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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000461/FIN-2018412-01 | 2018Q3 | Clinton | Security | |
| 05000346/FIN-2015502-01 | 2015Q2 | Davis Besse | Licensee-Identified Violation | The licensee-identified a finding of very low safety significance (Green) and an associated violation of 10 CFR 50.54 (q)(2) and 10 CFR Part 50.47(b)(14). Title 10 CFR 50.54(q)(2), requires, in part, that a holder of a license under this part, shall follow and maintain the effectiveness of an emergency plan that meets the requirements of Appendix E, of Part 50, and for nuclear power reactor licensees, the planning standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(14) requires, in part, that Periodic Exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Section 8.1.2.c.6 of the Davis-Besse Nuclear Power Station Emergency Plan, Revision 30, states, Semiannual Health Physics drills will be conducted which involve response to, and analysis of, simulated elevated airborne and liquid samples and direct radiation measurements in the environment. Contrary to the above, from mid-2013 to the end of 2014, the licensee failed to comply with the established drill and exercise program. Specifically, the health physics drill objectives were only being partially met during this time period. The drill scenarios were limited and did not provide an opportunity for the participants to complete sampling/analysis of liquid samples. As part of the corrective actions after the discovery of this issue, the licensees Emergency Response Staff conducted a drill on December 11, 2014, to ensure that all aspects of the Health Physics drill objectives were met. The performance deficiency was more than minor because the issue was associated with the Emergency Preparedness cornerstone and adversely affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the drill scenarios were limited in scope and health physics objectives specified in the emergency plan, were not carried out fully during exercises and drills, that were conducted in the middle of 2013 through the end of 2014. The NRC determined that this was a failure to comply with the licensees emergency plan and a degradation of a planning standard function in accordance with 10 CFR, Part 50.47(b)(14), and was a very low safety significance issue (Green) as indicated in Inspection Manuel Chapter 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. Because the finding is of very low safety significance (Green) and it was entered into the licensees Corrective Action Program as Condition Report, CR-2014-16715, this violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy. |
| 05000255/FIN-2014405-01 | 2014Q4 | Palisades | Security | |
| 05000255/FIN-2014405-02 | 2014Q4 | Palisades | Licensee-Identified Violation | |
| 05000237/FIN-2014408-01 | 2014Q4 | Dresden | Licensee-Identified Violation | |
| 05000237/FIN-2014408-02 | 2014Q4 | Dresden | Licensee failed to ensure data linked to potentially disqualifying information about an individual was retained in the shared database | The inspector reviewed AR 01505017, which described a Petition for Rulemaking filed by the Nuclear Energy Institute (NEI) on January 5, 2013. The petition requested that the NRC amend its regulations to limit the scope of third-party review of licensee decisions denying or revoking an employees unescorted access at their facility. The petition stated that a person who has been determined not to be trustworthy and reliable by a licensee and denied unescorted access to a nuclear power plant could have that determination overturned by a third party. On May 8, 2014, pursuant to an arbitrators ruling, the licensee removed data linked to potentially disqualifying information regarding an individual who the licensee had previously denied unescorted access from the shared database. This issue constituted a violation of NRC requirements, in that the licensee was required to ensure that data linked to potentially disqualifying information about an individual who applied for unescorted access authorization was retained in the shared database. In addition, on July 18, 2014, the NEI requested that the NRC endorse Revisio 4 to NEI 03-01, Personnel Access Requirements for Nuclear Power Plants. Revision4 to NEI 03-01 contained a process for reviewing denials of Unescorted Access, which would allow a third party to review the circumstances surrounding the denial but ensure that NRC access autho ization requirements were being met. Although NEI has requested to withdraw the Petition for Rulemaking, the NRC and the industry are still attempting to resolve the issue. The NRC concluded that the licensee made a good faith effort to resolve the issue prior to the arbitration and that it was not reasonable for the licensee to foresee and prevent the arbitrators ruling. Therefore, no performance deficiency associated with the violation was identified. The NRC performed a risk evaluation of the issue and determined it to be of very low security significance. Based on these facts, I have been authorized, after consultation with the Director, Office of Enforcement, and the Regional Administrator, to exercise enforcement discretion and refrain from issuing enforcement for this violation |
| 05000346/FIN-2014406-01 | 2014Q3 | Davis Besse | Security | |
| 05000341/FIN-2013408-01 | 2014Q1 | Fermi | Security | |
| 05000455/FIN-2011011-01 | 2011Q1 | Byron | Self-Revealing Failure of the 2A Diesel Generator Upper Lube Oil Cooler | A preliminary finding of low to moderate safety significance (White) and an apparent violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the 2A Diesel Generator (D/G) was required to be shutdown during routine monthly surveillance testing on November 17, 2010, when a flange connection on a spool piece connected to the upper lube oil cooler failed, resulting in a significant oil leak. The cause of the failure was that Work Order 1206254, Clean Tube Side of Lube Oil Coolers, did not contain appropriate acceptance criteria to ensure proper reassembly of the spool piece for the upper lube oil cooler following maintenance on January 17, 2010. Specifically, the work order package did not contain a final torque verification to ensure that the spool piece flange bolts were torqued to required values, which resulted in the leak. The licensee entered this issue into the correction action program as Issue Report (IR) 1141591, properly re-installed the spool piece, and returned the 2A D/G to service on November 21, 2010. The inspectors determined that this finding was more than minor, because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. The NRC assessed this finding through a Phase 3 Risk Evaluation of the Significance Determination Process and made a preliminary determination that it was an issue of low to moderate safety significance (White). The cause of this finding was related to the Work Practices component of the Human Performance cross-cutting area since licensee personnel proceeded in the face of uncertainty or unexpected circumstances during the upper lube oil cooler maintenance activity. |
| 05000440/FIN-2010001-01 | 2010Q1 | Perry | Failure to Perform Adequate Evaluation of Crane Support Structure Elements | A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for failure to provide adequate design control measures for crane support structure elements which included bridge crane rail, bridge crane rail clips, bridge crane rail clip studs, leveling plate and leveling plate anchors. Specifically, for evaluation of these structural elements, the licensee failed to demonstrate Seismic Category I compliance in accordance with their design and licensing basis and failed to evaluate the structural elements for resulting reaction forces from the Fuel Handling Building crane. The licensee documented these issues in CRs 11-88791; 11-90252; 10-86582; and 11-04124. The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency could lead to a more significant safety concern if independent spent fuel storage installation (ISFSI) loading was conducted. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Barrier Integrity cornerstone. Based on answering No to all the questions in the Barrier Integrity Cornerstone column of Table 4a, the finding was determined to be of very low safety significance (Green). The inspectors identified a Human Performance, Work Practices (H.4.c) cross-cutting aspect associated with this finding, in that the licensee did not ensure effective supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to have adequate oversight of design calculations and documentation for establishing structural adequacy of the rail, rail clips, rail clip bolts, leveling plate and leveling plate anchors |
