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05000443/FIN-2012005-03 + (Seabrook technical specification surveilla … Seabrook technical specification surveillance requirement 4.3.1.2, Reactor Trip System Instrumentation, requires that the reactor trip system features response time for each reactor trip function listed in Table 3.3-1 be verified to be within its limit at least once per 18 months. On September 25, 2012, NextEra identified that the full scope of response time testing for the reactor trip system function had not been completed since initial licensing because the implementing procedure did not verify the response time for both reactor trip methods. Testing had been completed on the under-voltage circuit, but no testing had been performed on the shunt trip circuit. The issue was determined to be a violation of Seabrook TS 6.7, Procedures and Programs, which requires that written procedures be established, implemented and maintained as recommended in RG 1.33, Revision 2, Appendix A, February 1978. RG 1.33, Appendix A, requires implementing procedures for each SR listed in TSs. Contrary to this requirement, since initial licensing, NextEras procedure for implementing TS SR 4.3.1.2 did not test the response time for reactor trip breaker function at least once per 18 months, which resulted in a violation of TS 3.1.2, Reactor Trip System Instrumentation, as described in LER 05000443/2012-002-00. The finding was associated with the Mitigating Systems cornerstone and was evaluated for significance using Exhibit 2 of IMC 0609, Appendix A. Since the finding was not a design or qualification deficiency, did not result in a loss of system safety function, did not result in loss of a single train for greater than its allowed outage time, and was not potentially risk significant due to external events, the finding was determined to be of very low safety significance (Green). The issue was entered into NextEras CAP as CR 1806525.s entered into NextEras CAP as CR 1806525.)
05000458/FIN-2010004-02 + (Second, 10 CFR 50. 75(f)(1) requires each … Second, 10 CFR 50. 75(f)(1) requires each power reactor licensee to periodically report to the NRC on the status of its decommissioning funding for each reactor or share of a reactor that it owns. The information in this report must identify any contracts upon which the licensee is relying pursuant to 10 CFR 50. 75(e)(1 )(v). Additionally, 10 CFR 50.9(a) requires that information provided to the NRC by a licensee be complete and accurate in all material respects. On March 30, 2009, Entergy submitted a decommissioning funding report for River Bend Station Unit 1, as required by 10 CFR 50. 75(f)(1). Contrary to the above, the decommissioning funding report Entergy submitted was not complete and accurate in all material respects in that it failed to report a contract (i.e., a power purchase agreement) with Entergy Texas, Inc., which Entergy has relied upon during the reporting period for financial assurance for the decommissioning of River Bend Station the decommissioning of River Bend Station)
05000387/FIN-2014005-06 + (Secondary Containment Door Found Ajar On F … Secondary Containment Door Found Ajar On February 12, 2014, PPL identified a secondary containment door (Door 612) between the HVAC room and central railroad bay wedged open by a door sign. In order for secondary containment to be operable in the as-found mode of operation, Door 612 had to be secured. PPL immediately secured the door, entered the condition into their CAP (2014-04709), and reported the condition under LER 50-387; 388/2014-002. Contrary to TS 5.4.1a, PPL did not secure the secondary containment door and maintain system operability in accordance with OP-134-002, RB HVAC Zones 1 and 3 after realignment of the secondary containment. The finding was more than minor because it adversely impacted the barrier performance attribute of barrier integrity and was determined to be of very low safety significance (Green) in accordance with IMC 0609, Appendix A, since the finding only represented a degradation of the radiological barrier function provided by standby gas treatment system. provided by standby gas treatment system.)
ENS 47516 + (Secondary containment pressure exceeded at … Secondary containment pressure exceeded atmospheric pressure which does not meet the surveillance requirement to have secondary containment vacuum greater than or equal to 0.25 inches of water vacuum. Secondary containment was declared inoperable and the Limiting Condition for Operation (LCO) Action Statement was entered. Manual control of the reactor building pressure control system was taken. Vacuum was less than 0.25 inches of water for approximately 1 minute. Secondary containment is (now) operable.There were no actual radiological releases associated with the event.Actual secondary containment integrity was not challenged. The secondary containment pressure excursion was a result of icing of the reactor building intake filters which caused the automatic reactor building pressure control system to function improperly.The licensee has notified the NRC Resident Inspector.e has notified the NRC Resident Inspector.)
ENS 46937 + (Secondary containment was declared inopera … Secondary containment was declared inoperable after transferring refuel floor supply fans. Secondary containment D/P (Differential Pressure) lowered to 0.17 inches of water vacuum which does not meet the surveillance requirement to have secondary containment vacuum greater than or equal to 0.25 inches of water vacuum. Refuel floor ventilation was restored back to the previous configuration and secondary containment D/P was restored back to greater than 0.25 inches of water vacuum. Vacuum was less than 0.25 inches of water for approximately 4 minutes.There were no actual radiological releases associated with the event.Actual secondary containment integrity was not challenged. The lowered secondary containment D/P was a result of a ventilation lineup change.The licensee has notified the NRC Resident Inspector and the State of Minnesota.dent Inspector and the State of Minnesota.)
ENS 48961 + (Secondary leak collection and release radi … Secondary leak collection and release radiation monitor RE19 A/B power supply was removed from service at 2058 EDT on 4/23/13 for planned maintenance activities. This radiation monitor is relied upon for emergency classifications. Expected duration of the maintenance activities is 72 hours.The licensee notified the State of Connecticut, the town of Waterford, and the NRC Resident Inspector.Waterford, and the NRC Resident Inspector.)
05000266/FIN-2015002-04 + (Section (b) of TS 5.7.1 requires, in part, … Section (b) of TS 5.7.1 requires, in part, that access toand activities ina high-radiation area be controlled by a radiation work permit or equivalent. Section (e) of TS 5.7.1 requires, in part, that entry into HRAs be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Contrary to the above, on April 14, 2015, an individual entered an HRA without being on a radiation work permit that allowed for HRA entry and was not made knowledgeable of the dose rates in the HRA. The licensee entered this issue into the CAP as AR 0204280. This violation is considered to be of very low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls; (2) there were no overexposures; (3) there was not a substantial potential for overexposures; and (4) the ability to assess dose was not compromised.bility to assess dose was not compromised.)
