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| Title | Quarter | Description | |
|---|---|---|---|
| 05000282/FIN-2000013-05 | the Failure to Determine During a Design Change of the Cooling Water Supply to the Auxiliary Feed Water System That the Change Could Have Affected the Operability and Reliability of the Auxiliary Feedwater System During a Design Basis Seismic Event | 2000Q4 | A Non-Cited Violation was identified during the review of a 1995 modification, installed in the cooling water system supply for the Auxiliary Feedwater Pumps. The design change review process did not consider the increased failure rate of the AFW system due to the increased probability that the AFW pump would trip on low suction pressure with the modification installed. Criterion III of 10 CFR Part 50, Appendix B requires that design changes be subject to design control measures commensurate with those applied to the original design, including verifying or checking the adequacy of the design by the performance of design reviews, calculations, or testing. The failure to determine the effect of a design change on the AFW system performance was identified as a violation of Criterion III of 10 CFR Part 50, Appendix B. |
| 05000282/FIN-2001011-02 | N/A | 2001Q2 | A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to take effective corrective action for a recurrent problem with equipment configuration control. The ineffective corrective action was more than a minor issue because if left uncorrected, the issue could become a more significant safety concern. However, since no specific cornerstone has been impacted, this finding is designated as No Color. |
| 05000282/FIN-2001014-01 | the Failure to Identify All of the Root Causes for and to Develop Corrective Actions to Preclude the White Finding Associated with the Inoperable Cooling Water Pumps | 2001Q2 | The inspectors determined that the licensee.s evaluation of the White finding did not identify all of the root causes for the finding and did not propose corrective actions to preclude recurrence. Specifically, the evaluation did not identify root causes associated with inadequate staff and management knowledge of the cooling water pump design and did not identify process and procedure inadequacies which allowed the condition to continue for 25 years. As a result, the licensee did not propose corrective actions for these root causes. In addition, the White inspection finding associated with the inoperable cooling water pumps will remain open. |
| 05000282/FIN-2006009-01 | Redundant Curcuit Not Entirely Protected | 2006Q3 | The inspectors identified a violation of 10 CFR Part 50, Appendix R, Section III.G.2, involving the licensees failure to ensure, in the event of a severe fire, that one redundant train of systems necessary to achieve and maintain hot shutdown conditions was free of fire damage. Specifically, the licensee failed to ensure, in the event of a fire in either one of the auxiliary feedwater (AFW) pump rooms (Fire Areas 31 and 32), that cables and circuits of one redundant train were adequately protected by a one-hour fire-rated barrier. This violation was entered into the licensees corrective action program (CAP) as 01045012, Appendix R Compliance Issues with Fire Area 31 and 32, dated August 17, 2006. The licensee initiated compensatory measures and will evaluate the violation during their transition to NFPA 805. The finding was more than minor because this failure could have affected the mitigating systems cornerstone objective and safe shutdown (SSD). Specifically, the licensees failure to physically protect the entire length of redundant cables required for SSD, in the event of a fire in the 10 CFR Part 50, Appendix R, Section III.G.2 fire area, left the SSD cables vulnerable to fire damage. Because the NRC identified violation was a circuit-related finding that was not associated with a finding of high safety significance, the inspectors evaluated the violation in accordance with the four criteria established by Section A of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation; (1) the licensee would have identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. As a result, the inspectors concluded that the violation met all four criteria established by Section A and the NRC is exercising enforcement discretion to not cite this violation in accordance with the NRCs Enforcement Policy. |
| 05000282/FIN-2007003-01 | Failure of the 12 Safety Injection Pump Breaker | 2007Q2 | 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. |
| 05000282/FIN-2007003-04 | Control of Very High Radiation Area Keys | 2007Q2 | 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 | Performance Indicator Accuracy for Occupational Radiation Safety | 2007Q2 | 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-2007005-02 | Potential Inadequate Corrective Actions to Prevent Unnecessary D5 Diesel Generator Unavailability (Section 1R15) | 2007Q4 | 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) |
| 05000282/FIN-2007006-01 | Evaluate TSC Operability During Time with Damper Disconnected | 2007Q4 | 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-2008002-01 | TSC Ventilation Issues Resulted in Inadequate Emergency Response Facility | 2008Q1 | The inspectors identified a NCV of 10 CFR 50.54(q), associated with 10 CFR 50.47(b)(8), for failing to maintain adequate emergency facilities to support emergency response. Specifically, the licensee failed to maintain control of the Technical Support Center ventilation system. As a result, the system was frequently found to be in a degraded condition that may not have provided adequate protection for emergency response personnel. This finding was more than minor because it was associated with the attribute of meeting the planning standards of 10 CFR 50.47(b). In addition, the finding affected the cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In accordance with the SDP Phase 1 Screening Worksheet of IMC 0609, the inspectors applied Appendix B, Emergency Preparedness Significance Determination Process, Section 4.8 and determined that this issue was of very low safety significance. Specifically, the Technical Support Center ventilation system was degraded for a period of longer than seven days from the time of original discovery. In addition, the degradation was to the extent that key emergency response organization members may not have been able to perform their assigned plan functions without compensatory measures. The finding was determined to be cross-cutting in the corrective action program aspect of the Problem Identification and Resolution crosscutting area because the licensee failed to thoroughly evaluate repeated problems with the Technical Support Center ventilation system such that the causes of the problems were identified and addressed (P.1(c)). (Section 1R15 |
| 05000282/FIN-2008002-03 | Worker not in compliance with TS 5.7.1.b received an ED dose-rate alarm when he inappropriately entered a HRA of the plant during steam generator set-up work. | 2008Q1 | A self-revealing finding of very low safety significance and an associated NCV were identified for the licensees failure to comply with Technical Specification 5.7.1.b for access control to high radiation areas of the plant. As a result of poor human performance, a contract radiation worker received an electronic dosimeter high doserate alarm while performing steam generator set-up activities, when he inappropriately entered a high radiation area of the plant on a non-high radiation area radiation work permit. As corrective actions, the licensee provided additional training to the individuals involved and reinforced the expectations for high radiation area access control. The finding was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and potentially affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation, in that the failure to implement controls for high radiation area entry may result in unplanned dose. The finding was determined to be of very low safety significance because the finding did not involve As-Low-As-Is-Reasonably-Achievable (ALARA) planning; it did not involve an overexposure; there was not a substantial potential for a worker overexposure; and the licensee=s ability to assess worker dose was not compromised. The cause of the finding is related to a cross-cutting aspect of human performance in work control (H.3(b)). (Section 2OS1 |
| 05000282/FIN-2008002-04 | Licensee-Identified Violation | 2008Q1 | Technical Specification 3.7.11, Condition A, required that an inoperable safeguards chiller water system (SCWS) loop be returned to operable status within 30 days. Technical Specification 3.7.11, Condition B, required that if Condition A was not met, both units must be in Mode 3 in 6 hours and in Mode 5 in 36 hours. As discussed in Section 4OA3.3 of this report and CAP 01120914, the 121 loop of SCWS was inoperable from October 1, 2007, through the time it was repaired on December 15, 2007, and neither unit was shut down. Since the leak was small enough not to affect functionality of the system, had not grown significantly, and was repaired in a timely manner once it was determined to be pressure boundary leakage, the violation was of very low safety significance |
