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05000247/FIN-2000006-01Apparent failure to augment the ERO in a timely manner - failure to meet planning standard 10 CFR 50.47(b)(2), Timely Augmentation of ERO2000Q2In response to the Alert of February 15, 2000, there was a failure to augment the ERO within 60 minutes of the declaration of the Alert contrary to the Indian Point 2 (IP2) E-Plan Figure 5.2-1. Followup inspection identified several program structure deficiencies or design problems that contributed to an apparent failure to meet NRC emergency planning standard 10 CFR 50.47(b)(2). This finding was an apparent violation of low to moderate safety significance because of the failure to meet an NRC emergency planning standard.
05000247/FIN-2000006-02Apparent failure to complete accountability in a timely manner - failure to meet planning standard 10 CFR 50.47(b)(lO), Protection of Radiation Workers2000Q2In response to the Alert of February 15,2000, there was a failure to account for onsite radiation workers within 30 minutes of initiation contrary to the IP2 E-Plan section 6.4.1 .d and E-Plan implementing procedure 1027 section 5.1.2.f. Followup inspection further identified several program deficiencies or design problems indicating an apparent failure to meet NRC emergency planning standard I0 CFR 50.47(b)(10) concerning accountability. This finding was an apparent violation of low to moderate safety significance because of the apparent failure to meet an NRC emergency planning standard.
05000247/FIN-2000006-03Improper dissemination of information to public and local official - failure to meet planning standard I0 CFR 50.47 (b) (7), public inform at i o n2000Q2In response to the Alert of February 15, 2000, there was a failure to properly disseminate information about the Alert conditions. As a result there was confusion in the public domain about whether there was a radiation release and its magnitude, and one local official was not notified in accordance with a pre-arranged agreement. This was contrary to the IP2 E-Plan section 5.2.3, which requires consistent information be disseminated. Followup inspection identified a number of program structure or design problems indicating an apparent failure to meet NRC emergency planning standard 10 CFR 50.47(b)(7) concerning dissemination of information. This finding was an apparent violation of low to moderate safety significance because of the failure to meet an NRC emergency planning standard.
05000247/FIN-2000007-06N/A2000Q2The control room operators did not enter significant plant items, such as event declaration and implementation of the emergency plan, in the control room logs, as required by Con Edison procedures. This procedure violation was a problem that was also noted for the August 31, 1999, loss of bus event. The failure to enter significant items into the control room logs was determined to be a non-cited violation. Although this issue does not affect any of the seven cornerstones (Attachment 1), it was considered important because prior corrective actions were not effective (Section 4OA2.3).
05000247/FIN-2000007-11N/A2000Q2In the operations and engineering support areas, corrective actions to resolve known problems were untimely or incomplete. While the problems were of very low risk significance, some of these procedure and equipment problems caused unnecessary challenges to the operators and delays in achieving cold shutdown after the event. These problems included difficult procedural guidance for aligning pressurizer spray flow, non-functional steam generator leak monitoring (N-16) recorder, high pressure steam dump system deficiencies, and the lack of gas turbine Nos. 2 and 3 remote start capability (Section 4OA5).
05000247/FIN-2005005-01Emergency Diesel Generator Building Flooding2005Q4On November 29, 2005, the inspectors reviewed the internal flood protection measures for the EDG building. All three 480 VAC EDGs are located in this building in a common area separated by installed fire barrier walls. The EDGs are approximately five feet above the concrete floor of the building and access to the EDGs is afforded by metal grate flooring. The major sources of potential flooding for the space are fire protection piping and the essential service water (ESW) system piping in the building. The building is designed such that water drains toward five shallow sumps that are connected to a common 12-inch-diameter drain line that discharges to the site drainage system. Each sump has two 3-inch-diameter openings that have backwater ball check valves to prevent back-leakage into the EDG building. Inspectors observed between 30 and 50 oil absorbent pads on the concrete floor underneath the EDGs. These pads were not fixed to the floor by any means, and would be free to migrate to the building sumps along with water during a flooding event. The pads were of sufficient size to effectively block the 3-inch holes in each of the building sumps as water level rises in the building. The IPEEE credits the building drains being sufficiently sized to prevent significant accumulation of water due to a break of fire protection piping in the room. This assumes that the function of these drains is not impeded by foreign material blockage. In addition, both the IPEEE and the IP2 Probabilistic Safety Assessment (PSA) state that a break of an ESW line is bounded by the fact that the EDGs are cooled by ESW and would be the only equipment negatively impacted by the flooding, and that this occurrence is analyzed by the total loss of service water event. There are inconsistencies between the IPEEE, dated 1995, and the PSA, which was completed in the 1998 time frame. The PSA does not account for the fire protection header as being a potential source of flooding for the EDG building, whereas the IPEEE does. Both analyses credit open ventilation louvers along the building north wall at grade level to drain water if the buildings installed drain capacity is insufficient. However, during the winter months these louvers are maintained shut. In addition, the IPEEE mentions an EDG building flood alarm in the control room and specific isolation procedures in the event of flooding. Neither the alarm, nor the specific isolation procedures, currently exist. Finally, the inspectors identified that 480 VAC normal feeder breaker control power exists in each EDG control cabinet. Flood water that reaches the bottom of the EDG control cabinets due to insufficient building drain capacity, and can not be relieved through closed building doors and closed ventilation louvers, could potentially render all three EDGs unavailable and trip the normal feeder breakers to all 480 VAC vital AC buses. In response to the , Entergy removed the oil absorbent pads from the EDG building and entered the issue into the corrective action program (CR-IP2-05-4868). This issue will be treated as a URI pending additional licensee evaluation and inspector review of the potential impact of flooding in the EDG building on the normal and emergency vital AC power sources
05000247/FIN-2005005-07Failure to Make a 10 CFR 50.72(B)(3)(XIII) Notification2005Q4A Severity Level IV violation of 10 CFR 50.72(b)(3)(xiii) was identified for not formally reporting a siren system problem that occurred on August 5, 2005. The inspectors noted that the duration of the siren system problem was short, the NRC was informally notified, the process for back-up route alerting was available, and the capability to actuate the sirens via a manual siren initiation method was not lost. Subsequent to this event, Entergy implemented corrective actions to formalize the manual siren system actuation method. Notwithstanding these circumstances, a formal notification to the NRC was required, because the normal processes for actuation of the sirens were not available and Entergy did not have formal procedures for, and had limited experience with, the manual siren initiation method. This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. Because this finding is of very low safety significance and has been entered into the corrective action program, it is being treated as an NCV.
