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05000416/FIN-2018003-022018Q3Grand GulfMinor ViolationMinor Violation: The licensee did not include any unplanned power changes as inputs for the Unplanned Power Changes per 7,000 Critical Hours performance indicator (PI) that was reported to the NRC for the second quarter 2016. Based on a plant event that took place on June 17, 2016, the inspectors noted that the PI data submitted by the licensee may not have been accurate. In response, the licensee submitted frequently asked question (FAQ) 17-01 to the reactor oversight process working group. This FAQ resulted in the determination that three unplanned power changes should have been reported associated with the event in question. Following resolution of the FAQ, the licensee reported the associated PI data. As required by 10 CFR 50.9, Completeness and accuracy of information, information provided to the NRC by a licensee shall be complete and accurate in all material respects. Contrary to the above, from July 2016 through May 2017, information provided to the NRC by the licensee was not complete and accurate in all material respects. Specifically, the data for the Unplanned Power Changes per 7,000 Critical Hours PI did not include any unplanned power changes for the second quarter 2016. Screening: The inspectors determined that this violation was of minor significance in accordance with the NRC Enforcement Policy, Section 6.9.d.11, since the PI data in question did not ultimately result in the PI changing from Green to White. Enforcement: The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2016-06028. The licensee took action to restore compliance by submitting an appropriate correction to the PI data. This failure to comply with 10 CFR 50.9 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes URI 05000416/2017001-02, Grand Gulf Unplanned Power Changes per 7000 Critical Hours Performance Indicator.
05000416/FIN-2018003-012018Q3Grand GulfFailure to Develop Adequate Work InstructionsA self-revealed, Green finding was identified when feedwater heater drain tank oscillations caused a feedwater perturbation which required a manual reactor scram. Specifically, the licensee failed to develop appropriate work instructions for filling and venting the feedwater heater 6A level transmitters.
05000416/FIN-2018002-082018Q2Grand GulfPerformance of Surveillance Testing Following Maintenance on Containment AirlockThe inspectors identified a Green non-cited violation of 10CFRPart50,AppendixB, Criterion XI, Test Control, for the licensees failure to perform surveillance testing of containment airlock seals under appropriate conditions. The licensee failed to appropriately control the sequence of maintenance and testing activities to ensure that surveillance testing was not performed subsequent to maintenance which could affect the validity of surveillance test results.
05000416/FIN-2018002-072018Q2Grand GulfLoss of Shutdown CoolingA self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes
05000416/FIN-2018002-062018Q2Grand GulfImproper Evaluation and Resolution of Intermediate Range MonitorNoise Leads to Manual Reactor ShutdownA self-revealed, Green non-cited violation of 10CFRPart50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement appropriate corrective actions related to intermediate range monitor (IRM) nuclear instrument (NI) electronic noise spiking. The failure to implement adequate corrective actions over the course of at least 5 years resulted in a plant shutdown due to declaration of multiple IRM channels inoperable while in Mode 2.
05000416/FIN-2018002-052018Q2Grand GulfFailure to Follow Procedure Requirements Resulting in Unplanned DoseA self-revealed, Green non-cited violation of Technical Specification 5.4.1 was identified when an individual alarmed a personnel contamination monitor upon exit from the radiologically controlled area. Specifically, the licensee failed to follow procedures to establish a decontamination plan or procedure, conduct a specific pre-job brief addressing appropriate contamination risk, and receive approval by radiation protection supervision prior to conducting decontamination activities on thereactor pressure vessel(RPV) O-rings
05000416/FIN-2018002-042018Q2Grand GulfHigh Radiation Area Boundary ViolationA self-revealed, Green non-cited violation of Technical Specification 5.7.1 was identified when an individual received a dose rate alarm when the individual failed to comply with established radiological barriers and protective measures and entered a high radiation area. Specifically, an individual leaned over a high radiation area barricade rope, thereby entering the high radiation area. The individuals radiation work permit (RWP) did not permit entry into a high radiation area.
05000416/FIN-2018002-032018Q2Grand GulfFailure to Adequately Test NUS Temperature SwitchA self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred.
05000416/FIN-2018002-022018Q2Grand GulfFailure to Follow ASME Requirements for Maintaining Inservice Inspection (ISI) Cycles and Perform ASME Required Inservice Inspections within the Scheduled ISI CycleThe inspector identified 15 examples of a Green non-cited violation (NCV)of 10 CFR 50.55(a)(g)(4)(ii), which requires that inservice examination of components classified as American Society of Mechanical Engineers (ASME), Section XI, Code Class 1, Class 2, and Class 3 be conducted during successive 120-month inspection intervals, and requires compliance with the requirements of the latest edition and addenda of the ASME Code (and all its paragraphs) applicable to the specific interval, including maintaining the 120-month inspection interval in accordance with the ASME Code, Section XI, Paragraph IWA-2430. Specifically, the licensee inappropriately adjusted its second inservice inspection 120-month cycle, and failed to perform VT-3 and MT examinations of 15 class 1, class 2, and class 3 components, including the high pressure core spray pump attachment weld and reinforcing band before the third inservice inspection cycle expired on November 30, 2017, as required by 10CFR50.55(a)(g)(4)(ii).
05000416/FIN-2018002-012018Q2Grand GulfFailure to Institute Effective Corrective Action to Preclude RepetitionAn NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective.
05000259/FIN-2017003-022017Q3Browns FerryFailure to Maintain Intake Building Flood BarrierAn NRC- identified NCV of Technical Specification (TS) 5.4.1, Procedures, was identified for the failure to follow procedure MCI -0-023- PMP003, Emergency Equipment Cooling Water (EECW) and Residual Heat Removal Service Water Pump (RHRSW) Removal and Reinstallation, Revision 22. The performance deficiency is more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective. A detailed risk evaluation by a regional SRA determined the finding was Green . The licensee entered the violation into the CAP as CR 1338684. The finding had a cross cutting aspect in the Avoid Complacency component of the Human Performance area because the maintenance staff chose to not refer to a previously related condition report (CR) (PER 599190) or the maintenance procedure that were corrective actions for a previous NRC finding. (H.12).
