the violation, the violation was processed through the traditional enforcement process and assigned a Severity Level IV. Specifically, the NRC Enforcement Policy states that a violation may be considered more significant than the underlying non-compliance if it includes indications of willfulness. As part of its corrective actions, the licensee re-examined the technicians prior UT examinations and found insignificant variation between re-examined UT examination results and the technicians original UT examination result | |
| 05000263/FIN-2010403-01 | 2010Q3 | Green | Licensee-identified | Security | ||
| 05000263/FIN-2011002-01 | 2011Q1 | Green | H.5 | Self-revealing | Inadequate System Isolation during Check Valve Maintenance | A finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to adequately implement the requirements of their fleet tagging procedure, a procedure affecting quality, during maintenance on the safety-related CST-88 B low pressure coolant injection (LPCI) fill line check valve. This failure resulted in an unintentional breach of the condensate service water (CSW) system and subjected workers to a potentially contaminated, pressurized water source. Additionally, at the time of the breach, the CSW system was one of the water sources being credited in support of the shutdown safety function of inventory control. The licensee entered this issue into the corrective action program (CAPs 1275935 and 1275963) and took immediate corrective actions to restore the check valve to its installed configuration to terminate the water leakage. At the time of this report, the licensee had assembled a team to perform a root cause evaluation. The inspectors determined that the licensees failure to adequately implement their tagging process to protect workers and equipment from the effects of breaching the pressurized CSW header during maintenance on a safety-related check valve was a performance deficiency because it was the result of the failure to meet a requirement, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because the performance deficiency could have reasonably been viewed as a precursor to a more significant event. In this instance, the performance deficiency resulted in an unintentional breach of the operating CSW system and subjected workers to a potentially contaminated, pressurized water source. Additionally, at the time of the breach, the CSW system was one of the water sources being credited in support of the shutdown safety function of inventory control. As a result, this finding was evaluated under the Initiating Events Cornerstone. The inspectors applied NRC IMC 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination, Attachment 1, to this finding. The finding was determined to have very low safety significance because it did not adversely affect core heat removal, inventory control, power availability, containment control, or reactivity guidelines. This finding has a cross-cutting aspect in the area of Human Performance, work control, because the licensee failed to appropriately plan work activities by incorporating job site conditions impacting plant systems and components (H.3(a)). |
| 05000263/FIN-2011002-02 | 2011Q1 | NRC identified | Calculation of Work Hours during Fatigue Rule Implementation | During this inspection, the inspectors identified a concern regarding the licensees implementation of fatigue rule requirements. Specifically, the inspectors reviewed an apparent cause evaluation (ACE) that the licensee had performed after identifying several violations of NRC fatigue rule requirements. The inspectors noted that one of the corrective actions developed and implemented in October 2010, as a result of this evaluation, involved tripling the period of planned shift turnover time on the front end of schedules of individuals in one department, to account for the turnover period on the back end of the shift. As a result of this action, the scheduled turnover period for personnel in this department was not consistent with NRC guidance on reasonable amounts of time for these activities. In addition, the inspectors noted that this turnover time period was applied to the front end of the schedules of all personnel in this department regardless of the amount of time spent performing actual turnover activities. This may potentially be in conflict with NRC regulations, specifically with respect to 10 CFR 26.205(b)(1), regarding calculation of work hours, 10 CFR 26.205(d) regarding work hour controls, and 10 CFR 26.203(b)(2) regarding implementation of fatigue rule procedures to ensure compliance with 10 CFR 26.205. The NRC inspectors plan to review actual turnover activities and associated records for the site as a whole to examine how the corrective action of concern has been put into practice. Pending NRC review of additional licensee information regarding site-wide practices for exclusion of shift turnover activities, as well as information on how the application of a fixed and potentially artificially long turnover period has affected actual work hours reported for individuals at the site, this issue will be treated as an Unresolved Item (URI) (URI) 5000263/2011002-02; Calculation of Work Hours during Fatigue Rule Implementation). | ||
| 05000263/FIN-2011002-03 | 2011Q1 | Green | H.8 | NRC identified | Failure to Control a Level 1 FME Area during New Fuel Receipt Activities | A finding of very low safety significance and associated NCV of Technical Specification 5.4, Procedures, was identified by the inspectors when the licensee failed to implement the requirements of their foreign material exclusion (FME) and control procedure during new fuel receipt activities. Specifically, the inspectors observed two operators exiting and re-entering a Level 1 FME area, without the knowledge of the FME monitor, at a point that was not being controlled by the FME monitor. When informed of the issue, the licensee took corrective actions to address the issue. The inspectors determined that the licensees failure to adequately implement the requirements of their FME control procedure during new fuel receipt activities to prevent the unmonitored access of two operators into a Level 1 FME area was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it impacted the human performance attribute of the Barrier Integrity Cornerstones objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 3 of the Table 4a worksheet to screen the finding. Since the finding only had the potential to impact the fuel barrier, it screened to be of very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel following procedures (H.4(b)). |