05000483/FIN-2009003-06 + (Section 20.1902(a) of Title 10 of the Code … Section 20.1902(a) of Title 10 of the Code of Federal Regulations states, in part, that the licensee shall post each radiation area with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION, RADIATION AREA. Contrary to the above, on October 27, 2008, during a walk down of the auxiliary building, the radiation area posting for room 1322 was found lying on the floor and was not conspicuously posted at the entrance to the room. Radiation levels within the room were as high as 30 millirem per hour. The violation was identified by a licensee individual who immediately notified the radiation protection department and the issue was corrected. This issue has been documented as Callaway Action Request 200811123. The finding was determined to be of very low safety significance because it did not involve ALARA planning and controls, did not involve an overexposure, did not have a substantial potential for overexposure, and did not result in an impaired ability to assess dosesult in an impaired ability to assess dose)
05000440/FIN-2008005-07 + (Section 20.1902(a) to Title 10 of the CFR … Section 20.1902(a) to Title 10 of the CFR requires, in part, that the licensee post each radiation area with a conspicuous sign or signs bearing the radiation symbol and the words Caution, Radiation Area. Contrary to the above, on January 15, 2008, a radiation area in the waste abatement and reclamation facility was not posted. The source of the radiation was a B12 box containing a scrap reactor water clean-up pump with dose rates of 80 millirem per hour at 30 centimeters. The violation was identified by licensee personnel and was documented in the licensees corrective action program as CR 08-33510. Immediate corrective actions were to properly post and control the area. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure, nor potential for overexposure, and the licensees ability to assess dose was not compromisedability to assess dose was not compromised)
05000382/FIN-2008004-06 + (Section 3.2 of licensee Refueling Procedur … Section 3.2 of licensee Refueling Procedure RF-005-002 states, that while fuel movement is in progress, PEER check/verifications are required for grapple operation of the refueling machine including verification of proper Z-coordinate, and weight verification when raising the hoist. Section 3.31 states that operation of fuel handling equipment with an interlock bypassed raises the risk of damaging fuel assemblies and equipment. Finally, a caution statement in Section 5.1.8 states that if the refueling machine load varies more than 100 pounds during withdrawal of a fuel assembly, fuel movement should be terminated. Contrary to this, on May 18, 2008, operators proceeded to lift a fuel assembly without verifying that the load did not vary by more than 100 pounds. The load varied by about 1400 pounds. The senior reactor operator (PEER) noted the initial load of the fuel assembly of approximately 1500 lbs, and failed to check the load again during travel. Both the refueling bridge operator and the senior reactor operator failed to note that the grapple and fuel assembly had rotated approximately 25 degrees out of position with the hoist box. At approximately 193 inches, the grapple actuator came in contact with the hoist box and began lifting the hoist box. The refueling machine weight gauge indicated an increase in weight to 2900 lbs. The finding was of very low safety significance (GREEN) because it did not represent an actual event that upset plant stability or damaged fuel cladding. The licensee entered the violation in their corrective action program as Condition Report CR-WF3-2008-2423ogram as Condition Report CR-WF3-2008-2423)
05000382/FIN-2008004-05 + (Section 3.33 of licensee Refueling Procedu … Section 3.33 of licensee Refueling Procedure RF-005-002, Refueling Equipment Operation, states that using the key override to move the refueling machine hoist in the outward direction with a fuel assembly in the core region would require entering the limiting condition for operation for Technical Specification 3.9.6. Technical Specification 3.9.6, requires suspending movement of fuel assemblies when the refueling mast overload cut off limit of less than or equal to 3350 pounds was unavailable. Contrary to the above, on May 18, 2008, the overload cut off limit was unavailable and operators placed the refueling machine in key override and moved a fuel bundle in the outward direction. The operators did not enter Technical Specification 3.9.6. The operators were in the process of moving fuel when the refueling machine computer had \"locked-up.\" In an effort to reboot the computer, licensee personnel placed the refueling machine in key override, which bypassed the refueling equipment interlocks. The intent was to place the mast in a position that would allow the computer to be rebooted. The personnel failed to realize that the actions were not permitted by Technical Specification and plant procedureschnical Specification and plant procedures)
ENS 41935 + (Section 4.1 of Appendix B of the Operating … Section 4.1 of Appendix B of the Operating License for Units 2 and 3 requires Southern California Edison (SCE) to report to the NRC within 24 hours any unusual or important environmental events, which includes unusual fish kills. Between August 19 and August 20, 2005, SCE removed an unusually large number of fish from the Units 2 and 3 intake structure. At approximately 1000 PDT on August 20, 2005, SCE estimated the quantity to be approximately 11,070 pounds (approximately 6420 pounds from Unit 2 and 4650 pounds from Unit 3). While the NRC has not specified a reporting limit for an unusual fish kill, SCE has internally defined this quantity as 4500 pounds. This unusual influx of fish is unrelated to plant operation and a heat treat of the intake structure was not being performed. However, there is a heat treat of the San Onofre Unit 2 intake structure scheduled for later today.The licensee stated that the fish kill was apparently the result of a large school of anchovies that swam to close to the intake.The licensee notified the NRC Resident Inspector.ensee notified the NRC Resident Inspector.)
ENS 40504 + (Section 4.1 of Appendix B of the Operating … Section 4.1 of Appendix B of the Operating License for Units 2 and 3 requires Southern California Edison (SCE) to report to the NRC within 24 hours any unusual or important environmental events, which includes unusual fish kills.Between February 2 and February 3, 2004, SCE removed an unusually large number of sardines from the Units 2 and 3 intake structure. At approximately 1100 on February 3, 2004, SCE determined the quantity to be approximately 13,590 pounds (approximately 6940 pounds from Unit 2 and 6650 pounds from Unit 3). While the NRC has not specified a reporting limit for an unusual fish kill, SCE has internally defined this quantity as 4500 pounds. SCE believes the unusualinflux of sardines may be related to current winter storm conditions.The NRC Resident Inspectors have been notified The licensee will also give a courtesy call to the San Diego Regional Water Quality Control BoardDiego Regional Water Quality Control Board)
05000247/FIN-2009003-02 + (Section 4OA5.2, on January 7, 2009, follow … Section 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 -00983-00062, -00069, -00077, -00137, and -00983)
05000338/FIN-2014005-01 + (Section 5.5, Programs and Manuals, of Nort … Section 5.5, Programs and Manuals, of North Anna TS stated, in part, that an offsite dose calculation manual shall be established, implemented and maintained. Section 6.2 of VPAP-2103N, Offsite Dose Calculation Manual (North Anna), Revision 23, required that radioactive liquid effluent monitoring instrumentation channels be maintained operable with a trip setpoint which will automatically isolate the discharge line in response to a high radiation condition. Contrary to section 6.2 of VPAP- 2103N, 1-LW-RM-111, the liquid radioactive effluent rad monitor was not maintained operable from May 13, 2014 to September 3, 2014. During the time period that 1-LW-RM-111 was inoperable the licensee was not aware of the situation due to inadequate procedure guidance for daily source checks required to verify operability as defined in 0-LOG-6A. This finding was identified by the licensee and entered ithe licensees corrective action program as CR558708, 1-LW-RM-111 not capable of performing design function from 5/13/14 to 9/3/14,and Apparent Cause Evaluation, ACE019800, 1-LW-RM-111 not capable of performing design functions from 5/13/14 to 9/3/14. The inspectors performed a significance determination using NRC Inspection Manual 0609, Appendix D, Public Radiation Safety Significance Determination Process, Section C, dated February 12, 2008. Because the licensee was able to monitor the radioactive effluent release with downstream radiation monitors on the circulating water line which have alarm capability thfinding was determined to be of very low safety significance (Green).e of very low safety significance (Green).)