05000259/FIN-2017003-012017Q3Browns FerryDegraded EDG Flood Door SealsAn NRC- identified non- cited violation (NCV of 10 CFR Part 50, Appendix B, Criterion V was identified for the licensee's failure to use appropriate procedural surveillance criteria to ensure the diesel generator buildings were protected against flood- water up to the design basis flood elevation. The annual door inspection procedure did not contain instructions with appropriate acceptance criteria to determine whether the diesel generator building doors would create a watertight seal when closed. The performance deficiency is more -than -minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective. A detailed risk evaluation by a regional Senior Risk Analyst ( SRA ) determined the finding was of very low safety significance (Green) . The licensee entered the violation into the corrective action program (CAP) as CR 1306268. The inspectors determined that the finding had a cross -cutting aspect in the Self -Assessment area of the Problem Identification and Resolution aspect (P.6), because recent self - assessments had not been self -critical of the external flood protection program and practices.
05000280/FIN-2017008-012017Q1SurryFailure to verify or check the adequacy of a design change in the Recirculation Spray Service Water Valve Pits.Green: The inspectors identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify or check the adequacy of the design of bulkheads in the recirculation spray service water motor operated valve pits. Specifically, the design allowed for unsealed penetrations in bulkheads and the licensee failed to demonstrate that the unsealed penetrations would not adversely affect the ability of the bulkheads to provide adequate train separation during a postulated pipe rupture. The licensee entered the issue into the CAP as Condition Report (CR) 1060189 and sealed the penetrations. This performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 3 capability to maintain train separation between the Recirculation Spray Service Water header motor operated valves was adversely affected due to the presence of degraded penetrations through the flood barriers. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.
05000281/FIN-2016004-022016Q4SurryInadequate Design Change Post Maintenance Testing Causes Water Intrusion into Station Service Transformer and a Reactor TripA self-revealing finding was identified because the test requirements section of the station service transformer (SST) design change (DC) was not comprehensive in that it did not test that the isolated phase bus ducting terminal boxes were constructed to prevent water intrusion into the boxes. This was discovered during a significant rainfall event partially caused by Hurricane Matthew, which filled up the A SST terminal box with water and eventually shorted the A phase of the main generator causing a Unit 2 main generator, main turbine, and subsequent reactor trip on October 9, 2016. As corrective action, sealant was applied to the SST terminal boxes on all seams and bolt holes; and weep holes with drain assemblies were installed on each box. This issue was documented in the licensees CAP as CR 1049987. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone, and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated October 7, 2016, the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because although the deficiency did cause a reactor trip, it did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the Operating Experience aspect of the Problem Identification and Resolution area, P.5, because the licensee did not evaluate and implement relevant external operating experience.
05000280/FIN-2016004-012016Q4SurryChange of Surveillance Frequency Caused the Charging Service Water Header to Become Biologically FouledA self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI was identified because the surveillance procedure frequency used to flush the service water (SW) piping in Mechanical Equipment Room (MER)-3 and MER-4 was changed from two weeks to four weeks without sufficiently considering the effects of river conditions on biological growth and without getting management permission to change the periodicity. As a result of the periodicity change, the B charging (CH) and main control room (MCR) SW header became blocked with biological growth and was declared inoperable on September 22, 2016, during the performance of 0-OSP-VS-012, High Flow Flush of SW Strainers and Piping in MER 3 and MER 4. As immediate corrective action, the licensee cleaned the clogged SW strainer and completed the backflushing of the SW header. The SW flushing periodicity was restored to a two week frequency to be seasonally and risk assessed and reduced as heavy fouling season ends. This issue was documented in the licensees corrective action program (CAP) as CR 1048251. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the performance deficiency (PD) was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not affect the design or qualification of the charging pump service water pump system and it did not represent a loss of system safety function. This finding has a cross-cutting aspect in conservative bias aspect of the human performance area, H.14, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowed.
05000390/FIN-2014003-032014Q2Watts BarFailure to Comply with Design Drawing Results in a Reactor TripAn NRC-identified finding was documented by the inspectors for the licensees failure to comply with a design drawing during a modification resulting in a trip of Unit 1 reactor. The inspectors determined that the licensees failure to properly implement Design Change Notice (DCN) 52295, complete bus differential wiring for main bus 2, as required by NPG-SPP-09.3, Revision 17, Plant Modifications and Engineering Change Control, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the objective of the Initiating Events cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correctly translate design drawings to implementing work order 08-816022-006 resulted in Unit 1 experiencing a 100 percent load rejection and reactor trip. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 1 - Initiating Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the resulting transient was within the design basis for Unit 1 and all plant systems functioned as required to place the unit in a stable, hot standby condition. The cause of the finding was directly related to the aspect of work management in the Human Performance cross-cutting area because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority.
05000390/FIN-2014003-012014Q2Watts BarFailure to Identify a Condition Adverse to QualityAn NRC-identified NCV of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion XVI, Corrective Action, was documented for the licensees failure to adequately identify a condition adverse to quality associated with the installation of 480 volt breaker 0-BKR-548-0021-S with non-conforming parts which was in service in safety-related 480 volt shutdown board 1B1. Immediate corrective action was to replace the non-conforming breaker. The inspectors determined that the licensees failure to adequately identify a condition adverse to quality associated with the installation of non-conforming parts as required by 10 CFR 50 Appendix B, Criterion XVI, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify the condition adverse to quality led to an additional six months that this non-conforming condition existed thus reducing the licensees ability to ensure the reliability and capability of plant safety systems. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the deficiency only affected the qualification of the breaker. The cause of the finding was directly related to the aspect of identification in the Problem Identification and Resolution cross-cutting area because the licensee did not identify this issue completely, accurately, and in a timely manner in accordance with the program.