| 05000346/FIN-2009007-01 | 2010Q1 | Davis Besse | Inappropriate Change of Fuel Transfer Tube Seal Configuration | A finding associated with two Apparent Violations of 10 CFR Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.71(e) was identified by the inspectors. Specifically, the licensee failed to implement design control measures which assured that the design basis, as specified in the license application, was correctly translated into specifications, drawings, procedures, and instructions and failed to correctly update the Updated Safety Analysis Report (USAR) to reflect the safety analyses associated with License Amendment 240. As a result of these failures, the current fuel transfer tube blind flange seal configuration was contrary to the licensing basis. The successful as-left local leak rate tests performed during the prior refueling outage (Refueling Outage 15) provided reasonable assurance for continued operation. The finding and apparent violations were entered into the licensees corrective action program. The inspectors assessed the preliminary significance of the finding using the traditional enforcement policy. The inspectors determined that had the information been complete and accurate at the time of amendment approval, the NRC would have reconsidered the regulatory position or initiated substantial further inquiry. This finding has a cross-cutting aspect in the area of Human Performance Resources, because the licensee did not have complete, accurate and up-to-date design documentation, procedures, and work packages. This cross-cutting aspect is considered reflective of current performance because the procedures in place at the time of this inspection, in addition to the procedures in place during the 1999-2000 timeframe, did not provide adequate guidance. (H.2(c)) This is a Severity Level III problem (Supplement VII). |
| 05000456/FIN-2009007-01 | 2009Q4 | Braidwood | Failure of Containment Sump Suction Valve 1SI8811B to Stroke Open | The inspectors identified a finding of substantial safety significance and an associated apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to prevent water from entering the motor operated valve actuator for valve 1SI8811B that resulted in corrosion of the torque switch. This resulted in the valve failing to stroke full open on June 24, 2009. The licensee determined that water entered the valve actuator through a flexible conduit penetration and pooled in the actuator limit switch box. This caused corrosion of the torque switch and minor corrosion of the limit switch. As part of the corrective actions for this event, the licensee sealed the susceptible conduit. Also, to address extent of condition, the licensee subsequently performed successful valve strokes of the 1SI8811A and 2SI8811A/B valves as part of previously scheduled maintenance windows. Additionally, the licensee performed a walkdown of the other SI8811 valves on both Units. Open conduit terminations were identified on all three remaining valves. The 2SI8811B valve was found to have the same susceptible conduit/cable tray configuration while the 1SI8811A and 2SI8811A valves had horizontal conduit terminations that were less susceptible to water intrusion. As a result, the licensee sealed the 2SI8811B valve open conduit termination. The inspectors determined that the finding was more than minor due to impacting the Equipment Performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems the respond to initiating events to prevent undesirable consequences. The finding associated with this apparent violation was assessed using a Phase 3 analysis in accordance with NRC Inspection Manual Chapter 0609 Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and is preliminary determined to have substantial significant safety significance (Yellow). The inspectors determined that this issue is associated with the Corrective Action Program component of the Problem Identification and Resolution cross-cutting area. (P.1(a)) Specifically, licensee staff was aware for several years of water leakage from the overhead areas around the SI8811 valves. Several corrective action documents were generated previously but the licensee did not adequately evaluate the potential safety significance of the water leakage and did not correct the issue. (Section 1R22.1.b |
| 05000306/FIN-2009010-02 | 2009Q3 | Prairie Island | Failure to Ensure Design Measures Were Appropriately Established for the Unit 2 Component Cooling Water System | An inspector identified apparent violation of 10 CFR Part 50,Appendix B, Criterion III, Design Control, was identified due to the licensees failure to establish design control measures to ensure that the design basis for the Unit 2 CCW system was correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure that the safety-related function of the CCW system was maintained following initiating events (such as high energy line break, seismic or tornado events) in the turbine building. This issue has been preliminarily determined to be of low to moderate safety significance (White).This issue was entered into the licensees corrective action program as corrective action document 1145695. Upon identifying this issue, the licensee immediately declared the Unit 2 CCW system inoperable and entered Technical Specification 3.0.3. The Technical Specification was exited following the closure of several system isolation valves approximately 2 hours later. The closure of the isolation valves prevented the Unit 2 CCW system from being vulnerable to failure following events in the turbine building. This finding was determined to be more than minor because it impacted the design control and external events aspects of the Mitigating Systems Cornerstone. The finding also impacted the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The initiating events in the turbine building could cause the CCW piping to fail. Loss of CCW inventory affects both trains of CCW based on the piping arrangement. The loss of both trains of CCW required a phase 3 significance determination. The results of the phase 3 assessment showed a delta core damage frequency of 3.2E-6, White. The cause of this finding was related to the cross-cutting element of Human Performance, Decision Making because the licensee failed to make safety-significant and risk-significant decisions using a systematic process to ensure that safety was maintained (H.1(a)). Since both the Unit 1 and Unit 2 cross-cutting aspects are from the same performance deficiency and are separated based on the risk determination, the aspect of H.1(a) counts as one cross-cutting aspect in this report. |
| 05000249/FIN-2009009-01 | 2009Q3 | Dresden | Inadvertent Control Rod Movement While Shutdown | A finding that has preliminarily been determined to be White, a finding with low to moderate safety significance, was self-revealed on November 3, 2008, when the licensee failed to prevent inadvertent and uncontrolled control rod withdrawal by non-licensed operators. After the finding was self-revealed, the control rods were returned to the full-in position to ensure there was no immediate safety concern and the licensee implemented corrective actions, including conducting a prompt investigation. The finding is also associated with five apparent violations of NRC requirements specified by 10 CFR 50.54(j), Technical Specification 3.1.1, and Technical Specification 5.4.1. The performance deficiency was determined to be more than minor because licensed operators did not maintain configuration control of the control rods when non-licensed operators were able to inadvertently cause control rods to move. Because probabilistic risk assessment tools were not well suited for this finding, the criteria for using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, were met. Based on the additional qualitative circumstances associated with this finding, regional management concluded the finding was preliminary low to moderate safety significance (preliminary White). The performance deficiency was determined to have resulted from several causes; however, the primary cause was determined to involve the ineffective use of operating experience. This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the licensee did not effectively implement and institutionalize operating experience through changes to station processes, procedures, and training programs. (P.2(b)) |
