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05000424/FIN-2015003-012015Q3VogtleFailure to Maintain Requalification Examination IntegrityAn NRC-identified Non-cited Violation (NCV) of 10 CFR 55.49, Integrity of examinations and tests, was identified for the licensees failure to adhere to requirements of NMP-TR-424, License Operator Continuing Training Exam Development, Version 3.1. NMP-TR-424 was the procedure that the licensee used to implement industry standard ACAD 07-001, Guidelines for the Continuing Training of Licensed Personnel. ACAD 07-001 is a methodology which can be used to fulfill 10 CFR 55.59(c), Requalification program requirements and 10 CFR 55.4, Systems approach to training (SAT). This violation has been entered into the licensees corrective action program (CAP) as condition report (CR) 10115484. The inspectors determined that the licensees failure to adhere to overlap standards in NMP-TR-424 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective in that the failure to adhere to examination overlap standards adversely affected the quality of the administration of the operating exams. The finding was determined to be of very low safety significance (Green) because there was no evidence that a licensed operator had actually gained an unfair advantage on an examination required by 10 CFR 55.59. The finding was directly related to the cross-cutting aspect of procedure adherence of the cross-cutting area of Human Performance because the training staff did not follow the guidance for all licensed operators 2014 annual operating exam.
05000424/FIN-2015003-022015Q3VogtleNRC Biennial Written Examinations did not Meet Qualitative StandardsAn NRC-identified finding was identified when between 20 and 40 percent of the written examination questions administered to licensed operators during the biennial requalification examination did not meet the requirements of NMP-TR-424, Licensed Operator Continuing Training Exam Development, and NUREG-1021, Operator Licensing Examination Standards For Power Reactors, Revision 10. The inspectors determined that the failure to ensure that biennial written examinations met the qualitative standards for written examinations was a performance deficiency (PD). The PD was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that the quality of biennial written examinations potentially impacted the licensees ability to appropriately evaluate licensed operators. The significance of the finding was determined to be Green because between 20 and 40 percent of the questions reviewed did not meet the standard. No cross-cutting aspect was identified that would be considered a contributor to the cause of the finding.
05000424/FIN-2015003-042015Q3VogtleLicensee-Identified ViolationTechnical Specifications 5.4.1.b, Procedures, required, in part, that written procedures shall be established, implemented, and maintained for the emergency operating procedures (EOPs) required to implement NUREG-0737, Clarification of TMI Action Plan Requirements, and Supplement 1 to NUREG-0737. Contrary to this requirement, as of August 15, 2007, the licensee failed to maintain EOPs required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1. The licensee failed to maintain EOP 19100-C, ECA-0.0 Loss of All AC Power, version 39, consistent with revised Pressurized Water Reactor Owners Group/Westinghouse Owners Group emergency response guidelines that restricted the RCS cooldown rate to less than 100 degrees Fahrenheit per hour to prevent thermal shock to the reactor coolant pump (RCP) seals following a loss of all alternating current power (SB) event. The licensee entered this violation into the CAP as CR 10066747 and revised EOP 19100-C consistent with the updated guidance. A bounding detailed risk evaluation was performed by an NRC regional senior risk analyst (SRA) who determined the finding to be of very low risk significance (Green). The dominant result was a grid-related Loss of Offsite Power that then proceeds to an SBO event and RCP seal failure due to thermal shock.
05000424/FIN-2015003-052015Q3VogtleLicensee-Identified Violation10 CFR 55.21, Medical examination, states, in part, that a licensee shall have a medical examination by a physician every two years. Contrary to the above, on March 24, 2015, the licensee identified that a licensed operator did not complete the required biennial NRC medical examination by February 2015, which was the two year due date. The licensed operators requirement to have a medical examination was incorrectly removed from the licensees learning management system (LMS) database when the operator entered the initial license training program to upgrade to a senior operator. The inspectors determined that the violation was not greater than very low safety significance (Green) because the licensed operator was not actively performing licensed duties in the control room. This issue was entered in the licensees corrective action program as CR 10045159.