05000390/FIN-2014003-022014Q2Watts BarFailure to Identify a Condition Adverse to QualityA NRC-identified non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was documented for the licensees failure to adequately identify a condition adverse to quality associated with the installation of relief valve 1-RFV-67- 1026D, Upper Containment Cooler 1D, an ASME Class III component. The licensee entered the issue into the corrective action program and performed an operability determination which concluded the cooler was operable. The inspectors determined that the licensees failure to adequately identify a condition adverse to quality associated with the non-conformance of relief valve 1-RFV-67-1026D, as required by 10 CFR 50 Appendix B, Criterion XVI, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct the condition adverse to quality in a timely manner led to an additional 4 years that this non-conforming condition existed prior to evaluation thus reducing the licensees ability to ensure the reliability and capability of plant safety systems. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because there existed an additional relief valve in the IST program that could protect the piping and cooler from over pressurization with appropriate compensatory measures. The cause of the finding was directly related to the aspect of evaluation in the Problem Identification and Resolution cross-cutting area because the licensee did not adequately evaluate this issue to ensure that an adequate resolution addressed the condition commensurate with its safety significance.
05000390/FIN-2014003-042014Q2Watts BarLicensee-Identified Violation10 CFR 50 Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, states, in part, that measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. Contrary to this requirement, the licensee did not ensure that a non-conforming component was properly identified and segregated as non-conforming to prevent possible use in a safety-related application. On March 22, 2014, during receipt inspection of a refurbished 6.9 kV circuit breaker per WO 115282290, breaker maintenance personnel determined that it was not in serviceable condition and initiated service requests 862089 and 862145 to document the non-conforming conditions. On March 31, 2014, the non-conforming circuit breaker was returned to inventory with two serviceable circuit breakers and made available for issue, rather than being identified and segregated as non-conforming, as required by procedure NPG-SPP-6.11. On April 4, 2014, discussions between warehouse personnel and maintenance personnel revealed that one of three 6.9 kV circuit breakers recently placed in inventory was unusable. All three circuit breakers were placed on hold and tagged as non-conforming. Once further investigation determined which circuit breaker was non-conforming, it was removed from inventory. This low safety significant adverse condition was captured in the licensees corrective action program as PER 872906.
05000390/FIN-2014003-052014Q2Watts BarLicensee-Identified Violation10 CFR 50 Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, states, in part, that measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. Contrary to this requirement, the licensee did not ensure that a non-conforming part was properly controlled to prevent its use in a safety-related application. On January 22, 2013, non-conforming main contactors were installed in a safetyrelated 480 volt breaker which was then installed in the plant on January 25, 2013. The breaker remained in service until May 20, 2014, when it was replaced. This low safety significant adverse condition was captured in the licensees corrective action program as PER 810826. (See Section 4OA2 for additional details)
05000390/FIN-2013004-012013Q3Watts BarContribution of Potential Current Transformer Imbalance to Reactor TripThe inspectors monitored the licensee performing a trouble shooting plan following the reactor trip. The licensee attributed the cause of the reactor trip to a loose phase A connection on a digital fault recorder. The inspectors continued to monitor the root cause development following reactor startup to determine the validity of the cause and review the associated LER for closure. On June 28, 2013, an A-phase high impedance ground fault occurred on the Roane 500kV transmission line approximately 22 miles from Watts Bar Nuclear Unit 1. Concurrently, the licensee experienced a reactor trip due to the actuation of the 1A Main Bank Transformer Feeder Differential Relay 187TF. The 500kV transmission line fault was caused by a tree that fell onto the A phase of the transmission line. The tree was cut by a local land owner. Operations personnel stabilized the plant using the Auxiliary Feedwater (AFW) System and the main steam dump valves. The secondaryside steam generator (SG) atmospheric relief valves, SG power operated relief valves (PORVs) and SG safety valves were not challenged during the transient. The Reactor Coolant System (RCS) responded to the initial plant transient as expected without actuating Pressurizer PORVs or initiating Safety Injection signals. Per design, a differential relay, such as the 187TF relay, should not trip due to an event occurring outside of the relays zone of protection because the input amperage subtracts from the output amperage equaling zero so no amperage is available to trip the relay. Specifically, the 187TF relay zone of protection covers the bus network between the two main generator output breakers and the 1A main bank transformer, which is within the plants switch yard, whereas the fault was 22 miles from the plant site. Initially, the licensee tried to verify that the differential circuits current transformers (CTs) were electrically balanced over their range of operation by injecting increasing levels of test amperage into the circuit. The CTs measure the amperage of the 500KV power and feed that measurement to the 187TF relay so verifying that the input CT amperage properly subtracts from the output CT amperage would validate the CTs characteristics. Because the technicians injecting the amperage did not disable the 86 relay, all of the remaining circuit breakers connected to the X bus opened. The 86 relay detects if any breaker on the bus fails to open when called upon. The 86 relay operation had no significant effects on the shutdown of the plant, but because the technicians did not secure the circuit properly, TVA management decided to stop the verification of the CTs. Alternatively, the technicians tried to verify circuit connections by physically moving the wiring. While handling an A phase wire connected to one of the Digital Fault Recorders, a technician stated that he noticed about 1/8 inch of movement and heard a click. The click is the locking mechanism that ensures the connection remains secure; however the design of the connector ensures electrical connection over one inch of movement. Before the 86 relay tripped the remaining bus relays, the amperage injection had passed approximately 2 amps through this portion of the circuit, which would have detected a loose connection. Inspectors observed that TVA stopped assessing other electrically significant reasons that could have tripped the 187TF relay such as CT imbalances. The root cause team determined that the click heard by the technician was the cause of the relay trip even though subsequent bench testing could not support it. In response to inspector questions, the licensee hired a 3rd party consultant, which also discounted the connector as an obvious cause of the 187TF relay trip. Both the inspectors and the 3rd party consultant believe that neither of the licensees troubleshooting techniques nor their root cause analysis has adequately addressed the cause of the 187TF relay trip, which can continue to challenge the reactor protection system on subsequent high impedance ground faults outside of the plant. The licensee plans to disable the 187TF relay during the next shutdown for refueling, in the spring of 2014, in order to measure the current sensed by the relay as the main generator load is decreased for shut down. This item is identified as unresolved item (URI) 050000390/2013004-01, Contribution of Potential Current Transformer Imbalance to Reactor Trip.