05000400/FIN-2016002-01 + (Section 50.48 of 10 CFR, Fire Protection, … Section 50.48 of 10 CFR, Fire Protection, states that a fire protection program that is maintained to the requirements of National Fire Protection Association (NFPA) standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, is an acceptable method for complying with the requirements of Section 50.48. Section 3.8.1 of NFPA 805 states, in part, that alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble alarms to the control room or other constantly attended location from which required notifications and response can be initiated. Contrary to the above, from December 2015, to May 17, 2016, neither the licensees design reviews nor post-modification tests identified that the fire protection system installed on the 286-ft elevation of the turbine building did not transmit trouble alarms to the Harris main control room. Following installation and testing, the newly-installed Protecta WireTM system and fire detection panel, 1-SFD-E144, were placed in service in late December 2015. On May 17, 2016, while performing maintenance periodic test, MPT-I0052, Turbine Building Local Fire Detection Control Panel LFDCP-10 Test and 1-SFD-E144 Test of the fire detection system, the technicians performing the test recognized that the remote trouble alarm function would not cause an alarm in the control room. The licensee entered the issue concerning the inadequate remote alarm function into the corrective action program via AR 2030427 and implemented actions to incorporate and test the remote trouble alarm function into the EC package. The licensee also initiated corrective actions via AR 2033716 and AR 2038682 to address issues in the design review process. Using IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green) since the reactor would still be able to achieve and maintain safe shutdown.ble to achieve and maintain safe shutdown.)
05000400/FIN-2016002-02 + (Section 50.48(c) of 10 CFR and NFPA 805, 2 … Section 50.48(c) of 10 CFR and NFPA 805, 2001 Edition, Section 2.4.2.2.2(b), Common Enclosure Circuits, require that those circuits which share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required component, shall be identified to prevent propagating fires outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables. Contrary to the above, from October 1986 to September 2014, the licensee failed to meet the requirements of 10 CFR 50.48(c) and NFPA 805, Section 2.4.2.2.2(b), in that, the licensee failed to identify and provide adequate electrical fault protection for the turbine emergency oil pump control cables 11376C and 11376D. The cables could have created a common enclosure fire hazard under postulated situations which could have resulted in a secondary fire in other fire areas and could have adversely affected the capability to achieve safe and stable plant conditions. A fire-induced failure could have caused the loss of the required safe shutdown components. This violation was determined to be of very low safety significance (Green) based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. A detailed risk evaluation was performed in accordance with NRC IMC 0609 Appendix F, and NUREG/CR6850 Rev. 0 and 1, using inputs from the licensees NFPA 805 Fire PRA. The major analysis assumptions included a one-year exposure interval, and secondary fires occurring between the power supply and the fire induced hot short. The dominant sequence was a fire in the main control board causing a secondary fire in the B cable spreading room which if unsuppressed could result in the inability to achieve safe shutdown resulting in core damage. The quantitative screening approach resulted in a calculated delta core damage frequency of less than 1E-06, which screened this violation to Green (very low safety significance). This violation was documented in the licensees corrective action program as Condition Report 692766.action program as Condition Report 692766.)
05000261/FIN-2014003-02 + (Section 50.49 of 10 CFR, Environmental Qua … Section 50.49 of 10 CFR, Environmental Qualification of electric equipment important to safety for nuclear power plants, states that each licensee shall establish a program for qualifying specified electric equipment. Section (a)(3) of 10 CFR 50.49 specifies the environmental qualification requirements for post-accident monitoring equipment. Section (f) of 10 CFR 50.49 requires, in part, that each item of electric equipment important to safety must be qualified by testing an identical item of equipment under identical conditions. Contrary to the above, since May 1992, the licensee failed to maintain the qualification of the limit switches for CVC-204B, letdown line isolation, in accordance with the tested configuration of the equipment which rendered the Post Accident Monitoring Instrumentation function inoperable. The licensee documented this condition in AR 640902 and AR 633207. The cause was determined to be associated with a human performance event in which the licensee failed to use the proper heat shrink insulators per procedure CM-309, Sealing Low Voltage Electrical Splices for Environmentally Qualified or Safety Related Splices. Following discovery of this condition, the licensee replaced the non-environmental qualified splice and returned the equipment to the test configuration. Using IMC 0609, Appendix A, issued June 19, 2012, The SDP for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours.ce rule program for greater than 24 hours.)
05000400/FIN-2017003-03 + (Section 50.54(q)(2) of 10 CFR requires, in … Section 50.54(q)(2) of 10 CFR requires, in part, that a licensee shall follow and maintain the effectiveness of an emergency plan which meets the planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR Part 50, Appendix E . Section 50.47(b)(4) of 10 CFR requires that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, from April 2010 to May 2017, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, the licensee's emergency classification scheme action levels for Category F Fission Product Barrier EAL , contained declaration threshold values for the containment high range radiation monitor , which were lower than the correct values due to an improper methodology used in calculating the loss of fuel clad barrier and potential loss of containment barrier threshold values and rendered the EALs ineffective. The licensee implemented compensatory actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented until a permanent change can be made to the license. The issue was entered into the licensees CAP as NCR 02123373. The inspectors evaluated this issue as an ineffective EAL per IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process , Figure 5.4 -1. The inspectors concluded that the violation was of very low safety significance (Green). Although the incorrect EAL would alone render an early EAL classification of a General Emergency (GE) based upon the specific radiation monitor, other EALs would provide a GE classification in an accurate and timely manner aligned with the incorrect threshold values of the containment high range radiation monitor .containment high range radiation monitor .)