05000425/FIN-2015003-032015Q3VogtleUnauthorized Entry into a High Radiation AreaA self-revealing NCV of Technical Specification (TS) 5.7.1, High Radiation Area, for an unauthorized entry into a high radiation area (HRA). The radiological aspects were not discussed in the pre-job brief, the health physics (HP) technician in containment did not challenge the crew as to whether or not they received their HRA briefing, and the crew did not follow adequate radiological safety practices, such as reading instructions on the HRA posting prior to entry and not leaning against piping. The licensee entered this issue into the CAP as CR 870060 The entry into a HRA without meeting the entry requirements specified in T.S. 5.7.1 was a performance deficiency. This performance deficiency was more than minor because it was associated with the Occupational Radiation Safety cornerstone attribute of Human Performance and adversely affected the cornerstone objective in that workers who enter HRAs with inadequate knowledge of current radiological conditions could receive unintended occupational exposures. The finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green). This finding does not involve a cross-cutting aspect because it is not current license performance.
05000261/FIN-2011003-012011Q2RobinsonRainstorm Results in Flooding of the Power BlockOn May 27, 2011, a heavy rainstorm was not successfully managed by the sites engineered rainwater management features. This resulted in water run-off into the protected area, backing up of storm drains and water intrusion into the power block, Auxiliary Building and other support buildings. Additional review by the NRC is required following the completion of the licensee\\\'s root cause investigation. The review will determine whether this issue represents a performance deficiency. This issue is identified as URI 05000261/2011003-1, Rainstorm Results in Flooding of the Power Block.
05000261/FIN-2011003-032011Q2RobinsonRefueling Water Storage Tank Inoperable While On PurificationThe inspectors identified a NCV of Technical Specification (TS) 3.5.4 Refueling Water Storage Tank (RWST), which required the RWST to be operable in modes 1 through 4. The licensee failed to comply with the TS Action Statements when the RWST was rendered inoperable by placing the non-seismically qualified purification loop in operation. Upon discovery the licensee promptly restored the RWST to operable status by removing the purification loop from service, put administrative controls in place to prevent use of the purification loop, and initiated Action Request (AR) 452093 to evaluate the event. Use of the non-seismically qualified Spent Fuel Pool Demineralizer System for purification of the Refueling Water Storage Tank was determined to be a performance deficiency. This action rendered the RWST inoperable and the licensee failed to comply with the required action statement for an inoperable RWST. The finding is more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, during a seismic event the purification piping could break and cause a loss of inventory in the RWST. Significance Determination Process (SDP) Phase 1 screening determined that this finding was within the mitigating systems cornerstone and was potentially risk significant due to a seismic external event and therefore required a Phase 3 SDP analysis. A phase 3 risk assessment was performed by a regional SRA using the NRC SPAR model. An exposure period of 213 days was utilized as this represented the worst case one year exposure period determined using the RWST purification history data. No recovery credit was assumed in the analysis. The non-seismic RWST purification piping and the dedicated shutdown diesel generator were assumed to fail at the same seismic input as that assumed for a loss of offsite power. The dominant sequence was a seismically induced loss of offsite power leading to a station blackout with failure of the emergency power system and failure to recover offsite power or the emergency diesel generators. Subsequent battery depletion and operator failure to control the turbine driven auxiliary feedwater pump would lead to core damage. The risk was mitigated by the low probability of a seismic event and the failure probability of the emergency diesel generators. The analysis determined that the risk increase of the performance deficiency was an increase in core damage frequency less than 1E-6/year a GREEN finding of very low safety significance. The cause of the finding was directly related to the conservative assumptions aspect in the Decision Making component of the Human Performance area because during a previous review of this evolution the licensee did not demonstrate the proposed action was safe in order to proceed. Instead the licensee could not find a requirement to show it was unsafe and concluded placing the RWST on purification was acceptable.