05000400/FIN-2012008-012012Q2HarrisB AND C MSIVs FAIL TO CLOSE DURING SURVEILLANCE TESTINGThe inspectors identified an URI associated with issues in the licensees MSIV maintenance and testing. These issues were potential contributing causes to the April 21, 2012, B and C MSIV failure to stroke close. Description: Several issues were identified regarding the licensees MSIV maintenance and testing. Some of the issues identified were: FnIn the last two refueling intervals, maintenance was making minor adjustments to the actuator hydraulic speed control system to decrease the time needed to shut the valves as a result of increasing stroke test closure time results. FnBeginning in 2001, work deficiency documents were initiated due to the MSIVs experiencing difficulty in opening during refueling outage cycling. There had not been any corrective maintenance conducted requiring valve internal disassembly and the licensee had not developed any periodic PMs to visually inspect the condition of valve internals. FnThe valve vendor manual recommended weekly valve partial exercising ten percent of its total stroke in order to assure that the actuator and valve was properly functioning. Prior to 2000, this partial exercising was being performed quarterly. In 2000, the licensee revised their IST program requirements to discontinue quarterly exercising in lieu of the 18-month cold shutdown TS stroke testing that was currently being conducted. FnPrior to the current MSIV failures; the MSIVs had never been tested as part of the licensees AOV program. Summary: The licensees root cause investigation was not completed at the conclusion of the special inspection; the determination as to whether these issues represented performance deficiencies was not completed. Pending completion of the licensees root cause evaluation (RCE) and subsequent NRC review to determine if a performance deficiency exists, disposition of these issues will be tracked via Unresolved Item (URI) 05000400/2012008-01, B and C MSIVs Fail to Close During Surveillance Testing.
05000321/FIN-2010003-032010Q2HatchFailure to follow procedure while in shutdown cooling to record corrected reactor water levelThe NRC identified a NCV of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe in procedure 34GO-OPS-015-2, Maintaining Cold Shutdown or Refueling Condition, appropriate documented instructions for recording and verifying reactor water level when reactor vessel level is greater than 60 inches and instrument 2B21-R605 is unavailable. To address this issue the licensee performed the immediate corrective action of initiating CR 2010104615 and has generated an action item to upgrade procedure 34GO-OPS-015-2. This performance deficiency is more than minor because it is associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability of systems (ability of operators to monitor, trend, and maintain reactor water level) to prevent undesirable consequences. Because this finding is associated with the safety of a reactor while the unit was in cold shutdown and on residual heat removal shutdown cooling, NRC IMC 0609, Attachment 4, directs using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, to determine the significance of this finding. In Appendix G, Attachment 1, Checklist 6 was used because during the time period of this finding the unit was in cold shutdown, with a time to boil < 2 hours, and reactor coolant system level < 23 feet above the top of the reactor vessel flange. Each item in Appendix G, Attachment 1, Checklist 6 was determined to have been met, therefore per Figure 1 of Appendix G this finding screened as GREEN significance because a Qualitative Assessment was not required by Checklist 6. This finding has a cross-cutting aspect in the Work Control component of the Human Performance area, because the licensee did not plan and coordinate work activities consistent with nuclear safety including planned contingencies, compensatory actions, or abort criteria. Specifically, the licensee did not plan and coordinate the activity of transitioning the reference leg for reactor water level instrument 2B21-R605 with contingencies, compensatory actions, or abort criteria addressed to ensure measurable reactor water level was available to control room operators.
05000321/FIN-2010003-012010Q2HatchFailure to maintain safety related cables in a nonsubmerged environmentThe NRC identified a NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure implement measures to assure that safety-related cables remained in an environment for which they were designed. Safety-related cables purchased and installed in underground electrical pull boxes at Hatch Nuclear Plant have been subjected to submergence, a condition for which they are not designed. To address this issue the licensee has performed the immediate corrective action of increasing the frequency of measuring water level and pump down of the pull boxes. The licensee initiated CR 2010104298 to address this issue. This performance deficiency is more than minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it is reasonable to conclude the cables may be in a degraded condition where the continued reliability of the cable cannot be ensured because: 1) the licensee does not have a cable testing/monitoring program to detect degradation of inaccessible or underground power cables; 2) the cables have been subject to a submerged physical environment which is outside the cables design parameters; and 3) there have been documented failures of cables throughout the nuclear industry due to degradation caused by submergence in water. Because the finding affects the safety of an operating reactor, the significance of this finding was screened using the Phase 1 of the SDP in accordance with NRC IMC 0609, Attachment 4, Table 4a. The finding screened as Green, because the finding is a design or qualification deficiency confirmed not to result in loss of operability or functionality. This finding has a cross-cutting aspect in the Work Control component of the Human Performance area, because the licensee did not appropriately coordinate activities by incorporating actions where maintenance scheduling is more preventive than reactive. Specifically, the licensee did not schedule performance of procedure 52PM-Y46-001-0, Inground Pull Box and Cable Duct Inspection for Water, at a frequency that prevented safety related cable submersion
05000321/FIN-2010003-062010Q2HatchLicensee-Identified ViolationTS 3.4.3 requires 10 of 11 SRVs to be operable during Modes 1, 2 and 3. Contrary to the above, on March 11, 2010 on Unit 1 it was identified during bench testing that five safety relief valves failed to lift at the required TS setpoint. The cause was found to be corrosion induced bonding between the pilot disc and seating surface. This condition was documented in CR 2010103338. This finding is of very low safety significance because a previous evaluation performed by the licensee bounds this condition and RCS pressure would be maintained below the TS safety limit.
05000321/FIN-2010003-052010Q2HatchLicensee-Identified ViolationTS 5.4.1.a requires written procedures be established, implemented and, maintained covering the activities specified in Regulatory Guide 1.33, Appendix A. Items 2g and 4a of Appendix A requires procedures for general power operation and operation of the reactors recirculation system to be established and implemented. Contrary to the above, Unit 2 operated at a core flow higher than that allowed on the power/flow map described in licensee procedure 34GO-OPS-005-2, Power Changes. This issue was documented in the licensees corrective action program as CR 2009108237. Because the finding is associated with the fuel barrier and sufficient fuel thermal limit margin was maintained during the time core flow was outside the bounds of the power/flow map, this finding is of very low safety significance.
