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05000341/FIN-2013003-012013Q2FermiNot Following Plant Documents for Switchyard ModificationsThe inspectors identified a finding of very low safety significance for the licensees failure to follow the augmented quality program (AQP), nuclear plant operating agreement (NPOA), and Updated Final Safety Analysis Report (UFSAR) for plant modifications installed in the 345-kilovolt (kV) and 120-kV switchyards by the International Transmission Company (ITC) around September 2011. Specifically, the ITC liaison did notify his counterpart at Fermi of the planned installation of new equipment in the switchyards, but no condition assessment resolution document (CARD) was issued or other communication made to Fermi 2 plant support engineering to conduct the required evaluation of proposed design modifications. In addition, no 10 CFR 50.59 review was performed of proposed changes to a modification. The finding was determined to be more than minor because the inspectors did not see a similar example in IMC-0612, Appendix E, Examples of minor issues. Further, because the licensee (nor ITC) had performed any design evaluation to assure the proposed activity would not have an adverse impact on the plant, the inspectors concluded that if left uncorrected this failure to perform a systematic design process in accordance with the AQP, NPOA, and UFSAR could lead to more significant safety concerns. Therefore, the issue screened as being more than minor. The inspectors evaluated the significance of the finding using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings at-Power, Exhibit 1 Initiating Events Screening Questions, and answered no to the Transient Initiators question, Does the finding contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions will not be available? Therefore, the issue screened as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not properly coordinate with ITC on the switchyard work to ensure the requirements of the AQP, NPOA, and UFSAR were met.
05000341/FIN-2013003-032013Q2FermiFailure to Maintain Configuration Control During Plant OperationThe inspectors identified a finding of very low safety significance for the licensures failure to maintain configuration control during plant operations. Specifically, the inspectors identified multiple instances concerning the improper storage of equipment and control of scaffolding from January 1 through June 30, 2013. These instances did not meet the requirements of several licensee programs and management expectations. The multiple instances constitute a programmatic issue with configuration control. This issue is more than minor because if left uncorrected would lead to a more significant safety concern and is similar to Inspection Manual Chapter (IMC) 0612, Appendix E, Section 4, Example a, in that the licensee routinely failed to perform procedurallyrequired engineering evaluations on similar issues. Specifically, multiple examples were identified where the licensee placed items in the plant without proper engineering evaluation. The inspectors evaluated the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at-Power, Exhibit 1 Initiating Events Screening Questions, and answered no to the Transient Initiators question, Does the finding contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions will not be available? Therefore, the issue screened as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee either failed to follow established procedures or removed the controls from applicable procedures.
05000341/FIN-2013002-012013Q1FermiLicensee-Identified ViolationTitle 10 CFR 50.65, Maintenance Rule, section (b)(2)(i), states, in part, that the scope of the monitoring program...shall include...nonsafety-related structures, systems, and components...that are...used in the plant emergency operating procedures. Contrary to the above, system C9600, the integrated plant computer system (IPCS), was not incorporated into the scope of the Maintenance Rule until CARD 11-31237 identified the need to perform maintenance rule scoping for the system. The IPCS provides safety parameter display system (SPDS) information to the operators in the main control room and the emergency response facilities, including the technical support center, emergency operations facility, operations support center, and the virtual private network appliance which transmits a subset of the SPDS information to the NRC. On January 30, 2012, system C9600, IPCS had been re-scoped as a nonsafety-related system that was explicitly utilized in the emergency operating procedures, and therefore, C9600 was incorporated into the Fermi maintenance rule program. The inspectors determined this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). This finding was determined to be of very low safety significance because all the screening questions in IMC 0609, Attachment 04, Table 4a, for the Mitigating Systems Cornerstone were answered no.
05000341/FIN-2012005-012012Q4FermiFailure of E4150F002A finding of very low safety significance was self-revealed for failing to adequately inspect and identify, and then correct severe degradation of the motor operator for E4150F002 (HPCI turbine steam supply inboard containment isolation valve), which failed on July 23, 2012, when operators were attempting to place the high pressure coolant injection (HPCI) system into standby. The failure analysis of the motor identified the severe degradation. The apparent cause evaluation team identified three apparent and contributing causes for the severe degradation: first, prolonged moisture from steam leaks or other water sources; second, improper end ring coatings; and third, failing to identify a degraded condition during a video probe inspection. A finding of very low safety significance was self-revealed for failing to adequately inspect and identify, and then correct severe degradation of the motor operator for E4150F002 (HPCI turbine steam supply inboard containment isolation valve), which failed on July 23, 2012, when operators were attempting to place the high pressure coolant injection (HPCI) system into standby. The failure analysis of the motor identified the severe degradation. The apparent cause evaluation team identified three apparent and contributing causes for the severe degradation: first, prolonged moisture from steam leaks or other water sources; second, improper end ring coatings; and third, failing to identify a degraded condition during a video probe inspection. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, appropriate corrective actions aspect because the licensee failed to adequately inspect and identify, and then correct severe degradation of the motor operator for E4150F002.