05000261/FIN-2011003-022011Q2RobinsonInadequate Seismic Analysis for Installation of Safety Related Cable Trays and ConduitThe inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to perform an adequate seismic analysis during the plant modification of the 125VDC Battery Chargers. Specifically, the interface evaluation for installation of the safety-related, Battery Charger, cable tray and conduit failed to consider the seismic interaction with the adjacent air-handling unit structure. Subsequent review and analysis determined that the modification introduced a degraded/nonconforming condition which does not affect operability. The licensee documented the issue in Nuclear Condition Report 458971 and initiated actions for a plant modification. The failure to perform an adequate seismic analysis for the installation of the safetyrelated cable trays and conduit is a performance deficiency. This performance deficiency is associated with the design control attribute of the Mitigating System Cornerstone. It is more than minor since it is similar to Inspection Manual Chapter 0612, Appendix E, Example, 3.a, in that the seismic analysis for the cable trays and conduits require revision and modification to the air handling unit structural supports to correctly resolve the seismic concerns. In accordance with IMC 0609 (Table 4a), Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency which resulted in a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the performance deficiency occurred in 1991 and does not represent current licensee performance.
05000261/FIN-2011002-042011Q1RobinsonRefueling Water Storage Tank Operability While On PurificationAn Unresolved Item is being opened to provide for additional inspection in response to an NRC identified issue regarding Refueling Water Storage Tank (RWST) operability with the purification loop in operation. The inspectors noted on March 8, 2011, that the RWST purification loop had been in operation for approximately 14 hours. The piping and components of the purification loop are shown on plant drawings to be beyond the seismic qualification boundary for the RWST. The licensee had previously reviewed this issue using AR 422778 in late 2010 and determined it was acceptable to place the RWST on purification without declaring the RWST inoperable. The inspectors questioned the basis for that conclusion. The licensee removed the RWST from purification and put administrative controls in place to prevent use of the purification loop until the issue is resolved. The licensee is continuing to evaluate the use of the RWST purification loop and the impact on operability of the RWST. Additional review by the NRC is required following the completion of the licensees evaluation. This review will also determine whether this issue represents a performance deficiency. The issue will be identified as URI 05000261/2011002-4, Refueling Water Storage Tank Operability While On Purification
05000335/FIN-2010003-012010Q2Saint LucieInadequate Operations Procedure Results in Loss of 1B 125 v DC Bus

A self-revealing NCV of Technical Specification 6.8.1 was identified for an inadequate operating procedure which resulted in the loss of the 1B Direct Current (DC) vital electrical bus and unplanned entry into Technical Specification Action 3.9.8.2.a. for losing operability of one train of shutdown cooling. Subsequently, the Unit 1 daily shutdown risk assessment changed from a low risk to a high risk condition for electric power availability

The failure to provide adequate procedural guidance for operating the 125 volt (v) DC vital bus is a performance deficiency. This finding was considered more than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone and adversely affected the objective of limiting the likelihood of a loss of the 125 v DC bus and a loss of shutdown cooling (SDC) event. If left uncorrected, the condition has the potential to become a more significant safety concern such as a loss of SDC while the reactor coolant system is open and the time to boil could be less than 2 hours. This finding was also determined to potentially have greater significance per IMC 06909, Appendix G, Attachment 1, Check List 3 due the increase in the likelihood that a loss of SDC will occur and the licensees ability to cope with a loss of off-site power was degraded. The phase 1 screening resulted in a need to perform a phase 2 and phase 3 evaluation due to the finding resulting in the loss of mitigating function, specifically the ability to perform decay heat removal. The finding occurred while the plant was shutdown and required entry into IMC-0609 Appendix G. A phase 2 analysis was performed by a regional project engineer and was sent to the regional SRA for review. In accordance with the guidance of NRC Inspection Manual Chapter 0609 Appendix G, the analysis was given to headquarters analysts to perform a detailed phase 3. The significance determination process phase 3 risk evaluation resulted in a risk increase for the finding <1E-6 for core damage frequency (CDF) and <1E-7 for large early release frequency (LERF). The initiators evaluated were loss of inventory (LOI), loss of offsite power (LOOP), and loss of residual heat removal (LORHR). The dominant sequences involved the LOOP initiator, failure of the DC B train resulting in the failure of RHR B, and the failure of the A train to provide a means to perform feed and bleed given the loss of RHR A. The analysis assumed the DC B train was nonrecoverable. Due to the short time to boil, gravity feed was not credited. The finding was characterized as of very low safety significance (Green). This characterization was due to the very short exposure time and that the deficiency was evaluated as a condition assessment rather that as an event assessment. This finding was related to the complete procedures aspect of the Resources component in the Human Performance crosscutting area.