05000321/FIN-2010003-042010Q2Hatch1A EDG fuel oil return line failure

An unresolved item (URI) was opened related to CR 2010104391, fuel oil return line fitting failure on the 1A EDG. As of the end of this inspection period the licensee had not completed their investigation into this issue. The determination of a performance deficiency cannot be made until the licensee completes and documents their inspection efforts in this area

On April 1, 2010 a fitting leak on the 1A EDG fuel oil return line was identified by the licensee during a monthly surveillance test run. CR 2010104391 documents that an attempt was made to tighten the fitting but the leak continued and that the leak needed to be repaired after the monthly surveillance test run was complete. The leak on this fitting was added to existing WO 1092436001 and scheduled for the 1A EDG system outage in May 2011. On June 3, 2010 during the monthly surveillance test run, the tubing associated with this fitting failed and diesel fuel oil was identified spraying onto the 1A EDG exhaust. The surveillance test run was terminated and the 1A EDG was secured by local operators in order to prevent a fire from starting. The failure of the 1A EDG fuel oil tubing was documented in CR 2010107248. URI 05000321/2010003- 04, 1A EDG fuel oil return line failure, will be identified to track this issue pending review of the investigation conducted under CR 2010107248 to evaluate whether this issue constitutes a performance deficiency.

05000321/FIN-2010003-022010Q2HatchFailure to follow corrective action program procedure and prevent recurrence of severity level 2 root causeA self-revealing NCV of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to follow their corrective action program procedure, NMP-GM-002, Ver. 4.0, that required severity level 1 and 2 condition reports (CR) to have corrective actions that prevent recurrence. From May 2006 to April 2010 licensee procedure NMP-GM-002, Corrective Action Program, Ver. 4.0, was not followed because corrective actions to prevent recurrence were not implemented prior to failure of Analog Transmitter Trip System (ATTS) card 1B21-N641C. The licensees immediate corrective actions were to replace the failed card, 1B21-N641C, the adjacent card 1B21-N690C and the high drywell pressure trip cards 1E11-N694A and C. The licensee initiated CR 2010105161 to address this issue. The performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the Initiating Events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure of ATTS card 1B21-N641C resulted in a spurious Loss of Coolant Accident (LOCA) signal that started Emergency Core Cooling System (ECCS) equipment and resulted in a power reduction to approximately 85%. Due to this finding affecting the safety of an operating reactor, the significance of this finding was screened using NRC IMC 0609, Attachment 4, Table 4a. Because the finding contributed to the likelihood of a reactor scram, but did not affect mitigation equipment availability, the finding screened as Green. The inspectors concluded that the performance deficiency does not have an associated cross-cutting aspect because the performance deficiency occurred in 2006 and is not indicative of the licensees current performance in the area of root cause investigations.
05000321/FIN-2009005-012009Q4HatchSubmerged safety-related medium voltage cablesAn unresolved item (URI) was opened related to underground pull box inspections which revealed a safety-related 4160 volt cable located in two pull boxes was submerged under water. The determination of a performance deficiency cannot be made until further information is provided by the licensee to support that the cables are designed, qualified, and acceptable for operation in a wetted and/or submerged environment. On December 10, 2009 during inspection of underground bunkers subject to flooding, the inspectors identified that safety-related 4160 volt cable, R22-S005-ES1- M08, located in pull boxes PB1-BF and PB1-BB was submerged. This issue was captured in the licensees corrective action program as CR 2009111808. The inspectors require documentation supporting the cables design, qualification, and testing history to evaluate whether this issue constitutes a performance deficiency. URI 05000321,366/2009005-01, Submerged safety-related medium voltage cable was identified to track this issue
05000341/FIN-2009005-012009Q4FermiFailure to Adhere to Self-imposed Maintenance Rule Procedural RequirementsA finding of very low safety significance (Green) was identified by the inspectors for the licensees failure to adhere to self-imposed maintenance rule procedural requirements. Contrary to maintenance rule monitoring requirements, while performing a maintenance rule functional failure evaluation of a diesel fire pump functional failure, licensee personnel inappropriately determined the failure was not maintenance preventable and assigned an incorrect causal code to the event prior to completing and documenting a cause analysis. These actions led to a delay in licensee recognition that the fire protection system had exceeded its performance criteria; therefore, presentation of the systems condition to the maintenance rule expert panel for a(1) consideration was delayed by several months. No violation of NRC requirements occurred. The licensee entered this item into their corrective action program (CAP) as condition assessment and resolution document (CARD) 09-28649. The licensees immediate actions included correction of the causal code for the maintenance rule functional failure and initiation of the a(1) evaluation process for the affected system. The system was placed in a(1) status on January 12, 2010. The finding was determined to be more than minor in accordance with IMC 0612, Appendix E, Example 4.a, in the not minor section, because procedural noncompliance in the maintenance rule area continues to be an issue at Fermi. In addition, if left uncorrected, the failure to adhere to procedures could have the potential to lead to a more significant safety concern. Specifically, not following procedural requirements specified in the Fermi Maintenance Rule Conduct Manual has lead to a failure to monitor degraded equipment in an a(1) status as required. The finding affected the Mitigating Systems cornerstone and was determined to be of very low significance (Green), because the finding was a procedural compliance issue that was confirmed not to result in loss of operability or functionality; it did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events. This finding has a cross-cutting aspect in Human PerformanceWork Practices human error prevention techniques due to a failure to use error prevention techniques during the assessment of the failure (H.4(a))
05000341/FIN-2009003-032009Q2FermiFailure to Adequately Control Potential Debris Source Term in Primary ContainmentThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion III (Design Control), for the failure to control debris source term inside the drywell. The licensee installed ty-wraps inside the drywell as part of a design modification without performing a debris transport and loading analysis of the emergency core cooling system (ECCS) suction strainers in the torus. Once identified, the licensee performed the analysis and replaced the ty-wraps with ones of an acceptable material. The finding was determined to be more than minor because the failure to control the debris source term inside the primary containment could lead to loss of the ECCS during an accident condition. Specifically the debris could be transported from the drywell to the torus and cause the ECCS strainers to become blocked causing degradation in the ECCS flow during the accident and, therefore, affected the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance because the engineering analysis determined the ECCS flow rates would remain above the values assumed in the safety analysis and the debris loading did not exceed the structural limits of the strainers. There were no cross-cutting aspects associated with this finding since the deficiency was not reflective of current licensee performance