05000247/FIN-2005006-02Failure to Adequately Evaluate and Correct Nitrogen Gas Migration and Accumulation in Portions of the Safety Injection System2005Q2An apparent violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Action) and station procedures were identified associated with the failure to evaluate and correct a condition adverse to quality. Specifically, the condition adverse to quality involved the leakage of water from the No. 24 safety injection accumulator past several closed valves, allowing water containing absorbed nitrogen to reach other portions of the safety injection emergency core cooling system (including the common suction supply piping for the safety injection pumps and the 23 safety injection pump casing). As the water moved from a higher to lower system pressure, the nitrogen gas was released from the water, thereby challenging the performance of the safety injection pumps. In addition, Entergys initial evaluation of this condition did not appropriately consider available industry operating experience relative to gas migration into emergency core cooling system piping. This issue is greater than minor because it is associated with the Equipment Performance attribute of the Mitigation Systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The Significance Determination Process (SDP) Phase 1, Phase 2, and Phase 3 were used to determine that this issue represented a finding with preliminarily low to moderate safety significance. The analysis used the NRCs best functionality estimates for the three safety injection pumps over a 17-day period when it was judged that adverse gas accumulation conditions existed. Specifically, the 23 safety injection pump was not functional due to the pump casing being filled with gas. The team concluded that the 21 and 22 pumps, given the accumulated gas in the pump suction piping, would not have functioned 75% of the time (assigned a 75% failure probability) for high flowrate and low discharge pressure conditions in response to a medium break loss of coolant accident; and 25% of the time for low flowrate and high discharge pressure conditions in response to other initiating events. The Phase 1 screening identified that a Phase 2 analysis was needed because the 23 safety injection pump train was not functional for longer than the technical specification allowed outage time of 72 hours. Given the uncertainty in the Phase 2 analysis, a Phase 3 analysis was necessary to improve the accuracy of the result. The Phase 3 analysis for internal and external initiating events, using the above assumptions and licensee risk information, identified an increase in core damage frequency of approximately 1 in 900,000 years of operation (low E-6 per year range); and an increase in large early release frequency of approximately 1 in 3,000,000 years of operation (low E-7 per year range).
05000247/FIN-2006005-01Inadequate Containment Closure Equipment2006Q4The inspectors identified a Severity Level IV NCV of 10 CFR 50.59, Changes, Tests and Experiments, for failure to obtain a license amendment pursuant to 10 CFR 50.90 prior implementing a change to alter the requirements of a shutdown fission product barrier. The inspectors reviewed Safety Evaluation 04-0732-MD-00-RE R1, Installation of a Temporary Roll-up Door on the Containment Equipment Hatch, to determine if the conclusion that a licensee amendment was not required was correct. Entergy concluded that the roll-up door was equivalent to the closure plate and, therefore, adequate to close containment as required by the action statement. The inspectors found that the door was not designed to be air-tight; therefore, any radioactive release inside containment would bypass the roll-up door. The inspectors concluded that the roll-up door did not meet the design or licensing basis of the closure plate as described in the Updated Final Safety Analysis Report (UFSAR) and previously approved license amendments. Consequently, Entergy incorrectly concluded that a license amendment pursuant to 50.90 was not required prior to implementing the change. Entergy entered the issue into their corrective action program to evaluate and correct. The inspectors determined that Entergy changed the requirements for the shutdown fission product barrier (containment) prior to receiving NRC approval. As a result, traditional enforcement was used to evaluate the issue because the deficiency affected the NRC\'s ability to perform its regulatory function. The severity level of the violation was determined to be Severity Level IV in accordance with example D.5 of Supplement 1 of the NRC Enforcement Policy. Additionally, the issue was determined to be of very low safety significance (Green) based on the low decay heat levels at the time the roll-up door was credited in accordance with the significance determination process described in Inspection Manual Chapter (IMC) 0609 Appendix H, Containment Integrity.
05000247/FIN-2006005-03Assess Reliability / Unavailability of the Gas Turbine System and Impacts on Functionality2006Q4The inspectors identified the URI during a routine Maintenance Rule inspection on the gas turbine system. Gas turbines 1 and 3 (GT-1 and GT-3) are credited in Entergy's analysis to cope with station blackout and Appendix R fire scenarios to ensure safe shutdown of the reactor. The system is classified as risksignificant in accordance with Entergys Maintenance Rule program. This system has been in a category (a)(1) monitoring status since the inception of the Maintenance Rule in 1996. An (a)(1) action plan was established to improve overall system performance. However, Entergy may not have provided justifiable (a)(1) goals for maintenance preventable functional failures (MPFFs) and Entergy may not have appropriately classified repeat maintenance preventable functional failures (RMPFFs). Specific to reliability, the goal was set as less than or equal to five MPFFs and no RMPFFs in a 24 month rolling cycle. The number of allowable MPFFs was calculated under the assumption that there would be, on average, 82 start demands during the 24 month cycle. The inspectors reviewed the operating history over the last three years and determined that the number of start demands averaged 38 during the 24 month cycle. The inspectors need more information to evaluate Entergy's goals for MPFFs to determine their adequacy. Additional information is required to evaluate Entergy's implementation of the Maintenance Rule as it pertains to the gas turbine system. Actual unavailability and reliability information is needed to evaluate the gas turbine system performance and to assess whether performance of the system is bounded by the Station Blackout / Appendix R commitments, and assumptions in the design basis. This issue will be treated as a URI pending additional licensee input and inspector evaluation of gas turbine system performance.
05000247/FIN-2006201-01Security2006Q3Security
05000247/FIN-2007002-04Containment Sump Modification Missing Weld Data2007Q1During the Spring 2006 outage, Entergy completed a partial modification to install upgraded sump strainers in response to Generic Safety Issue 191, which was associated with debris-induced clogging of pressurized water reactor sumps. Prior to restart from the Spring 2006 outage, Entergy identified several instances where weld data sheets were missing for the sump modification. Entergy formed a reconstitution engineering team to recover the missing data sheets or disposition the missing data through engineering evaluation. This effort was completed and Entergy determined that the sump was operable prior to restart. On January 22, 2007, Entergy learned that additional weld records for the sump strainer installation were potentially missing, and initiated an independent review into eight of the 63 completed work packages associated with the strainer modification. The review identified additional missing weld records which were lost, misplaced, or discarded, but which had not been identified or evaluated during the previous reconstitution effort. Entergy initiated CR IP2-2007-00699 on February 8, 2007, to document the results of the independent review and initiate corrective actions. Entergy completed an engineering review of the newly identified missing information and concluded that the sumps remained operable. Additional actions planned by Entergy include a review of the remaining containment sump work packages and a visual inspection of safety related welds with missing weld data. This issue is unresolved pending the completion of Entergys review and NRCs subsequent evaluation.