05000335/FIN-2010003-022010Q2Saint LucieUntimely Corrective Actions to Resolve Seat Leakage of Containment Spray Valves 2-MV-07-3/4

The inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure of the licensee to take timely and effective corrective actions to prevent seat leakage past containment spray isolation valves 2-MV-07-3 and 2-MV-07-4 resulting in long standing Reactor Coolant System (RCS) inventory perturbations while in reduced inventory operations and a long term operator workaround

The finding was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening. Specifically, if left uncorrected the condition has the potential to become a more significant safety concern such as a loss of shutdown cooling while in mid-loop operations when the time to boil could be 15 minutes or less. Using the NRC Manual Chapter 0609, ASignificance Determination Process,@ Appendix G, Shutdown Operations Significance Determination Process, Checklist 3, the finding was determined to be of very low safety significance because Core Heat Removal, Inventory Control, Power Availability, Containment Control, and Reactivity Guidelines were all met. This finding was related to the appropriate and timely corrective actions aspect of the corrective action program (CAP) component in the problem identification and resolution crosscutting area.

05000335/FIN-2010003-032010Q2Saint LucieFailure to Take Timely and Effective Corrective Actions to Prevent RCS Pressure Boundary Leakage through the RCP Seal Lines

A self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified when a reactor coolant pump (RCP) seal line weld failure resulted in RCS pressure boundary leakage in July 2009. Specifically, the licensee failed to prevent the recurrence of RCS pressure boundary leakage, a significant condition adverse to quality, caused by conditions of low stress, high-cycle fatigue affecting RCP seal line welds. Licensee personnel shutdown the reactor and entered reduced inventory operations to perform repairs. The issue was entered into the CAP

The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the finding was associated with repeated RCP weld failures and affected the integrity of the RCS pressure boundary. The finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that the cause of this finding was related to the appropriate and timely corrective actions aspect of the CAP component in the problem identification and resolution crosscutting area.

05000335/FIN-2009004-012009Q3Saint LucieSeat leakage of Containment Spray Valves 2MV-07-3/4While reviewing condition report 2007-41688, the inspectors determined that seat leakage past containment spray system isolation valves 2-MV-07-3 and 2-MV-07-4 dates back to 1990. The valves are Pacific 12 inch gate valves with SB-0 Limitorque motor operators. Leakage has been as high as 3.37 gpm measured in 2004. Repairs on the valve seats and wedges have been ineffective. The repair activities have mainly consisted of lapping the valve seating surfaces and performing a satisfactory blue dye check. The valves are not containment isolation valves and require no periodic in-service test. The licensee has repeatedly planned replacement of the valve with an updated flexible wedge style valve during multiple refueling outages. However, as the refueling outages approach, the repair is cancelled due to scheduling conflicts or to further evaluate the condition. This has been noted by the inspectors during the last two refueling outages as the operator work around remains active. The inspectors determined that in 1996, the licensee developed a compensatory measure and procedure change to install a temporary hose from a drain valve downstream of 2-MV-07-3/4 to allow the seat leakage to drain to the floor drain system in the auxiliary building vice leaking into the containment spray system and discharging down into containment during shutdown cooling operations. Essentially, the licensee has created an operator work around that could have an adverse effect on shutdown cooling operations and reactor coolant system inventory while in midloop conditions which requires frequent makeup to the RCS to maintain reactor vessel inventory and adequate net positive suction head (NPSH) on the operating pump. This issue is unresolved pending completion of NRC review and analysis of licensee actions associated with the operator work around and is identified as Unresolved Item (URI) 05000389/2009-004-01, Seat Leakage of Containment Spray Valves 2 MV-07-3 and 2 MV-07-4