| 05000263/FIN-2011002-04 | 2011Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification LCO 3.0.4 states, in part, that when an LCO is not met, entry into a MODE in the applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the applicability for an unlimited period of time. Technical Specification LCO 3.6.1.3, Primary Containment Isolation Valves (PCIVs), states, in part, that each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE in MODES 1, 2, and 3, when associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation. Technical Specification LCO 3.3.6.1 states, in part, that the primary containment isolation instrumentation for Function 1, Main Steam Line Isolation, shall be OPERABLE for the Reactor Vessel Water Level Low Low, Main Steam Line Flow High, and Main Steam Line Tunnel Temperature High functions in MODES 1, 2, and 3. Contrary to the requirement of TS LCO 3.0.4, on November 22, 2010, the inboard and outboard main steam line PCIVs were not operable (unable to automatically close on a primary containment isolation signal due to an electrical isolation) prior to entering Mode 2, and the associated actions to be entered did not permit continued operation in Mode 2 for an unlimited period of time. Once identified, the licensee restored electrical power to the PCIVs and entered the issue into the corrective action program as CAP 01259879. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. Using the Table 4a worksheet, the inspectors answered Yes to Question 3 and applied IMC 0609, Appendix H, Containment Integrity Significance Determination Process. Per IMC 0609, Appendix H, the finding was considered a Type B finding; that is, a finding that has potentially important implications for integrity of containment without affecting the likelihood of core damage. Table 6.2, Phase 2 Risk Significance Type B Findings at Full Power, provided the risk significance for this finding. For BWR Mark I reactor types, the significance of Type B findings for less than three days duration is Green. | |
| 05000263/FIN-2011002-05 | 2011Q1 | Severity level Enforcement Discretion | NRC identified | Containment Overpressure Not Ensured in the Appendix R Analysis | The licensee issued Licensee Event Reports (LER 05000263-2009-001-00 and LER 05000263-2009-001-01) regarding the licensees failure to consider the spurious opening and venting of the primary containment, via purge and vent valves, in the event of a fire in the main control room or cable spreading room. Both LER revisions were closed in Inspection Report 05000263/2009004 and documented as a violation of NRC requirements. Because the licensee was transitioning to NFPA 805 and the violation met the criteria established by the NRC Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR Part 50.48(c)) for licensee in NFPA 805 transition, the NRC exercised enforcement discretion to not cite the violation in accordance with the NRCs Enforcement Policy. On December 22, 2010, the licensee provided an update to LER 05000263-2009-001-02 to reflect their withdrawal of the letter of intent to voluntarily implement 10 CFR 50.48(c) at the MNGP. On May 14, 2009, the NRC issued EGM 09-002, Enforcement Discretion for Fire Induced Circuit Faults, dated May 14, 2009, which authorized enforcement discretion for non-compliance issues associated with fire induced multiple circuit cable faults, providing that the licensee identified the non-compliances, entered them into their CAPs, and instituted appropriated compensatory measures until the issues were corrected, within the six month period following a planned revision to RG 1.189, Fire Protection for Nuclear Power Plants. Regulatory Guide 1.189, Revision 2, issued in October 2009, provided a method acceptable to the NRC to evaluate and resolve multiple fire induced circuit faults. After the six month period designated for the identification of non-compliances, the EGM further authorized enforcement discretion for an additional 30 month period, for licensees to resolve the identified multiple fire-induced circuit fault issues. The inspectors screened this violation and determined that because the violation was associated with multiple fire induced circuit faults and was identified during the discretion period as described in EGM 09-002, the NRC is exercising enforcement discretion for this violation in accordance with EGM 09-002. | |
| 05000263/FIN-2011003-01 | 2011Q1 | Green | H.2 | Self-revealing | Poor Maintenance Practices Result in CV-3490 Failing Shut | A finding of very low safety significance was self-revealed when, on two separate occasions, CV-3490 (12 reactor feedwater pump recirculation to the condenser) failed closed while the 12 reactor feedwater pump was being placed in service. The cause of each failure was directly related to poor maintenance practices while performing work on CV-3490s valve positioner. Additionally, each failure resulted in an automatic trip of the 12 reactor feedwater pump. The licensee entered this issue into their corrective action program, corrected the mechanical issues, and performed an extent-of-condition review. The inspectors determined that the performance deficiency affected the cross-cutting area of Human Performance, having work practice components, and involving aspects associated with ensuring that supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. (H.4(c)) The finding was more than minor because it impacted the configuration control attribute of the Initiating Events Cornerstone objective of limiting those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 1 of the Table 4a worksheet to screen the finding. The inspectors answered No to the question does the finding contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available and, therefore, the finding was screened to be of very low safety significance. |
| 05000263/FIN-2011003-02 | 2011Q1 | Severity level IV | NRC identified | Failure to Update USAR for Cask Lift Height Restrictions | A Severity Level IV non-cited violation (NCV) of 10 CFR 50.71(e),Periodic Update of the Final Safety Analysis Report and an accompanying Green finding were identified by the inspectors for the licensees failure to update the Updated Safety Analysis Report (USAR) with the cask maximum lift height restrictions imposed by Nuclear Regulatory Commission (NRC) staff. As a result, the licensee had not adequately evaluated whether the plant licensing basis necessitated retention of cask lift height limitations when transitioning from the use of the 25 ton NFS-4 or 25 ton NAC-1 spent fuel shipping cask and 70 ton IF-300 spent fuel shipping cask to the heavier 105 ton NUHOMS cask. The licensee entered this issue into its corrective action system. The inspectors determined that the failure to update the USAR with the cask lift height restrictions for the 25 ton and 70 ton spent fuel cask was contrary to 10 CFR 50.71(e) and was a performance deficiency warranting a significance evaluation. Violations of 10 CFR 50.71 (e) are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying finding is evaluated under the SDP to determine the significance of the violation. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because, if left uncorrected, the performance deficiency could have led to a more significant safety concern. Specifically, the inspectors could not readily conclude that the absence of lift height limitations would not require additional calculational analyses and/or require a license amendment. The inspectors determined that the finding was of very low safety significance following a qualitative significance determination review. Specifically, the inspectors determined that only seismic events exceeding the level of an Operational Basis Earthquake (OBE) of 0.03g could impact core damage frequency (CDF). The licensee supplied information that the median annual probability of exceeding the peak ground acceleration for the OBE at Monticello was approximately 7.0E-4/yr. In addition, the predicted shipping cask lifts was 19.2/yr with an average lift duration of 30 minutes. Thus, the frequency of exceeding the OBE while lifting a shipping cask was estimated to be 7.7E-7/year. This value is a bounding frequency estimate for delta-CDF in that it does not imply with certainty that there will be a cask drop during an earthquake nor does it imply with certainty of core damage during an earthquake given a cask drop. The Senior Reactor Analyst (SRA) concluded that the risk due to simultaneous occurrence of an OBE or greater seismic event during use of the reactor building crane for shipping cask lifts was best characterized as very low (Green). The inspectors determined that this finding did not reflect current performance because it was a legacy issue with the failure to properly update the USAR occurring almost 30 years ago and, therefore, there was no cross-cutting aspect associated with this finding. | |
| 05000263/FIN-2011003-03 | 2011Q1 | Green | NRC identified | Failure to Update USAR for Cask Lift Height Restrictions | A Severity Level IV non-cited violation (NCV) of 10 CFR 50.71(e),Periodic Update of the Final Safety Analysis Report and an accompanying Green finding were identified by the inspectors for the licensees failure to update the Updated Safety Analysis Report (USAR) with the cask maximum lift height restrictions imposed by Nuclear Regulatory Commission (NRC) staff. As a result, the licensee had not adequately evaluated whether the plant licensing basis necessitated retention of cask lift height limitations when transitioning from the use of the 25 ton NFS-4 or 25 ton NAC-1 spent fuel shipping cask and 70 ton IF-300 spent fuel shipping cask to the heavier 105 ton NUHOMS cask. The licensee entered this issue into its corrective action system. The inspectors determined that the failure to update the USAR with the cask lift height restrictions for the 25 ton and 70 ton spent fuel cask was contrary to 10 CFR 50.71(e) and was a performance deficiency warranting a significance evaluation. Violations of 10 CFR 50.71 (e) are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying finding is evaluated under the SDP to determine the significance of the violation. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because, if left uncorrected, the performance deficiency could have led to a more significant safety concern. Specifically, the inspectors could not readily conclude that the absence of lift height limitations would not require additional calculational analyses and/or require a license amendment. The inspectors determined that the finding was of very low safety significance following a qualitative significance determination review. Specifically, the inspectors determined that only seismic events exceeding the level of an Operational Basis Earthquake (OBE) of 0.03g could impact core damage frequency (CDF). The licensee supplied information that the median annual probability of exceeding the peak ground acceleration for the OBE at Monticello was approximately 7.0E-4/yr. In addition, the predicted shipping cask lifts was 19.2/yr with an average lift duration of 30 minutes. Thus, the frequency of exceeding the OBE while lifting a shipping cask was estimated to be 7.7E-7/year. This value is a bounding frequency estimate for delta-CDF in that it does not imply with certainty that there will be a cask drop during an earthquake nor does it imply with certainty of core damage during an earthquake given a cask drop. The Senior Reactor Analyst (SRA) concluded that the risk due to simultaneous occurrence of an OBE or greater seismic event during use of the reactor building crane for shipping cask lifts was best characterized as very low (Green). The inspectors determined that this finding did not reflect current performance because it was a legacy issue with the failure to properly update the USAR occurring almost 30 years ago and, therefore, there was no cross-cutting aspect associated with this finding. | |
| 05000263/FIN-2011003-04 | 2011Q1 | Green | H.5 | Self-revealing | Failure to Maintain Radiation Exposure ALARA During Inboard Main Steam Isolation Valve Repair | A finding of very low safety significance (Green) was self-revealed due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees as-low-as-is-reasonably-achievable (ALARA) planning and work control program. Specifically, the licensee failed to properly incorporate ALARA strategies or insights while planning and executing a maintenance activity on the C inboard main steam isolation valve. This issue resulted in the expansion of collective exposure for this work from 4.044 person-rem to 9.654 person-rem. The licensee entered this issue into their corrective action program as CAP 1281395. The finding was more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone. Additionally, this issue affected the cornerstone objective of ensuring the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Also, the finding was similar to Example 6.i in Appendix E of IMC 0612, in that it resulted in a collective exposure of greater than 5 person-rem and exceeded the outage goal by greater than 50 percent. The inspectors determined that this finding was of very low safety significance because Monticello Nuclear Generating Plants (MNGPs) current three-year rolling average collective dose is 136.266 person-rem, less than the 240 person-rem per unit standard. This finding had a cross-cutting aspect in the area of Human Performance, related to the cross-cutting component of work control, in that the outage plan did not adequately incorporate actions to address the impact of work on different job activities. |