05000247/FIN-2007005-04Impact of incorrect jacket water and lube oil control elements on EDG performance2007Q4The inspectors identified an unresolved item (URI) concerning incorrect temperature control valve elements installed on the 21 and 22 EDG jacket water and lube oil systems. The original EDG design required 170F temperature control elements in the jacket water system three-way temperature control valve and 180F temperature control elements in the lube oil system three-way temperature control valve to maintain EDG jacket water and lube-oil systems within the required temperature bands. The respective three-way valves control the inlet temperatures to the jacket water cooler and the lube oil cooler by sending or bypassing jacket water or lube oil to each systems cooler. The EDG jacket water and lube oil coolers are cooled by the service water system. In 1989, the EDG design was modified by DER-1691, Engineering Evaluation of Increasing Overloading Capacity on the Emergency Diesel Generators, which specified, in part, that 180F temperature control elements be installed in the jacket water system and 195F temperature control elements be installed in the lube oil system to account for an EDG power up-rate and a 10F increase in design basis ultimate heat sink temperature. The 180F and the 195F control elements assured EDG operability during a 30 minute period at a rating of 2300 kilowatt (kw) and a higher design basis service water temperature of 95F. The original 170F and 180F control elements were designed for a maximum short-term loading of 1950kw and a maximum service water temperature of 85F. Following completion of the EDG upgrades, on October 26, 2002, the original 170F jacket water control elements and 180F lube oil control elements were incorrectly installed on the 22 EDG under WO 02-33401. The incorrect jacket water and lube oil control elements were also installed on the 21 EDG on February 27, 2003, under WO-01-22824 Based on the available information, the inspectors were unable to verify the impact on EDG performance with the incorrect control elements installed due to a number of other upgrades that were also made to the EDGs under DER-1691. Some of these included upgraded heat exchangers on the jacket water and lube oil systems, an upgraded exhaust manifold, and upgrades to the EDG ventilation system. Entergy has contracted with a vendor to perform an analysis to determine the actual impact on past operability of having the original 170F control elements in the jacket water system and the 180F control elements in the lube-oil system based on actual service water temperatures that have been observed. The EDGs are currently operable because service water temperatures are substantially below the original service water design temperature of 75F. In addition, Entergy initiated actions to install the correct control elements in the two affected EDGs prior to service water temperatures exceeding 75F in 2008. The impact of incorrect jacket water and lube-oil control elements on the 21 and 22 EDGs will be an unresolved item pending NRC review of Entergys analysis of past operability.
05000247/FIN-2007007-03Use of Motor Control Center Methodology for Periodic Verification of the Design Basis Capability of Safety- Related MOV2007Q1The team identified an unresolved item (URI) concerning the adequacy of the motor control center (MCC) testing methodology used for periodic verification of the design bases capability of safety-related MOVs. Entergy implemented MCC testing in 2004 as a method of implementing periodic verification in addition to the previously NRCreviewed method of taking stem thrust and torque measurements at the valve. The MCC method uses motor current, voltage, and winding resistance measured at the MCC to calculate motor torque of the valves motor operator. The calculated motor torque is then compared to motor torque target and limit values based on 1) packing loads, 2) thrust required to close the valve, 3) stall motor torque, and 4) valve or actuator structural limits. Entergy Report IP-RPT-04-00890, Technical Basis for Using MCC Technology for Periodic Verification Testing at IP 2 and IP 3, states that this methodology would be used initially on MOVs with generally low safety significance and high operating margin, but also states that the report applies to all safety related MOVs at IP 2 and IP 3. Since 2004, Entergy has used the MCC methodology for periodic verification on nine safety-related MOVs: three high risk, three medium risk, and three low risk MOVs, where risk significance is defined as the combined effects of MOV risk of failure and safety significance. Based on the available information, the team was unable to verify that the MCC method had been appropriately validated. Specifically, there did not appear to be a justified correlation between the MCC methodology calculated motor torque and actual stem thrust and torque. It was also unclear whether the MCC methodology had adequate allowances to compensate for its uncertainties in establishing MOV design basis capability (such as uncertainties related to stem friction coefficient, load sensitive behavior, and actuator efficiency) since stem thrust and stem torque are not directly measured. MCC testing was performed in 2004 as a periodic verification test on MOV 747, the No. 21 RHR heat exchanger discharge valve, a high risk valve. The team identified that this test was invalid because this was not performed in accordance with IP-RPT-04-00890. Specifically, MOV 747 was tested using the Motor Torque Method of MCC testing which, according to IP-RPT-04-00890, is only valid for motors whose torque is between 2 and 60 foot-pounds. The motor on MOV 747 is an 80 foot-pound. motor and use of the Correlated Thrust/Torque Method was required. As a result, Entergy exceeded the sixyear periodic verification test interval for MOV 747 because the last at the valve valid performance verification test was performed in May 2000. Entergy has provided justification for the reasonable continued operability of the valve until its scheduled testing in 2008 based on successful in-service tests, stem lubrication and actuator preventive maintenance and inspection performed in 2006. The team reviewed Entergys basis for the operability of MOV 747 and determined that there was reasonable assurance of continued operability of the MOV. Entergy committed to follow the Joint Owners Group program for periodic verification of MOVs in their response to NRC Generic Letter 96-05. This periodic verification program established valve margin by measuring stem thrust and torque at the valve. Entergys response did not indicate that the MCC method would be used for MOV periodic verification. The MCC test method for periodic verification of MOVs is a departure from the NRCreviewed method, which is based on direct measurement of stem thrust and torque. The acceptability of the use of the MCC methodology for periodic verification of MOVs will be an unresolved item pending further NRC review. Included with this review will be a determination of whether the MOV performance testing conducted on MOV 747 constitutes a violation of NRC requirements.
05000247/FIN-2008002-01Failure of 21 Swp Due to Inadequate Maintenance Procedure2008Q1A self-revealing, non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide an adequate procedure for installing cable termination lugs on the 21 service water pump motor cables. As a result, Entergy maintenance personnel installed undersized terminal lugs for the 21 service water pump motor jumper cables on January 26, 2000, which resulted in a high resistance connection that degraded over time and eventually caused the cables to fail while the pump was in service on January 27, 2008. Entergy entered this issue into the corrective action program, replaced the jumper cables with insulated bus bars, tested the motor for damage, and changed Engineering Standard ENN-EE-S-008-IP, IPEC (Indian Point Energy Center) Electrical Cable Installation Standard, to ensure the use of correctly sized terminal lugs in the future. Entergy also plans to perform an extent-of-condition review that includes thermography and visual inspections of other safety related motor cable terminations. The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone; and, it affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to provide adequate procedural steps to ensure that the 21 service water pump was installed with appropriate electrical connectors. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At- Power Situations, and determined that it was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train for greater than its Technical Specification allowed outage time; and it did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events
05000247/FIN-2008003-01Failure to Follow Site Procurement Procedure for EDG Temperature Control Valve Elements2008Q2The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings because Entergy personnel did not implement the requirements of procedure SAO-270, Procurement Program, for the procurement of safety related temperature control valve (TCV) elements for the emergency diesel generators (EDGs). Specifically, Entergy did not perform a technical evaluation as required for the TCV elements which resulted in the purchase and installation of incorrect TCV elements on the 21 and 22 EDGs between 2002 and 2003. The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At- Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because the installation of incorrect TCV elements represented a design deficiency that was confirmed not to result in a loss of operability of the EDGs. Specifically, engineering analysis verified past EDG operability was maintained based on analysis that assumed the highest observed service water temperature over the past three years. Entergy entered this issue into the corrective action program and installed the correct TCV elements in 21 and 22 EDGs.