| 05000282/FIN-2008002-05 | Licensee-Identified Violation | 2008Q1 | Part 50.54(q) of 10 CFR required that the licensee follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b). Part 50.47(b)(8) of 10 CFR required that adequate emergency facilities and equipment to support the emergency response are provided and maintained. Prairie Island Nuclear Generating Plant Emergency Plan, Revision 36, Section 7.1.1.B required that the TSC have a shielding and ventilation cleanup system to provide habitability under accident conditions. As reported by the licensee in Event Notification 44057, on March 12, 2008, the licensee completed a preliminary review of a revised dose analysis for the TSC that showed that the whole body dose acceptance criteria of NUREG-0696 could have been exceeded during an event and compensatory measures might be required. The licensee entered the issue into its corrective action program as CAP 01130807. By the same logic as used for another TSC finding discussed in Section 1R15 of this report, the TSC function was considered to be degraded for more than seven days from time of discovery, but still functional, and the violation was determined to be of very low safety significance |
| 05000282/FIN-2008003-01 | 11 TURBINE-DRIVEN Auxiliary Feedwater Pump Inoperable During Startup from Outage 1R25 | 2008Q2 | On March 23, 2008, the licensee tested the auxiliary feedwater system using Surveillance Procedure (SP) 1103, 11 Turbine-Driven Auxiliary Feedwater Pump Once Every Refueling Shutdown Flow Test. During the test, the 11 TDAFW pump was secured due to the turbine outboard bearing temperature reaching 220 degrees Fahrenheit (oF). Operations personnel declared the 11 TDAFW pump inoperable. The licensee conducted troubleshooting efforts and determined that the high bearing temperature was due to improper installation of the turbine insulation following maintenance. The licensee properly re-installed the insulation and the pump was restored to an operable status on March 24, 2008. The inspectors reviewed the Unit 1 operator logs and the results of multiple auxiliary feedwater system tests conducted between March 16 and March 23, 2008, to determine whether the licensee had a prior opportunity to identify the outboard bearing temperature issue. This review included an assessment of turbine outboard bearing temperature trends captured by the Emergency Response Computer System (ERCS) system. The inspectors ascertained that on March 16, 2008, operations personnel performed the following tests concurrently: SP 1102 11 Turbine-Driven Auxiliary Feedwater Pump Monthly Test; SP 1330 11 Turbine-Driven Auxiliary Feedwater Turbine/Pump Bearing Temperature Test; and SP 1376 Auxiliary Feedwater Flow Path Verification Test After Each Cold Shutdown. Operations personnel began the series of tests by performing SP 1102. This test was performed with the 11 TDAFW pump operating and supplying approximately 35 gpm of water to the steam generators. Once this condition was established, the operators transitioned into SP 1376 which required that the 11 TDAFW pump supply 70 gpm of water to the steam generators for approximately 20 minutes. The inspectors performed a detailed review of the turbine outboard bearing temperature data provided by ERCS and identified that the bearing temperature reached approximately 216oF and was continuing to increase when the 20 minutes had elapsed. Operations personnel considered the results of SP 1376 satisfactory even though the bearing temperature was within 4 degrees of the vendor specified limit. Operations personnel then reduced the 11 TDAFW pump flow rate to establish the conditions needed to perform SP 1330. The reduction in feedwater pump flow resulted in a corresponding reduction in bearing temperature. The inspectors reviewed SP 1330 and found that the procedure could be considered satisfactory when three successive turbine outboard bearing temperature readings taken ten minutes apart varied by less than 3 degrees. The inspectors reviewed the ERCS data attached to SP 1330 and found that the operators concluded that the 11 TDAFW pump outboard bearing temperature was stabilized at 211oF. The inspectors questioned this conclusion because the outboard bearing temperature recorded during SP 1376 was approximately 5 degrees higher. The operators also initiated CAP 1131305 to document that the outboard bearing temperature exceeded the alert limit of 203oF. However, no further actions were taken to address or evaluate whether the 11 TDAFW pump outboard bearing temperature would remain less than 220oF during post-accident conditions. The inspectors sampled the licensees corrective action system to determine whether any previous bearing issues had occurred due to the improper installation of insulation following maintenance. The inspectors found that the 11 TDAFW pump outboard turbine bearing failed during the performance of SP 1103 on June 6, 2006. The licensee determined that the bearing failed due to improper bearing installation during maintenance. However, improper insulation installation following maintenance contributed to the increased bearing temperatures. Corrective actions for this event consisted of replacing the bearing, installing the insulation correctly, and providing additional oversight of maintenance activities conducted on the 22 TDAFW pump in 2007. The inspectors questioned engineering and maintenance personnel to determine if written procedural guidance regarding the installation of the turbine insulation following maintenance had been provided to the insulators. The inspectors were informed that written procedural guidance had not been provided because insulation installation was considered to be a skill of the craft activity. At the conclusion of the inspection period, based upon the information known to date, the inspectors have not concluded whether the high temperatures experienced by the 11 TDAFW bearings caused the pump to be inoperable. As a result, the inspectors were unable to fully evaluate this issue for potential performance deficiencies and safety significance. Therefore, this issue was considered an unresolved item (URI 05000282/2008003-01) pending further review by the inspectors. The inspectors needed to evaluate how the operability of the 11 TDAFW was impacted by: (1) opportunities for the licensee to have identified the high bearing temperatures during the testing completed March 16 though 23, 2008, and (2) the timeliness and adequacy of corrective actions taken for past issues related to the improper installation of the TDAFW insulation following maintenance, specifically, whether or not the insulation installation was skill of the craft |
| 05000282/FIN-2008003-02 | USAR Not Updated to Include Analyses | 2008Q2 | The inspectors identified an Non-Cited Violation of 10 CFR 50.71, Maintenance of records, making of reports, for the licensees failure to adequately update the Prairie Island Nuclear Generating Plant Updated Safety Analysis Report (USAR) to include analyses performed in response to Generic Letter (GL) 2004-02. Title 10 CFR 50.71(e) requires, in part, that the USAR be revised to include the effects of all analyses of new safety issues performed by or on behalf of the licensee at Commission request. The Commission, through GL 2004-02, requested that licensees perform an evaluation of the Emergency Core Cooling Systems and its associated recirculation functions and, if appropriate, take additional actions to ensure system function. The licensee, in response to GL 2004-02, performed analyses of debris generation and transport, chemical effects, downstream effects, upstream effects, and strainer and other structural analysis, but did not update the safety analysis report to include those analyses. This issue potentially impacted the NRCs ability to perform its regulatory function and therefore, it was evaluated using the traditional enforcement process. The inspectors determined that the finding was more than minor because of the potential to impact the regulatory process by using IMC 0612, Appendix B, Issue Screening, dated September 20, 2007. Specifically, the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. This finding has a cross-cutting aspect in the area of human performance, work practices (H.4(c)) because the licensee did not ensure supervisory and management oversight of work activities such that nuclear safety was supported. Corrective actions included revising the USAR to reflect the analyses and submitting the updated information to the NRC. (Section 4OA5.1.c |