05000261/FIN-2016002-01 + (Section 50.55a(h)(2) of 10 CFR states in p … Section 50.55a(h)(2) of 10 CFR states in part, for nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of Institute of Electrical and Electronic Engineers (IEEE) Std. 6031991 and the correction sheet dated January 30, 1995. The Robinson FSAR (current licensing basis) Section 3.1.2.20, states in part that, reactor protection is designed to meet all presently defined reactor protection criteria and is in accordance with the proposed Institute of Electrical and Electronic Engineers (IEEE) 279 Standard for Nuclear Plant Protection Systems, August 1968. IEEE-279, Section 4.2, requires that any single failure within the protection system shall not prevent proper protection system action when required. Contrary to this requirement, from initial startup, until April 13, 2016, when using a FRBV (i.e., FRBV in the open position in Modes 1, 2, and 3), and a MSLB occurred, the protection system would not provide the proper system protection action. Specifically, with a single failure of the FRBV to close, the protective system action to isolate feedwater could not be accomplished. This would cause an increase in secondary mass available for release in containment structure, resulting in a higher peak containment pressure that would challenge the containment design pressure. As corrective actions, the licensee implemented a standing instruction and placed caution tags on the FRBVs to ensure the valves remain closed/isolated while operating in Modes 1, 2, and 3. Additionally, the licensee completed an engineering change to update the containment analysis and licensing basis. The licensee entered this issue into the CAP as CRs 2012658, 2020495, and 2018710. The failure to meet the single failure criterion for feedwater isolation following a main steam line break inside containment was a performance deficiency (PD). Significance Determination Process (SDP) screening in accordance with NRC IMC 0609.04 determined that the PD affected the secondary short term heat removal safety function of the mitigating systems cornerstone. The finding was determined to represent a loss of function and a detailed risk assessment was performed per NRC IMC 0609 Appendix A. The bounding analysis assumed a conditional core damage probability of 1.0, a 14 day exposure period estimated from surveillance and outage schedules, and main steam line break inside containment (MSLBIC) initiating event probability and main feedwater regulating valve bypass (MFWRVBV) failure to close probabilities from the NRC SPAR model data. The dominant sequence was an MSLBIC with a failure to close of the MFWRVBV which was assumed to lead to core damage and large early release. The risk was mitigated by short exposure period and the low likelihood of the MSLBIC and the failure to close of the MFWRVBV. The bounding analysis determined that the PD represented a risk increase of < 1.0E-7/year, a GREEN finding of very low safety significance for both core damage frequency and large early release frequency. frequency and large early release frequency.)
05000282/FIN-2011003-11 + (Section 50.65 (a)(iv) of Title 10 of the C … Section 50.65 (a)(iv) of Title 10 of the Code of Federal Regulations requires that licensees assess and manage the increase in risk that may result from proposed maintenance activities prior to performing maintenance. Contrary to the above, on May 10 and 11, 2011, the licensee failed to properly assess and manage the risk associated with the 12 diesel driven cooling water pump which was rendered unavailable when expanding a Unit 1 outage-related clearance order. This resulted in Unit 2 entering an unplanned orange risk condition. This issue was documented in CAPs 1282986 and 1285097. The licensee determined that this issue occurred because the risk model used by the work week managers and operations personnel had not been updated to reflect recent plant changes. Corrective actions included opening two valves, which restored cooling water supply to the pump, and ensuring personnel were aware of risk model changes that should be considered when performing risk assessments. The inspectors determined that the failure to properly assess plant risk in accordance with 10 CFR Part 50.65(a)(iv) was a performance deficiency that required an SDP evaluation. The inspectors consulted IMC 0609 to assess the impact on both the outage (Unit 1) and online (Unit 2) units. Specifically, Checklist 4 of Appendix G, Shutdown Operations, was utilized by inspectors to assess the impact on Unit 1. The inspectors concluded that all checklist attributes were met and Phase 2 or 3 evaluations were not required. Additionally, the redundancy in cooling water pumps available at the time of occurrence did not cause a risk change to Unit 1. For Unit 2, the inspectors utilized IMC 0609, Appendix K, Maintenance Rule Risk Assessment Significance Determination Process, and determined the risk deficit was not greater than 1E-06. Consequently, the inspectors concluded that this finding was of low safety significance (Green).ng was of low safety significance (Green).)
05000282/FIN-2010003-05 + (Section 50.65 (a)(iv) of Title 10 of the C … Section 50.65 (a)(iv) of Title 10 of the Code of Federal Regulations requires that licensees assess and manage the increase in risk that may result from proposed maintenance activities prior to performing maintenance. Contrary to the above, on May 12, 2010, the licensee failed to properly assess and manage the risk associated with establishing the RCS as intact, releasing the containment airlock operator from duties, and the removal of equipment hatch from the Unit 2 containment. This resulted in Unit 2 entering an unplanned orange shutdown safety assessment path for the containment closure function. This issue was documented in CAP 1232396. Corrective actions included re-establishing the RCS as intact, closing the equipment hatch, re-instating the airlock operator, developing a procedure to clearly state the requirements to be met to declare the RCS intact, and a review of other outage activities to ensure that they were governed by specific procedures appropriate to the circumstance. The inspectors determined that the failure to properly assess plant risk in accordance with 10 CFR Part 50.65(a)(iv) was a performance deficiency that required an SDP evaluation. The inspectors consulted Inspection Manual Chapter (IMC) 0609, Appendix K, Maintenance Rule Risk Assessment Significance Determination Process, and found that this appendix could not be used due to the qualitative nature of shutdown safety assessments. Appendix K suggested that qualitative risk assessment issues be evaluated through a management review performed in accordance with IMC 0609, Appendix M. The inspectors were concerned with this approach since Unit 2 was shut down at the time this finding occurred. The inspectors consulted a Region III Senior Reactor Analyst (SRA) for additional assistance. Using IMC 0609, Appendix G, Significance Determination Process for Shutdown Conditions, the SRA determined that Unit 2 was in plant outage state #2. The SRA also found that the shutdown SDP stated that IMC 0609, Appendix H, Containment Integrity Significance Determination Process, should be used for shutdown findings related to containment issues. Using Section 4.0 of Appendix H, the SRA determined that this finding was a type B finding since it was related to a condition that had potentially important implications for the integrity of containment without affecting the likelihood of core damage. The SRA then used Section 6.2, Approach for Assessing Type B Findings at Shutdown, and determined that this finding was of low safety significance (Green) because it occurred during the late time window of the outage.during the late time window of the outage.)
05000346/FIN-2013008-01 + (Section 6.4.2.a of Procedure DB-FP-00005, … Section 6.4.2.a of Procedure DB-FP-00005, Fire Brigade, Revision 17, permitted table-top exercises to be credited for fire drills. License Condition 2.C(4) requires the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report (USAR). Section 9.5.1 of the USAR states that the FHAR is incorporated in the USAR by reference. In Appendix D, Compliance Matrix to Appendix A, of the FHAR, the licensee indicated that they complied with Position B5.b of NRC-Guidelines of Appendix A to Branch Technical Position 9.5-1 with respect to fire drills. Position B5.(b) stated fire brigades must operate as a team, drills provide fire brigades the opportunity to test itself in the major areas of the plant, drills should include the simulated use of equipment, and drills permit supervising personnel to evaluate the effectiveness of communications. The use of table-top exercises for drills did not allow for training as a team, testing themselves in major areas of the plant, simulating use of equipment in each area, and evaluating the effectiveness of communications. The inspectors determined that this issue was licensee identified because, at the time of the inspection, the licensee was in the process of reviewing the conduct of fire drills as documented by CR-2013-03331, Review of NRC Inspection Report issued for Kewaunee Power Station related to Fire Brigade Drills, dated March 7, 2013. The inspectors determined that the issue was of very low safety significance using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, because the licensee took only minimal credit for table-top exercises. Specifically, in the prior year, 2012, the licensee took credit for only one table-top exercise involving three individuals for a drill. Each of the shift crews had participated in drills in the plant each quarter during 2012. The licensee had entered this issue into their Corrective Action Program as CR-2013-04610, NRCFP2013: NRC Observations on Fire Drills and Pre-Fire Plans and planned to revise their procedure to be consistent with their licensing basis. be consistent with their licensing basis.)