05000341/FIN-2012004-022012Q4FermiInadequate Evaluation of Steam Dryer/Steam Separator Lifting DeviceA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the NRC inspectors for the failure to ensure the adequacy of the steam dryer/steam separator lifting device design. Specifically, the inspectors identified four examples where the licensee failed to perform adequate evaluations of the structural elements The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the purpose of the lifting device design requirements was to limit the likelihood of a structural component failure, and hence, to ensure safe load handling of heavy loads over the reactor core or over safety-related systems. The inspectors determined the finding was of very low safety significance following a qualitative significance determination process review performed by the Region III Senior Risk Analyst. The inspector did not identify a cross-cutting aspect associated with this finding because the concern was related to a calculation from the 1980s, and thus was not necessarily indicative of current licensee performance.
05000341/FIN-2012004-032012Q4FermiFailure to Perform ASME Inservice Testing Comprehensive Pump Test RequirementA finding of very low safety significance and an associated NCV of 10 CFR 50.55a(f), Inservice testing requirements, and 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, was identified by the NRC inspectors. Specifically, the licensee failed to perform a required comprehensive pump test for division 1 and 2 emergency equipment cooling water makeup pumps within 2 years of the start of the third inservice testing interval. The third inservice testing interval commenced on February 17, 2010, and included a requirement to perform a comprehensive pump test for the division 1 and 2 emergency equipment cooling water makeup pumps within two years and every two years thereafter. The required comprehensive pump tests were not performed prior to February 17, 2012. The finding was determined to be more than minor because the finding was associated with the configuration control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective of ensuring the capability of systems to prevent undesirable consequences (i.e., core damage). This finding was determined to be of very low safety significance because, following IMC 0609, Appendix E, Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, all questions were answered no. This finding has a cross-cutting aspect in the area of Human Performance, Decision Making, supervisory and management oversight aspect because the licensee failed to appropriately oversee the development and implementation of the comprehensive pump testing.
05000341/FIN-2012004-042012Q4FermiFailure of Control Rod 10-35 to Fully Scram during Scram Time TestingA self-revealed finding of very low safety significance and an associated NCV of 10 CFR 50 Appendix B, Section V, Instructions, Procedures, and Drawings, was identified for the failure to adequately prevent foreign material from entering the hydraulic control unit for control rod 10-35, which caused control rod 10-35 to fail to fully insert on October 24, 2010. Subsequently, on November 18, 2011, control rod 10-35 again failed to fully insert during scram time testing. The root cause team identified the presence of foreign organic material and concluded it had been present for a long time, i.e., at least since or prior to 2006, and this material was the cause of the deficient operation of control rod 10-35 in October 2010 and November 2011. The inspectors determined this finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective of ensuring the capability of systems to prevent undesirable consequences (i.e., core damage). This finding was determined to be of very low safety significance because, following IMC 0609, Appendix E, Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, all questions were answered no. There was no cross-cutting aspect for this finding and NCV because the foreign material entered hydraulic control unit 10-35 sometime prior to 2006; and, therefore, the foreign material exclusion program inadequacies do not represent current performance.
05000341/FIN-2012004-052012Q4FermiLicensee-Identified ViolationFermi TS 5.7.1(b) for high radiation area (HRA) controls states, in part, Entry into such areas...may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them. Contrary to the above, on April 6, 2012, two individuals entered an HRA in the torus room without being briefed on their radiological conditions. This issue was documented in the licensees corrective action program in CARD 12-22833. Immediate corrective actions included briefing the workers involved on their radiological conditions and verifying their individual accumulated radiological exposure. The finding was determined to be of very low safety significance because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised.
05000341/FIN-2012004-012012Q4FermiInspection Procedure for Reactor Pressure Vessel Head Strongback and Steam Dryer/Separator Lifting Device Omitted Testing RequirementsA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified by the NRC inspectors. Specifically, the licensee failed to perform dimensional testing of the reactor pressure vessel head strongback and the steam dryer/steam separator lifting device required by American National Standards Institute (ANSI) N14.6-1978. In addition, the license failed to perform nondestructive testing of steam dryer/steam separator lifting device major load carrying welds and critical areas required by ANSI N14.6-1978. These issues were entered into the licensees corrective action program. The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the purpose of the dimensional testing of reactor pressure vessel head strongback and steam dryer/steam separator lifting device and nondestructive testing of the steam dryer/steam separator lifting device major load carrying welds and critical areas is to limit the likelihood of a reactor pressure vessel head strongback or steam dryer/steam separator lifting device structural component failure, and hence, to ensure safe load handling of heavy loads over the reactor core or over safety-related systems, structures and components. The inspectors determined the finding was of very low safety significance following a qualitative significance determination process review performed by the Region III Senior Risk Analyst. The inspector did not identify a cross-cutting aspect associated with this finding because the concern was related to licensing basis established in the 1980s, and thus was not necessarily indicative of current licensee performance.