| 05000282/FIN-2008003-03 | Failure to Test Check Valve SI-9-5 Under Suitable Environmental Conditions | 2008Q2 | An inspector identified finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, due to the licensees failure to ensure Check Valve SI-9-5 was tested under suitable environmental conditions. Specifically, the licensee preconditioned SI-9-5 prior to testing by increasing reactor pressure and tapping on the valve with a hammer. The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone. The finding affected the cornerstone objective of limiting the frequency of those events that upset plant stability and challenge critical safety functions. The inspectors concluded that the finding was of very low safety significance because it was not a primary system loss of coolant accident or transient initiator. Additionally, the finding did not screen as potentially risk significant due to a fire, seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the corrective action program component of the Problem Identification and Resolution cross-cutting area because the licensee failed to evaluate the cause of the 2008 SI-9-5 valve test failures to ensure that the resolution addressed the cause and extent of condition (P.1(c)). The corrective actions for this issue included restoring the valve to an operable but degraded condition, providing training on preconditioning, providing training on the use and implementation of the operability determination process, and improving the thorough evaluation of equipment related deficiencies. (Section 4OA5.3 |
| 05000282/FIN-2008003-04 | Licensee-Identified Violation | 2008Q2 | Title 10 CFR Part 50, Appendix B, Criterion V, required, in part, that activities affecting quality be accomplished in accordance with instructions, procedures, and drawings appropriate to the circumstance. Procedure D63, Installation Guidelines for Threaded Fasteners, Section 1.0 stated that this procedure was to be employed by maintenance personnel when removing and installing threaded fasteners. Procedure D63, Step 7.3.7 required that bolts/studs be torqued to the proper stress value. Contrary to the above, on May 15, 2008, licensee personnel tightened safety related fasteners without employing the proper stress values included in Procedure D63. This violation was of very low safety significance because the failure to torque the fasteners to the required value did not impact the operability of the specific components. The licensee initiated CAP 1137836 to document this issue. Corrective actions included a review of the torquing practices to ensure that the fasteners are appropriately installed |
| 05000282/FIN-2008003-05 | Licensee-Identified Violation | 2008Q2 | Title 10 CFR Part 55.49, stated, in part, that station personnel shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. The integrity of a test or examination was considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. This included activities related to the preparation and certification of license applications and all activities related to the preparation, administration, and grading of the tests and examinations required by this part. Contrary to this, on one occasion, a trainer directed an operator to validate a biennial written examination required by 10 CFR 55.59(a)(2) that was intended for use later in the examination cycle. The examination contained three questions that appeared on the operators own biennial written examination. Because the operator saw questions that he was administered on his own examination, a violation of 10 CFR 55.49 requirements occurred. The violation was of very low safety significance because the operator correctly answered the questions both times the operator encountered them, if the questions were removed from the examination, the examination was still a valid examination, and the operators grade on the examination was still a passing grade. Immediate actions were taken by the licensees training department included verifying that no other examination compromise situations occurred during the examination validations. The licensee entered this condition into the corrective action program as CAPs 1115206 and 1115320. The licensees training personnel were briefed concerning examination security requirements and the need to comply with examination security procedures |
| 05000282/FIN-2008004-01 | Failure to Maintain Staff Respiratory Qualifications Including Personnel Qualifications Necessary for Emergency Response Duties As Required By Station Procedures. | 2008Q3 | The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q) for the failure to maintain staff respiratory qualifications, including personnel qualifications necessary for emergency response duties, as required by station procedures. Specifically, the inspectors identified multiple instances over the last several years where station personnel, including those required to maintain their respiratory readiness necessary for emergency response functions, failed to maintain their qualifications current. The most recent instances being a fire brigade member standing duty without the necessary respiratory fit test and a reactor operator standing duty without the necessary respiratory protection training. Planned corrective actions included periodic reviews to identify respiratory protection qualification issues prior to expiration to ensure that impacted departments maintained compliance with station procedures until the next scheduled periodic review. The issue was more than minor because it was chronic in nature and associated with the facilities/equipment attribute of the Emergency Preparedness Cornerstone. The inspectors determined that the issue affected the cornerstone objective to ensure adequate protection of plant emergency workers (and consequently the health and safety of the public in the event of a radiological emergency) should the workers be called upon to use the equipment. Since the finding did not represent a functional failure of the Planning Standard, and the workers who were required to use respiratory protective equipment were not qualified and/or trained to use that equipment, the finding was determined to be of very low safety significance (Green). The inspectors determined that this finding was cross-cutting in the area of Problem Identification and Resolution because the licensee failed to take appropriate corrective actions once the issue was identified (P.1(d)). (Section 2OS3.4) |
| 05000282/FIN-2008004-02 | Load Sequencer Test Procedure Conflicts With Vendor Manual Information | 2008Q3 | The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V for the failure to ensure that the surveillance procedures used to test the safety-related load sequencers included appropriate qualitative acceptance criteria. Specifically, the acceptance criteria specified in the procedure conflicted with vendor manual information and was less conservative. Corrective actions for this issue included revising the surveillance procedures to include the vendor manual information and implementing a comprehensive preventive maintenance program to improve the availability and reliability of the load sequencers. This finding was more than minor because it was associated with the procedure quality and equipment performance attributes of the Mitigating Systems Cornerstone. In addition, the finding affected the cornerstone objective of ensuring the availability and reliability of equipment to respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding was of very low safety significance because it was not a design issue resulting in loss of operability or functionality, it did not result in a loss of safety function, it did not result in loss of safety function for a single train for greater than the allowed outage time, and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that this finding was cross-cutting in the Human Performance, Decision Making area because the licensee failed to use conservative assumptions during the February 2007 decision that led to making the load sequencer surveillance procedure non-conservative (H.1(b)). (Section 4OA2.3 |
| 05000282/FIN-2008004-03 | 11 Turbine-Driven Auxiliary Feedwater Pump Inoperable Due to Improperly Installed Insulation | 2008Q3 | The inspectors identified an apparent violation (AV) of Technical Specification 5.4.1 for the failure to establish, implement and maintain procedures governing the installation of insulation on the 11 turbine-driven auxiliary feedwater (TDAFW) pump. The failure to establish and implement adequate instructions resulted in the 11 TDAFW pump being inoperable for 10 days due to improper insulation installation during the March 2008, Unit 1 refueling outage. This issue has the potential to have low to moderate safety significance; however, this may change pending the completion of the SDP. Corrective actions for this issue included correctly installing the insulation, exploring the installation of a different insulation package that was easier to install, and performing an internal inspection to determine if mechanical clearances inside the turbine were contributing to the increased turbine bearing temperatures. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 11 TDAFW pump would not have been able to perform its safety function during the worst case, post-accident conditions. This finding was not an immediate safety concern because the licensee took immediate action to declare the pump inoperable once the condition of the insulation was identified. The inspectors determined that this finding was cross-cutting with respect to the Decision Making aspect of the Human Performance area because the licensee failed to use conservative assumptions when determining the need to establish and implement instructions for installing the turbine insulation (H.1(b)). (Section 4OA5.3 |