05000272/FIN-2008002-04 + (Section 9.1 of the Salem Hope Creek Physic … Section 9.1 of the Salem Hope Creek Physical Security Plan states that the Access Authorization Program implements regulatory requirements utilizing the provisions in NEI 03-01 (Revision 1). Section 6.5, Item 4, of NEI 03-01 (Revision 1) states, If the individual has not been covered by a licensee BOP (Behavior Observation Program) and random drug and alcohol testing program from 6 30 days following the individuals last period of UAA (Unescorted Access Authorization), the licensee shall subject the individual to random selection for pre-access drug and alcohol testing. Contrary to this requirement, PSEG identified that between August 2007 and January 2008, seven individuals were not covered by a licensee BOP and random drug and alcohol testing program for the 6 to 30 day period following their last UAA. These individuals subsequently had their UAA reinstated without being subjected to random selection for preaccess drug and alcohol testing prior to granting them UAA. This was identified in PSEGs corrective action program in notification 20352044.e action program in notification 20352044.)
05000461/FIN-2007003-01 + (Section F of Clinton Power Stations operat … Section F of Clinton Power Stations operating license NPF-62, states that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the USAR. The USAR required that the fire protection program follow the requirements of Branch Technical Position APCSB 9.5-1, Appendix A, Plants Under Construction and Operating Plants. Branch Technical Position APCSB 9.5-1, Appendix A, requires that floors, walls and ceilings enclosing separate fire areas be sealed or closed to provide a fire resistance rating at least equal to the fire barrier itself. On November 2, 2006, the licensee identified two open, unsealed, 12\" x 12\" penetrations in the floor of the Division 3 switchgear room. The penetrations were under the main feed and reserve feed breakers to the 4kV switchgear for Division 3 . The inspectors determined that the failure to seal two penetrations between separate fire zones was a performance deficiency warranting a significance determination. The inspectors performed a Phase 2 evaluation using IMC 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that a credible fire scenario existed in that an energetic fault in the 4 kV Division 3 switchgear located directly above the open penetrations could ignite a non-safety related cable tray located directly below the open penetrations. A fire could then propagate horizontally along the non-safety related cable tray and then involve a Division 1 cable tray. The inspectors conservatively assumed that only Division 2 equipment would be available in such a scenario. Based on four vertical cabinet sections as being potential ignition sources, a 30 minute fire propagation time to reach the Division 1 cable tray, and remaining mitigating Division 2 equipment available, the inspectors determined that the issue was of very low safety significance.issue was of very low safety significance.)
05000354/FIN-2012201-01 + (Securit)

05000395/FIN-2010403-01 + (Security 95001 Supplemental Inspection follow up from Inspection Report 2009404 Dated October 26, 2009 Not publically available)
ENS 51483 + (Security condition in the Owner Controlled … Security condition in the Owner Controlled Area outside of the Protected Area. (There is) an unknown vehicle located on the South side of the Intake Canal. The vehicle is locked, the engine is not running, and the parking lights are on. Security is performing an inspection of the vehicle for explosives or other contraband in conjunction with local law enforcement.The Unusual Event was declared based on EAL HU-1.The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies.Notified (via phone and E-mail): DHS SWO, FEMA Ops Center, and NICC Watch Officer. Notified (via E-mail): FEMA NWC and NuclearSSA.* * * UPDATE FROM WILLIAM RAYBURN TO DONALD NORWOOD AT 2255 EDT ON 10/20/15 * * *At 2229 EDT, the Unusual Event for a Security Condition at Davis-Besse Nuclear Power Station was terminated. An inspection of the vehicle in question was performed and it was determined that no threat existed to the site at any time.The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies.Notified R3DO (Daley), IRD (Stapleton), NRR (Morris) and ILTAB (Tucker).Notified (via phone and E-mail): DHS SWO, FEMA Ops Center, and NICC Watch Officer. Notified (via E-mail): FEMA NWC and NuclearSSA.ied (via E-mail): FEMA NWC and NuclearSSA.)
ENS 44367 + (Security notified Limestone County Sheriff … Security notified Limestone County Sheriffs dispatch at 0340 CDT that Browns Ferry had experienced a fire and arc-over event at the area of cooling tower switchgear 'C'.At about 0225 CDT a fault occurred in the cable tray between the 161 kV supply transformer and the 4 kV cooling tower 'C' switchgear resulting in clearing of the 161 kV line and a subsequent fire. The fire was extinguished at 0236 CDT by on-site responders. At 0340 CDT, Limestone County Sheriff department was contacted by site security and informed of the event and asked to increase patrols in the area of the county road adjacent to Browns Ferry. It was known that the event was likely due to operational issues and not sabotage but the final determination had not been made at that time. Limestone County Sheriff was contacted again at 0520 CDT and informed that the problem was operational and that their assistance is not required.There are no safety related plant equipment reportability issues resulting from this event. The reportability is solely due to notification of an outside agency.This condition is reportable within 4 hours according to 10 CFR 50.72(b)(2)(xi). Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials.Licensee has notified NRC Resident Inspector.ensee has notified NRC Resident Inspector.)
ENS 49362 + (Security personnel reported to the Main Co … Security personnel reported to the Main Control Room that at time 1437 CDT, an alarm indicated that a secondary containment door was open beyond the normal delay time allowed for entry and exit. Security personnel responded and found the door open and unattended with the dogs extended meaning that the door was unable to be closed. Security personnel secured the door at time 1441 CDT. No deficiencies were found with the door. The fact that the door was open and unattended beyond the time allowed for normal entry and exit violates Technical Specification 3.6.4.1, 'Secondary Containment- Operating,' because surveillance requirement SR 3.6.4.1.3 was not met. This surveillance requires that doors be closed except during normal entry and exit.In summary, an SSC was inoperable in a required mode as a result of a personnel error and no redundant equipment in the same system was operable. Thus, the condition is reportable under 10CFR50.72(b)(3)(v)(D).The licensee notified the NRC Resident Inspector.ensee notified the NRC Resident Inspector.)