| 05000282/FIN-2008009-01 | 2009Q1 | Prairie Island | Radioactive Material Shipment Package Radiation Levels Exceeded | A self-revealing finding with an apparent violation of regulatory requirements was identified involving a failure of the licensee to properly radiologically characterize, prepare, and ship a package containing radioactive material in a manner that assured, under conditions normally incident to transport, conformance with Department of Transportation (DOT) radiation level limitations specified by 49 CFR 173.441(a), (i.e., 200 millirem per hour (mrem/h)) on any external surface of the package as required by 10 CFR 71.5 (and 49 CFR 173.441(a)). Additionally, the licensee did not provide nor ensure that the individuals involved in preparing this shipment were trained and qualified for the task as specified by 49 CFR 172.704, Training Requirements. The finding involved an October 29, 2008, radioactive material shipment, via an exclusive-use open transport vehicle that was determined to have radiation levels of 1630 mrem/h on the external surface of a package upon receipt at the shipping destination. As immediate corrective actions, the licensee suspended all radioactive shipment activities. The licensee entered this performance deficiency in their corrective action program; initiated a root cause evaluation; and initiated corrective measures, including various process improvements to prevent recurrence. This finding is more than minor since it was associated with the Public Radiation Safety Cornerstone program and process attribute and affected the cornerstone objective to ensure adequate protection of the public from exposure to radioactive materials given that package radiation levels were elevated. Preliminarily, the significance of this finding is considered as having a substantial safety significance (Yellow), since the radiation level was greater than five times the limit (1000 mrem/h) but less than ten times the limit (2000 mrem/h) specified by the DOT regulatory requirement. Although the surface of the package with elevated radiation levels would not be routinely accessible to a member of the public during transport, that aspect was fortuitous and not the result of design nor package preparation by the licensee. The condition had the potential to adversely affect personnel who would normally receive the package or respond to an incident involving the package, with a reasonable expectation that the package conformed to DOT radiation limitations. Additionally, the cause of this finding had a cross-cutting aspect in the area of Human Performance. Specifically, the licensee failed to appropriately plan the work activity by incorporating risk insights and job site conditions, including conditions which may impact radiological safety (H.3 (a)). This finding is documented within the licensees corrective action system as RCE 1157726. (Section 2PS2 |
| 05000282/FIN-2008009-02 | 2009Q1 | Prairie Island | Failure to Perform Formal Job Planning to Evaluate the Radiological Hazards | An NRC-identified finding of very low safety significance with an associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was identified in the area of occupational radiation safety associated with the licensees failure to perform adequate job planning to evaluate the radiological hazards, as required by station procedures. Specifically, the licensee failed to properly assess the radiological hazards to workers associated with the decontamination, demobilization and packaging of fuel sipping equipment on the refuel floor. This issue has been entered into the licensees corrective action program and implemented corrective actions that include changes to procedures to include a holistic risk-based review of radiologically significant work. The finding is more than minor because, given the radiological uncertainty of working with fuel handling equipment, if left uncorrected the finding could become a more significant safety concern. The finding was determined to be of very low safety significance because it did not involve unintended collective dose (ALARA planning); there was no overexposure, nor potential for overexposure; and the licensees ability to assess dose was not compromised. Additionally, the cause of this finding had a cross-cutting aspect in the area of Human Performance. Specifically, the licensee failed to appropriately plan the work activity by incorporating risk insights and job site conditions, including conditions which may impact radiological safety (H.3 (a)). (Section 2OS2 |
| 05000255/FIN-2008011-02 | 2008Q4 | Palisades | Failure to Implement Effective Radiological Controls for Working with Equipment in contact with Failed Fuel | An NRC-identified finding of very low safety significance and associated NCV of 10 CFR 20.1501 was identified for failure to perform adequate radiological evaluations necessary to properly assess the radiological hazards and prescribe appropriate radiological controls necessary to minimize dose to workers associated with failed fuel. Fuel reconstitution, a planned activity for the refueling outage, had a high potential to result in discrete radioactive particles and alpha contamination from the degraded fuel pins. The licensee failed to anticipate these radiological hazards and to implement appropriate controls to minimize exposure to radiation. As corrective actions, the license revised all radiation work permits (RWPs) associated with work in the spent fuel pool. The finding is more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the licensee did not implement radiological controls necessary to minimize dose to workers. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, the NRC could not identify an overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised. The cause of this deficiency had a cross-cutting aspect in the area of Human Performance. Specifically, the licensee failed to appropriately plan the work activity by incorporating risk insights and job site conditions, including environmental conditions, which may impact radiological safety (H.3(a)). (Section 4OA5 |
| 05000255/FIN-2008011-03 | 2008Q4 | Palisades | Failure to Post and Control Access to High Radiation Area | A self-revealed finding of very low safety significance and associated NCV of Technical Specification 5.7.1 was identified for the failure to post and control an area with dose rates greater than 100 millirem/hour as a high radiation area. Specifically, the area of the refuel floor that contained the fuel reconstitution equipment was not posted as a high radiation area. Dose rates of approximately 450 millirem/hour were measured 30 centimeters (cm) from the equipment after three workers received electronic dosimeter alarms (dose rate). As corrective actions, the licensee corrected the radiological posting and controls for the area. The finding is more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, job specific radiological surveys failed to identify elevated dose rates around the spent fuel pool during fuel reconstitution demobilization. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised. This finding appeared to be caused by inadequate coordination of work activities between the radiation protection staff and the contractors. Consequently, the cause of this deficiency had a cross-cutting aspect in the area of Human Performance. Specifically, the licensee failed to appropriately coordinates work activities by incorporating actions to communicate, coordinate, and cooperate with each other during activities in which inter-departmental coordination is necessary to assure plant and human performance (H.3(b)). (Section 4OA5 |
| 05000255/FIN-2008002-12 | 2008Q1 | Palisades | Licensee-Identified Violation | Technical Specification 5.7.2 requires areas with dose rates greater than 1000 millirem/hour to be posted and controlled to prevent inadvertent entry. Contrary to this, on September 22, 2007, a steam generator platform worker left the work area before the steam generator hand hole covers were in place. This configuration allowed inadvertent access to an area where rates exceeded 1000 millirem/hour. This was identified in the licensees corrective action program as CR-PLP-2007-04304 and was reported as a Performance Indicator occurrence. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromise |
| 05000255/FIN-2008002-11 | 2008Q1 | Palisades | Licensee-Identified Violation | Technical Specification 5.7.1 requires areas with dose rates greater than 100 milirem/hour to be posted and controlled as a High Radiation Area. Contrary to this, on September 20, 2007, and other dates, the high radiation area posting and barricade was found altered and ineffective on the 590 elevation of containment. This was identified in the licensees corrective action program as CR-PLP-2007-04236. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised |
| 05000255/FIN-2008002-10 | 2008Q1 | Palisades | Failure to Comply with TS 3.5.2 B and C (Section 4OA3) | A self revealing NCV of TS 3.5.2 B and C was identified for the inability of an automatic valve in the Emergency Core Cooling System (ECCS), CV-3047, to reposition fully closed on an actuation signal. As a result, one train of ECCS was inoperable for longer than allowed by technical specifications. When the licensee identified that the valve would not fully close, the licensee took the actions required by TS and repaired the valve. The inspectors determined the failure to take required actions in accordance with TSs is more than minor because the finding impacts the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the objective to ensure availability, reliability and capability of systems which respond to initiating events. During the injection phase of an accident, more flow would bypass the core with the valve approximately 18 percent open, than if the valve had been fully closed. The inspectors determined the finding is of very low safety significance (Green) because the finding is not associated with a loss of safety function for the ECCS system. The finding includes a cross-cutting aspect in the area of human performance in that the licensee did not adequately coordinate work activities to address the impact of actions needed to ensure the valve was closed when the valve was declared inoperable (H.3(b)). (Section 4OA3 |
| 05000255/FIN-2008002-09 | 2008Q1 | Palisades | Failure to Comply with TS 3.8.4 B and C (Section 4OA3) | A self revealing NCV of TS 3.8.4 B and C was identified for failing to recognize that battery cell parameters were not within TS limits and for failing to take actions in accordance with TS for an inoperable battery. Specifically, cell 43 of the right train safety-related battery (ED02) was below technical specifications for individual cell voltage without recognition by the site staff. As a result, compensatory actions and a plant shutdown required by TSs were not completed as required. As an immediate action, the licensee completed the required actions required by TS including restoration of the battery to an operable status. The inspectors determined the failure to take required actions in accordance with TSs is more than minor because the finding impacts the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the objective to ensure availability, reliability and capability of the systems which respond to initiating events. The inspectors determined the finding is of very low safety significance (Green), because the finding did not cause a loss of safety function for the right train battery. The finding includes a cross-cutting aspect in the area of human performance in that human error prevention techniques (H.4(a)), in this case an adequate pre-job brief, were not effective in ensuring prompt notification of the shift manager. (Section 4OA3 |
| 05000255/FIN-2008002-05 | 2008Q1 | Palisades | Failure to Use, to the Extent Practical, Process or Other Engineering Controls to Control the Concentration of Radioactive Material in Air (Section 2OS1) | A self-revealed finding of very low safety significance and associated NCV of 10 CFR 20.1701 was identified for the failure to use, to the extent practical, process or other engineering controls to control the concentration of radioactive material in air. On September 12, 2007, the licensee failed to implement effective engineering controls in the reactor containment to reduce the levels of radioactive iodine gases. The failure resulted in elevated levels of airborne radioactivity and the intakes of radioactive material by the licensees staff. As corrective actions, the licensee conducted a root cause evaluation and has entered the problem in the corrective action program as CR-PLP-2007-04002. The finding is more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not implementing adequate engineering controls resulted in unplanned exposures to radioactive material. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised. The engineering controls comprised of a charcoal filtration ventilation system that were planned to be used to control the concentration of radioactive material in air were either depleted soon after placed in service or installed backwards. Consequently, the cause of this deficiency had a cross-cutting aspect (H.3(a)) in the area of Human Performance related to work control. Specifically, the licensee failed to plan and coordinate work activities with planned contingencies and compensatory actions. (Section 2OS1.2 |
| 05000255/FIN-2008002-04 | 2008Q1 | Palisades | Failure to Maintain Procedures for the Maintenance of PAPR Batteries (Section 2OS1) | The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 20.1703(c) for the failure to implement written procedures to ensure batteries for powered air purifying respirators (PAPRs) are adequately charged before use. As of January 16, 2008, the licensee failed to maintain procedures that provided adequate instructions concerning the charging of PAPR batteries, which resulted in two failures of a PAPR unit to properly function and in the intake of radioactive material on September 9, 2007. As corrective actions, the licensee revised procedures and replaced the battery chargers with a model that indicates battery charge condition. The licensee entered the issue into the corrective action program as CR-PLP-2007-04149 and CR-PLP-2008-00229. The finding is more than minor because it impacted the equipment and instrumentation attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not providing adequate procedures for control of PAPR battery charging resulted in an unplanned exposure to radioactive material. The finding was determined to be of very low safety significance because it was not an As Low As Reasonably Achievable (ALARA) planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised. The inspectors did not identify a cross-cutting aspect associated with this finding. (Section 2OS1.2 |
| 05000255/FIN-2008002-02 | 2008Q1 | Palisades | Failure to Monitor the Feedwater System Under 10 CFR 50.65a(1) (Section 1R12) | The inspectors identified a Green NCV of Title 10, Code of Federal Regulations (CFR) 50.65 for the failure to include a B feed regulating valve deficiency to close during startup operations as a functional failure in the maintenance rule program. The inspectors noted that the failure should have placed the feedwater system into maintenance rule 10 CFR 50.65a(1) status in the fourth quarter of 2007. This caused a lapse in the determination of appropriate system monitoring and goal setting to maintain system reliability. This issue was entered into the licensee\'s corrective action program as CR-PLP-2008-00562 and the licensee placed the system in a(1) status. The finding is more than minor because, in accordance with Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues (example 7b) and Enforcement Manual Section 8.1.11, Maintenance Rule a(1) and a(2) violations are not minor because they involve structures, systems, and components (SSCs) that have demonstrated some degraded performance or condition. The finding is of very low safety significance because there was no design deficiency, the finding did not represent an actual loss of a safety function, nor does this involve a risk significant system for mitigating fire, flood, seismic, or severe weather events. This finding also has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program (P.1(c)) because the licensee failed to thoroughly evaluate the cause and extent of condition of the failed feed regulating valve. (Section 1R12) |
| 05000255/FIN-2008002-08 | 2008Q1 | Palisades | Main Feed Pump Trip due to Inadequate Configuration (Section 4OA3) | A self-revealed finding occurred on January 13 when the B Main Feed Pump failed. The failure occurred due to improper maintenance on the lube oil pump associated with the Main Feed Pump that resulted in a loss of lube oil flow and trip of the Main Feed Pump. The failure was not a violation of NRC requirements. The licensee manually tripped the reactor in accordance with procedures and repaired the Main Feed Pump. The licensee entered the issue into the corrective action program as Condition Report (CR)-PLP-2008-0151 and repaired the pump. The inspectors concluded that this finding is more than minor in accordance with Inspection Manual Chapter 0609 because the finding is associated with the increase in the likelihood of an initiating event. Specifically, the improper pump assembly led to a partial loss of feed and subsequent plant trip. The inspectors determined the finding is of very low safety significance, Green, in accordance with the phase one screening checklist because the finding did not affect a mitigating system in addition to being a transient initiator. The finding does not represent a violation of NRC requirements; however, it does represent a failure to meet self imposed requirements to provide task instructions commensurate with the complexity of the work and qualifications of the workers. The finding includes a cross-cutting aspect in the area of Human Performance, resources due to an inadequate work package (H.2(c)). (Section 4OA3 |