| 05000282/FIN-2008004-04 | Licensee-Identified Violation | 2008Q3 | Operating License Condition 2.C.(4) of Prairie Island Nuclear Generating Plant Unit 1 Operating License DPR-42 and Prairie Island Nuclear Generating Plant Unit 2 Operating License DPR-60 required that Nuclear Management Company implement and maintain in effect all provisions of the approved fire protection program as described and referenced in the USAR and as approved in Safety Evaluation Reports dated February 14, 1978; September 6, 1979; April 4, 1980; December 29, 1980; July 28, 1981; September 12, 1984; June 25, 1985; October 27, 1989; and October 6, 1995. USAR Section 10.3.1.4, Fire Protection System Inspection and Testing, states that the fire protection surveillance and test requirements are identified in Operations Manual F5, Appendix K, Fire Protection Systems Operability Requirements. Operations Manual F5, Appendix K, Fire Protection Systems Operability Requirements, Section 8.8, required that each fire hose station be visually inspected each month and hydrostatically tested every three years. Contrary to the above, on August 13, 2008, the licensee identified that they had failed to visually inspect the in-plant hose reels on a monthly basis. In addition, the hose reels had not been hydrostatically tested every three years as required by procedure. The inspectors assessed the significance of this finding using IMC 0609, Appendix F. The inspectors determined that this finding was related to fixed fire protection systems. The inspectors assigned a low degradation rating to this finding because alternate hose stations, which were properly tested, were available in the plant for use in fire fighting activities. Based upon the guidance provided by IMC 0609, Appendix F, this finding was of very low safety significance. The licensee initiated CAP 1147453 to document this issue. Corrective actions for this issue included replacing the hose reels and changing the preventive maintenance program to ensure that the hose reels were tested as required |
| 05000282/FIN-2008004-05 | Licensee-Identified Violation | 2008Q3 | Title 10 CFR Part 50, Appendix B, Criterion V, required that activities affecting quality be prescribed by documented instructions, procedures, and drawings appropriate to the circumstance. Fleet Procedure FP-OP-COO-01, Conduct of Operations, Step 3.5, required operations personnel to control equipment configuration such that the status of plant equipment was known at all times. In addition, Operating Procedure C18.1, Engineered Safeguards Equipment Support Systems, Section 5.25, required operations personnel to consult engineering if the temperatures on the 695-foot elevation of the auxiliary building approached 104 degrees. Contrary to the above, on July 2, 2008, operations personnel failed to control the plant configuration on the 695-foot elevation of the Unit 1 auxiliary building prior to the indicated temperature exceeding 104 degrees. As a result, engineering was not consulted as directed by Operating Procedure C18.1. This violation was of low safety significance because it did not result in the inoperability of any safety-related equipment. The licensee initiated CAP 1142946 to document this issue. Corrective actions for this issue included revising the operator rounds procedure to ensure that auxiliary building temperatures were trended on a periodic basis, improving the instrumentation used to determine the auxiliary building temperatures, and improving the guidance provided in Operating Procedure C18.1 |
| 05000282/FIN-2008005-01 | Failure of Contractors to Follow Welding Procedures | 2008Q4 | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, in September 2008 for the failure of contractor welders to adhere to welding procedures during structural weld overlay (SWOL) repairs on a pressurizer surge nozzle. A review of the weld records indicated that the welders either failed to utilize the correct travel speeds in performing the SWOL or to accurately document relative travel speed settings as required by procedure, in order to ensure that the correct heat input (a welding essential variable) was maintained. The inspectors also identified that the welders failed to input the correct welding parameters into the welding controller for a portion of the overlay as required by procedure. This resulted in the heat input parameters being exceeded. Corrective actions for this issue included the removal and repair of the weld. This finding was more than minor because if left uncorrected, it would have become a more significant safety concern. Specifically, the failure to control the heat input could have reduced the impact toughness of the pressurizer weldment such that it would be susceptible to brittle fracture. The finding was of very low safety significance (Green) because the contractor subsequently addressed the programmed versus actual travel speed discrepancies and determined that the resulting heat inputs were bound by the welding procedure specifications (WPS) parameters. Furthermore, the contractor repaired the surge nozzle as a result of using the incorrect welding parameters before returning Unit 2 to service. The inspectors determined that this finding was cross-cutting in the Human Performance, Work Practices area because licensee personnel failed to ensure supervisory and management oversight of contractor activities such that nuclear safety was supported |
| 05000282/FIN-2008005-02 | Component Cooling Water Susceptible to High Energy Line Break Interaction | 2008Q4 | On July 29, 2008, the licensee initiated CAP 1145695 to document that CCW piping located in the turbine building, and used to supply water to the chemistry cold lab, passed directly underneath high energy piping from the 15A and 15B feedwater heaters. As part of the CAP review, operations personnel requested that an operability review be completed to evaluate the impact that a failure of this high energy piping could have on the continued operability of the CCW system. The licensee initially determined that a failure of this piping would have some impact on the Unit 2 CCW system because the 2A train of CCW normally supplied water to the chemistry cold lab. There was no impact on the operation of the Unit 1 CCW system. As a result, this condition was expected to have little impact on overall plant operation because each unit maintained its ability to achieve its required operating condition following a Unit 1 high energy piping failure. Over the next few days, the licensee continued to review the high energy and CCW piping configurations in the Unit 1 and Unit 2 turbine buildings. On July 31, 2008, the licensee identified that a failure in a Unit 2 turbine building high energy line could impact the continued operability of the Unit 2 CCW system. The licensee conducted an operability review and determined that the Unit 2 CCW system was inoperable due to the potential interaction between the Unit 2 turbine building high energy piping and the Unit 2 CCW system. The licensee also determined that the operators ability to bring Unit 2 to a cold shutdown condition following this type of piping failure may be impacted. Operations personnel immediately entered TS 3.0.3 to address this concern. The operators closed several valves in the CCW system to isolate the CCW piping in the turbine building from the other CCW piping. By closing these valves, operations personnel eliminated the potential that the Unit 2 CCW system would become inoperable following a Unit 2 turbine building high energy piping failure. The Unit 2 CCW system was restored to an operable status following the valve closures. The CCW piping to the turbine building remained isolated at the conclusion of the inspection period. At the conclusion of the inspection period, the inspectors were reviewing the high energy and CCW piping configurations in the Unit 1 and Unit 2 turbine buildings to ensure that the piping failures discussed in the licensees CAP documents and Licensee Event Report (LER) 05000306/2008-01 were the most limiting locations. In addition, the inspectors were waiting for information regarding the actual times that each units CCW system was aligned to the chemistry cold lab to determine whether the impacts of a Unit 1 high energy piping failure on the continued operability of the Unit 1 CCW system and the shutdown cooling function of the RHR system needed further review. Finally, the inspectors needed to review several older CAPs and operating experience related to the issue to determine whether the issue was within the licensees ability to foresee and correct (i.e., is it a performance deficiency). As a result, this item was considered unresolved pending the receipt and review of the above information (URI 05000306/2008005-02 |