| 05000263/FIN-2011003-05 | 2011Q1 | Green | H.14 | Self-revealing | Inadequate C.1 Startup Procedure Review | A finding of very low safety significance and non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when an unexpected recirculation pump runback occurred during the performance of Reactor Dynamics Testing. The event was the result of the licensee failing to adequately assess the operational impact of a recent revision to Procedure C.1, Startup Procedure, which resulted in operating the plant in a manner that challenged feedwater pump protective features. The licensee entered this issue into their corrective action program (CAP 01288070) and initiated corrective actions to address the issue. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having decision-making components, and involving aspects associated with the licensee conducting effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions, identify possible unintended consequences, and determine how to improve future decisions. (H.1(b)) The finding was more than minor because it impacted the procedure quality 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. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 1 of the Table 4a worksheet to screen the finding. The inspectors answered No to the questions associated with loss-of-coolant accident (LOCA) initiators, transient initiators, and external events initiator, and screened the finding to be of very low safety significance. |
| 05000263/FIN-2011004-01 | 2011Q3 | Green | NRC identified | NOED for Emergency Diesel Generator Load Rejection Surveillance Requirement 3.8.1.7 | On September 27, 2011, during an engineering self-assessment, the licensee identified a potential issue associated with the testing methodology used to demonstrate each EDGs capability to withstand the rejection of an electrical load that is equivalent to the single largest post-accident electrical load. On September 29, 2011, the licensee verified that their existing surveillance test OSP-ECC-0566, Low Pressure ECCS (emergency core cooling system ) Automatic Initiation and Loss of Auxiliary power Test, Revision 8, did not ensure that the load rejection test was performed with sufficient load to satisfy the requirements of SR 3.8.1.7 (Verify each EDG rejects a load greater than or equal to its associated single largest post-accident load and, following load rejection, the frequency is less than or equal to 67.5 Hz.). On September 29, 2011, at approximately 1700, the licensee declared both 11 and 12 EDGs inoperable and entered the Action for TS 3.8.1.E, Two EDGs Inoperable. At approximately 2200, the licensee requested enforcement discretion to extend the Action Completion Time for TS 3.8.1.F, from twelve hours to five days, to allow time to perform the required EDG load rejection testing. At approximately 23:58, the Agency granted NOED 11-3-001. The inspectors evaluation of the issue included a review of the technical documents associated with the issue and several meetings with the licensee management and technical staff. The initial information gained by the inspectors and their assessment of the issue was communicated to senior agency managers well in advance of the licensees NOED request, significantly contributing to the Agencys understanding and appropriate disposition of the issue. Additional information associated with the inadequate surveillance procedure and EDG operability is documented in Section 1R15 of this report. | |
| 05000263/FIN-2011004-02 | 2011Q3 | Green | H.11 H.12 | NRC identified | Failure to Follow Emergency Diesel Generator Quarterly Surveillance Procedure | The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to follow the quarterly emergency diesel generator (EDG) surveillance procedure during testing of the EDG air start system. Specifically, the licensee failed to follow a procedural step that involved in-service testing of a check valve in the EDG air start system that, if degraded, could allow air to bleed out of the starting air tanks which are required for diesel generator operability. The licensee entered this issue into their corrective action program (CAP), and corrective actions for this issue included suspension of the test, performance of a Human Performance Investigation Team review, and disqualification of the individual performing the test. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having work practices components, and involving aspects associated with using human error prevention techniques during performance of work activities. (H.4(a)) The inspectors determined that the licensees failure to follow their EDG surveillance procedure was a performance deficiency, because it was the result of the failure to meet a requirement; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because the performance deficiency was associated with the Human Performance attribute of the Mitigating Systems Cornerstone and affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). As a result, this finding was evaluated under the Mitigating Systems Cornerstone. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 2 of the Table 4a worksheet to screen the finding. The finding was determined to have very low safety significance because the inspectors answered No to all five questions. |
| 05000263/FIN-2011004-03 | 2011Q3 | Green | NRC identified | Shipping and Transportation of a Radioactively Contaminated Condensate Demineralizer Vessel | On July 14, 2011, it was reported to the licensee by the driver of the vehicle that there was a puncture in the side of a container package on radioactive material shipment number 11-127. The package was a Sealand box inside an enclosed conveyance. The Sealand box contained a radioactively contaminated condensate demineralizer vessel and the puncture was a nominal 4 by 6 inch hole. There was no spread of contamination as a result of the compromised package. The inspectors initial review determined that a performance deficiency exists, in that, the shipping container contents was inappropriately braced and blocked for transport. Regulations require that licensees ensure that loads not shift under conditions normally incident to transportation. The inspectors will review the additional information provided by the licensee and determine the significance of the performance deficiency. | |