05000247/FIN-2008003-02Station BLACKOUT/APPENDIX-R Diesel Generator Post Modification Test Deficiencies2008Q2The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the requirements of EN-DC-117, Post Modification Testing and Special Instructions, to control revisions to the station blackout/Appendix R diesel generator (SBO/App-R DG) post modification test, or to review and approve the test results. Specifically, the SBO/App-R DG post modification test was not sufficient to demonstrate the SBO/App-R DG could perform its intended design functions. As a corrective measure, Entergy subsequently performed additional testing to demonstrate system operability. The inspectors determined the finding was more than minor because it was associated with the procedure quality 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 post modification test deficiencies represented reasonable doubt regarding the operability of the SBO/App-R DG. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events. The finding had a cross-cutting aspect in the area of human performance because Entergy\'s supervisory and management oversight of work activities was not adequate to ensure testing was properly performed. (H.4(c)
05000247/FIN-2008003-03Inadequate Operating Procedure for Station BLACKOUT/APPENDIX-R Diesel Generator2008Q2The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because the SBO/App-R DG operating procedure 2-SOP-27.6, Appendix-R Diesel Generator Operation, was not adequate. Specifically, the procedure could not be performed as written, and was not sufficient to ensure operators could start the SBO/App-R DG, and energize an electrical bus within the required time of one hour. Entergy subsequently revised the procedure to correct the most critical deficiencies, and pre-staged equipment to reduce the time needed to energize a bus. As an interim corrective measure, Entergy relied upon operator training for other deficiencies, pending final corrective actions. The finding was more than minor because it was associated with the procedure quality 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 procedure deficiencies resulted in a reasonable doubt whether the SBO/App-R DG could be started and aligned in a timely and correct manner, as required by design. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance (Green) because it was not a design or qualification deficiency; it did not represent a loss of system safety function of a single train; and it did not screen as potentially risk significant due to external events. The finding had a cross-cutting aspect in the area of human performance because Entergys procedure for the SBO/App-R DG was not adequate to assure nuclear safety in implementing necessary operator actions for a SBO. (H.2(c)
05000247/FIN-2008003-04Inadequate Seismic Design Control Associated with a Temporary Modification to Emergency Diesel Generator Service Water Return Piping2008Q2The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control because Entergy did not adequately analyze, document, or translate seismic considerations for temporary service water hoses installed on the 21 and 23 emergency diesel generator (EDG) heat exchangers during the March 2008 refueling outage. Entergy entered the issue into the corrective action program, evaluated past operability concerns, and added restraints to the temporary service water hoses. The inspectors determined that this finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of the EDG system during a Seismic Class I design basis event. This finding was evaluated using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs (Pressurized Water Reactors) and BWRs (Boiling Water Reactors). The finding was determined to be of very low safety significance (Green) because the finding did not degrade the equipment, instrumentation, training or procedures needed for any shutdown safety function. Entergy performed a subsequent operability evaluation which provided reasonable assurance that the EDGs would have performed the safety function during a design basis seismic event. The finding had a cross-cutting aspect in the area of human performance because Entergy personnel made non-conservative assumptions regarding the seismic adequacy of the temporary hose modification. Specifically, Entergy personnel did not perform an engineering analysis to validate their assumptions that the temporary service water hoses would not adversely impact the seismic qualification of the EDGs. (H.1(b)
05000247/FIN-2008003-05Failure to Follow Plant START-UP Procedure Regarding MBFP Turbine Runback Arm/Defeat Switch2008Q2A Green, self-revealing non-cited violation (NCV) of Technical Specification 5.4.1, Administrative Controls - Procedures, was identified, because Entergy did not implement the requirements of plant startup procedure 2-POP-1.3, Plant Startup from Zero To 45% Power. Specifically, operators performed a step out of sequence in the plant operating procedure that was not warranted by plant conditions, and resulted in a main turbine runback followed by a manual reactor trip initiated by control room operators. Entergy entered this issue into the corrective action program, initiated procedural enhancements, performed a post-trip evaluation, and a root cause evaluation. The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using the Phase 1 analysis of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At- Power Situations, and determined it to be of very low safety significance because it did not contribute to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The finding had a cross-cutting aspect in the area of human performance because Entergy staff utilized work practices that did not support effective human error prevention techniques by proceeding in the face of uncertainty and unexpected circumstances, when they prematurely positioned the arm/defeat switch contrary to plant procedures and conditions. (H.4(a)
05000247/FIN-2008003-06Failure to Follow Camera Controls Procedure Resulting in RFI Induced MBFP Runback and Subsequent Manual Reactor Trip2008Q2A Green, self-revealing finding was identified because Entergy did not implement procedural requirements to evaluate flash photography in the vicinity of sensitive control cabinets. Specifically, Entergy did not implement procedure EN-NS-214, Camera Controls for Access and Use, and evaluate the potential impact of flash photography on sensitive control circuitry. Radiofrequency interference (RFI) from the digital camera during flash photography resulted in a main boiler feed pump runback which required a subsequent manual reactor trip. Entergy entered the issue into the corrective action process, performed site-wide training regarding the potential impacts of RFI from digital cameras on digital plant equipment and reinforced expectations to site personnel regarding procedural compliance. The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and impacted the objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. 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 the likelihood that mitigation equipment or functions would not be available. The inspectors determined that this finding has a cross-cutting aspect in the area of human performance because Entergy did not effectively communicate expectations regarding procedural compliance and personnel did not follow the applicable procedures. (H.4(b)
05000247/FIN-2008003-07Failure to Maintain Quality Records for Containment Sump Modification2008Q2The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, because Entergy did not maintain sufficient records to furnish evidence that a safety-related containment sump modification was performed in accordance with the design documentation. Specifically, nine of 63 work orders completed during the 2R17 refueling outage for the modification were missing data or missing entirely due to being lost, misplaced, or contaminated during implementation of the project. Entergy entered the issue into the corrective action process, evaluated the operability impact of the missing data, and performed visual inspections of accessible safety-related welds during the 2R18 refueling outage. The inspectors determined that this finding was more than minor because it was associated with the design control 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. The inspectors evaluated this finding using the Phase 1 analysis in IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that this finding was of very low safety significance because the finding did not represent a design or qualification deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to external events initiating events. Entergy performed inspections during 2R18 and completed technical evaluations of missing data that provided reasonable assurance of sump operability. The finding had a cross-cutting aspect in the area of human performance because Entergy did not appropriately coordinate work activities to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination was necessary to assure plant and human performance. (H.3(b)