| 05000255/FIN-2008002-07 | 2008Q1 | Palisades | Failure to Control the Release of Radioactive Material (Section 2PS3) | A self-revealed finding of very low safety significance and associated NCV of 10 CFR 20.1501 was identified for failure to perform an adequate radiological survey to assure compliance with 10 CFR 20.1802, which requires that the licensee control and maintain constant surveillance of licensed material that is in a controlled area or unrestricted areas and that is not in storage. On January 17, 2008, the NRC notified the licensee that radioactive material was identified by another NRC licensed facility when workers arrived following Palisades refueling outage 1R19. That licensee identified six pairs of footwear and other personal items with radioactive contamination levels between 6,000 and 20,000 disintegrations per minute, which had been improperly released from the Palisades site. As immediate corrective actions, the affected materials were confiscated by the other site. Additionally, the licensee identified two earlier occurrences of inappropriate surveys that were performed early in the refueling outage that resulted in the inadvertent release of radioactive material. As corrective actions, the licensee planned to implement new procedure documents, and the issue was entered into the licensees corrective action program as Condition Reports CR-PLP-2007-04338 and CR-PLP-2008-01180. The finding is more than minor because it impacted the program and process attribute of the Public Radiation Safety Cornerstone and it adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive material released into the public domain, in that inadequate surveys resulted in the failure to control radioactive material. The finding was determined to be of very low safety significance because it was a radioactive material control finding, it was not a transportation finding, and it did not result in public dose greater than 0.005 rem. The finding was caused by the decision to allow manual release surveys of a large number of workers that alarmed the personal contamination monitor, which overwhelmed the ability of the radiation protection staff to conduct effective monitoring of personnel. Consequently, the cause of this deficiency had a cross-cutting aspect (H.1(a)) in the area of Human Performance related to decision making. Specifically, the licensee failed to make risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. (Section 2PS3.1 |
| 05000255/FIN-2008002-06 | 2008Q1 | Palisades | Failures to evaluate the shallow (skin) dose to three workers involved in tool disassembly and failure to barricade and conspicuously post each entryway to a high radiation area (Section 2OS1) | The inspectors identified an URI concerning events that occurred on October 4, 2007, when three contract workers received electronic dosimeter dose rate alarms when they disassembled tools used for fuel reconstitution on the 649 level of the Auxiliary Building near the spent fuel pool. Radiation surveys performed after stainless steel inserts were placed in a box identified gamma dose rates greater than 100 millirem/hour and highly elevated beta radiation levels. At the time of the inspection, the licensee had not completed an evaluation of the radiological hazards of the work performed. As a result, the shallow dose for workers involved in the work evolution was unknown. Similarly, the inspectors could not evaluate the consequence of the apparent improper radiological posting for the area. Therefore, this issue remains under review by the NRC and is categorized as an URI (URI 05000255/2008002-06) |
| 05000255/FIN-2008002-03 | 2008Q1 | Palisades | Inadequate General Operating Procedure for Mode Transition (Section 1R20) | The inspectors identified a NCV of Technical Specification (TS) 5.4.1 for the failure to have adequate procedure guidance for the general operating procedures for mode transition to power operations. Specifically the general plant operating procedure for mode transition did not have adequate guidance to ensure the actions required by TS 3.0.4 were completed for failure of a radiation monitor required by TS prior to mode transition. Prior to the mode transition, the licensee completed the required action based on the inspectors concerns and wrote a CR. The inspectors determined the failure to have adequate procedures for mode transition in accordance with TS is more than minor because, if left uncorrected, this and other mode transitions could have occurred with less than the required equipment operable or appropriate actions completed, which could become a more significant safety concern. The inspectors determined the finding is of very low safety significance (Green), because the actual mode transition occurred only after completion of the required actions based on the response to the inspectors concerns. The finding includes a cross-cutting aspect in the area of human performance in that licensee did not adequately use conservative assumptions in decision-making to demonstrate the proposed action was safe (H.1(b)). (Section 1R20 |
| 05000255/FIN-2008002-01 | 2008Q1 | Palisades | Failure to Ensure Fire Door Was Closed (Section 1R05) | The inspectors identified a Green NCV of License Condition 2.C.(3), Fire Protection, for failure to ensure a fire door between an emergency diesel generator room and a vital switchgear room was closed. This partially open door degraded the fire containment capability assumed in the fire hazards analysis. The fire door was closed and this issue was entered into the licensee\'s corrective action program as CR-PLP-2008-00075. The finding is more than minor because it is associated with the protection against external factors (fires) attribute of the mitigating system cornerstone and affected the objective to maintain the reliability and capability of systems that respond to events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609, Appendix F, Fire Protection SDP, the inspectors conducted a Phase I SDP screening. The inspectors determined the finding is of very low safety significance (Green), because the fire areas had fully functional, automatic waterbased fire suppression which provided adequate coverage in both rooms and no transient combustible loads were present in either room. The finding includes a cross-cutting aspect in the area of human performance in that human error prevention techniques (H.4(a)), in this case adequate self checking, were not effective in ensuring this door was closed after use. (Section 1R05 |
| 05000282/FIN-2007006-01 | 2007Q4 | Prairie Island | Evaluate TSC Operability During Time with Damper Disconnected | The inspectors reviewed CAP 01110686 that documented that operators found an actuating rod for a TSC emergency ventilation damper disconnected. The actuator had been disconnected without procedural guidance. With the actuator disconnected, the damper would not function as designed. The actuating rod was connected following initiation of the CAP. While the licensee identified the issue with a TSC ventilation damper, the licensee failed to evaluate the issue for past operability and regulatory impact until questioned by NRC inspectors. Once the licensee recognized the need to evaluate this issue, the inspectors noted inconsistencies between the information provided by various licensee departments, particularly Emergency Preparedness, Engineer and Licensing. Furthermore, these inconsistencies indicated shortcoming in the licensees management oversight associated with this issue. Therefore, this issue is considered an unresolved item (URI) pending the inspectors review of the operability of the TSC while the damper was disconnected. (URI 05000282/2007006-01; 05000306/2007006-01) |
| 05000282/FIN-2007005-02 | 2007Q4 | Prairie Island | Potential Inadequate Corrective Actions to Prevent Unnecessary D5 Diesel Generator Unavailability (Section 1R15) | For the D5 DG crankcase pressure issue, the inspectors noted that this had been a longstanding problem with the D5 and D6 DGs. Licensee corrective actions may have been inadequate and/or not completed in a timely manner, resulting in unnecessary increased unavailability of the DGs. The licensee was performing a root cause evaluation of the high crankcase pressure problem technical issues, but it was not complete at the end of this inspection period. The licensee did complete a root cause evaluation of the management and organizational issues leading to the potential inadequate corrective actions. That evaluation indicated that there were wide-ranging organizational, management, leadership, and cultural issues contributing to the problem. Thus there may have been performance deficiencies which resulted in a violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. This issue was considered a URI pending the inspectors review of the technical root cause report. The inspectors review will determine whether failure to take adequate and timely corrective actions led to unnecessary DG unavailability and whether that unavailability resulted in a more than minor increase in risk. (URI 05000306/2007005-02) |