| 05000282/FIN-2008005-03 | Control Rod Bent due to Contractors Failure to Follow Procedures | 2008Q4 | A finding of very low safety significance and an associated NCV of TS 5.4.1 was self-revealed on October 9, 2008, due to the failure of contractor staff to follow procedures during refueling activities. This failure to follow procedures resulted in the insertion of a plug in a local leak rate testing port on the fuel transfer tube flange. The plug subsequently contacted a control rod located in a new fuel assembly and damaged the control rod while lifting the fuel assembly to a vertical position. Corrective actions for this issue included removing the plug, inspecting the fuel bundle and refueling equipment for damage, verifying the clearances between the fuel transfer tube flange and the upender basket, establishing a minimum design clearance between the fuel transfer tube flange and the top of a control rod, and using underwater cameras to ensure that clearances were maintained during fuel movement activities. The inspectors determined that this finding was more than minor because if left uncorrected, the failure to follow procedures during refueling activities could lead to the unknown installation of other equipment and increase the potential of damaging reactor fuel and/or plant equipment; therefore become a more significant safety concern. The inspectors reviewed IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined that this type of finding was unable to be evaluated using this Appendix. As a result, the inspectors submitted the finding for management evaluation using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. NRC Management reviewed the details of this issue and concluded that this finding was of very low safety significance because the insertion of the plug, and the subsequent contact between the plug and the control rod, did not result in damage to irradiated fuel. The inspectors determined that this finding was cross-cutting in the Human Performance, Work Practices area because the licensee failed to ensure supervisory and management oversight of work activities, including contractors, was maintained such that nuclear safety was supporte |
| 05000282/FIN-2008005-04 | Operator Manipulates Incorrect Component due to Failure to Follow Procedures | 2008Q4 | One self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1 was identified on October 13, 2008, due to an operators failure to follow procedures during refueling activities. The failure to follow procedures resulted in a loss of seal injection flow to the 11 reactor coolant pump due to the manipulation of a Unit 1 seal injection valve rather than a Unit 2 seal injection valve. Corrective actions for this issue included communicating this event to all Operations personnel, resetting the operations departments event free clock and providing additional training of the use of human performance tools. The inspectors determined that this finding was more than minor because if left uncorrected, a continued failure to follow procedures could lead to the incorrect operation of additional plant equipment and become a more significant safety concern. The inspectors determined that this issue was of very low safety significance because the finding would not result in leakage that exceeded any TS limit and because the finding would not have affected other mitigation equipment. Specifically, the reactor coolant pumps were designed to be able to operate without seal injection flow for several hours as long as the component cooling water supply to the thermal barrier heat exchanger remained within allowable ranges. The inspectors concluded that this finding was cross-cutting in the Human Performance, Decision Making area because the operator failed to use the systematic process for implementing procedures when deciding which valve needed to be manipulate |
| 05000282/FIN-2008005-05 | Decrease in Reactor Power due to Failure to Follow Procedures | 2008Q4 | A finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion V, was self-revealed on November 6, 2008, due to instrumentation and controls technicians failing to follow procedures during calibration of the power range nuclear instruments. The failure to follow procedures resulted in the uncontrolled movement of the Unit 2 control rods and a six percent reduction in reactor power. Corrective actions for this issue included removing the technicians qualifications, conducting remedial training, performing a site-wide stand down to reinforce procedure use and adherence, and providing additional oversight of control room activities for several days. The inspectors determined that the finding was more than minor because it caused a plant transient and if left uncorrected, it would become a more significant safety concern that could result in additional plant transients, testing errors, and the failure to properly operate equipment. The inspectors determined that this finding was of very low safety significance because it did not contribute to both the likelihood of a reactor trip and that mitigating systems equipment would not be available. The inspectors concluded that this finding was cross-cutting in the Human Performance, Decision Making area because the technicians failed to use the systematic process for implementing procedures to ensure that nuclear safety was maintaine |
| 05000282/FIN-2008005-07 | Respirator Qualification Deficiency Results in Non-Compliance with 10 CFR Part 50, Appendix R | 2008Q4 | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix R, Section J, on December 30, 2008, due to the licensees failure to ensure that an alternate safe shutdown access path was provided with emergency lighting units that contained at least an 8-hour battery power supply. Corrective actions for this issue included ensuring that all personnel on-shift were respirator qualified so that alternate safe shutdown access pathways would not need to be used. The inspectors determined that this issue was more than minor because if left uncorrected, the failure to properly evaluate alternative safe shutdown access paths against regulatory requirements could become a more significant safety concern due to its potential impact on safely shutting down the plant following a fire. The inspectors determined that this finding was of very low safety significance due to its low exposure time and low degradation rating. The inspectors concluded that this finding was cross-cutting in the Human Performance, Decision Making area because the licensee failed to make this safety-significant/risk-significant decision using a systematic process that included a review of the safe shutdown analysis timeline and input from fire protection personne |
| 05000282/FIN-2008005-08 | Licensee-Identified Violation | 2008Q4 | 10 CFR 50.65(a)(4) requires in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities before performing maintenance. Contrary to the above, on September 22, 2008, licensee personnel failed to adequately manage the risk associated with the initiation of planned testing activities after emergent equipment failures. Specifically, at approximately 11:00 a.m., on September 22, 2008, Unit 2 Train A RHR failed and shortly thereafter, the breaker to the valve isolating auxiliary feedwater to SG 22 was opened resulting in an unplanned Orange path entry. This condition existed for approximately two hours and 13 minutes before licensee personnel recognized the condition and restored power to the AFW isolation valve. The licensee entered this issue into their corrective action program as CAP 1162470 |
| 05000282/FIN-2008007-01 | Failure to Perform a 10 CFR 50.59 Evaluation for Bulk Hydrogen Storage Facility | 2008Q4 | The inspectors identified a Severity Level IV NCV, having very low safety significance, of 10 CFR 50.59, Changes, Tests, and Experiments, for the licensees failure to perform a safety evaluation associated with installation of a bulk hydrogen storage facility. Specifically, the licensee had not evaluated the adverse affects on the Circulating Water System from a postulated hydrogen tank explosion in the bulk storage facility located directly above buried Circulating Water System return lines. The licensee stopped work on the installation of the bulk hydrogen facility and documented the NRC identified issues in the corrective action system. The inspectors concerns also prompted the licensee to identify above ground Cooling Water System pipe in the nearby Turbine Building, which had not been evaluated in the hydrogen blast analysis. The finding was more than minor because the inspectors could not reasonably determine that this change would not have ultimately required prior approval from the NRC. This finding was categorized as Severity Level IV because the underlying technical issue for the finding was determined to be of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situation. Specifically, the inspectors answered No to the Mitigating Systems screening questions in the Phase 1 Screening Worksheet because the licensee had not yet filled the bulk storage facility with hydrogen, so no possibility of explosion and damage to plant equipment existed. The cause of the finding is related to the cross-cutting element of Human Performance Decision Making, because the licensee failed to make conservative assumptions in decision making associated with the effects of a postulated hydrogen tank explosion (IMC 305, Section 06.07.c, Item H.1(b)). (Section 1R17.1.b |