| 05000263/FIN-2011005-01 | 2011Q4 | Green | H.5 | NRC identified | E Condensate Demineralizer Alarm Response Procedure Limits Exceeded | The inspectors identified a finding of very low safety significance and non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures, when the operators did not take conservative action to address a high differential pressure condition on an inservice condensate demineralizer vessel. Specifically, operators allowed the E condensate demineralizer to exceed differential pressure operating limits prescribed in Alarm Response Procedure 80-DPAH-2215, Vessel T-7E D/P High, and remain above those prescribed limits for approximately a shift before taking action to correct the abnormal condition. Specific corrective actions taken by the licensee to address this issue included updating the applicable alarm response procedures and operating procedures to reflect current system limitations; engineering management reinforcing the expectation that informal processes are not acceptable when communicating technical guidance to operations staff; and site management reinforcing the expectation that, once a degrading trend is recognized, actions must be taken in sufficient time to prevent crossing established operating limits. The inspectors determined that the licensees failure to maintain the E condensate demineralizer differential pressure within prescribed operational limits was a performance deficiency because it was the result of the failure to meet a requirement or a standard; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it impacted the Human Performance 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. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 1 of the Table 4a worksheet to screen the finding. For transient initiators, the inspectors answered no to the question, Does the finding contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment of functions will not be available, and determined the finding to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having Work Control components, and involving aspects associated with the licensee planning and coordinating work activities, consistent with nuclear safety, specifically the need for planned contingencies, compensatory actions, and abort criteria |
| 05000263/FIN-2011005-02 | 2011Q4 | Green | H.10 | Self-revealing | Inadequate Completion of CAPRs Associated with 2RS to 2R Feeder Cable Testing | A finding of very low safety significance and NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was self-revealed following a reactor scram, which was the direct result of an electric plant realignment caused by a faulted feeder cable and lockout of the stations 2R transformer. Specifically, annual testing to monitor the performance of the 2R feeder cables, which was put in place as a corrective action to prevent recurrence to address issues identified subsequent to a similar event in 2008, had not been performed since the cables were placed back in service following that event. To address the identified material deficiencies, the licensee replaced and tested the electrical cables between 2RS and 2R in their entirety, employing a new route designed to avoid cable submergence. Additional corrective actions were put in place to strengthen the licensees planned maintenance deferral process and their cable condition monitoring program. The inspectors determined that the licensees failure to perform annual testing of the 2R transformer feeder cables, as required by the stations planned maintenance program, was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. The inspectors determined that the issue was more than minor because it impacted 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. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 1 of the Table 4a worksheet to screen the finding. Because the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would not be available, the Region III Senior Reactor Analyst (SRA) performed a Phase 3 analysis, and screened the finding to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having decision-making components, and involving aspects associated with the licensees making safety-significant or risk-significant decisions using a systematic process to ensure safety is maintained |
| 05000263/FIN-2011005-03 | 2011Q4 | Green | H.12 | Self-revealing | Rod Worth Minimizer Inoperable During Reactor Plant Startup | A finding of very low safety significance and NCV of TS 3.3.2.1, Control Rod Block Instrumentation, was self-revealed to the operating crew, when normal startup testing could not be accomplished due to improperly configured equipment. Specifically, the operating crew transitioned from Mode 4 to Mode 2, with the rod worth minimizer (RWM) mode switch in the BYPASS position. With the RWM mode switch in the BYPASS position and the required actions of 3.3.2.1(c) not met, the requirements of TS 3.3.2.1, that the RWM be operable in Mode 1 and Mode 2 when thermal power is less than or equal to 10 percent rated thermal power, could not be met. Actions taken by the licensee in response to this event included declaring the event a reactivity management event; making an NRC notification under 50.72(b)(3)(v)(D); resetting their site event clock; providing additional training for the applicable operating crew; and revising procedures associated with this event to clarify the sequencing of key activities associated with the transition between Mode 4 and Mode 2. The inspectors determined that the licensees failure to properly control the configuration of the RWM prior to entering an operating mode that required its operability was a performance deficiency, because it was the result of the failure to meet a requirement or a standard; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors determined that the issue was more than minor because it impacted 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. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors answered No to the questions associated with transient initiators and screened the finding to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, having work practices components, and involving aspects associated with personnel work practices that support human performance, specifically in the areas of pre-job briefing, self and peer checking, and proper documentation of activities |
| 05000263/FIN-2011005-04 | 2011Q4 | Green | P.5 | Self-revealing | Failure to Properly Block and Brace a Radioactive Package for Transport | The inspectors reviewed a self-revealed finding of very low safety significance and an associated NCV of 10 CFR 71.5. Specifically, the licensee failed to appropriately block and brace a radioactively contaminated condensate demineralizer vessel within a transport package, such that, the package contents would not compromise and penetrate the transport package. The issue has been entered into the licensees corrective action program as CR (condition report) 01294652. Corrective actions were implemented to address supervisions responsibilities during shipment preparation regarding appropriate blocking and bracing of package contents. The finding was more than minor because the performance deficiency could be reasonably viewed as a precursor to a significant event, in that, the penetration of the transportation package by its contents could lead to the inadvertent spread of radioactive contamination in the public domain. Using IMC 0609, Attachment D, for the Public Radiation Safety SDP, the inspectors determined the finding to be of very low safety significance. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution (operating experience) |