05000341/FIN-2012005-022012Q4FermiInadequate Implementation of Overhaul Post-maintenance Testing and Operation of South Reactor Feed Pump TurbineA self-revealed finding of very low safety significance and associated NCV of Technical Specification 5.4.1.a was identified for the licensees failure to establish and implement procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, the licensee failed to control the three factors identified by the root cause evaluation team within their refueling outage (RF)-15 south reactor feed pump turbine (SRFPT) overhaul maintenance instructions and post-maintenance testing instructions; and within the operating procedures for the reactor feed pumps during synchronizing the main generator to the electrical grid following recovery from repairs performed on main unit transformer 2B. The south reactor feed pump (SRFP) catastrophically failed, and as a result, the reactor was shut down because of decreasing condenser vacuum. The inspectors determined the failure to control the presence of three factors in concert: (1) no turbine diaphragm alignment with tight clearances; (2) automatic admission of steam with challenging thermal properties; and (3) less than adequate post-maintenance testing, was a performance deficiency that required evaluation using the SDP. The inspectors determined this finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability. This finding was determined to be of very low safety significance because, following IMC 0609, Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded the finding did not require quantitative assessment. Therefore, the finding was determined to be of very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, supervisory and management oversight aspect because the licensee failed to appropriately oversee the overhaul of the SRFPT by a vendor, and the post maintenance testing and operation of the SRFPT during and after RF-15.
05000341/FIN-2012005-032012Q4FermiLicensee-Identified ViolationTitle 10 CFR 50.65, Maintenance Rule, section (a)(1) requires, in part, that holders of an operating license shall monitor the performance or condition of structures, systems, or components within the scope of the rule as defined by 10 CFR 50.65 (b), against licensee established goals, in a manner sufficient to provide reasonable assurance such structures, systems, or components are capable of fulfilling their intended functions. Contrary to the above, the system engineer for system T2300 primary containment (torus-to-reactor vacuum breakers) failed to perform evaluations of various CARDs that documented as-found conditions outside the torus-to-reactor vacuum breaker acceptance criteria to determine whether maintenance rule functional failures had occurred. The maintenance rule expert panel had determined the T2300 system should be monitored as (a)(1) at the time. CARD 11-30255 was issued for this concern, and the functional failure evaluations were performed. This finding was determined to be of very low safety significance because all the screening questions in IMC 0609, Attachment 04, Table 4a, for the Mitigating Systems Cornerstone were answered no.
05000341/FIN-2012005-042012Q4FermiLicensee-Identified ViolationA finding of very low safety significance (Green) and associated violation of 10 CFR, Part 50, Appendix B, Criterion III, Design Control was identified by the licensee for the failure to ensure the ECCS mode of operation of RHR would be capable of performing its mitigating function in mode 3 following RHR realignment from its shutdown cooling mode of operation. Specifically, the operability requirements of RHR in mode 3, as defined by TS 3.5.1, were not translated into applicable procedures or specifications of the system in that neither the procedures nor the design prevented the condition that would lead to steam void formation during a loss of coolant accident that initiates at this mode resulting in steam binding of the systems pumps and/or an adverse water hammer. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. A Phase II SDP was conducted using IMC 0609, Appendix G. The finding screened as very low safety significance. The licensee entered this concern into its corrective action program as CARD 12-24503 and initiated a Condition Evaluation for TS 3.5.1 ECCS Operating may be non-conservative. In the interim, the licensee has implemented actions to declare the division of RHR inoperable when used in the shutdown cooling mode of operation in mode 3. The safety function is maintained by the other division of RHR. The licensee plans to evaluate the BWR Owners Group analysis of the postulated mode 3 loss of coolant accident scenario and implement permanent procedural, design, and/or licensing basis changes as necessary.
05000341/FIN-2012003-022012Q2FermiEnergizing Bus 65E with Ground Truck Installed and Subsequent Loss of Shutdown CoolingA self-revealed Green finding and associated NCV of 10 CFR 50 Appendix B, Section V, Instructions, Procedures, and Drawings, for failure to follow procedures when the licensee energized a safety-related electrical bus with a ground truck installed in bus 65E breaker position E4. This resulted in the loss of the safety-related bus and a temporary loss of shutdown cooling. The licensee failed to comply with sequence step 61 of Safety Tagging Record 2012-001122, which had connected a ground truck in bus 65E position E4 and installed a red danger tag. The Operations Conduct Manual, Chapter 12 (MOP12), 3.6.2 specifies that red tagged equipment is not to be operated. The licensee entered this item into their corrective action program as CARD 12-23118. The inspectors determined this finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). This finding was determined to be of very low safety significance because, following IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklist for both PWRs and BWRs, concluded the finding did not require quantitative assessment. Therefore, the finding was determined to be of very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, supervisory and management oversight aspect because the licensee failed to appropriately oversee the proper clearance of Safety Tagging Record 2012-001122
05000341/FIN-2012003-032012Q2FermiFailure to Monitor Reactor Pressure during Reactor Pressure Vessel Hydrostatic TestA self-revealed Green finding and associated NCV of Technical Specification (TS) 5.4.1.a was identified for the licensees failure to establish and implement procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, the licensee failed to control reactor pressure in the band specified in the reactor pressure vessel hydrostatic test procedure. A valid high pressure reactor scram actuation was received after operators failed to recognize that the reactor pressure vessel pressure instrument being monitored became inaccurate. Immediately after the scram, operators stabilized the plant at approximately 600 psig and reset the reactor scram. The licensee entered this issue into their corrective action program as CARD 12-23824. The inspectors evaluated the finding using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process - Phase 1 Operational Checklists for Both Power Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The inspectors consulted Checklist 8, BWR Cold Shutdown or Refueling Operation; Time to Boil > 2 Hours: RCS Level < 23\' Above Top of Flange. The inspectors determined the finding did not adversely impact any shutdown defense-in-depth or mitigation attributes on the checklist, nor did it meet any of the checklist specific requirements for a Phase 2 or Phase 3 SDP analysis. Consequently, the finding was determined to be of very low safety significance. This finding has a cross-cutting aspect in the area of human performance, work practices component, because the licensee failed to use human error prevention techniques commensurate with the risk of the assigned task, such that activities are performed safely. Specifically, the licensee failed to monitor the specified primary instrumentation for critical plant parameters.