05000335/FIN-2009004-022009Q3Saint Lucie2B2Reactor Coolant Pump Failed Seal Injection LineOn July 8, 2009, during mid-shift operation, a Reactor Coolant System (RCS) Inventory Balance was performed and a 0.065 gpm RCS unidentified leak rate was calculated. Additionally, a Containment Atmosphere Particulate Radiation Monitor indicated an upward trend. The licensee performed a robotic inspection of containment in an attempt to identify any RCS leaks before shutting the unit down on July 13, 2009. Further investigation verified RCS pressure boundary leakage at the lower cavity piping J-groove weld of the 2B2 Reactor Coolant Pump (RCP). The licensee entered this issue into the Corrective Action Program (CAP) as CR 2009-19624. The licensees immediate corrective action included replacing seal packages to reset fatigue usage at the J-groove welds, flange removal, cutting and capping of the upper cavity lines and replacing middle cavity piping between the flange and next piping flange. To conduct repairs, the licensee entered into a higher risk Plant Operating Status (POS) of mid-loop configuration with reduced inventory. The inspectors noted that potentially similar weld failures took place in August 2007, on the 2B1 RCP pipe-to-elbow weld on the outboard side of the first flanged coupling of the 2B1 RCP seal injection 34 inch diameter line; in December 2007, on the 2B2 RCP weld connecting the 34 inch diameter seal injection line with the seal housing; and in January, 2009, on the 2B1 RCP pipe-to-flange weld on the outboard side of the first flanged coupling of the upper cavity pressure sensing line. The inspectors remained concerned whether licensee corrective actions associated with the previous weld failures were appropriate considering the repetitive failures. This issue is unresolved pending completion of NRC review and analysis of the final root cause evaluation and is identified as URI 05000389/2009004-02, Reactor Coolant Pump Failed Seal Injection Line. This LER is open
05000321/FIN-2008005-012008Q4HatchFailure to Report a Reportable ConditionA NRC-identified violation of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73, Licensee Event Report System, was identified when the licensee did not recognize the loss of all three main control room (MCR) air handling units (AHUs) was a reportable condition. Consequently, the licensee failed to make an eight hour report as required by 10 CFR 50.72 and submit a licensee event report (LER) within 60 days as required by 10 CFR 50.73. This violation does not apply to Unit 1 because it was in a refueling outage and the AHUs were not required to be operating. This violation has been entered into the licensees CAP as CR 2008111957. Failure to recognize the loss of the MCREC system safety function was reportable is a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function of event assessment. The inspectors determined this finding was a SL IV violation because the failure to report this condition did not substantively impact the Agency\'s regulatory responsibilities and the Agency would not have responded in a significantly different manner had the information been properly reported. This finding had the cross-cutting aspect of evaluating for reportability in the area of Problem Identification and Resolution (P.1(c)) because the licensee evaluated reportability only for the entry into TS LCO 3.0.3. (Section 4AO5)
05000321/FIN-2008005-032008Q4HatchLicensee-Identified ViolationTechnical Specification 5.7.1.a requires, in part, that each high radiation area, in which the intensity of radiation is > 100 mrem/hr but < 1000 mrem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on October 14, 2008, the licensee was transferring Unit 1 condensate phase separator resin to the vendors equipment for receiving the resin; however, the licensee did not barricade nor conspicuously post the areas that contained the pipes used for transferring the resin as a high radiation area. Licensee evaluations performed after the event showed that the intensity of radiation was >100 mrem/hr but <1000 mrem/hr measured at 30 cm from the pipe surfaces in those areas. This finding was entered in the licensees corrective action program as Condition Report 2008110421. This finding is of very low safety significance because there was no evidence of unauthorized worker entry into the area and no unexpected /unintended radiation exposures to licensee personnel
05000321/FIN-2008005-022008Q4HatchLicensee-Identified Violation10 CFR 50.73(a)(2)(i)(B) requires in part that the licensee shall report any condition which was prohibited by technical specifications. Contrary to this, on May 19, 2008, the licensee determined that pressure boundary leakage resulting from a weld failure in an instrumentation sensing line was discovered on March 8, 2005, and not reported. This issue was entered in the licensees corrective action program under CR 2008103067. This finding is of very low safety significance because the leak was very small and within the RCS leakage accident analysis