05000247/FIN-2008004-01Auxiliary Feedwater System Configuration Control Deficiencies2008Q3The inspectors identified a Green NCV of Technical Specification 5.4.1, Administrative Controls - Procedures, because Entergy did not implement the Auxiliary Feedwater (AFW) operating procedures required by Regulatory Guide 1.33 Appendix A. Specifically, the inspectors identified an AFW drain valve that was not in the required position and an AFW isolation valve that was in the correct position but was not locked as required. Entergy evaluated the as-found configuration of the valves and determined that the AFW system operability was not impacted. Entergy also performed system alignment verifications of AFW and other safety-related systems as part of an extent-ofcondition review. The inspectors determined the finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. This finding was similar to the more-than-minor example 3.c found in IMC 0612 Appendix E in that more than one valve was unlocked or out of its required position. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external events initiating events. Specifically, the inspectors determined that the as-found configuration of the identified components did not adversely impact system operability. The finding had a crosscutting aspect in the area of human performance because operators did not use adequate self and peer checking techniques when shutting an open drain valve or when attaching a locking device to an isolation valve. (H.4(a)) (Section 1R04)
05000247/FIN-2008004-02City Water Tank Below Required Level Due to Inadequate Design Change Implementation2008Q3The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, when Entergy did not implement on-line leak repair procedures to repair a steam leak on valve MS-2A. Specifically, Entergy performed multiple leak sealant injections on valve MS-2A without engineering controls described in station on-line leak repair procedures. Corrective actions planned included reviewing this issue with the planning and component engineering departments and determining if training on the online leak sealing procedures is warranted. The finding was more than minor because, if left uncorrected, inadequate control of leaksealant injections would become a more significant safety concern. The inspectors determined the significance of the finding using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function. Specifically, Entergys operability evaluation concluded that the sealant that was injected extruded back out of the leak path and likely did not reach the valves seat or hinge. The finding had a cross cutting aspect related to work control in the area of Human Performance. Entergy personnel did not appropriately plan work activities to conduct online leak repairs on a safety related component. Specifically, Entergy did not identify necessary engineering procedures to adequately perform leak seal repairs on MS-2A during the planning process. These procedures provide necessary limitations, contingencies, and abort criteria. (H.3.(a)) (Section 1R18)
05000247/FIN-2008004-03On-Line Leak Repairs Made Without Use of Proper Procedures2008Q3The inspectors identified a non-cited violation of Technical Specification 5.4.1, Procedures, because Entergy did not implement portions of an engineering change package for an alarm setpoint change following modification to the city water tank minimum required water volume calculation. As a result, city water tank level dropped below the minimum water level required by the Technical Requirements Manual. Corrective actions included updating plant procedures and training of personnel. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the Cornerstones objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the significance of the finding using a phase 1 analysis described in Inspection Manual Chapter 0609 Appendix F, Fire Protection Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the degradation rating was determined to be low. The finding had a cross-cutting aspect related to formally defining the authority and roles for decisions affecting nuclear safety in the area of Human Performance in that Entergy management did not ensure that roles and responsibilities were communicated clearly to a member of the engineering change team responsible for implementing Operations procedure changes. As a result, the proper procedure changes were not made to plant procedures and logs which ultimately led to unmitigated low levels in the city water tank. (H.1(a)) (Section 1R15)
05000247/FIN-2008005-01Licensee-Identified Violation2008Q4Technical Specification (TS) 3.6.6 requires, in part, a minimum of three trains of containment fan cooler units (FCU) to be operable in mode 1. Contrary to this requirement, the required number of FCUs was not operable in mode 1, assuming a single failure of one emergency diesel generator (EDG) to supply power to the 5A bus during a design basis accident, coincident with a loss of offsite power. Specifically, on March 28, 2008, Entergy identified an incorrect wiring configuration that would have prevented the automatic start of the 23 FCU. This condition, coincident with a single failure of an EDG during a design basis accident, coincident with a loss of offsite power, would have prevented the automatic start of at least three FCUs. This issue was corrected on March 28, 2008 and entered into Entergys corrective action program as IP2-2008-01482. The finding was more than minor because it was associated with the mitigating systems cornerstone and impacted the objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding was of very low safety significance (Green) based on IMC 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, because the finding did not represent an actual loss of safety function of the FCUs for greater than their allowed outage time.
05000247/FIN-2008012-01Inadequate Design Control of Internal Recirculation Pumps2008Q3The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of the internal recirculation pump minimum flow rates. Specifically, Entergy did not verify the adequacy of the pump minimum flow rates for sustained operation under low flow rate conditions or for strong-pump to weak-pump interactions which could result in dead-heading the weaker pump during parallel pump operation. Following identification of the issue, Entergy revised the Emergency Operating Procedures (EOP) to not start a second internal recirculation pump during conditions of high head recirculation, submitted a licensee event report (LER) for each generating unit, and entered the issue into the corrective action program. The finding was determined to be more than minor because it is associated with the design control attribute of the Mitigating Systems (MS) Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. On Unit 2, the team determined the finding was of very low safety significance because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. On Unit 3, the finding was determined to be of very low safety significance based on a Significance Determination Process (SDP) Phase 3 risk assessment. Also, the Unit 3 finding had a crosscutting aspect in the area of Problem Identification and Resolution because Entergy did not implement operating experience information through changes to station processes, procedures, and equipment. (IMC 0305 aspect P.2 (b)
05000247/FIN-2009002-01Failure to Identify Open Louvers in 11 Fire Pump House2009Q1The inspectors identified a finding of very low safety significance because Entergy personnel did not adequately implement procedure EN-LI-102, Corrective Action Process, and promptly identify a condition adverse to quality associated with open louvers in a fire protection pump room following pump testing on January 14, 2009. The open louvers resulted in freezing conditions in fire protection piping located in the room and cracked two six-inch header isolation valves on January 17, 2009. Entergy entered the issue into the corrective action program and performed a site-wide extent-of condition walkdown of louvers. The finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and it affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated using Phase1 of IMC 0609 Appendix F, Fire Protection Significance Determination Process. The inspectors determined the issue was of very low safety significance (Green) because the cracked valves were easily isolated and did not pass sufficient water to render the fire header non-functional (low degradation rating). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance related to work practices - human error prevention techniques. Specifically, Entergy personnel that routinely tour the 11 fire pump house did not question the abnormally cold room temperatures. (H.4(a) per IMC 0305