05000341/FIN-2012003-012012Q2FermiControl Rod 10-35 Failure to ScramOn October 24, 2010, control rod 10-35 failed to insert upon actuation of an automatic reactor scram caused by loss of condenser vacuum (CARD10-29509). An emergent issue team was formed to investigate this event. The apparent cause was determined to be a hydraulic lock caused by blockage in the flow path between the control rod drive mechanism and the scram discharge volume. The investigation never found any foreign material, but postulated that the likely foreign material was discharged into the scram discharge volume, ultimately ending up in the torus room sump. As a corrective action, for cycle 15 the licensee increased the frequency of performing TS surveillance SR 3.1.4.2 scram time testing to every 100 days, adjusted the representative sample size to assure all rods would be tested during cycle 15, and included control rod 10-35 in each quarterly scram time testing sample. On November 18, 2011, the control rod failed to fully insert during scram time testing. The rod was fully inserted and remained there for the rest of the cycle. The inspectors are waiting for the licensees evaluation of this event, specifically their conclusions regarding the foreign material found, and their evaluation of how the foreign material could have been present, causing the first event, but migrated to allow successful scram time testing on November 11, 2010, and the first three quarters of 2011 before finally causing the failure identified on November 18, 2011. Because the licensee had not completed their evaluation, this issue is being treated as an unresolved (URI) item.
05000346/FIN-2012002-042012Q1Davis BesseInadequate Control of Locked High Radiation Area KeyTS 5.7.2(a)(1) requires that High Radiation Areas with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation be provided with a locked or continuously guarded door, gate, or other barrier that prevents unauthorized entry, and in addition, that the door and/or gate keys to these areas be maintained under the administrative control of the shift supervisor, radiation protection manager, or his/her designee. Contrary to this requirement, on February 15, 2012, licensee personnel failed to properly control the key to a Locked High Radiation Area vault storing a high integrity container loaded with primary resin. Specifically, a Radiation Protection (RP) technician checked out the subject key at the beginning of the work shift in order to access the Locked High Radiation Area vault for a planned evolution. At the end of the shift, the RP technician failed to return the key to the appropriate secure key storage cabinet, instead leaving it in an unsecured desk drawer. Several hours later when the key was identified as being missing, the RP technician, who had left the plant, was contacted and the key was recovered. At no point during the time the key was uncontrolled was the Locked High Radiation Area vault, which can only be accessed by the removal of a twenty-two ton cover, opened and improperly accessed. The objective of the Occupational Radiation Safety Cornerstone of Radiation Safety is to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. A key attribute of this objective is human performance, and specifically, procedure use and adherence. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. The licensees failure to appropriately control the key to a Locked High Radiation Area vault storing a high integrity container loaded with primary resin per established plant procedures resulted in the potential for unauthorized access to a High Radiation Area with a dose rate greater than 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation. The licensee had entered this issue into their CAP as CR 2012-02489. Corrective actions planned or completed by the licensee include the performance of a formal apparent cause evaluation, enhancements to procedural controls for Locked High Radiation Area keys, and additional training for RP personnel.
05000341/FIN-2012002-012012Q1FermiLicensee-Identified ViolationTS 5.4.1 requires the licensee to establish, implement, and maintain applicable written procedures for the safety-related systems and activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Section 9.a, Procedures for Performing Maintenance, of Regulatory Guide 1.33, Revision 2, Appendix A, further states, in part, that: Maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Revision 20 of the licensees Work Conduct Manual, MWC10, Work Package Preparation, describes requirements for configuration control. Step 4.10.2.3.e states, Upon completion of the maintenance activity or prior to completing the work package (work activity), all temporary alterations shall be removed and the equipment/SSCs shall be returned to the As-Designed condition. Contrary to the above, on February 22, 2012, the licensee failed to properly restore the configuration of Division 1 H2O2 sample pump following maintenance. Specifically, the pump discharge tubing hoses were left unconnected causing the system to trip when it was attempted to be restarted. The inspectors reviewed this issue using the guidance contained in Appendix B, Issue Screening, of Inspection Manual Chapter 0612, Power Reactor Inspection Reports. The inspectors determined the violation was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Configuration Control and affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the as-found condition of the H2O2 sample pump discharge tubing potentially introduced a leakage path from the primary containment to the secondary containment. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, using the Phase 1 Significance Determination Process worksheet for the Barrier Integrity Cornerstone. The finding screened as very low safety significance (Green) because the inspectors answered No to the screening questions under the Containment Barrier column of Table 4a. Specifically, because maintenance had installed Swagelok fittings on the ends of the discharge tubing, an actual open pathway in the physical integrity of the primary containment did not exist. The licensee had entered this issue into their corrective action program as CARD 12 21428. A local leak rate test was performed for the as-found condition which measured the leakage at 28.1 scfh (standard cubic feet per hour), which added a small amount to the existing primary containment total leakage rate (70.44 scfh). The total leakage rate remained below the TS 3.6.1.1 limit of La (296.3 scfh).