| 05000282/FIN-2008008-01 | 11 TDAFWP Inoperable for a Time Period Which Significantly Exceeded Time Allowed by TS | 2008Q4 | A self-revealing apparent violation of Technical Specifications was associated with the licensees failure to adequately control the position of a valve that could isolate the 11 TDAFWPs discharge pressure switch. Because of the valve being closed, the 11 TDAFWP failed to run as required, subsequent to a reactor trip. The manifold isolation valve was determined to have been shut for 138 days, rendering the 11 TDAFWP inoperable for a time period that significantly exceeded the Technical Specification allowed outage time (72 hours) for the pump. This issue has been preliminarily determined to be of low to moderate safety significance (White) for Unit 1. This issue was entered into the licensees corrective action program (CAP 01146005). The licensee took prompt corrective actions to restore the mispositioned valve to its normal (open) position; perform valve lineups to verify correct equipment configurations for the remaining auxiliary feedwater pumps; and perform appropriate surveillance testing on the 11 TDAFWP to verify the components operable status. This finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it impacted the configuration control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of the systems that respond to initiating events to prevent undesirable consequences. The cause of this finding was related to the cross-cutting element of human performance for resources (H.2.(c)). (Section 4OA3.3 |
| 05000282/FIN-2008009-01 | Radioactive Material Shipment Package Radiation Levels Exceeded | 2009Q1 | 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 | Failure to Perform Formal Job Planning to Evaluate the Radiological Hazards | 2009Q1 | 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 |
| 05000282/FIN-2009002-01 | Failure to Protect Fire Protection Equipment from Effects of Extreme Cold Temperatures | 2009Q1 | The inspectors identified a finding of very low safety significance on January 13, 2009, due to the fire protection system pumps being unable to auto start, as designed, in response to a low fire header pressure condition. Corrective actions for this issue included unthawing the sensing line, verifying the screenhouse ventilation systems configuration, revising the normal screenhouse ventilation procedure to ensure that it provided guidance on shutting down the exhaust fans, and repairing several normal screenhouse ventilation system equipment deficiencies. This finding was more than minor because if left uncorrected, the failure to protect mitigating systems equipment from the effects of extreme cold temperatures could result in the system failing to function when needed. The inspectors determined that this finding was of very low safety significance because it was assigned a low fire degradation rating as specified in the Fire Protection Significance Determination Process. This finding was determined to be cross-cutting in the Human Performance, Resources area because the licensee failed to have a complete and accurate normal screenhouse ventilation procedure to ensure that operation of the system would not result in the freezing of mitigating systems equipment during extreme cold weather conditions (H.2(c)). No violations of NRC requirements occurred because the fire pumps could have been started manually if needed and because the normal screenhouse ventilation system was nonsafety-related. (Section 1R01.1 |
| 05000282/FIN-2009002-02 | Failure to Follow Procedures During Performance of Operability Evaluations | 2009Q1 | The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement Procedure FP-OP OL-01, Operability Determination, to assess the capability of the 122 Control Room Chilled Water Pump to meet its mission time following the discovery of increased pump vibrations. Corrective actions for this issue included revising the operability recommendation and repairing the degraded pump. This finding was more than minor because, if left uncorrected, failure to adequately implement the operability procedure could result in safety-related components been incorrectly declared operable rather than inoperable or operable, but non-conforming (a more significant safety concern). This finding was of very low safety significance because the finding did not represent an actual loss of safety function of a single train for longer than its Technical Specification allowed outage time. The inspectors concluded that this finding was cross-cutting in the Human Performance, Decision Making area because the licensee failed to validate the underlying assumptions made when determining the continued operability of a safety-related component (H.1(b)). (Section 1R15.1 |
| 05000282/FIN-2009002-03 | Failure to Follow Procedure During D5 Post-Maintenance Testing | 2009Q1 | The inspectors identified a finding of very low safety significance on February 25, 2009, due to operations and maintenance personnel failing to identify a turbocharger coolant vent line fretting condition during a D5 emergency diesel generator post-maintenance test or during previous D5 operations. The lack of identification resulted in D5 operating with degraded conditions prior to the fretting issue being evaluated in the corrective action program. Corrective actions for this issue included performing an ultrasonic examination of the fretted area in support of an evaluation to determine whether the pipe needed to be replaced prior to declaring the diesel generator operable. The licensee also documented the untimely identification of the issue within its corrective action program. This finding was more than minor because if left uncorrected, the failure to identify, evaluate, and correct equipment issues could result in returning safety-related equipment to service with deficiencies that impact the ability of the equipment to perform its safety function (a more significant safety concern). The inspectors determined that the finding was of very low safety significance because it was not associated with an actual loss of safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors considered the finding to be cross-cutting in the Problem Identification and Resolution, Corrective Action Program area because operations and maintenance personnel failed to identify this issue in a timely manner commensurate with its safety significance (P.1(a)). No violations of NRC requirements occurred because D5 was not operable at the time this issue was identified and corrective actions were taken before it became operable. |
| 05000282/FIN-2009002-04 | Failure to Adhere to Licensed Power Level Specified in Operating License | 2009Q1 | A self-revealed finding and Non-Cited Violation of Prairie Island Nuclear Generating Plant Operating License DPR-42, Section C.1, was identified on January 2, 2009, due to the failure to maintain Unit 1 reactor power below the thermal power limitations stated in the facility operating license. Corrective actions for this issue included revising all associated operating procedures to ensure that operations personnel take action to lower reactor power if power levels exceed the licensed thermal power limitations. The inspectors determined that this issue was more than minor because if left uncorrected the operation of the reactor beyond the limits specified in the operating license could become a more significant safety concern. The inspectors determined that this issue was of very low safety significance because the finding was only associated with the fuel aspect of the Barrier Integrity Cornerstone and no core thermal limits were violated. The inspectors determined that this finding was cross-cutting in the Human Performance, Resources area because the licensee failed to have complete, accurate and up-to-date procedures regarding the maintenance of licensed thermal power levels (H.2(c)). |
| 05000282/FIN-2009002-05 | Failure to Follow Procedure Use and Adherence Procedure Following Receipt of Abnormal Operating Procedure Entry Condition | 2009Q1 | The inspectors identified a finding of very low safety significance and a Non Cited Violation of Technical Specification 5.4.1 due to operations personnel failing to implement abnormal operating procedures following an unexpected control rod insertion on November 6, 2008. Corrective actions for this issue included revising licensed operator training and providing guidance to operations personnel on the need to enter abnormal operating procedures following the receipt of an entry condition. The inspectors determined that this finding was more than minor because the failure to enter abnormal operating procedures to respond to unexpected conditions could result in incorrect actions being taken following a plant event (a more significant safety issue). The inspectors concluded that this issue was of very low safety significance because the finding was not a loss of coolant accident initiator, was not an external events initiator, and would not have resulted in both the likelihood of a reactor trip and that mitigating systems equipment would not have been available. The inspectors determined that this finding was cross-cutting in the Human Performance, Work Practices area because the licensee had not effectively communicated expectations regarding procedural compliance following the receipt of an abnormal operating procedure entry condition (H.4(b)). |