| 05000263/FIN-2011005-05 | 2011Q4 | Severity level IV | P.2 | NRC identified | Failure to Make a Required 60 Day Event Report Per 10 CFR 50.73(a)(2)(vii)(A-D) | The inspectors identified a Severity Level IV NCV and associated finding of very low safety significance of 10 CFR 50.73(a)(2)(vii)(A-D), Licensee Event Report System, for the failure to report an event to the NRC within 60 days, where a single cause or condition caused two independent trains to become inoperable in a single system designed to help maintain safe reactor shut down, remove residual heat, control radioactive releases, or mitigate accidents. Specifically, on September 29, 2011, the licensee identified that the surveillance test procedures being used to demonstrate load reject capabilities of both EDGs had never contained the correct load rejection testing requirements from the applicable design documents. As a result, the surveillances were considered never met, and both EDGs were declared inoperable. During their evaluation and subsequent reporting of the issue, the licensee failed to recognize that the inoperability of both diesel generators caused by a single common cause was reportable to the NRC within 60 days under the 50.73 common cause criterion. The licensee entered this issue into their corrective action program (CAP 1318116). Corrective actions for this issue included plans to revise their existing licensee event report (LER) and to perform an apparent cause evaluation to further evaluate the issue. The inspectors determined that the failure to report required plant events or conditions to the NRC in accordance with reporting requirements was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. In addition, it had the potential to impede or impact the regulatory process. As a result, the NRC dispositions violations of 10 CFR 50.73 using the traditional enforcement process instead of the SDP. However, if possible, the underlying technical issue is evaluated using the SDP. In this case, the inspectors determined that the licensee failed to develop and implement adequate Emergency Diesel Generator (EDG) testing procedures during their transition to the Improved Technical Specifications in 2006, which resulted in both EDGs being declared TS inoperable, but available for use. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attributes of Human Performance and Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding had very low safety significance because they answered No to all five questions contained in Column 2 of the Table 4a worksheet. As a result, the inspectors determined that the finding had very low safety significance (Green). In accordance with Section 6.9.d.9 and 6.9.d.10 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was an example where the licensee failed to make a report required by 10 CFR 50.73; it represented a failure to identify all applicable reporting codes on an LER that may impact the completeness or accuracy of other information submitted to the NRC; and the underlying technical issue was evaluated by the SDP and determined to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency affected the cross-cutting area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with properly classifying and evaluating for reportability conditions adverse to quality (P.1(c)). |
| 05000263/FIN-2011005-06 | 2011Q4 | Green | P.2 | NRC identified | Failure to Make a Required 60 Day Event Report Per 10 CFR 50.73(a)(2)(vii)(A-D) | The inspectors identified a Severity Level IV NCV and associated finding of very low safety significance of 10 CFR 50.73(a)(2)(vii)(A-D), Licensee Event Report System, for the failure to report an event to the NRC within 60 days, where a single cause or condition caused two independent trains to become inoperable in a single system designed to help maintain safe reactor shut down, remove residual heat, control radioactive releases, or mitigate accidents. Specifically, on September 29, 2011, the licensee identified that the surveillance test procedures being used to demonstrate load reject capabilities of both EDGs had never contained the correct load rejection testing requirements from the applicable design documents. As a result, the surveillances were considered never met, and both EDGs were declared inoperable. During their evaluation and subsequent reporting of the issue, the licensee failed to recognize that the inoperability of both diesel generators caused by a single common cause was reportable to the NRC within 60 days under the 50.73 common cause criterion. The licensee entered this issue into their corrective action program (CAP 1318116). Corrective actions for this issue included plans to revise their existing licensee event report (LER) and to perform an apparent cause evaluation to further evaluate the issue. The inspectors determined that the failure to report required plant events or conditions to the NRC in accordance with reporting requirements was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. In addition, it had the potential to impede or impact the regulatory process. As a result, the NRC dispositions violations of 10 CFR 50.73 using the traditional enforcement process instead of the SDP. However, if possible, the underlying technical issue is evaluated using the SDP. In this case, the inspectors determined that the licensee failed to develop and implement adequate Emergency Diesel Generator (EDG) testing procedures during their transition to the Improved Technical Specifications in 2006, which resulted in both EDGs being declared TS inoperable, but available for use. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attributes of Human Performance and Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding had very low safety significance because they answered No to all five questions contained in Column 2 of the Table 4a worksheet. As a result, the inspectors determined that the finding had very low safety significance (Green). In accordance with Section 6.9.d.9 and 6.9.d.10 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was an example where the licensee failed to make a report required by 10 CFR 50.73; it represented a failure to identify all applicable reporting codes on an LER that may impact the completeness or accuracy of other information submitted to the NRC; and the underlying technical issue was evaluated by the SDP and determined to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency affected the cross-cutting area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with properly classifying and evaluating for reportability conditions adverse to quality (P.1(c)). |