05000341/FIN-2012002-022012Q1FermiLicensee-Identified ViolationTS 3.3.1.2, Table 3.3.1.2-1 requires functional testing of TS 3.3.1.2. Table 3.3.1.2-1 requires functional testing of the source range monitors (SRMs) to be conducted within 12 hours following shutdown. Enclosure A, Section 2.E.1 of MWC 13, Outage Nuclear Safety specifies requirements that, in Mode 4, at least three SRM channels are maintained operable. Enclosure D, Risk Assessment, of MWC 13 requires written risk management actions to operate with less than three operable SRMs. Contrary to the above, on March 26, 2012, the licensees operators failed to properly assess the risk impact of losing three SRMs in Mode 4 due to a failure of the SRM drive-in pushbutton in accordance with Title 10 CFR 50.65(a)(4). Specifically, the licensee did not properly recognize the risk impact on the outage defense-in-depth requirements of declaring three SRMs inoperable that led to orange nuclear safety risk for reactivity management. The inspectors reviewed this issue using the guidance contained in Appendix B, Issue Screening, of IMC 0612, Power Reactor Inspection Report. The inspectors determined the violation was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Configuration Control and was similar to the not-minor-if statement of example 4.e of IMC 0612, Appendix E, Examples of Minor Issues. Specifically, the inoperability of three SRMs placed the overall plant risk in a higher licensee-established risk category. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding screened as very low safety significance (Green)
05000346/FIN-2012002-032012Q1Davis BesseAdditional Emergency Diesel Generator Inoperability Caused by Inadequate Maintenance Procedure InstructionsTS 5.4.1(a) requires the licensee to establish, implement, and maintain applicable written procedures for the safety-related systems and activities recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A. Section 9.a, Procedures for Performing Maintenance, of RG 1.33, Revision 2, Appendix A, further states, in part, that: Maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to this requirement, on January 26, 2012, licensee personnel failed to properly connect a strip chart recording device needed to support a planned TS surveillance on EDG No. 2. Specifically, the improper connection on the recording equipment caused test data essential to the completion of the TS surveillance to be lost, which resulted in the need to perform the surveillance a second time. This additional performance of the surveillance added significant time to the periods of inoperability and unavailability for EDG No. 2, and caused the licensee to make an unplanned entry into an elevated (i.e., Orange) plant risk awareness state. Upon investigation into the matter, the licensee identified that the applicable maintenance procedure controlling the connection of the strip chart recording equipment only contained detailed connection instructions for the test connections on the EDG itself; the proper configuration for the test connections on the recording equipment was not specified within the procedure, but instead was left to the skill and knowledge of the technician performing the equipment setup. The objective of the Mitigating Systems Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A key attribute of this objective is human performance, and specifically, procedure quality. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. The licensees failure to provide technically adequate written procedures and instructions for the connection of the strip chart recording device needed for the EDG No. 2 TS surveillance resulted in the need to perform that surveillance a second time and added significant time to the periods of inoperability and unavailability for EDG No. 2. The licensee had entered this issue into their CAP as CR 2012-01367. Corrective actions planned or completed by the licensee include revision to the EDG TS surveillance procedure to provide enhanced details on the proper connection of the strip chart recording device.
05000346/FIN-2012002-022012Q1Davis BesseFailure to Maintain SAFETY-RELATED DC Systems Design ControlThe inspectors identified a finding, with two examples, of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to maintain the electrical separation of the redundant safety-related direct current (DC) systems in compliance to the design and licensing bases. The licensee initiated corrective actions including opening the breakers to the non-safety-related loads inside containment and setting the automatic transfer switches (ATSs) to prevent auto-transfer of loads. The performance deficiency was determined to be more than minor because the issue was associated with the Mitigating Systems Cornerstone attribute of design control, 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 failure to address the impact of high-impedance ground faults in non-safety equipment on safety-related DC sources and the failure to maintain compliance to RG1.6 when installing ATSs between redundant DC power sources impacted the reliability of the DC power system. The inspectors evaluated the finding to be of very low safety significance (Green) using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations. Using the Phase 1 SDP worksheet for the Mitigating Systems Cornerstone, the inspectors answered no to all five screening questions. Based on the date of occurrence of this violation (more than 20 years old), the inspectors did not identify a cross-cutting aspect as the finding was not representative of current performance.