05000341/FIN-2009003-022009Q2FermiFailure to Adequately Dedicate a Commercial Grade Item for Safety Related UseA Green self-revealing finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, was identified for the failure to adequately dedicate a commercial grade item for use in a safety-related application. The vendor supplied a mismatched stem and locknut in a valve rebuild kit which was procured as a commercial grade item and dedicated by the licensee for use in a safety-related application. The valve later failed when the locknut fell off the stem which caused the system to be inoperable. The finding was determined to be more than minor because the finding was associated with the design control attribute and affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance from a Phase 1 SDP because it only affected the loss of function of one division of non-interruptible air supply system (NIAS) for less than the Technical Specification allowed outage time. There were no cross-cutting aspects associated with this finding since the deficiency was not reflective of current licensee performance
05000341/FIN-2009003-042009Q2FermiLicensee-Identified ViolationAs described in Section 4OA2.3 of this report, during routine surveillance testing on February 1, 2009, the Division 1 RHR torus suction isolation valve, E1150F004A, would not stroke properly from the control room. Troubleshooting identified loose wires in the MCC bucket that were the cause of the failure. The licensee secured the loose wires, ensured there were no additional loose wires, completed a stroke test of the valve, and returned it to service the following day. The licensee entered this issue into their CAP as CARD 09-20637 and completed an apparent cause evaluation. The licensee concluded that the wires most likely came loose during previous diagnostic testing post-maintenance activities which did not incorporate steps to ensure that leads disturbed during the test were tight prior to closing out the MCC bucket. The licensee revised the applicable maintenance procedure to include such an inspection. 10 CFR 50,Appendix B, Criterion V requires, in part, that activities affecting quality be properly preplanned and performed in accordance with procedures appropriate to the circumstances. Contrary to the above, licensee procedure 35.306.009 was not appropriate to the circumstances because it did not ensure that potentially disturbed leads were tight and secure following maintenance. The licensee entered this issue into their CAP as CARD 09-20637. This issue screened as Green because there was no loss of safety function
05000369/FIN-2009006-012009Q2Mcguire
McGuire		Inadequate Procedure for RN System Flow BalancingThe team identified a finding of very low safety significance involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to provide adequate procedures for flow balancing of the service water (RN) system. The RN flow balance procedure was inadequate in that it made no provision in the acceptance criteria to limit or evaluate minimum flow control valve seat/disc clearance, and subsequent potential for increased flow obstruction, resulting from system flow balancing. The licensee entered this deficiency into their corrective action program (CAP) for resolution. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, changing position of the flow control valves without consideration of potential flow obstruction could impact the capability to adequately cool safety related equipment. The team assessed this finding for significance in accordance with the SDP for Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green), in that no actual loss of safety system function was identified. No cross-cutting aspect was identified because the performance deficiency did not reflect current performance
05000369/FIN-2009006-022009Q2Mcguire
McGuire		Failure to Correctly Translate Design Basis Information Related to the Isolation Time for Safety Related MOVs into Instructions and ProceduresThe team identified a finding of very low safety significance involving a NCV of 10 CFR50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that the applicable design bases were correctly translated into the in-service test (IST) acceptance criteria for safety-related motor operated valves (MOVs). Specifically, the licensees testing did not account for test inaccuracies associated with limit switch actuation or minimum EDG frequency into IST stroke time testing. The licensee entered this deficiency into their CAP for resolution. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Not accounting for test inaccuracies and EDG under frequency, the IST did not ensure that MOV isolation times referenced in the Updated Final Safety Analysis Report (UFSAR) were verified by testing. The team assessed this finding for significance in accordance with the SDP for Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green), in that no actual loss of safety system function was identified. No cross-cutting aspect was identified because the performance deficiency did not reflect current performance
05000369/FIN-2009006-032009Q2Mcguire
McGuire		Inadequate Verification of the Design Adequacy of the Control Circuit Voltage for 600 VAC Safety Related MotorsThe team identified a finding of very low safety significance involving a NCV of 10 CFR50, Appendix B, Criterion III, Design Control, for failure to establish measures to verify the design capability of the control circuit voltage for 600 VAC safety related motors fed from motor control centers. Specifically, there was no voltage drop calculation or cable configuration specification for the control circuits that established the adequacy of the control circuit to energize the safety related motors. The licensee entered this deficiency into their CAP for resolution. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Due to the lack of appropriate analysis, the 600V motor control circuit design basis accident capability was not assured and further evaluation was required to demonstrate that the equipment could perform its safety function. The team assessed this finding for significance in accordance with the SDP for Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green), because it was a design deficiency determined not to have resulted in the loss of safety function. No crosscutting aspect was identified because the performance deficiency did not reflect current performance
05000341/FIN-2009003-012009Q2FermiFailure to Completely Disassemble and Remove Scaffold from the Steam TunnelThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V (Procedures), for the failure to follow procedures. The licensee partially dismantled a scaffold and left the remaining scaffold poles in place which was contrary to the licensees scaffold procedure. Once identified, the licensee removed the scaffold materials and entered the issue into their corrective action program for resolution. The finding was determined to be more than minor because if left uncorrected, it would become a more significant safety concern. Specifically, the scaffold components represented potential high energy line break induced missiles which could have damaged components that were required to remain operable to mitigate the event and, therefore, affected the Mitigating Systems Cornerstone. This finding was determined to be of very low safety significance because the phase 3 SDP estimated the change in core damage frequency due to the finding was 3.8E-7/yr. This finding had a cross-cutting aspect in the area of human performance, work practices, because the licensee did not utilize human error prevention techniques H.4(a), such as self-checking and proper documentation of activities.