| 05000255/FIN-2007006-04 | 2007Q3 | Palisades | Internal Dose Assessment for O-RING Work | However, there was an event that occurred during the refueling outage where an individuals respiratory protection equipment failed during the removal of the reactor head o-ring. This event resulted in an intake of radioactive material. However, the internal dose assessment was not complete because the event occurred within an area where it was difficult to detect which radionuclides were present. In accordance with plant procedures, samples (i.e., area contamination and in vitro bioassay) were collected and sent to a contracted off-site facility to perform analysis for those difficult to detect radionuclides. The results from this analysis were not available during the inspection. Additionally, the cause(s) for the respiratory equipment failure was still under evaluation during the inspection. Therefore, this issue remains under review by the NRC and is categorized as an Unresolved Ite |
| 05000255/FIN-2007006-05 | 2007Q3 | Palisades | Increased Airborne Radioactivity in Containment | However, events occurred during the refueling outage that created airborne radioactivity areas within the containment building. Increased levels of noble gas were identified after the pressurizer manway was removed and increased again after the steam generator manways were removed, as part of the work that was scheduled during the refueling outage. Increased levels of iodine-131 were identified after the reactor head was lifted to support the refueling outage. The increased airborne radioactivity levels caused small, but measurable, intakes of iodine-131 to several hundred workers during the refueling outage. At the time of the inspection, the events were still under review with respect to the causes of the events, the extent of the personnel intakes, the adequacy of pre-job planning, and the adequacy of contingency actions to mitigate the conditions before allowing work to continue in the affected areas. Therefore, this issue remains under review by the NRC and is categorized as an UR |
| 05000282/FIN-2007003-04 | 2007Q2 | Prairie Island | Control of Very High Radiation Area Keys | The inspectors evaluated an issue concerning the licensees failure to maintain sufficient control over keys to posted VHRAs (i.e., the C-sump) during both the last Unit 1 (U1R24) and Unit 2 (U2R24) refueling outages; the licensee is potentially in violation of 10 CFR 20.1602 requirements. The requirements contained in 10 CFR 20.1602 for control of access to VHRAs requires, in part, that in addition to the requirements for 10 CFR 20.1601 (control of access to high radiation areas), the licensee shall institute additional measures to ensure that an individual is not able to gain unauthorized or inadvertent access to VHRAs. The licensees specific procedures instituted the additional controls necessary to ensure compliance with the requirements of 10 CFR 20.1602. These procedures included the requirements that the C-sump always be treated as a VHRA and that the VHRA keys be controlled by the shift supervisor and not be issued to anyone without the permission of the plant manager or his designee. The inspectors preliminary review of this issue determined that during the 2006 refueling outages (U1R24 and U2R24), the C-sump VHRA keys may have been signed out by RP supervision over multiple shifts. Subsequently, RP supervision transferred possession of the keys to containment radiation protection technician (RPT) Leads, who then transferred possession of the VHRA keys from RPT Lead to RPT Lead over a period of multiple shifts. The inspectors also noted that non-conservative control of VHRA keys had been previously identified in the licensees corrective action program, after the spring 2006 (U1R24) refueling outage and prior to beginning the fall 2006 (U2R24) refueling outage (CAP 01029886, dated May 2006). The licensee was in the process of reviewing its practices for issuance of VHRA keys and was evaluating the specific circumstances surrounding the control of C-sump keys during the two refueling outages in 2006. As a result, the licensee planned to provide the NRC with additional information to demonstrate compliance with 10 CFR 20.1602. The NRC will review the licensees assessment when it is completed. Therefore, this issue remains under review by the NRC and is categorized as URI 05000282/2007003-04; 05000306/2007003-04). |
| 05000282/FIN-2007003-05 | 2007Q2 | Prairie Island | Performance Indicator Accuracy for Occupational Radiation Safety | During the review of the Occupational Exposure Control Effectiveness PI, the inspectors identified one unreported PI occurrence as defined in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 4, associated with the December 4, 2006 movement and transport of High Integrity Container No. 129 in the radioactive waste barrel yard. The occurrence was not reported by the licensee in its PI submission. The specific circumstances associated with this activity have been previously reviewed and documented in inspection report 05000282/2007002; 05000306/2007002. Additionally, the issue concerning the licensees potential failure to maintain sufficient control over keys to posted VHRAs in both the last Unit-1 (U1R24) and Unit-2 (U2R24) refueling outages remained under review by the licensee. The NRC will review the licensees assessment and any additional information provided by the licensee to determine if these issues represented unreported PI occurrences. Consequently, the NRC will categorize the accuracy of the Occupational Exposure Control Effectiveness PI as an URI pending the inspectors review of additional information from the licensee, (URI 05000282/2007003-05; 05000306/2007003-05). |
| 05000282/FIN-2007003-01 | 2007Q2 | Prairie Island | Failure of the 12 Safety Injection Pump Breaker | On April 3, 2007, during the performance of SP 1322, Safeguards Busses Weekly Inspection - Operating, Revision 20, operators identified that the closing springs were discharged on Breaker 16-7. Breaker 16-7 is the 4.16 kilovolt (KV) alternating current breaker for the 12 SI pump, a safety-related and risk-significant mitigating system component. Operators declared the 12 SI pump inoperable and entered TS Limiting Condition for Operation (LCO) 3.5.2, Condition A, at 7:29 p.m. Operators verified that breaker control power was available and the breaker was racked into bus 16 with auxiliary contacts properly engaged. Based on the actions taken by the operators, the problem was suspected to be internal to the breaker; therefore, the licensee replaced Breaker 16-7 with a spare breaker. Operators demonstrated operability of the 12 SI pump, and exited LCO 3.5.2, Condition A, at 4:01 a.m. on April 4, 2007. The licensee entered the deficient condition into the corrective action program with CAP 01085806. Electrical maintenance personnel performed failure analysis of Breaker 16-7 under WO 323973. During initial bench testing, Breaker 16-7 operated sporadically. Electrical maintenance personnel measured the resistance of the closing spring charging motor and observed that resistance varied from several ohms to seven mega-ohms. On closer inspection of the closing spring charging motor, electrical maintenance personnel observed that one of the motor brushes demonstrated excessive wear. Signs of arcing and charring were also observed between the brush and commutator. The inspectors reviewed the licensees historical operability evaluation CAP 01085806, Action 07, with respect to determining when the failure occurred. The licensee concluded that the closing springs failed to charge the last time the breaker was closed. The operating logs indicated the 12 SI pump breaker was last closed on March 15, 2007, when the 12 SI pump was run for a routine surveillance test. The inspectors reviewed the breaker operating sequence in the technical manual and compared the sequence described to the logic used by the licensee in the determination of the time of failure. The inspectors found that the licensees conclusion was supported by the technical manual. Based on the evaluation of historical operability, the 12 SI pump would not have been available if required from March 15 through April 4, 2007. The inspectors reviewed all operating log records associated with the 12 SI pump from March 15 to April 3, 2007, noting that SP 1322 had been performed by operators on two previous occasions (March 20 and March 27, 2007) prior to the discovery on April 3, 2007, of the discharged closing spring on Breaker 16-7. Step 11.2.10 of SP 1322 directed the performer of the procedure to verify that the closing springs were charged for all bus 16 breakers that were in service. The inspector reviewed CAP 01085806, Action 10, that reviewed operator performance. Based on interviews of the operators that performed SP 