05000247/FIN-2009002-02Failure to Identify Damaged Components in EDG Ventilation Motor Control Center #22009Q1The inspectors identified a NCV of very low safety significance related to 10 CFR50, Appendix B, Criterion XVI, Corrective Actions, because Entergy did not promptly identify and correct an adverse condition related to an electrical fault. Specifically, personnel did not identify a safety-related cubicle had experienced an electrical fault prior to replacement of upstream fuses and restoration of power to the damaged cubicle. Entergy entered the issue into the corrective action program as IP2-2009-00342 andIP2-2009-00483, trained all operations personnel on the requirements to replace fuses and re-energize electrical equipment, and plans to revise the operations procedure for operating electrical equipment. This issue was more than minor because the finding was associated with the external factors attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power operations. The inspectors determined that the issue increased the likelihood of a fire in the emergency diesel generator (EDG) building. The condition was evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection Significance Determination Process. It was determined that in the event of a fire consuming the MCC, no transient would be placed on the plant and no components required to safely shutdown the plant would be impacted. As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green. The inspectors determined that a cross-cutting aspect was associated with this finding in the area of human performance related to conservative decision making. Specifically, Entergys decision-making was non-conservative related to its decisions on the process used to identify the source of the acrid odor; re-energize the damaged electrical equipment; and keep a damaged electrical component energized for 14 days prior to its removal from the MCC. (H.1(b) per IMC 0305
05000247/FIN-2009002-03Failure to identify and Promptly Correct Degraded 480 Volt Switchgear Room Fire Door2009Q1The inspectors identified a NCV of very low safety significance related to License Condition 2.K., fire protection program, because personnel did not promptly identify and correct a degraded three-hour rated fire door latch mechanism on the west entrance of the 480-Volt switchgear room. Specifically, inspectors identified the fire door in a nonfunctional state on several instances over the course of a month. Entergy personnel replaced the fire door latch mechanism on March 3, 2009. This issue was entered into the corrective action program as six condition reports spanning several weeks and included an extent of condition walkdown of site fire doors. The finding was more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. This fire door, when degraded, impacts the reliability of mitigating systems in the 480-Volt switchgear room that are relied upon during a postulated large fire in the turbine building, and vice versa. This finding was evaluated using Phase 1 of IMC 0609 Appendix F, Fire Protection Significance Determination Process. Since the area in question had a fire watch posted during the time the door was degraded for an unrelated issue, an adequate level of protection was maintained to compensate for the degraded door. As such, according to task 1.3.1, the inspectors determined the finding was Green. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution because Entergy personnel did not thoroughly evaluate a degraded fire door latch on several occasions, such that the resolution of the problems addressed the causes. (P.1(c) per IMC 0305
05000247/FIN-2009002-04Inadequate Maintenance Procedure for EDG Ventilation Motor Control Center #22009Q1The inspectors identified a NCV of very low safety significance related to TS5.4.1, Administrative Controls: Procedures, because Entergy did not maintain an adequate maintenance procedure for a safety-related electrical motor control center (MCC). Specifically, the eight-year maintenance procedure for the affected EDG ventilation MCC did not contain an adequate method to identify high resistance connections within the cubicle as was expected in the applicable preventative maintenance industry template. Subsequently, a high resistance connection within the MCC developed into a phase-to-phase electrical fault on January 28, 2009. Entergy entered the issue into the corrective action program, scoped the affected MCC and 21additional MCCs into the sites thermography program, and planned to revise the maintenance procedure. This issue was more than minor because the finding was associated with the external factors attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety systems during shutdown as well as power operations. Specifically, the high resistance connection degraded into a phase-to-phase fault and increased the likelihood of a fire in the EDG building. The condition was evaluated by a Senior Reactor Analyst utilizing Phase 2 of IMC 0609 Appendix F, Fire Protection Significance Determination Process. It was determined that in the event of a fire consuming the MCC, no transient would be placed on the plant and no components required to safely shutdown the plant would be impacted. As a result, in accordance with task 2.3.5 of Appendix F, the issue was screened to Green. The inspectors determined that the finding had a cross-cutting aspect associated with the area of problem identification and resolution related to the use of operating experience (OE). Specifically, Entergy personnel did not implement industry recommended practices, or an alternate equivalent method, for identifying high resistance connections in electrical switchgear. (P.2(b) per IMC 0305
05000247/FIN-2009002-05Failure to Include RWST Level Maintenance In Online Risk Assessment2009Q1The inspectors identified a NCV of very low safety significance related to 10 CFR50.65(a)(4), because Entergy personnel did not adequately assess the risk associated with the unavailability of the Refueling Water Storage Tank (RWST) level indication during planned maintenance on the level transmitters and instrumentation. Entergy entered the issue into the corrective action program (CR-IP2-2009-00342), updated the risk model to include the maintenance activity, assessed the risk, and appropriately coded the maintenance activity to ensure it would be risk assessed in the future. The inspectors determined that this finding was more than minor because it was a maintenance risk assessment issue in which personnel did not consider risk significant SSCs that were unavailable during maintenance. The RWST level indication is specifically listed in Table 2 of the plant specific Phase 2 SDP risk-informed inspection notebook. The inspectors determined the significance of this issue in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The inspectors determined that this finding was of very low safety significance because the Incremental Core Damage Probability Deficit was less than 1E-6.The inspectors determined that the finding had a cross-cutting aspect in the area of human performance related to work control. Specifically, Entergy personnel did not appropriately plan work activities by incorporating risk insights for affected plant equipment. (H.3(a) per IMC 0305
05000247/FIN-2009002-06Inadequate Test Acceptance Criteria for Auxiliary Component Cooling Check Valves2009Q1The inspectors identified a NCV of very low safety significance related to 10CFR 50.55a, Codes and standards, because Entergys procedure, 2-PT-Q031A for an auxiliary component cooling water pump, did not contain appropriate acceptance criteria for positively determining that safety-related check valves performed their safety function when required in accordance with the American Society of Mechanical Engineers(ASME) OM Code. Specifically, the test used reverse rotation of a parallel pump to verify that the pumps discharge check valve was closed although previous site-specific experience demonstrated that the pump impeller would not rotate backwards when the check valve was stuck open. Entergy entered this issue into their corrective action program as CR-2009-1312.The inspectors determined that the performance deficiency was greater than minor because it was associated with the procedure quality attribute of the Mitigating System cornerstone and it adversely affected the cornerstones objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the test criterion used in procedure 2-PT-Q013A did not ensure that valve755A reliably performed its safety function when tested as demonstrated by testing performed in January 2005. The inspectors determined that the performance deficiency was of very low safety significance (Green) IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings. Specifically, the inspectors determined that this finding was of very low safety significance because the finding did not result in a loss of safety function and did not screen as potentially risk-significant due to external events initiating events. The inspectors determined the finding had a cross-cutting aspect related to effective corrective actions in the corrective action program component of the problem identification and resolution area. Specifically, Entergy personnel did not implement effective corrective actions to resolve the testing inadequacy since 2005 and during subsequent quarterly testing. (P.1(d) per IMC 0305