05000361/FIN-2009002-012009Q1San OnofreFailure to Follow Procedures when Performing Maintenance on the Auxiliary Feedwater SystemA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the failure of maintenance personnel to follow maintenance order instructions to fully remove fuses to establish conditions necessary to perform valve testing on the auxiliary feedwater system. Instead of removing the fuse entirely from the fuse holder, maintenance personnel only removed one side of the fuse and left the other side inserted. This inappropriate maintenance practice caused plastic deformation on the associated side of the fuse holder, which impacted the design configuration of the auxiliary feedwater control system, and its ability to perform its required design function under all design basis accident conditions. This finding was entered into the licensees corrective action program as Nuclear Notification 200253911. The finding is greater than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the issue using the Significance Determination Process (SDP) Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barriers Cornerstones provided in Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that this finding represented a loss of the system safety function for the ability to provide secondary heat removal during a station blackout. This required that a Phase 2 estimation be completed. Because the Phase 2 assumptions significantly overestimated the risk related to this finding, the senior reactor analyst conducted a Phase 3 evaluation to provide a bestestimate risk assessment. The analyst calculated that a total ACDF of 4.4 x 10-7, therefore this finding is of very low risk significance (Green). The finding has a crosscutting aspect in the area of human performance associated with work practices because maintenance personnel did not comply with expectations regarding procedural compliance to follow the procedure as written without deviating from its intent (H.4(b))
05000341/FIN-2009002-012009Q1FermiInadequate Procedural Controls Over Construction of Storage Racks and Storage AreasA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to include criteria in procedures for evaluation of storage areas and storage racks built in the power block. Licensee procedure MOP11, Combustible Material, placed controls on the storages areas and storage racks to ensure that combustible loading remained acceptable but failed to incorporate adequate guidance for designating the storage area and constructing the racks to ensure nearby safety-related equipment would not be adversely affected during a plant transient or seismic event. After the issue was raised, modifications to the scaffold storage locations were completed, as needed. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control (plant modifications) and it adversely impacted the cornerstone objectives. As a result of not evaluating the storage areas, safetyrelated components, systems or structures could have been affected. This finding was determined to be of very low safety significance because it did not result in loss of operability or functionality. The inspectors determined that the finding had an associated cross-cutting aspect of Problem Identification and Resolution, Corrective Action Program, Corrective Action (P.1 (d).
05000361/FIN-2009002-022009Q1San OnofreFailure to Properly Implement the Operability Determination ProcessThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure of operations and engineering personnel to follow procedures and adequately evaluate degraded conditions to support operability decision-making. Specifically, operations and engineering personnel failed to adequately evaluate the operability of the Unit 2 component cooling water system Train B, when a tube leak was identified, and subsequently, when the tube exhibited a degrading trend. This finding was entered into the licensees corrective action program as Nuclear Notification 200289984. The finding is greater than minor because the degraded component cooling water heat exchanger is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not result in a loss of safety function of component cooling water Train B for greater than the technical specification allowed outage time. The finding has a crosscutting aspect in the area of human performance associated with decision-making because the licensee did not review past operability decisions to verify the validity of the underlying assumptions (H.1(b))
05000361/FIN-2009002-032009Q1San OnofreFailure to Properly Inspect Scaffolding in Safety-Related AreasThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of maintenance personnel to properly install and inspect scaffolding in safetyrelated areas in accordance with written procedural requirements. Four instances were found where the minimum separation distance between a scaffold and safety-related components was less than the minimum allowed by procedure and an approved engineering evaluation to justify the deviation was not performed. The licensee evaluated the scaffolds and modified them as necessary. This finding was entered into the licensees corrective action program as Nuclear Notification 200356209. The finding is greater than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The inspectors concluded this finding was associated with the Mitigation Systems Cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not affect both trains of any single mitigating system or represent an actual loss of a safety function. This finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee did not utilize appropriate human performance techniques to ensure that scaffold construction was performed safely (H.4(a)) (Section 1R18)
05000361/FIN-2009002-072009Q1San OnofreFailure to Follow Procedure for Aligning a Reactor Coolant System Ion ExchangerA self-revealing non-cited violation of Technical Specification 5.5.1.1 was identified for the failure of operations personnel to follow procedures to place Ion Exchanger ME074 in service which resulted in an interruption of letdown flow and diversion of approximately 160 gallons of reactor coolant to the radiological waste system. This finding was entered into the licensees corrective action program as Nuclear Notification 200319240. The finding is greater than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and affects 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. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee did not properly use human error prevention techniques (H.4(a))
05000361/FIN-2009002-062009Q1San OnofreFailure to Follow Procedures to Reassemble a Reactor Coolant System Pressure Retaining ComponentA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the failure of work control and maintenance personnel to follow the procedure requirements for work on a reactor coolant system pressure retaining component. Specifically, work control and maintenance personnel did not use work documents and procedures to reassemble the vent valve for the control element drive mechanism associated with control element Assembly 22, which resulted in a reactor coolant system leak during the fill and vent process. This finding was entered into the licensees corrective action program as Nuclear Notification 200323460. The finding is greater than minor because it is associated with the reactor coolant system equipment and barrier performance attribute of the Barrier Integrity Cornerstone and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process, Checklist 4, the finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory by more than two feet when not in a mid loop operation. This finding has a crosscutting aspect in the area of human performance associated with work control because the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of work on different job activities (H.3(b))
05000361/FIN-2009002-052009Q1San OnofreFailure to Properly Evaluate Boric Acid Leakage from the Reactor Coolant Pump Vapor SealThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of engineering personnel to properly perform an evaluation of reactor coolant pump vapor seal boric acid accumulation caused by a clogged vapor seal drain line, in accordance with boric acid corrosion control program procedures. Specifically, engineering personnel failed to follow the requirements of Procedures SO23-XV-85 and SO23-XV-8.15 to properly evaluate the impact of boric acid leakage to reactor coolant system pressure boundary components. This finding was entered into the licensee\'s corrective action program as Nuclear Notification 200258836. The finding is greater than minor because if left uncorrected, excessive boric acid buildup would have a potential to lead to a more significant safety concern. The finding is associated with the Initiating Events Cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding would not result in exceeding the technical specification limit for reactor coolant system leakage and would not have affected other mitigation systems resulting in a total loss of their safety function. The finding has a crosscutting aspect in the area of human performance associated with decision-making because engineering personnel did not use conservative assumptions to identify possible unintended consequences associated with the identified boric acid accumulation (H.1.(b))