1322 on March 20 and March 27, 2007, both operators accurately described what was thought to be the proper method to verify the charged status of the closing springs. A more detailed review of the operator actions is being evaluated as part of the root cause evaluation which is not yet complete. The root cause evaluation team evaluated the possibility that the closing spring was partially charged following operation of the 12 SI pump March 15, 2007. The potential exist that the operators performing the 12 SI pump breaker checks on March 20 and on March 27, 2007, observed the yellow coloring on the tags on the closing spring guides when viewed through the shutter door opening in the breaker door. Per SP 1322, the ability to see yellow on the tags on the closing spring guides would indicate the spring was in the charged state. The inspectors observed an operator check the state of the charge of a closing spring on a similar breaker and compared the observed action to the direction provided in Step 11.2.10 of SP 1322. The inspectors noted that operators could possibly conclude that the closing spring was charged when it was actually still in the discharged state. This observation was discussed with the shift manager and additional investigation was initiated by the licensee for this aspect of the issue. The licensee concluded that knowledge and methodology differences existed among operators performing checks of 4.16 KV closing spring breakers. The licensee entered this condition into their corrective action program with CAP 01098025 on June 20, 2007. This issue is considered an Unresolved Item (URI) 05000282/2007003-01 pending completion of the licensees root cause evaluation. The inspectors subsequent review of the evaluation will determine whether the 12 SI pump motor circuit breaker would have closed with the partially charged closing spring, and the adequacy of the procedure for performing the 4.16 KV closing spring surveillance. |
| 05000263/FIN-2007003-02 | 2007Q2 | Monticello | Licensee-Identified Violation | Technical Specification 3.9.2, Refuel Position One-Rod-Out Interlock requires immediately, with the refuel position one rod-out-interlock inoperable, that 1) control rod withdrawal be suspended, and 2) actions be initiated to fully insert all insert-able control rods in core cells containing one or more fuel assemblies. Contrary to this requirement, during control rod exercise testing on April 20, 2007, the reed switch which fed the one rod-out-interlock was inoperable for Control Rod 26-35 and immediate actions were not taken to suspend control rod withdrawal. Although unrecognized at the time by the operators - because the full core display position indication for Control Rod 26-35 did not change from green to amber in color, the one-rod-out interlock was not met and thus was inoperable. After completing testing for Control Rod 26-35 and while withdrawing the next control rod (26-31) in the test one notch (Position 02), the operators realized that the color change had not occurred for Control Rod 26-35. The operators immediately re-inserted Control Rod 26-31, stopped the test and notified management, and entered the issue into the corrective action program as CAP 01088836. This finding is of very low safety significance because at no time was more than one control rod withdrawn. |
| 05000237/FIN-2006010-05 | 2006Q3 | Dresden | DOA 1900-01, step D.1.c. Can Not Be Performed Under a Loss of AC Power Coincident with Loss of Coolant Accident (LOCA) Conditions | The inspectors identified an unresolved item regarding the performance of DOA 1900-01, Loss of Fuel Pool Cooling, Revision 14. DOA 1900-01, step D.1.c. can not be performed under a loss of AC power coincident with a loss of coolant accident (LOCA) conditions. On January 18, 2006, during testing of the 2A fuel pool cooling pump, per DOA 1900-01, heat exchanger tube side relief valves 2-1999-279 (A relief valve) and 2-1999-280 (B relief valve) lifted. On January 20, 2006, during testing of the 2B fuel pool cooling pump, per DOA 1900-01, both A and B heat exchanger tube side relief valves (2-1999-279 and 2-1999-280) lifted. The 2A fuel pool cooling pump was tested again on January 20, and both A and B relief valves lifted. Following each of the incidents, DOP 1900-01, Fuel Pool Cooling and Cleanup System Startup, was utilized to reseat the relief valves and return the system to a stable condition. The licensee concluded that after a fuel pool cooling pump trip, the pump can not be re-started without operator manual actions in the reactor building. On January 20, 2006, the licensee determined that DOA 1900-01, step D.1.c. can not be performed under a loss of AC power coincident with loss of coolant accident (LOCA) conditions. Step D.1.c. provides guidance on how to start a fuel pool cooling pump in case access to the reactor building is not possible. This condition affects Unit 2 and likely affects Unit 3. These events were documented in IR 444332. The inspectors challenged the licensee as to whether the condition of Unit 2 (and potentially Unit 3) fuel pool cooling system should be an operator workaround or challenge. The licensee initiated IR 528541 to address the inspectors concern. Also, the inspectors inquired as to whether any compensatory actions were in place and if there was an alternate success path to accomplish the re-start of the fuel pool cooling pumps under a loss of AC power coincident with loss of coolant accident (LOCA) conditions. The compensatory action in place directed operations personnel to take actions to ensure DOA 1900-01, step D.1.c. is not used on either unit until a solution to the problem is implemented. At the end of the inspection period, the licensee was still evaluating if there is an alternate success path to accomplish the re-start of the fuel pool cooling pumps. The inspectors considered this issue to be an unresolved item pending evaluation efforts. |
| 05000237/FIN-2006010-04 | 2006Q3 | Dresden | Full Flow Testing of the Diesel Driven Flood Pump at Design Conditions | The inspectors reviewed the details of the pump test and the licensees conclusion as documented. The inspectors noted that the pump capacity at the maximum pump speed was measured at 298 gpm at a discharge head of 114 psig. This was only 80.5 percent of the expected capacity of 370 gpm based on the manufacturers pump curve. The inspectors were concerned that the lower than expected flowrate may be an indication of pump degradation. In addition, the licensee tested the pump at only one point and assumed that the pump curve would follow the same pump curve established by the manufacturer. The inspectors questioned the method of testing and the method used to extrapolate the flowrate to generate a pump curve. Also during this inspection, the inspectors questioned licensee personnel to determine if any actions had been taken to address the cause of the 19.5 percent degraded pump test results, and whether the licensee initiated any actions to ensure the pump would not degrade further over time. The inspectors learned that no actions had been taken. Licensee personnel stated that the test result did not necessarily indicate that the pump was degraded. The inspectors concluded that the lack of additional points tested on the pump curve did not ensure the pump would provide adequate flow at design conditions. The inspectors concluded that the licensees corrective actions of IR 246038 were not fully effective. Specifically, the lack of a robust testing methodology to ensure performance of the emergency flood pump to the manufacturers pump curve resulted in the licensees planning to send the pump to an offsite facility for adequate testing. This is an unresolved item pending NRC review of the licensees planned corrective action to perform full flow testing of the diesel driven pump at design conditions. (URI 05000237/2006010-04; 05000249/2006010-04) |