05000346/FIN-2012002-012012Q1Davis BesseSeismic Instrumentation Unavailable for Emergency Event ClassificationThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 50.54(q) for failing to follow and maintain an emergency plan that meets the requirements of emergency planning standard 10 CFR 50.47(b)(4). Specifically, the licensee failed to maintain configuration control of seismic instrumentation necessary for the declaration of emergency events. The seismic instrumentation was out of service without the knowledge of the on-shift operating crew and no compensatory measures were in place. The licensee entered this performance deficiency into their corrective action program (CAP) as condition report (CR) 2012-01950 and CR 2012-01984. The inspectors determined that the issue was a performance deficiency as it was within the licensees ability to foresee and correct. This finding was determined to be more than minor because it was associated with the emergency response organization (ERO) performance attribute of the Emergency Preparedness Cornerstone. This finding affected the cornerstone objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding is of very low safety significance because it did not result in the loss or degradation of a risk significant planning standard. One Alert and one Notification of Unusual Event Emergency Action Level (EAL) initiating condition would have been rendered ineffective such that a seismic event would have been declared in a degraded manner. This finding was also associated with the cross-cutting area of human performance. Specifically, the licensees work control process failed to appropriately control work on the seismic monitoring system. This resulted in a loss of configuration control and of instrumentation necessary to classify a seismic event without compensatory measures in place.
05000341/FIN-2011005-012011Q4FermiFailure to Develop Appropriate Corrective Actions for a Maintenance Rule (a)(1) Monitored System.The NRC inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 50.65 for failure to develop appropriate corrective actions for an (a)(1) monitored system. The licensee failed to determine the cause of repeated SS-1 computer and printer lock ups in the D1100 process radiation monitor system. They determined the D1100 SS-1 computer should be monitored as (a)(1) status, and established (a)(1) monitoring goals, established a get-well plan, and implemented their plan. However, the get-well plan corrective actions failed to meet the (a)(1) monitoring goals and further inspection revealed the weaknesses in the causal determination and the ineffectiveness of the corrective actions. The inspectors determined this finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). This finding was determined to be of very low safety significance because all the screening questions in IMC 0609, Attachment 04, Table 4a, for the Mitigating Systems Cornerstone were answered no. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action, problem evaluation aspect because the licensee failed to appropriately evaluate the causes of the D1100 SS-1 computer problems.
05000341/FIN-2011005-022011Q4FermiPlacing a Fuel Bundle in the Wrong Cell During Fuel ShuffleA self-revealed finding of very low safety significance (Green) was identified by the inspectors for placing a fuel bundle in the wrong cell during a fuel shuffle in the spent fuel pool. The error was noted later in the fuel shuffle when another bundle was moved to the same location, and the operators noted that the cell was filled. Specifically, on November 1, 2011, movement of spent fuel in the Spent Fuel Pool was taking place in preparation for testing of boron concentration in the high density racks. While performing step 150 of the approved MES32003, Special Nuclear Material/Component Transfer Form, the presence of a fuel bundle already occupying the target location (4N-12) for step 150 was self-revealed. The Refuel Floor Coordinator was informed, and the bundle was returned to its original starting location. This issue was placed in the licensees corrective action program as CARD 11-29841, Fuel Move Error in Spent Fuel Pool. The inspectors determined that this finding was more than minor because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. This finding was determined to be of very low safety significance because all the screening questions in IMC 0609 Attachment 0609.04 Table 4a, Characterization Worksheet for IE, MS, and BI Cornerstones were answered no . This finding had a cross-cutting aspect in the area of human performance, work practices because the licensee failed to provide direct licensed operator oversight (H.4(c)) of fuel handling operations in the spent fuel pool.