| 05000263/FIN-2011005-07 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Criterion XI requires, in part, that A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to this requirement, on September 29, 2011, the licensee identified that they had failed to utilize a test program which incorporated all requirements from the applicable design documents to demonstrate that both EDGs would perform satisfactorily in service. Specifically, the test procedures being used by the licensee to demonstrate operability of the EDGs did not contain the correct load rejection testing requirements from the applicable design documents. As a result, the licensee determined that they had never demonstrated that they met load rejection surveillance requirement 3.8.1.7, and that both EDGs were inoperable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding had very low safety significance because they answered No to all five questions contained in Column 2 of the Table 4a worksheet. The licensee developed new test procedures which included the appropriate acceptance criteria and test methodologies, satisfactorily tested both EDGs, and entered this issue into their CAP as AR 01305683, CDBI FSA-Question on Definition of Post Accident Load in TS and AR 1306107, largest post-accident load greater than in test OSP-ECC-0566. | |
| 05000263/FIN-2011005-08 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated Severity Level IV NCV of 10 CFR 50.72(b)(3)(v)(D). Title 10 CFR 50.72(b)(3)(v)(D) requires, in part, that operating reactor licensees shall notify the NRC within eight hours of the occurrence of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. Contrary to this requirement, on November 1, 2011, the licensee identified that they had failed to make a required non-emergency notification within eight hours for a safety system functional failure of the CREF and CRV systems. Specifically, on October 21, 2011, following a reactor scram, the No. 11 EDG ESW pump was declared inoperable due to low cooling water pump flow. The loss of this pump resulted in the No. 11 EDG, \'A\' CREF, and \'A\' CRV being declared inoperable when the redundant B Division CREF and CRV systems were already out-of-service due to preplanned maintenance. As a result, the licensee entered TS 3.0.3 due to both CRV and CREF systems being inoperable. The licensee failed to recognize that this represented a potential loss of safety function at the time of the event. The inspectors determined that the failure to report required plant events or conditions to the NRC was a performance deficiency, and it had the potential to impede or impact the regulatory process. The NRC dispositions violations of 10 CFR 50.72 using the traditional enforcement process, and if possible, the underlying technical issue is evaluated using the SDP. The underlying technical issue was associated with both trains of CREF/CRV being inoperable and unavailable during a scram, resulting from a lockout of the 2R transformer. This issue was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, because both trains of the CREF/CRV system were inoperable and unavailable, the regional SRAs performed a Phase 3 risk evaluation to determine the risk significance of the issue. As a result of the SRAs evaluation, the inspectors determined that the finding had very low safety significance. Because the failure to make the required 50.72 report had the potential to impede or impact the regulatory process, the inspectors used the Traditional Enforcement process to disposition the issue. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV. The licensee entered the issue into their CAP as AR 1310956, Missed 8 Hour Report, and made the required 50.72 report. | |
| 05000263/FIN-2011008-01 | 2011Q3 | Green | NRC identified | Hydrogen Bottles Located Below RHR System Cables | The inspectors identified a finding of very low safety significance and associated NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to evaluate the impact of the installation of the hydrogen/oxygen analyzer system on safety-related residual heat removal (RHR) system cables. Specifically, the licensee failed to evaluate how a failure of the hydrogen bottles and the resulting fire or explosion could impact RHR cables located directly above the hydrogen bottles. The licensee entered this issue into their corrective action program to review the placement of the hydrogen bottles. The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance. | |
| 05000263/FIN-2011008-02 | 2011Q3 | Green | NRC identified | Failure to Inspect and Test the MCR Air Intake Smoke Detector | The inspectors identified a finding of very low safety significance and associated NCV of License Condition 2.C.4 for the licensees failure to inspect and test the main control room (MCR) air intake smoke detector. Specifically, the licensee failed to inspect and test the smoke detector between 2006 and 2011 as required by the preventative maintenance program. The licensee successfully tested the detector once the performance deficiency was identified and entered this issue into their corrective action program to evaluate the status of the detector. The inspectors determined that the finding was more than minor because if left uncorrected, the failure to inspect and test the MCR air intake smoke detector would become a more significant safety concern. Specifically, if the licensee continued not testing and maintaining the detector it would eventually fail to respond properly and result in a delayed notification to control room operators of a fire that could result in smoke entering the control room. This finding was of very low safety significance because the licensee successfully tested the detector. This finding did not have a cross-cutting aspect because the finding was not representative of current performance. | |
| 05000263/FIN-2011010-01 | 2011Q4 | Green | NRC identified | Failure to Follow Fire Water Aging Management Program Implementing Procedure | The inspectors identified a finding of very low safety significance (Green) involving the licensees failure to accomplish activities affecting quality in accordance with procedures. Specifically, the licensee failed to incorporate operating experience in accordance with procedures. This impacted the licensees ability to implement an effective aging management program for the fire protection system. No violation of NRC requirements was identified. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Appendix F, Fire Protection SDP, and the Monticello SPAR model, the inspectors determined that this finding had very low safety significance. The inspectors did not identify an associated crosscutting aspect for this finding. | |