ENS 52995 + (Security personnel reported to the Main Co … Security personnel reported to the Main Control Room that at time 1000 CDT (on 9/27/2017), an alarm indicated that a secondary containment door was open beyond the normal delay time allowed for entry and exit. Security personnel responded and found the door open and unattended with the dogs extended meaning that the door was unable to be closed. Security personnel secured the door at time 1004 CDT. No deficiencies were found with the door. The fact the door was open and unattended beyond the time allowed for normal entry and exit results in Technical Specification 3.6.4.1 'Secondary Containment-Operating,' not being met because surveillance requirement SR 3.6.4.1.3 is not met. This surveillance requires that doors be closed except during normal entry and exit. By definition in NUREG-1022, when Secondary Containment is inoperable, it is not capable of performing its specified safety function which in turn makes this condition reportable in accordance with 10 CFR 50.72(b)(3)(v)(D).The NRC Resident Inspector has been notified. NRC Resident Inspector has been notified.)
ENS 40231 + (Security weapon and holster left unattende … Security weapon and holster left unattended. Immediate compensatory actions taken upon discovery. Contact the HOO for details.The licensee will be notifying the NRC Resident Inspector and has notified the R1TAS (Greg Smith).* * * UPDATE AT 1300 EDT ON 10/10/03 FROM STEVE MORRISSEY TO DICK JOLLIFFE * * *The licensee is retracting this report based upon further review. Contact the HOO for details.The licensee will inform the NRC Resident Inspector. Notified R1DO (Ron Bellamy).t Inspector. Notified R1DO (Ron Bellamy).)
ENS 40226 + (See event # 40223, Part 21 reported on 10/ … See event # 40223, Part 21 reported on 10/04/03 by GE Nuclear Energy for background information.On October 5, 2003, Progress Energy Carolinas, Inc. received notification from General Electric of a Part 21 involving the potential for numerous, unexpected confirmation count (CC) resets in the event of an instability condition. These CC resets may result in the inoperability of TS Table 3.3.1.1-1,'Reactor Protection System Instrumentation, Function 2. f, OPRM Upscale.'All Unit 1 and Unit 2 OPRM channels have been declared inoperable. Technical Specification 3.3.1.1, Action I requires an alternate method of detecting and suppressing thermal hydraulic instabilities to be implemented within 12 hours. The alternate methods of detection and suppression are currently in place.Due to the implementation of Technical Specification required compensatory actions, there is minimal safety significance.The NRC Resident Inspector was notified of this event by the licensee.Also see similar events reported by NMP Unit 2 (event # 40217) and Fermi Unit 2 (event # 40215) # 40217) and Fermi Unit 2 (event # 40215))
ENS 45810 + (Seismic Recording System activation with g … Seismic Recording System activation with ground motion readily felt by control room personnel. Plant inspections performed to date show no observable damage to systems or structures." The seismic event was reported by U.S. Geological Survey as a magnitude 7.2 event occurring 26 km (16 miles) SW of Guadalupe Victoria, Baja California, Mexico. Initial ground motion analysis on site indicated 0.027 g acceleration in the transverse direction.The licensee has notified the NRC Resident Inspector.* * * UPDATE FROM ROBERT MEYERS TO DONG PARK AT 2055 EDT ON 4/4/2010 * * *The Notification of Unusual Event classification was terminated at 1725 PDT on 4/4/2010 after completion of walk downs with no equipment issues being found. Notified R4DO (Okeefe), NRR EO (Bahadur), IRD (Gott), DHS (Vestal), and FEMA (Via). IRD (Gott), DHS (Vestal), and FEMA (Via).)
ENS 47185 + (Seismic event occurred resulting in a Noti … Seismic event occurred resulting in a Notification of an Unusual Event. No plant impact to either unit. Both units are stable with safety systems functional.There were no personnel injuries and no reports of structural damage. The licensee has notified the NRC Resident Inspector.* * * UPDATE FROM DAN WILLIAMSON TO HOWIE CROUCH AT 1819 EDT ON 8/23/11 * * *Limerick is terminating from the Unusual Event (HU5) based on the conditions that originally presented entry to the Unusual Event (HU5) no longer exist and it is unlikely that plant conditions will deteriorate. No emergency conditions exist at this time for either unit. Both Unit 1 and Unit 2 are operating at 100% power and are stable. A walk down of the facility has been completed with no deficiencies identified. No indication of system degradation has been detected.The licensee has notified the NRC Resident Inspector and has notified state and local authorities.Notified R1 IRC (Dentil), IRD (Gott), DHS (Bean), FEMA (Via), USDA (Kraus), and DOE (Turner).EMA (Via), USDA (Kraus), and DOE (Turner).)
05000413/FIN-2012009-02 + (Self-revealing findings were identified fo … Self-revealing findings were identified for the licensees failure to follow EDM-141, Procurement Specifications for Services. The licensee did not identify the need for the blocking feature for the instantaneous underfrequency protective function in both the vendor specification and the supporting information provided to the vendor. The offsite power supply to Unit 1 would have been lost anytime there was a generator trip from high power without this blocking feature. This finding resulted in an apparent violation (AV) of Technical Specification (TS) 3.8.1, AC Sources Operating, for Unit 1 and TS 3.8.1, AC Sources Operating, and TS 3.8.2, AC Sources Shutdown, for Unit 2 because the installed modification resulted in inoperability of the offsite power source for both units. Unit 2 was impacted whenever offsite power was provided from Unit 1. The finding does not represent an immediate safety concern because the licensee corrected the blocking function prior to unit restart. The violation was placed in the licensees corrective action program as PIP C- 12-3403. The performance deficiency (PD) was more than minor because it affected the availability and reliability of the Equipment Performance attribute and adversely affected the Mitigating Systems cornerstone objective in that an offsite power supply would not have been available to mitigate expected operational transients and design basis events. For Unit 1, the significance was preliminarily determined to be within the range for a finding of substantial safety significance (Yellow). For Unit 2, the significance was preliminarily determined to be within the range for a finding of greater than very low safety significance (Greater than Green). The safety significance will be designated as To Be Determined (TBD) because the safety characterization is not final. The PD was directly related to the aspect of accurate design documentation in the component of Resources in the cross-cutting area of Human Performance in that the engineering design procedures were not complete because there was no requirement for verification of non safety-related design changes.tion of non safety-related design changes.)