05000247/FIN-2009002-07Failure to Follow Radiation Protection Procedures2009Q1The inspectors identified a NCV of very low safety significance related to Technical Specification 5.4.1.a, Procedures, because Entergy personnel did not generate condition reports or investigation paperwork for multiple high dose-rate alarms as required by station procedures. Specifically, personnel did not generate the required condition reports and adequately document the investigations for six instances of unplanned or un-briefed electronic dosimeter alarms that occurred between January2009 and March 2009. The performance deficiency resulted in workers receiving unanticipated dose rate alarms with no formally-documented investigation prior to returning to work in a Radiologically Controlled Area. Entergy entered the finding into the corrective action program as condition report CR-IP3-2009-01253 and 01318.The finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone attribute of programs and process, and adversely affected the objective to ensure adequate protection of worker health and safety from exposure to radiation. Moreover, the inspectors identified a programmatic deficiency to maintain and implement programs to keep exposures as low as reasonably achievable, because multiple examples were identified regarding the failure to satisfy station radiation protection procedures. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure of an individual, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The inspectors determined that the finding had a cross-cutting aspect related to procedural adherence in the work practices component of the human performance area. Specifically, Entergy personnel did not follow procedures to generate condition reports and document investigations when high dose-rate alarms were received by workers.(H.4(b) per IMC 0305
05000247/FIN-2009003-01Inadequate Design Change Package for Installation of Main Boiler Feed Pump Control System Tubing2009Q2The inspectors documented a self-revealing finding of very low safety significance because Entergy engineers did not provide adequate guidance in a design change package for installation of tubing in the 21 main boiler feedwater pump (MBFP) control system that eventually led to the tubing failure and an unplanned trip of the reactor plant. Entergys design change procedure required that instructions delineating installation precautions be provided in the design change package. Entergys corrective actions included repairing the affected tubing, identifying and replacing similar tubing on the 22 MBFP, and examining Unit3 MBFPs to identify the extent of the condition. Entergy staff placed this issue into the corrective action program and performed a root cause analysis. The finding was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that affect plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, the incorrectly installed MBFP control tubing resulted in a loss of the 21 MBFP and, ultimately, a reactor trip due to low steam generator water level. The inspectors determined that the finding was of very low safety significance (Green) using the Phase 2 Indian Point Unit 2 risk-informed inspection notebook, in accordance with IMC0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined there was no cross-cutting issue associated with the finding because the performance deficiency did not reflect current licensee performance. Specifically, the performance deficiency occurred several years ago and was outside the current assessment period, and procedures have since been improved in the design control, work control and vendor control processes that reduced the likelihood of vendors working one quipment without sufficient training or work instructions.
05000247/FIN-2009003-02Licensee-Identified Violation2009Q2Section 4OA5.2, on January 7, 2009, following installation and post work testing of an additional backup nitrogen supply to the ADVs, Entergy personnel identified that surveillance tests for the nitrogen backup supplies to the ADVs were never performed contrary to TS surveillance requirement 3.3.4.2.The inspectors determined this constituted a violation of TS 3.3.4, Remote Shutdown, which includes the TS surveillance requirement to verify that the nitrogen backup supply control circuit and transfer switch to the steam generator ADVs are capable of performing their intended function. Contrary to this requirement, Entergy personnel did not verify the functionality of the control circuitry associated with the nitrogen backup supply to the ADVs. The inspectors determined this issue was of very low safety significance (Green) per SDP Phase 1 screening because the safety function of the ADVs was not lost. Specifically, the inspectors determined the remote shutdown function for the steam generator requires only one ADV to be operable. All four ADVs were capable of being operated with the normal station air supply. Entergy personnel entered the issues into the corrective action program as CR-IP2-2009-00062, -00069, -00077, -00137, and -00983
05000247/FIN-2009005-01Incomplete Licensed Operator Medical Examinations2009Q4An NRC-identified Severity Level IV Violation of 10 CFR 50.9, Completeness and accuracy of information was identified because Entergy submitted inaccurate medical information for licensed operators. The inspectors identified submittals to the NRC were inaccurate due to the omission of a tactile test (test performed to ensure that operators can distinguish among various shapes of control knobs and handles by touch) from the required licensed operator medical examinations. The inspectors determined that Entergy's medical physician did not adequately test all licensed operators (both initial and renewal licensees) in accordance with 10 CFR 55.21 and 10 CFR 55.33 with respect to ANSI/ANS-3.4 1983. However, Entergy had submitted medical information, as required by 10 CFR 55 for licensed operators and applicants that stated the testing had been performed satisfactorily. Following identification of the issue, Entergy entered the issue into the corrective action program (CR-IP3-2009-04487) and completed corrective actions to develop and administer an appropriate test. The inspectors noted that all licensed operators passed this new test and no new license conditions were required. Entergy's failure to provide complete and accurate information to the NRC could have resulted in an incorrect licensing action and is a performance deficiency because the licensee is required to comply with 10 CFR 50.9. Because this violation of 10 CFR 50.9 is considered to be a violation that potentially impedes or impacts the regulatory process, it is dispositioned using the traditional enforcement process. The finding was more than minor because documents which provided the information to the NRC were signed under oath by the company medical physician and the Site Vice President.
05000247/FIN-2009005-02Siren Test Failure2009Q4A self-revealing NCV of very low safety significance of 10 CFR 50.47(b)(5) was identified because Entergy personnel did not ensure the alert and notification system (ANS) sirens remained available for notification of the populace within the plume exposure pathway emergency planning zone (EPZ). Specifically, Entergy personnel did not use procedures, step lists, or checklists while performing maintenance on the ANS siren system which caused approximately 8% of the siren system to be degraded for 56 days. The siren technicians did not use a detailed written procedure or work instruction to perform siren file updates, but instead relied on performing the task from memory. As a result, on September 16, 2009, Entergy conducted a full volume siren test during which a total of 18 sirens indicated a failure to function. Entergy entered the siren failures into their corrective action process for resolution and performed a root cause of the event to determine the short and long term corrective actions. The finding was more than minor because it was associated with the Emergency Preparedness (EP) cornerstone attribute of facilities and equipment, and impacted the cornerstone objective of ensuring that Entergy is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated using (MC 0609 Appendix B, Emergency Preparedness Significance Determination Process (SOP) and was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect associated with the area of Human Performance because Entergy did not ensure adequate supervisory and management oversight of work activities performed by siren technicians
05000247/FIN-2009005-03Partial Loss of Control Room Indication during Nl 41 Recorder Replacement2009Q4A self-revealing non-cited violation (NCV) of very low safety significance of 10 CFR 50, Appendix B Criterion V Instructions, Procedures, and Drawings, was identified because Entergy personnel did not perform work regarding replacement of a control room digital recorder. As a result, during performance of the work, personnel inadvertently shorted a live wire resulting in a partial loss of control room indications and alarms related to the safety relief valve acoustic monitor flow indications, low range steam and feed flow indications, and inadvertent control rod movement. Entergy personnel reset the breakers to restore control room indications and entered this issue into the corrective action program as CR-IP2-200904860. Personnel subsequently replaced the digital recorder with the circuit breaker opened to eliminate the electrical hazard. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the grounded recorder power supply resulted in a loss of control room indications and alarms that could have impacted operations response to an event. The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance related to work practices. Specifically, Entergy personnel did not follow procedures during the replacement of a control room digital recorder.