05000327/FIN-2007005-022007Q4SequoyahImproper Information Provided for MSPIThe inspectors reviewed the importance weighting ratios for both the unavailability and unreliability portions of the five different MSPI indicators as delineated in the MSPI basis document. The inspectors noted that, for Emergency AC Power, a separate ratio was specified for each EDG on each unit so that when calculating MSPI for Unit 1 there was one ratio for EDG 1A, one ratio for EDG 1B, one ratio for EDG 2A, and one ratio for EDG 2B. The importance of the Unit 1 EDGs was higher for the Unit 1 indicator than the Unit 2 EDGs with each Unit 2 EDG having an identical importance for Unit 1. The opposite was true for Unit 2. The Unit 2 EDGs were more important. However, when reviewing the derivation reports for the Emergency AC Power indicator, the inspectors noted that the same importance ratios were used on each EDG for each unit so that each EDG was equally important to each unit. The inspectors determined that the basis document was correct but the importance ratios had been improperly entered into the CDE database that calculated the Emergency AC Power MSPI. In addition, while reviewing the indicator as part of addressing inspector questions, engineering personnel determined that three previous failures not had not been classified properly. The licensee had originally classified a failure of EDG 2A on October 3, 2005, which involved a broken sight glass on one of the generator bearings, as a demand failure based on the inability of the EDG to complete its function. After reviewing it further the licensee realized that the EDG would have started but would not have been able to complete its mission because the bearing would have failed due to loss of oil. Therefore they reclassified the failure of October 3, 2005, as a failure to run. The licensee also determined that two previous failures on July 20, 2006 and August 7, 2007, were not actually failures because the affected equipment was outside the boundary of the system. The as-reported numbers for the Emergency AC Power MSPI for the quarter ending September 2007 were -5.3E-7 for Unit 1 and -5.3E-7 for Unit 2. The effect of the improper use of the importance measures was to change the unreliability portion of the indicator from a negative to positive number while the total indicator remained negative. After adjusting for importance measures the numbers would have been -2.0E-7 for Unit 1 and -7.0E-8 for Unit 2. With the additional classification changes to the failure data the numbers became -1.12E-6 for Unit 1 and -1.17E-6 for Unit 2. While these numbers remained in the green band, the changes also affected earlier time periods. In two previous quarters, June 2006 and March 2007, the Unit 2 indicator was 1.04E-6. Because this information involved licensee failure to provide complete and accurate information concerning a ROP performance indicator, the inspectors determined that it had the potential to impact NRC ability to perform its regulatory function. Enforcement: 10 CFR 50.9 requires that information provided to the NRC be complete and accurate in all material respects. Contrary to this, from July 1, 2006 until December 31, 2007, the licensee provided information regarding the Emergency AC Power MSPI indicator that was inaccurate. Specifically, the importance ratios for both the unavailability and unreliability portions of the indicator were improperly entered into the calculation for determining the indicator resulting in inaccurate reporting of the MSPI for Emergency AC Power. However, this item will remain unresolved pending NRC review of the previous data for the indicator and is identified as URI 05000327,328/2007005-02, Improper Information Provided for MSPI. This item has been entered into the licensees corrective action program as PER 135288.
05000259/FIN-2006002-022006Q1Browns FerryFailure to Report a Safety System Functional Failure Per 10 CFR 50.73A Severity Level IV non-cited violation (NCV) of 10 CFR 50.73(a)(2)(v)(D) and (vii)(D) was identified by the inspectors for the licensees failure to submit a licensee event report for a safety system functional failure of the Unit 2 residual heat removal pressure suppression chamber containment isolation valves. This issue was documented in the licensees corrective action program as Problem Evaluation Report 99193. In Section IV of the NRC Enforcement Policy, the significance of violations involving the failure to make required reports is not dispositioned using the Reactor Oversight Programs Significance Determination Process. The licensees failure to provide a written event report does potentially impact the NRC\'s ability to carry out its regulatory function. However, because this failure to report per 10 CFR 50.73 did not actually impede or influence regulatory action, and the condition that required reporting under 10 CFR 50.73 was previously determined to be of very low safety significance in inspection report 05000260/2005003, the NRC has characterized the significance of this reporting violation as a Severity Level IV in accordance with Section IV.A.3 and Supplement I of the NRC Enforcement Policy.
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