05000327/FIN-2011006-04 + (Sequoyah Fire Protection License Condition … Sequoyah Fire Protection License Condition 2.C.(16) for Unit 1 requires that TVA implement and maintain in effect all provisions of the approved fire protection program referenced in the Sequoyah Nuclear Plants Final Safety Analysis Report (FSAR) as described in the Fire Protection Report (FPR). Part III of the FPR states that SQN must comply with 10 CFR 50 Appendix R, Section III.G. Section III.G.2 requires that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, on January 28, 2009, the licensee identified that cables associated with the Unit 1 6900-volt power supply in fire areas FAA 1 and 29 would not be free of fire damage to support safe shutdown (see Section 40A3). This violation is of very low safety significance (Green). This issue was determined to be of very low safety significance based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase II Quantitative Screening Approach. This violation was documented in the licensees CAP as PER 162189.mented in the licensees CAP as PER 162189.)
05000327/FIN-2001007-02 + (Sequoyah Physical Security Plan, paragraph … Sequoyah Physical Security Plan, paragraph 5.3.1, Personal Searches, establishes personal search requirements for individuals entering the protected area. The licensee implements personal search requirements through Sequoyah Nuclear Plant Physical Security Instruction PHYSI-32, Security Instructions for Members of the Security ForcePHYSI-32, Rev. 24, Step 3.3.C, required that individuals entering the protected area shall be subjected to a personal search, including processing through the metal detector. If an alarm is received on the metal detector, the individual who caused the alarm shall be asked to ensure that all metal is removed (including shoes) and to process through the metal detector again. Should the individual alarm the detector again, the member of the security force shall physically search the individual.Contrary to the above, on April 19, 2000, the licensee deliberately failed to follow PHYSI-32 during the personal search of an individual entering the protected area. Specifically, a senior licensee official received an alarm from the metal detector while entering into the protected area, and a security officer did not ask him to ensure that all metal, including his shoes, was removed. The contract security officer physically searched the official instead of requesting that he remove his shoes and process through the metal detector again. process through the metal detector again.)
ENS 42910 + (Seven pairs of Tc-99m sources were receive … Seven pairs of Tc-99m sources were received in syringes in lead shielding, placed in their holders, and inventoried at 0800. At 0830 one source (15 milliCuries) was found missing. A search was conducted visually and with a radiac. The missing source was not found.THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIALSources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks.re close to it for a period of many weeks.)
05000321/FIN-2009002-04 + (Several failures of the LOCA/LOSP timer ca … Several failures of the LOCA/LOSP timer cards were identified over the period July 2008 through March 2009. The licensee determined that probable causes for these failures were degraded power supplies to these cards and potential defects identified on the timer cards. These issues were captured in the licensees corrective action program as CRs 2008107899, 2009101880 and 2009102221. The licensee performed an extent of condition review and all affected timer cards were identified, refurbished and tested satisfactorily prior to returning to them to service. The licensee is performing a root cause investigation for the LOCA/LOSP timer card failures and the contributing causes surrounding these events. The inspectors will review the root cause and determine if a performance deficiency existed and evaluate past operability of these timer cards. This issue is unresolved pending NRC review of the licensees completed root cause investigation and is designated as URI 05000366/2009002-04, Failures of theUnit-2 EDG LOCA/LOSP Timer Cardsres of theUnit-2 EDG LOCA/LOSP Timer Cards)
05000315/FIN-2002004-04 + (Several findings associated with the imple … Several findings associated with the implementation of the corrective action program were identified within the mitigating system and public radiation cornerstone areas. The inspectors determined that the six findings identified in the past 12 months indicated an adverse performance trend and had a common casual factor associated with the failure to promptly and effectively resolve conditions adverse to quality.Although the individual findings highlighted were of very low safety significance (Green) the number of findings were determined to be a substantive cross-cutting issue indicative of an adverse performance trend pertaining to implementation of the corrective action program.entation of the corrective action program.)
05000440/FIN-2008002-07 + (Severity Level IV The inspectors identifie … Severity Level IV The inspectors identified a non-cited violation of 10 CFR Part 50.72(b)(2)(iv)(B), \\\"Four Hour Reports.\\\" The inspectors determined that the licensee failed to report a manual actuation of the reactor protection system when it was not part of a preplanned sequence. Specifically, on June 22, 2007, the \\\'B\\\' reactor recirculation pump failed during a plant shutdown sequence and the licensee inserted a manual scram above preplanned power levels and not in accordance with the preplanned sequence. Licensee operators decided to insert the manual scram earlier than planned due to the unexpected loss of flow in the \\\'B\\\' reactor recirculation system loop. (Section 4OA1.b.1circulation system loop. (Section 4OA1.b.1)
05000315/FIN-2003006-04 + (Severity Level IV Violation. On May 16, 20 … Severity Level IV Violation. On May 16, 2003, the NRC issued a Notice of Violation to the licensee associated with an incident that occurred at the D. C. Cook Nuclear Power Plant on January 28, 2002. The incident involved an employee of the Framatome Corporation, a contractor at the D. C. Cook plant, that failed to follow the instructions of a radiation protection technician and subsequently failed to immediately exit the work area in the Unit 2 Containment Building when the employee's electronic dosimetry alarmed. The NRC Office of Investigations investigated the matter and concluded that the individual deliberately failed to follow radiation protection requirementsSince the violation was determined to be deliberate, the NRC did not assign a significance to the violation using the Significance Determination Process. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violation was categorized at Severity Level IV.tion was categorized at Severity Level IV.)
05000261/FIN-2017001-02 + (Severity Level IV. An NRC -identified seve … Severity Level IV. An NRC -identified severity level IV (SL IV) NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, was identified for the licensees failure to provide complete and accurate information in a license amendment request (LAR), dated November 19, 2015, requesting extension of the containment leak rate test frequencies required by various containment technical specifications (TS s). In this LAR, the licensee incorrectly stated that they had revised their ASME BPVC, Section XI, Subsection IWE program to include visual examinations of the test connections in the leak -chase channel penetration pressurization system ( PPS) , when in fact, the program had not been revised and the examinations had not been performed . This information was material to the NRC because it was used, in part, as the basis for the approval and issuance of License Amendment 247, dated October 11, 2016, extending the TS containment leak rate test frequencies. The licensees corrective actions included conducting the visual examinations of the test connections in the leak -chase channel PPS during the ongoing refueling outage in March 2017 and initiating actions to add the visual examination requirements to their Subsection IWE program. This issue was entered into the licensees CAP as NCR 02110516. The failure to provide complete and accurate information in accordance with 10 CFR 50.9(a) for the LAR associated with License Amendment 247 is a violation of NRC requirements . This violation was screened against the ROP guidance in IMC 0612, Appendix B, Issue Screening, and no associated ROP finding was identified. The inspectors evaluated this issue using the Traditional Enforcement process because it had the potential to impact the NRCs ability to perform its regulatory function. Specifically, the violation impacted the regulatory process, in that the inaccurate information was material to the NRCs review and acceptance of licensee actions to address the industry -wide operating experience discussed in NRC IN 2014- 07. Based on licensee inaccurate information that they had addressed IN 2014 -07 by revising their containment ISI program to perform visual inspections of accessible tubing in the containment leak -chase channel PPS system, the NRC staff concluded that the licensee was properly implementing the ASME BPVC, Section XI, Subsection IWE program. In accordance with the guidance in Sections 2.2 and 6.9 of the NRC Enforcement Policy, the inspectors determined this is an SL IV violation, because had the information been complete and accurate at the time provided, it likely would have resulted in the need for further clarification of the licensees actions to address NRC IN 2014- 07 , but would not have caused the NRC to change its decision to issue the license amendment or resulted in substantial further inquiry . Also, on March 23, 2017, the licensee completed the visual examinations of the subject tubing in the leak -chase channel system and did not identify any significant degradation. In accordance with IMC 0612, Appendix B, traditional enforcement issues are not assigned a cross -cutting aspect. are not assigned a cross -cutting aspect.)