| 05000282/FIN-2009002-06 | Licensee-Identified Violation | 2009Q1 | 10 CFR Part 50, Appendix B, Criterion V, requires in part, that activities affecting quality shall be accomplished in accordance with procedures appropriate to the circumstance. Contrary to the above, on February 12, 2009, licensee personnel failed to perform surveillance testing on the 12 Containment Spray Pump in accordance with the surveillance procedure. Specifically, operations personnel failed to adhere to procedural requirements regarding a 30 minute full flow time restriction for the 12 Containment Spray Pump. In addition, operations personnel did not obtain vibration readings at the specified reference points. These procedure compliance failures resulted in the surveillance exceeding the 30 minute restriction by approximately 1.5 minutes. Additionally, horizontal and axial vibration readings were taken in an alternate location due to accessibility issues resulting from a scaffold. Corrective actions for this issue included a procedure change and an evaluation of the vibration data. The licensee entered this issue into the corrective action program as CAP 1169248 |
| 05000282/FIN-2009003-01 | Potential Turbine Building Flooding Issue | 2009Q2 | One unresolved item was identified due to the potential that internal flooding following a random pipe break or high energy line break (HELB) event in the turbine building could result in the loss of safety-related mitigating systems equipment. As part of the ongoing review of the potential interaction between high energy lines and the component cooling water system (discussed in NRC IR 05000282/2008005; 05000306/2008005 and 05000282/2009010; 05000306/2009010), a potential internal flooding issue was identified. Specifically, the licensee postulated that a turbine building HELB could result in the subsequent failure of cooling water piping and actuation of the fire protection sprinklers such that an unlimited supply of water could be introduced into the turbine building. This unlimited supply of water could potentially result in an internal flooding event that impacted the availability of redundant safety-related equipment that was required to respond to an event. The inspectors discussed the internal flooding licensing basis with regulatory assurance and engineering personnel. The licensing basis stated that a rupture of a high energy pipe cannot directly or indirectly result in a loss of redundant safety equipment required to mitigate the event. The licensing basis also stated that the potential for flooding safety-related equipment due to a HELB event must be evaluated. Lastly, NRC design requirements state that failures of non-safety related systems must not result in the complete failure of safety-related equipment. The licensee conducted a review and determined that flooding effects due to a HELB or random pipe break had not been analyzed. As a result, it was not clear whether a postulated HELB or random pipe break event could result in internal flooding of the turbine building that would impact the availability of safety-related/mitigating systems equipment. The licensee was conducting several evaluations at the conclusion of the inspection period. As discussed in Section 1R15 of this inspection report, the licensee performed an operability evaluation of this potential flooding concern and determined that compensatory measures were needed to ensure that the water resulting from the potential internal flooding event would not accumulate in the turbine building and result in the unavailability of mitigating systems equipment. However, the potential for safety-related equipment to be impacted by internal flooding following a HELB or a random pipe break was considered unresolved pending a review of the licensees analyses (URI 05000282/2009003-01; 05000306/2009003-01) |
| 05000282/FIN-2009003-02 | Failure to Positively Control Compensatory Measures | 2009Q2 | The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V on April 28, 2009, for failure to have adequate procedures to control compensatory actions for degraded/nonconforming conditions. Specifically the failure to implement positive controls for the Unit 2 roll-up door as a compensatory measure for an operability determination invalidated the determination. The door was discovered less than the 18-open requirement which supported the flooding evaluation. Corrective actions for this issue included opening the Unit 2 turbine building roll-up door to greater than 18 inches open, implementing positive configuration controls for the compensatory measures, and revising the operability determination procedure to require the implementation of positive controls. The inspectors determined that this finding was more than minor because if left uncorrected the failure to properly control compensatory measures could result in rendering equipment inoperable (a more significant safety concern). This finding was of very low safety significance because it was not a design or qualification deficiency, did not result in a loss of system safety function or the loss of a single train for greater than the Technical Specification allowed outage time, and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event since the roll-up door was 14 inches open and would have provided some mitigation following an internal flooding event. The inspectors determined that this issue was cross-cutting in the Human Performance, Resources area because the licensee failed to ensure that the operability determination procedure was adequate in regards to the control of compensatory measures (H.2(c)). (Section 1R15.1b(1) |
| 05000282/FIN-2009003-03 | Failure to Maintain Control of Unit 1 Containment Personnel Airlock Configuration | 2009Q2 | A self-revealed finding of very low safety significance and an Non-Cited Violation of Technical Specification 5.4.1 were identified on March 25, 2009, due to the failure of licensed operators to maintain control of the Unit 1 containment personnel airlock outer door. This resulted in the Unit 1 containment personnel airlock being unknowingly inoperable for approximately 45 minutes. Corrective actions for the issue included returning the Unit 1 containment airlock outer door to an operable status, developing a case study for inclusion during licensed operator training, and developing a procedure on operating the containment airlock doors. This finding was determined to be more than minor because if left uncorrected the failure to fully understand and control the configuration of plant equipment could become a more significant safety concern. The inspectors determined that this finding was of very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool; the finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; the finding did not represent an actual open pathway in the physical integrity of reactor containment due to the inner airlock door being fully closed; and the finding did not involve an actual reduction in function of the hydrogen igniters in the reactor containment. The inspectors concluded that this finding was cross-cutting in the Human Performance, Work Practices area because licensee personnel failed to follow procedures regarding the requirement to maintain an awareness of the configuration of plant equipment at all times (H.4(b)). (Section 1R15.1b(2) |
| 05000282/FIN-2009003-04 | Failure to Control Maintenance Activities to Ensure Plant Equipment is Not Unnecessarily Challenged | 2009Q2 | A self-revealed finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V were identified on March 19, 2009, due to the failure to have adequate procedures to control maintenance activities to ensure that plant equipment was not unnecessarily challenged. Specifically, the failure to adequately control maintenance on the 12 diesel-driven cooling water pump resulted in the unplanned automatic start of the 121 motor-driven cooling water pump during post-maintenance testing activities. Corrective actions for this issue included adding instructions to the post-maintenance testing procedure to ensure that it properly referenced the procedure used to realign the 121 motor-driven cooling water pump. The licensee planned to complete a review of safety-related preventive maintenance procedures to ensure that proper procedure referencing and branching was utilized. Lastly, the licensee will add additional staff to assist with the procedure upgrade program and the coordination of preventive maintenance activities. The inspectors determined that this finding was more than minor because if left uncorrected the failure to properly control maintenance activities could become a more significant safety concern. In addition, the inspectors determined that the identification of this issue in conjunction with several other procedure upgrade project issues is reflective of a significant programmatic deficiency in coordination of maintenance and operations procedures. This finding was determined to be of very low safety significance because it was not a design deficiency, did not result in a loss of system safety function, was not an actual loss of safety function for greater than the Technical Specification allowed outage time, and did not screen as a potentially significant seismic, flooding, or severe weather issue. The inspectors determined that this finding was cross-cutting in the Human Performance, Resources area because the licensee did not have complete, 3 Enclosure accurate and up to date procedures regarding testing of the 12 diesel-driven cooling water pump and realignment of the 121 motor-driven cooling water pump (H.2(c)). (Section 1R19.1 |