05000346/FIN-2011012-012011Q3Davis BesseUnable to Locate Fatigue Analysis for Class I ValvesThe fatigue monitoring program is an existing program which, when enhanced, will be comparable to Section X.M1, Fatigue Monitoring, of the GALL Report. The fatigue monitoring program manages fatigue of select primary and secondary components, including the reactor vessel, reactor internals, pressurizer, and steam generators by tracking thermal cycles. The program provides an analytical basis for confirming the actual number of cycles does not exceed the number of cycles used in the design analysis and the cumulative usage is maintained below the allowable limit or that appropriate corrective actions are taken to maintain component cumulative fatigue usage below the allowable limit during the period of extended operation. During the application review, NRR staff raised concerns related to document discrepancies; the most-limiting locations may not have been analyzed; actual transient severity versus design transient severity; use of an inspection program versus a preventative action program; establishing program acceptance criteria; review of operating experience; and cycle counting. The concerns were documented in RAIs B.2.16-1 through B.2.16-7. In Letter L-11-166, which responded to RAIs B.2.16-1 through B.2.16-7, the applicant answered NRR RAIs and committed to enhancements to evaluate additional plant-specific component locations that may be more limiting than those considered in NUREG/CR-6260; provide for updates of the fatigue usage calculations if the allowable cycle limit is approached; and establishing acceptance criteria for maintaining the cumulative usage below the Code design limit of 1.0 throughout the period of extended operation. The inspectors reviewed revised program documentation and verified these enhancements were included. The inspectors also reviewed other program documentation, the existing program (including the transient status log from the most-recent outage), condition reports, and interviewed the applicants staff responsible for the program. The inspectors noted NRR, during their review of the application, had identified the licensee could not locate the fatigue analysis for Class I valves. This issue is a current licensing basis issue and is considered an Unresolved Item pending further review (URI 05000346/2011012-01) Unable to Locate Fatigue Analysis for Class I Valves. While not directly part of the fatigue monitoring program, during their review of the program, the inspectors noted that in LRA Table 3.3.1, Item 3.3.1-01, the applicant stated that, for steel cranes, fatigue analysis is a TLAA, and further evaluation is documented in LRA Section 3.3.2.2.1. In LRA Section 3.3.2.2.1, the applicant indicated that fatigue TLAA evaluations are addressed in Section 4. However, there is no discussion of fatigue TLAAs of steel cranes in LRA Section 4. The applicant issued OIN-378 to track completion of the changes required to the LRA and license renewal program documents in order to document disposition of steel crane cycles as TLAAs. While reviewing the existing program, the inspectors noted that in Attachment 3 of the program, for transient 30, the column Estimated Date to Reach Limit is marked as N/A. This is incorrect, as Pressurizer Spray Nozzle is a monitored transient. The applicant issued SAP Notification 600704256 to track the correction of this error. The inspectors concluded that implementation of the enhanced Fatigue Monitoring Program will provide reasonable assurance that the aging effects due to cyclic fatigue will be managed so that the program components will continue to perform their intended function, consistent with the current licensing basis, for the period of extended operation.
05000346/FIN-2011004-052011Q3Davis BesseCode Surface Examination Requirements Not Applied to Closure Head Stud HolesThe RVCH is a single piece forging fabricated to the SA 508 material standard with an ASME NPT stamp to document that this pressure boundary part was fabricated to the requirements of the 1989 Edition of the ASME Code Section III. The requirements for examination of forgings are contained in the ASME Code Section III, Article NB 2540 Examination and Repair of Forgings and Bars. Specifically, NB-2541(a) requires in part that, In addition, all external surfaces and accessible internal surfaces shall be examined by a magnetic particle (MT) method (NB 2545) or a PT method (NB-2546). Also, NB-4121.3 Repetition of Surface Examinations After Machining required If, during the fabrication or installation of an item, materials for pressure containing parts are machined, then the Certificate Holder shall re-examine the surface of the material in accordance with NB-2500 when: (a) the surface was required to be examined by the MT or liquid penetrant method in accordance with NB-2500; and (b) the amount of material removed from the surface exceeds the lesser of 1/8 inch or 10 percent of the minimum required thickness of the part. For the 60, 7-inch diameter stud holes drilled through the vessel head flange, no surface examinations (e.g., MT or PT) were conducted on the interior bore surfaces of the stud holes. The inspectors observed a licensee demonstration of the potential accessibility of the flange stud holes for MT examination. Specifically, a licensee MT qualified examiner positioned an AC yoke used for MT examinations on the interior bore surfaces of an RVCH flange stud hole. Based on this demonstration, the inspectors estimated that it would be possible to perform an MT exam for accessible portions of the interior bore surfaces for a depth of about 2 inches from the top and bottom flange faces for each of the 60 stud holes. Because this accessible interior surface on the RVCH forging had not been examined by MT or PT, the inspectors were concerned that the RVCH did not meet the requirements of NB-2541(a) and NB-4121.3 discussed above. In response to the inspectors questions, the licensee established a position that accessible interior surfaces of the RVCH stud holes did not require a surface examination. The licensee position was based on Code Interpretation III-1-77-162, which states in part that drilled holes are not considered to be material form surfaces and the requirement for examination of holes (if any) resides in NX-4000 and NX-5000. The licensee concluded that the reexamination of machined surfaces as discussed in NB-4121.3 did not apply to the accessible interior surfaces of the flange stud holes because they were not material form surfaces. This issue is considered an unresolved item pending completion of an NRC staff review to determine an Agency position on the licensees interpretation of these Code requirements. The licensee documented this issue in CR 2011-01739.