05000285/FIN-2003009-01 + (Severity Level IV. Several examples of a v … Severity Level IV. Several examples of a violation of Technical Specification 5.8.1.a for the failure to follow radiation protection procedure requirements were identified. Fourteen different security officers deliberately violated applicable radiation protection procedural requirements on 62 occasions by not signing in on the required radiation work permit (RWP) 02-004 and not obtaining an electronic alarming dosimeter when assigned to the Alpha 1 security post during the period of April 27 through October 8, 2002. This violation is being treated as a Severity Level IV violation consistent with the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR- 200303574.orrective action program as CR- 200303574.)
05000390/FIN-2016003-01 + (Severity Level IV. The NRC identified a Se … Severity Level IV. The NRC identified a Severity Level IV violation of 10 CFR 50.9 Completeness and Accuracy of Information, for the failure to maintain continuous compensatory fire watch information that was complete and accurate in all material respects. The licensees actions of creating falsified fire watch completion records for the 713 elevation of the Auxiliary Building was a performance deficiency. The licensee entered this issue into the corrective action program as CR 1019953 and took remedial action against the involved individuals commensurate with the circumstances. The NRC evaluated this issue under the traditional enforcement process because it involved willfulness. In consideration of the fact that the individuals were contract fire watch personnel with minimal supervisory responsibilities, and that the underlying safety significance of the missed fire watch was low, the NRC concluded that this violation should be characterized at Severity Level IV in accordance with Section 2.2.1.d of the Enforcement Policy. Furthermore, because this violation involved willfulness and lack of supervisory oversight, the non-cited violation criteria of paragraph 2.3.2.a.4.(c) was not satisfied, such that this violation will be cited. This violation was evaluated under the traditional enforcement process and thus does not have a cross cutting aspect.thus does not have a cross cutting aspect.)
05000305/FIN-2003005-02 + (Severity Level IV. The inspector identifie … Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that the facility licensee, between January 2, 2000, thorugh August 26, 2002, submitted to the NRC, NRC Forms 396 for 13 individuals applying for an initial operator's license and 18 licensed operators applying for renewal of their operator licenses, that were not accurate in all material respects. Specifically, the NRC Forms 396 certified that each applicant and licensed operator met the medical requirements of ANSI/ANS 3.4-1983. In fact, all the applicants and licensed operators were not adequately examined for all medical tests as required to meet the minimum standards of ANSI/ANS 3.4-1983.he minimum standards of ANSI/ANS 3.4-1983.)
05000305/FIN-2003005-04 + (Severity Level IV. The inspector identifie … Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that on or about August 13, 2002, a senior facility licensee representative submitted to the NRC, NRC Forms 398 for three individuals, each applying for an initial operator's license, that were not accurate in all material respects. The facility licensee provided inaccurate information by certifying on the NRC Form 398 that the initial operator license applicaitons for three individuals had appropriately met the minimum training requirements for reactivity manipulations on the refrenced facility simulator in accordance with 10 CFR 55.31(a)(5) and 10 CFR 55.46(c)(2).10 CFR 55.31(a)(5) and 10 CFR 55.46(c)(2).)
05000382/FIN-2006009-01 + (Severity Level IV. The inspector identifie … Severity Level IV. The inspector identified a violation of 10 CFR 50.9, with two examples, for the failure to provide accurate information to the NRC associated with the Safety System Unavailably High Pressure Injection and Residual Heat Removal Performance Indicators. The performance indicator information was inaccurate because the licensee improperly concluded that the Train B high pressure safety injection and Train B containment spray systems were still available during an extended period when the containment safety injection sump suction valve was partially open. The inspector found that the licensee had underestimated the size of valve (SI 602B) opening when assessing system availability and failed to address inconsistencies between their field data, diagnostic test data and their own informal calculations. Further, a second analysis performed by a contractor (to determine the as-found valve position) was inadequate, as it contained several errors and inappropriate assumptions. The licensee also provided inadequate contractor oversight with respect to this effort. The erroneous valve position determination resulted in the licensee reporting system availability information that caused the performance indicators to be Green when the High Pressure Safety Injection System Unavailability Performance Indicator should have been Red and the Residual Heat Removal System Unavailability Performance Indicator should have been Yellow. The failure to provide accurate information to the NRC in accordance with 10 CFR 50.9 requirements was a performance deficiency. The issue had more than minor significance in that, had the information been accurate, two performance indicators would have changed color. Per the NRC Enforcement Policy, Section IV.A.3, these issues are not subject to the Significance Determination Process. The Enforcement Policy, Supplement VII, specifies that a Severity Level III violation would be appropriate for these issues. However, considering: 1) the NRC's recently implemented Mitigating Systems Performance Index program, which would have resulted in the subject performance indicators returning to the Green threshold; and 2) the risk associated with the underlying valve performance issue was of very low safety significance (Green), the NRC determined that a Severity Level IV violation was more appropriate. This finding had problem identification and resolution crosscutting aspects, in that the implementation of the licensee's Corrective Action Program did not result in a thorough evaluation of the identified condition such that information reported to the NRC was verified to be complete and accurate. was verified to be complete and accurate.)
05000255/FIN-2007002-07 + (Severity Level IV. The inspectors identifi … Severity Level IV. The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR 50.59, \\\"Changes, Tests, and Experiments,\\\" for a failure to seek a license amendment. Specifically, when Setpoint Change 96-012 involving the low suction pressure trip of the auxiliary feedwater pumps was implemented, no safety evaluation was performed. When the evaluation was performed in December 2006 the licensee failed to evaluate known deficiencies. Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the significance determination process. The performance deficiency met Supplement I.D.5, \\\"Violations of 10 CFR 50.59 that result in conditions evaluated as having very low safety significance by the SDP,\\\" for a Severity Level IV Violation. (Section 4OA5.4)erity Level IV Violation. (Section 4OA5.4))

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