| 05000263/FIN-2011010-02 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 through a planned surveillance test for the failure to implement and maintain in effect all provisions of their approved fire protection program. Specifically, the installation of the intake structure pre-action sprinkler system did not comply with NFPA 13 (1983) section 3-11.1.1, which requires that all sprinkler pipe and fittings shall be so installed that the system may be drained and resulted in the plugging of the sprinkler system. This prevented water from flowing through sprinkler heads and caused the system to be non-functional. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The Region III Senior Risk Analyst (SRA) used the risk assessment tools of IMC 0609, Appendix F, Fire Protection SDP, and performed bounding analyses using the Monticello Standard Plant Analysis Risk (SPAR Model), Version 8.15. The SRA also reviewed and discussed the licensees bounding risk assessment documented in PRA Memo 11-01-, Revisions 0 and 1, Risk Assessment of Intake Fire Suppression System Plugging. The finding was determined to be of very low safety significance (green) because the risk increase using bounding assumptions was below 1E-6. The licensee entered this issue into their corrective action program as AR 01305183, Intake Fire Sprinkler Configuration Discrepancy, and restored the functionality of the sprinkler system by flushing the piping and replacing system components. The licensee further planned to modify the system to allow proper drainage in accordance with the design requirements. | |
| 05000263/FIN-2011010-03 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 for the failure to implement and maintain in effect all provisions of their approved fire protection program. This includes adhering to the 10 CFR 50, Appendix B Quality Assurance Program requirements for the design, procurement, installation, testing and administrative controls for the fire protection program. Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as deficiencies are promptly identified and corrected. Contrary to the above, from August 21, 2007 until August 26, 2011, the licensee failed to promptly identify and correct a condition adverse to quality that resulted in the plugging of the intake structure sprinkler system. Specifically, the licensee failed to perform corrective actions (work order 342675-02) to flush the intake structure sprinkler system following a blockage event in the EDG rooms in 2007. The performance deficiency was determined to be more than minor because the plugging in the intake structure pre-action sprinkler system was left uncorrected for four years and became a more significant safety concern. The inspectors concluded that this finding was associated with the Mitigating Systems cornerstone. The Region III SRA used the risk assessment tools of IMC 0609, Appendix F, Fire Protection SDP, and performed bounding analyses using the Monticello Standard Plant Analysis Risk (SPAR Model), Version 8.15. The SRA also reviewed and discussed the licensees bounding risk assessment documented in PRA Memo 11-01-, Revisions 0 and 1, Risk Assessment of Intake Fire Suppression System Plugging. The finding was determined to be of very low safety significance (Green) because the risk increase using bounding assumptions was below 1E-6. The licensee flushed the system, restored functionality, and wrote AR 01303860 to document the multiple rescheduling. | |
| 05000263/FIN-2011010-04 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 for the failure to implement and maintain in effect all provisions of their approved fire protection program. This includes adhering to the 10 CFR 50, Appendix B Quality Assurance Program requirements for the design, procurement, installation, testing and administrative controls for the fire protection program. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions procedures, or drawings. Contrary to the above, on April 30, 2009, the license failed to follow procedure FP-OP-OL-01 Operability/Functionality Determination, when assessing identified blockage in the intake structure fire protection sprinkler piping. Specifically, the assessor failed to justify assumptions, perform an extent of condition, and obtain additional condition bounding information to ensure an accurate assessment of the condition. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Appendix F, Fire Protection SDP, and the Monticello SPAR model, the inspectors determined that this finding had very low safety significance. The licensee entered this issue into their corrective action program as AR 01304353, Inaccurate functionality assessment for CAP 1180222, in order to perform further evaluation of the deficiency. | |
| 05000263/FIN-2011010-05 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | \ The licensee identified a finding of very low safety significance (Green) and associated NCV of License condition 2.C.4 for the failure to implement and maintain in effect all provisions of their approved fire protection program. This includes adhering to the 10 CFR 50, Appendix B Quality Assurance Program requirements for the design, procurement, installation, testing and administrative controls for the fire protection program. Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that test results shall be documented and evaluated to assure that test requirements have been satisfied. Contrary to this requirement, on April 30, 2009, the licensee failed to document and evaluate the results of a PMT that did not meet all of its acceptance criteria. Specifically, when a step in the PMT required flow through the inspector test valve was not accomplished, the PMT was not annotated as failure and the PMT work order was signed off as complete without further evaluation. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Appendix F, Fire Protection SDP, and the Monticello SPAR model, the inspectors determined that this finding had very low safety significance. The licensee entered this issue into their corrective action program as AR 01304348, Failed PMT results not captured in PMT WO, in order to perform further evaluation of the deficiency. | |
| 05000263/FIN-2011201-01 | 2011Q4 | Green | H.13 | NRC identified | Security | |
| 05000263/FIN-2011404-01 | 2011Q2 | Green | Licensee-identified | Licensee-Identified Violation |