05000247/FIN-2009005-04Transient Combustibles Stored on the ISFSI pad2009Q4An NRC-identified Severity Level IV, NCV of 10 CFR 72.212(b)(2)(ii), was identified because Entergy personnel did not evaluate a change to the written evaluation described in its Holtec Updated Final Safety Analysis Report (UFSAR) prior to implementing the change. Specifically, inspectors identified that Entergy personnel were storing combustible material on the Independent Spent Fuel Storage Installation (ISFSI) pad which was contrary to the Holtec UFSAR and the Entergy 72.212 Evaluation Report which stated that transient combustibles will not be stored on the ISFSI pad. Following the inspectors' questions, Entergy personnel determined the required evaluation in accordance with the requirements of 10 CFR 72.48(c) was not performed. Entergy personnel entered the issue into their corrective action program and verified that all combustibles had been removed from the pad. The Reactor Oversight Process (ROP) was not used for this finding because inspections of ISFSI activities are covered under NRC Manual Chapter 2690 and are not incorporated in the reactor safety cornerstones in the ROP's Significance Determination Process (SOP). It was determined that the failure to evaluate a change to the written evaluation required by 10 CFR 72.212 using the requirements of 10 CFR 72.48(c) was a performance deficiency that was reasonably within Entergy's ability to foresee and prevent. The finding was determined to be a Severity Level IV violation based on Supplement VI, Example 0.2 of the NRC Enforcement Policy. A cross-cutting aspect was not assigned since the performan.ce deficiency was not applicable to evaluation in accordance with the ROP.
05000247/FIN-2009007-01Failure to evaluate the impact on breaker coordination for the Westinghouse Amptector type LSG trip unit discriminator feature.2009Q3The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that Entergy did not verify the adequacy of design because they did not evaluate the impact of the installed Amptector discriminator instantaneous trip feature on breaker coordination. Following identification Entergy entered the issue into the corrective action program and performed an operability assessment and extent-of-condition review. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the 480Vac bus to respond to initiating events to prevent undesirable consequences. Specifically, load center Bus 6A (and 2A, 3A and 5A) would be incapable of meeting the design basis function when required if the incoming line breaker to the load center bus were to trip due to lack of coordination for a fault on a non-Class 1E circuit during a design basis accident. The finding was determined to be of very low safety significance because the design deficiency was confirmed not to result in loss of operability or functionality. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current performance
05000247/FIN-2009007-02Failure to ensure that the CCW pump hydraulic performance test procedures had acceptance criteria that incorporated the limits from applicable design documents.2009Q3The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that Entergy did not to ensure that the component cooling water pump hydraulic performance test procedures had acceptance criteria which incorporated applicable design limits sufficient to ensure continued pump operability. Specifically, if the pump flow rate had degraded to the lower limit of the acceptance band, as listed in the test acceptance criteria, the pump would not have been able to meet the design basis flow requirements at the minimum acceptable differential pressure listed in the test procedure. In addition, the test acceptance criteria for design basis flow rate and differential pressure had no allowance for measurement uncertainty of the test instruments. In response to this deficiency, Entergys short-term corrective actions included initiation of a corrective action condition report and completion of an operability determination for the affected equipment. The finding was more than minor because it was associated with the design control attribute of the Mitigating Cornerstone and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the test acceptance criteria did not ensure that the No. 23 component cooling water pump remained capable of performing its safety function under design basis conditions. The finding had very low safety significance because it was not a design or qualification deficiency, did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because Entergys initial operability review, issue prioritization, and subsequent evaluation did not adequately assess actual pump performance. (P.1(c))
05000247/FIN-2009007-03Failure to identify several degraded city water system pipe supports in the utility tunnel.2009Q3The team identified a finding of very low safety significance because Entergy did not identify or evaluate material deficiencies of the city water system, as required by ENLI- 102, Corrective Action Process. Specifically, Entergy did not identify or evaluate several degraded pipe supports on city water system piping in the utility tunnel, which represented reasonable doubt on system operability. The city water system provides a backup water supply for the condensate storage tank, fire fighting water supply, and provides alternate cooling to selected safety-related and risk significant pumps. The finding was not a violation because the city water piping, in the utility tunnel, is not safety-related, and the utility tunnel is not a safety-related or seismic structure. Entergy entered this issue into the corrective action program, assessed operability and extent-ofcondition, and repaired one of the non-functioning pipe supports to restore additional margin. The finding was more than minor because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the piping system could have potentially collapsed if additional pipe supports became degraded. The team determined the finding was of very low safety significance because it was not a design or qualification deficiency, did not represent of an actual loss of safety function of a single train, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because Entergy did not adequately implement the corrective action program with a low threshold for identifying issues. (P.1(a)
05000247/FIN-2009402-01Security2009Q4
05000247/FIN-2009402-02Security2009Q4
05000247/FIN-2009402-03Security2009Q4
05000247/FIN-2009402-04Licensee-Identified Violation2009Q4
05000247/FIN-2009402-05Licensee-Identified Violation2009Q4
05000247/FIN-2009402-06Licensee-Identified Violation2009Q4
05000247/FIN-2010002-01Isolation of Service Water to All Emergency Diesel Generators2010Q1A self-revealing NCV of Technical Specification (TS) Limiting Condition of Operation (LCO) 3.8.2 was identified when Entergy personnel did not maintain service water (SW) cooling to the emergency diesel generators (EDGs) when the reactor was in cold shutdown. Specifically, on March 13, 2010, Entergy personnel isolated cooling water flow to the EDGs for a period of three minutes. This condition was corrected after an alarm in the control room alerted the operators to the condition and the operators promptly directed the restoration of cooling water to the EDGs. The inspectors determined that the isolation of cooling water flow to the standby EDGs was a violation of TS LCO 3.8.2, which requires Two EDGs to be capable of supplying two safeguards power trains of the onsite AC electrical power distribution subsystem(s) required by LCO 3.8.10. Inadequate SW cooling to the EDGs, if left uncorrected, could have caused the EDGs to fail from a lack of cooling. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the objective to assure the availability, reliability and capability of systems that respond to initiating events to prevent core damage. The finding was determined to be of very low safety significance (Green) because further analysis by Entergy staff determined that the EDGs could have operated without cooling water for the period of three minutes. The finding has a cross-cutting aspect in the area of human performance related to work practices. Entergy personnel did not incorporate actions to address the impact of work on different job activities, and did not plan work activities to support equipment reliability by limiting safety systems unavailability and reliance on manual actions (H.3.b).
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