| 05000282/FIN-2009003-05 | 23 Inverter Rendered Inoperable During Training Activities | 2009Q2 | A self-revealed finding of very low safety significance and a Non-Cited Violation of Technical Specification 5.4.1 were identified on February 27, 2009, due to operations personnel failing to adequately implement procedures which control safety-related equipment. Specifically operations personnel, unintentionally, rendered the 23 instrument inverter inoperable during the performance of on-the-job training activities. Corrective actions for this issue included returning the 23 instrument inverter to an operable status, providing additional training on the use of human error prevention techniques to the apprentice plant attendant, and providing additional training on the instrument inverters. The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was determined to be of very low safety significance because it was not a design deficiency, did not result in a loss of system safety function, was not an actual loss of safety function of one train of equipment for greater than the Technical Specification allowed outage time, and did not screen as a potentially significant seismic, flooding, or severe weather issue. The inspectors concluded that this finding was cross-cutting in the Human Performance, Work Practices area because human error prevention techniques were not used to ensure that an on-the-job training activity was performed safely (H.4(a)). (Section 4OA3.2 |
| 05000282/FIN-2009003-06 | 22 Battery Charger Rendered Inoperable During Maintenance on 22 Inverter | 2009Q2 | A self-revealed finding of very low safety significance and a Non-Cited Violation of Technical Specification 5.4.1 were identified on April 26, 2009, due to maintenance personnel failing to implement procedures which control safety-related equipment. Specifically maintenance personnel did not comply with work order instructions or procedures, rendering the 22 battery charger inoperable during the performance of maintenance on the 22 instrument inverter. Corrective actions for this issue included issuing a stop work order and remediating the maintenance workers on human performance tool use. The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was determined to be of very low safety significance because it was not a design deficiency, did not result in a loss of system safety function, was not an actual loss of safety function of one train of equipment for greater than the Technical Specification allowed outage time, and did not screen as a potentially significant seismic, flooding, or severe weather issue. The inspectors concluded that this finding was cross-cutting in the Human Performance, Work Practices area because maintenance personnel did not follow procedures during this maintenance activity (H.4(b)). (Section 4OA3.3 |
| 05000282/FIN-2009003-07 | 122 Air Compressor Rendered Non-Functional During Clearance Order Activities | 2009Q2 | A self-revealed finding of very low safety significance was identified on April 30, 2009, due to operations personnel failing to implement procedures which 4 Enclosure control plant equipment. Specifically operations personnel operated the incorrect component, rendering the 122 air compressor non-functional during the performance of independent verification activities. Corrective actions for this issue included restoring the 122 air compressor to a functional status and briefing operations personnel on the details/lessons learned from this event. The inspectors determined that this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was determined to be of very low safety significance because it was not a design deficiency, did not result in a loss of system safety function, was not an actual loss of safety function for one or more non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as a potentially significant seismic, flooding, or severe weather issue. The inspectors concluded that this finding was cross-cutting in the Human Performance, Decision Making area because the operator failed to use conservative assumptions when making the decision regarding the need to operate breaker 121E-6, 1A2-B4 (H.1(b)). No violation of NRC requirements was identified because the air compressor was non-safety related. (Section 4OA3.4 |
| 05000282/FIN-2009003-08 | Failure to Ensure Turbine Valve Testing Procedure was Adequate | 2009Q2 | A self-revealed finding of very low safety significance was identified on May 9, 2009, due to operations personnel failing to ensure that procedures used to test the Unit 2 turbine stop valves provided adequate guidance regarding the valve position limiter setting. The failure to ensure that adequate guidance was provided prior to performing the turbine stop valve test resulted in a reactor coolant system transient and a seven percent reduction in reactor power. Corrective actions for this issue included revising the test procedure to ensure that guidance regarding the valve position limiter setting was adequate, providing additional training on the digital electro-hydraulic control system to operations personnel, and re-enforcing the human performance fundamentals. The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone. In addition, the finding affected the cornerstone objective of limiting the likelihood of events that upset plant stability during power operations. The inspectors concluded that this finding was of very low safety significance because it did not result in exceeding the Technical Specifications limit on reactor coolant system leakage, did not result in a total loss of safety function of a mitigating system, did not contribute to both the likelihood of a reactor trip and that mitigating systems equipment would not be available, and it did not increase the likelihood of a fire or flood. The inspectors determined that this finding was cross-cutting in the Human Performance, Decision Making area because operations personnel failed to use conservative assumptions in deciding how the valve position limiter operated. In addition, operations personnel failed to demonstrate that their proposed actions regarding the valve position limiter setting was safe (by reviewing design basis or training documents and/or requesting assistance from additional personnel) prior to performing the test (H.1(b)). No violation of NRC requirements was identified because the turbine stop valves are non-safety related. (Section 4OA3.5 |
| 05000282/FIN-2009003-09 | Licensee-Identified Violation | 2009Q2 | Title 10 CFR Part 50, Appendix B, Criterion V, requires, in part, that activities affecting quality be prescribed by instructions, procedures, and drawings appropriate to the circumstance. Contrary to the above, on May 22, 2009, the licensee determined that work instructions used to install the 21 CC pump bearings were not appropriate to the circumstance. The finding was determined to be of very low safety significance, because the 21 CC pump was not inoperable for greater than the TS allowed time. The licensee documented this issue in CAP 1179272. Corrective actions for this issue included replacing the 21 CC pump inboard bearing and revising the bearing installation instructions |
| 05000282/FIN-2009003-10 | Licensee-Identified Violation | 2009Q2 | Title 10 CFR Part 50, Appendix B, Criterion V, requires, in part, that activities affecting quality be prescribed by instructions, procedures, and drawings appropriate to the circumstance. Contrary to the above, on March 29 and April 18, 2009, the licensee failed to have procedures appropriate to the circumstance to ensure that periodic maintenance on the D5 and D6 EDG ventilation dampers was performed. The finding was determined to be of very low safety significance, because the dampers did not cause the EDG to be inoperable for greater than the TS allowed time. This issue was documented in 49 Enclosure CAPs 1175563 and 1178658. Corrective actions for this issue included returning the ventilation dampers to service and implementing preventive maintenance procedures to address damper alignment issues |