05000361/FIN-2009002-042009Q1San OnofreInadequate Procedure for Reactivity ManipulationsThe inspectors identified a non-cited violation of Technical Specification 5.5.1.1 for the failure of operations personnel to follow procedures for performing reactivity manipulations. Specifically, a procedure modification performed to Procedure SO23-3-2.19.2, Control Element Assembly Exercise and Troubleshooting, was inaccurate and incomplete to appropriately control reactivity manipulations, and thus, an adequate procedure was not in hand as required by Procedure SO123-O-A1, Conduct of Operations, to appropriately control the control element assembly manipulations by a licensed operator. This finding was entered into the licensees corrective action program as Nuclear Notification 200339686. The finding is greater than minor because it is associated with procedure quality attribute of the Initiating Events Cornerstone and affects 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. Using Manual Chapter 0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process, Checklist 4, the finding is determined to have very low safety significance because the finding did not increase the likelihood of a loss of reactor coolant system inventory, degrade the ability to terminate a leak path, or degrade the ability to recover decay heat removal. This finding has crosscutting aspect in the area human performance associated with work control because the licensee did not appropriately plan a work activity (H.3.(a))
05000341/FIN-2009002-022009Q1FermiCore Spray Pump Motor Oil Leak Not Proptly Identified and CorrectedA finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed for the failure to promptly identify and correct an oil leak that subsequently rendered a safety-related pump inoperable. Maintenance staff discovered an oil leak near a safety-related pump and informed Operations staff of the leak but the licensee failed to identify the source of the leak for 5 days and, therefore, failed to take prompt corrective actions. Once identified, the licensee repaired the damaged instrument tube and restored the pump to service. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance 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 oil leak ultimately rendered the pump inoperable. The finding was determined to be of very low safety significance because a Phase 2 SDP determined the risk to be very low. This finding had an associated cross-cutting aspect of Problem Identification and Resolution, Corrective Action Program, Issue Identification (P.1 (a))
05000341/FIN-2008005-032008Q4FermiInadequate Degraded Voltage Protection SchemeAs documented in Inspection Report 05000341/2008008, the inspectors identified a concern with respect to the adequacy of the degraded voltage protection scheme. In correspondence dated June 11, 1981, the licensee described the time delay settings for the degraded voltage relays; however, they did not discuss the impact of receiving a safety injection signal during degraded voltage conditions. At that time, the NRC did not identify this oversight, and had subsequently determined the degraded voltage scheme was acceptable. As documented in Inspection Report 05000341/2008008, the inspectors determined that the current degraded voltage protection scheme is inadequate in that the time delay relay settings for the degraded voltage relays for both divisions could impact the emergency core cooling system injection timing requirements. Additionally, for a short period of time under degraded voltage conditions, voltage could be too low for proper operation of safety-related motors but high enough to prevent EDG start. After further review, the inspectors determined that the provisions of 10 CFR 50.109(a)(4), were applicable and that a modification was necessary to bring a facility into compliance with the rules or orders of the NRC. The licensee was requested to respond with a description of the intended actions to address the noncompliance including a proposed schedule to complete those actions. In a letter dated August 4, 2008, from J. Plona (ADAMS Accession No. ML0822502561), the licensee stated the installation of the modification which will bring the plant into full compliance with General Design Criteria 17 will be completed by the end of the fourteenth refueling outage in 2010. As stated in Inspection Report 05000341/2008008, this issue was not dispositioned as a violation because the NRC had accepted the inadequate degraded voltage protection scheme in 1981 and in 1985. However to ensure actions are completed to correct the condition, this issue is considered open pending completion of the licensees modification. (VIO 05000341/2008005-03
05000341/FIN-2008005-012008Q4FermiInadequate Heat Exchanger inspection FrequencyA finding of very low safety significance was identified by the inspectors for the failure to test and/or inspect the safety-related non-interruptible air supply (NIAS) control air compressor (CAC) aftercoolers in accordance with Generic Letter (GL) 89-13 commitments. The licensee inspected the heat exchangers every 10 to12 years which was not in accordance with their GL 89-13 commitment to frequently inspect them. This finding was entered into the licensees corrective action program (CAP) as condition assessment and resolution document (CARD) 08-27672. Corrective actions planned included changing the frequency to comply with the licensees GL 89-13 commitment. No violation of regulatory requirements occurred. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because the finding was confirmed not to result in a loss of operability or functionality. No cross-cutting aspect was assigned because this issue is not indicative of current plant performance. (Section 1R07.1(1)
05000341/FIN-2008005-022008Q4FermiInadequate Control Air Compressor Capacity Test ProgramA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, was identified by the inspectors for the failure to perform adequate testing for both the Division 1 and 2 CACs. Specifically, the licensee failed to incorporate appropriate acceptance criteria and failed to include appropriate test methodology in the test procedures for the safety-related CACs. Corrective actions planned included revising and/or re-performing the test procedures as necessary. The finding was more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because the finding was confirmed not to result in a loss of operability or functionality. No cross-cutting aspect was assigned because this issue is not indicative of current plant performance. (Section 1R22.1(1)
05000341/FIN-2008004-032008Q3FermiInadequate 10 CFR 50.59 Evaluation for Reactor Building Missile ProtectionThe inspectors identified a Green (Severity Level IV) NCV for an inadequate 10 CFR 50.59, Changes, Tests, and Experiments, evaluation resulting in failure to receive prior NRC approval for changes in licensed activities associated with protection of safety-related equipment against tornado generated missiles. Specifically, the licensee failed to demonstrate that the proposed change did not result in an increase in the probability of a malfunction of equipment important to safety previously evaluated in the Updated Final Safety Analysis Report (UFSAR). As part of the corrective actions, the licensee installed missile shields and initiated a study to determine the appropriate long-term corrective actions. The finding was greater than minor because the change had the potential for impacting the NRCs ability to perform its regulatory function as the inspectors determined the change would have required prior NRC approval. Based on a phase 3 significance determination, the senior risk analyst determined the finding was of very low safety significance because the change in core damage frequency for this finding was calculated to be less than 1.0E-7. This was determined to be a Severity Level IV NCV of 10 CFR 50.59(a)(2)(i) (1989). (Section 4OA2.3)
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