| 05000237/FIN-2006010-03 | 2006Q3 | Dresden | Adequacy of Ground/Well Waterborne Monitoring to Satisfy Radiological Effluent Technical Specification Surveillance Requirements | No findings of significance were identified. However, the inspectors questioned the basis for the licensees waterborne ground/well offsite sampling locations and corresponding compliance with the RETS surveillance requirement specified in Chapter 12.5 of the ODCM. Specifically, Table 12.5-1 of the RETS, Radiological Environmental Monitoring Program, requires that quarterly ground/well waterborne samples be collected and analyzed from three sources only if likely to be affected. Waterborne sources likely to be affected are defined in Table 12.5-1 as those that are tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination. The licensee has historically sampled from two offsite wells located to the south and west of the Dresden site, both south of the Illinois River. Liquid radwaste effluents are discharged to the Illinois River which flows in a westerly direction. As determined during the licensees most recent land use census, several residential wells of varying depths located on the northern banks of the Illinois River downstream of the licensees liquid effluent (radwaste) discharge point are used as potable sources and/or for irrigation purposes and potentially may be affected sources if the hydraulic gradient or recharge properties are suitable. However, the technical basis for limiting the well water sampling program to the two wells historically sampled versus other offsite wells including residential wells downstream of the stations liquid effluent discharge into the Illinois River on the northern banks of the river could not be provided by the licensee. Consequently, compliance with Table 12.5-1 of the RETS could not be determined. As documented in issue report (IR) 532766, the licensee is contemplating plans to further evaluate Dresden site hydrologic data, including several existing hydrogeology studies, to validate its historical well sampling activities and to assess compliance with Table 12.5-1 of the RETS. Pending the outcome of the licensees evaluation and the NRCs review of that information, this issue is categorized as an Unresolved Item (URI 050-237/2006010-03; 050-249/2006010-03). |
| 05000456/FIN-2006012-01 | 2006Q2 | Braidwood | Failure to Perform Surveys to Assure Compliance with 10 CFR 20.1301, Which Limits Radiation Exposure to 0.1 Rem | 10 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Pursuant to 10 CFR 20.1003, survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. 10 CFR 20.1301 requires the licensee to conduct operations so that the total effective dose equivalent to individual members of the public from the licensed operation does not exceed 0.1 rem (1 mSv) in a year. Between November 1996 and March 2005, the licensee did not make surveys to evaluate the potential hazards and to assure compliance with 10 CFR 20.1301, which limits radiation exposure to members of the public from licensed operations to 0.1 rem. Specifically, in November 1996, December 1998, and November 2000, failed vacuum breakers in the licensees radioactive waste blowdown line resulted in large volumes of liquid contaminated with licensed material to leak in an uncontrolled manner to the unrestricted areas. Following the identified releases of radioactive material, the licensee failed to perform an adequate radiological survey to identify the extent of radiation levels, to evaluate the potential hazards associated with the radioactive material, and to ensure that the dose to the public did not exceed the levels specified in 10 CFR 20.1301. (AV 05000456, 457/2006008-01). Technical Specification 6.8.4.e.5 requires that the licensee maintain and implement a program to determine the cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year in accordance with the methodology and parameters in the Offsite Dose Calculation Manual (ODCM) at least once per 31 days. Between November 1996 and March 2006, the licensee failed to determine the cumulative dose contributions from liquid effluents that inadvertently leaked into onsite and offsite groundwater (resulting from failed vacuum breakers along the circulating water blowdown line in 1996, 1998, and 2000) in accordance with the methodology and parameters in the ODCM within 31 days. Specifically, an estimated 250,000 gallon leak from Vacuum Breaker No.1 in November 1996 released water with radioactive material to the groundwater pathway; however, the licensee did not determine the dose from the release. In December 1998, an estimated 3 million gallon leak from Vacuum Breaker No. 3 released water with radioactive material to the groundwater pathway; however, the licensee did not determine the dose from the release. In November 2000, an estimated 3 million gallon leak from Vacuum Breaker No. 2 released water with radioactive material to the groundwater pathway; however, the licensee did not determine the dose from the release. (AV 05000456, 457/2006008-02). Technical Specification 6.9.1.6 requires that the Annual Radiological Environmental Operating Report include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period and that the material shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) and in 10 CFR Part 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. 10 CFR Part 50, Appendix I, Section IV.B.2 states the licensee shall establish an appropriate surveillance and monitoring program to provide data on measurable levels of radiation and radioactive materials in the environment to evaluate the relationship between quantities of radioactive material released in effluents and resultant doses to individuals from principal pathways of exposure. Between November 1996 and March 2006, the licensee did not establish an appropriate surveillance and monitoring program to evaluate the relationship between quantities of radioactive material released in effluents and resultant doses to individuals from principal pathways of exposure. Specifically, the unplanned radioactive material released in 1996, 1998, and 2000 from the circulating water blowdown line vacuum breakers constituted new principal pathways of exposure (i.e., the groundwater pathway) which the licensee had not adequately evaluated with the existing Radiological Effluent Monitoring Program (REMP). (AV 05000456, 457/2006008-03). After considering the information developed during the inspection, the NRC has concluded that the inspection finding is appropriately characterized as White. The NRC also determined that the inspection finding involved three violations of NRC requirements, as cited in the attached Notice of Violation (Notice). The three violations involved your staffs failure to: 1) perform adequate radiological surveys, as required by 10 CFR 20.1501; 2) adequately implement a program to assess the cumulative dose contributions, as required by Technical Specification 6.8.4.e.5; and 3) conduct an adequate environmental monitoring program to provide data on measurable levels of radiation and radioactivity in the environment resulting from the releases, as required by Technical Specification 6.9.1.6. The circumstances surrounding the violations are described in detail within NRC Inspection Report 05000456/2006008; 05000457/2006008 (DRS). In accordance with the NRC Enforcement Policy, the Notice of Violation is considered an escalated enforcement action because it is associated with a White finding |
| 05000266/FIN-2005013-06 | 2005Q4 | Point Beach | Multiple Examples of the Failure to Notify the NRC within 8 Hours as Required by 10 CFR 50.72 | The inspectors identified a finding of very low safety significance (with three examples) for the licensees failure to notify the NRC within 8 hours in accordance with 10 CFR 50.72(b)(3)(ii)(B), following the identification that the nuclear power plant was in an unanalyzed condition that significantly degraded plant safety. Each occurrence was reported by the licensee following repeated questioning by the inspectors that occurred in April, September, and November 2005. Following the November occurrence, the inspectors reviewed the licensees previous causal evaluations and corrective actions. The inspectors noted that while the licensee had appropriately evaluated and initiated corrective actions for the technical issues in April and September 2005, the licensee had not appropriately evaluated or developed any corrective actions to address the failure to adequately report these issues to the NRC in a timely manner. Because this issue affected the NRCs ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The finding has been reviewed by NRC management and is determined to be a Green finding of very low safety significance. Because the licensee entered the issue into their corrective action program (CAP068938), this violation is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee has taken actions to perform a causal evaluation and address the knowledge and procedural aspects of this finding. The cause of the finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to appropriately evaluate and take adequate corrective actions for the reportability aspect of these issues. |
| 05000266/FIN-2002014-01 | 2003Q2 | Point Beach | Decreased an Emergency Plan Commitment Without Prior NRC Approval | In October 1998, the licensee decreased its Emergency Plan's effectiveness without prior NRC approval due to an inadequate 10 CFR 50.54(q) review of six Emergency Response Organization (ERO) positions, which the licensee recategorized from being 30 minute response positions to be 60 minute response positions. These six positions were reestablished as 30 minute response positions in late January 2003. This Severity Level IV violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. |