05000346/FIN-2011004-012011Q3Davis BessePlant Transient During HPI Flow Instrument String ChecksOn September 15, 2011, instrumentation and controls (I&C) technicians replaced the HPI 3A and 3B flow instrument signal monitors with refurbished modules. Upon insertion of the module into the cabinet, the control room received an unexpected alarm for ICS Input Mismatch. The alarm immediately cleared and was attributed to a slight disruption in voltage when the modules were inserted. A decision was made to continue replacement activities. On September 16, 2011, I&C technicians commenced PMT of the signal monitors. During the string check of the HPI flow instrument alarms, annunciator alarm 14-4-E, ICS Input Mismatch, was received. The alarm initially cleared, then returned. Coincident with ICS Input Mismatch alarm, the plants ICS began reducing reactor power without any operator input. On-watch plant operators entered procedure DB-OP-02526, Primary to Secondary Plant Upset, and went through actions of placing ICS stations in manual control. The I&C technicians performing the HPI flow instrument signal monitor refurbishment were directed to stop their activities. Reactor power initially dropped to approximately 95 percent before operators stabilized the plant, and then returned reactor power to approximately 100 percent using manual controls. The refurbished HPI flow instrument signal monitor modules were removed from the system and taken to the I&C shop for inspection and testing, while the original signal monitor modules were reinstalled. Inspection and testing of the refurbished modules in the I&C shop did not reveal any issues. The modules have been sent to the licensees testing laboratory for further analysis. The inspectors continued to review the circumstances surrounding the event to determine if the issue was within the licensees ability to foresee and correct and should have been prevented. Pending further review of the licensees cause analysis, the issue is considered an unresolved item.
05000341/FIN-2011004-012011Q3FermiFailure to Maintain Separation of Metal Containers and Combustible RadwasteA finding of very low safety significance and an associated NCV of the Fermi 2 Facility Operating License Condition 2.C(9),for the fire protection program, was identified by the inspectors for the licensees failure to ensure combustible radwaste was not stored with spent charcoal filter material and HEPA filters. Specifically, the licensee failed to ensure the radwaste combustible material for the cleanup of the December 2010 resin spill was not in the same storage area as the metal containers in the on-site storage facility as required by Updated Final Safety Analysis (UFSAR) Chapter 11, Radwaste Waste Management, Section 7.2.2.4, Onsite Storage Facility, Fire Protection. This issue was placed in the licensees corrective action program as CARD 11-28704, NRC Issue with Resin Storage in the Offsite Storage Facility. The site has taken action to separate the material as required by the UFSAR. The finding was more than minor because if left uncorrected, the storage of the combined material in bay 1 and bay 4, could lead to a more significant safety concern in that the potential for an unplanned radiation release was possible. The licensee was using the area for storage of the metal containers and normal combustible radwaste. A fire in this area of the plant has the potential to affect radioactive material. The finding affected the Public Radiation Safety Cornerstone, Radioactive Material Control Program. Screening under IMC 0609, Appendix D, Public Radiation Protection Significance Determination Process was required. Based on a review of Appendix D, the inspectors concluded that the exposure received would be less than 0.005 rem total effective dose equivalent. Therefore, the finding screened to very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Control, because the licensee failed to coordinate work activities between Radiation Protection and Fire Protection groups to ensure combustible material was not stored with the metal containers in accordance with the UFSAR.
05000346/FIN-2011004-022011Q3Davis BesseFailure to Control ECCS Room Cooler Valve PositionA finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified by the inspectors for the licensees failure to control the configuration of the emergency core cooling system (ECCS) room cooler service water (SW) outlet valves in accordance with procedures. Specifically, the licensee failed to update procedures used to set the appropriate throttle position for the valves, and by using information tags to control valve position, failed to follow plant status control procedures. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems Cornerstone attributes of Design Control and Configuration Control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, an incorrect throttle position of the ECCS room cooler outlet valves could have an effect on the reliability or availability of ECCS train 2 equipment. A past operability review determined that the as-found flowrate to ECCS room coolers 1 and 2 was reduced with outlet valves SW87 and SW103 mispositioned, however, the flow was sufficient to not affect the operability of ECCS room coolers 1 and 2. Using the Phase 1 SDP worksheet for the Mitigating Systems Cornerstone, the finding screened as very low safety significance (Green) because the inspectors answered No to the screening questions in Table 4a. Specifically, the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of Human Performance, Resources component, because the licensee did not ensure that personnel, equipment, procedures, and other resources are available and adequate to assure nuclear safety. Specifically, the licensee did not process a document change request to update procedures used to verify SW valve alignments.
05000346/FIN-2011004-032011Q3Davis BesseFailure to Take Timely Corrective ActionsA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Procedures, Instructions, and Drawings, were identified by the inspectors for the licensees failure to correct deficiencies, deviations, and/or nonconformances associated with safety-related systems, structures, and components (SSCs) in a timely manner, as required by the licensees Quality Assurance Program Manual (QAPM) and CAP implementing procedure. Specifically, the inspectors identified a trend on the part of the licensee to leave certain low significance/low priority corrective actions for various safety-related SSCs completely unscheduled and unaddressed, in some cases for extensive periods of time that ranged up to 8 years. The licensee initiated their own review to determine the full extent of condition of this issue, and entered the issue into their CAP as CR 2011-00385. The finding, which was associated with the Mitigating Systems Cornerstone, was determined to be of more than minor significance because the issue represented a programmatic deficiency associated with the licensees CAP that if left uncorrected would have the potential to lead to a more significant safety concern. Using the Phase 1 SDP worksheet for the Mitigating Systems Cornerstone, the inspectors determined that the finding was of very low safety significance because each of the SSC deficiencies, deviations, and/or nonconformances identified by the inspectors represented an issue that did not result in the loss of operability or functionality. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, for certain deficiencies, deviations, and/or nonconformances associated with safety-related SSCs the licensee took no corrective actions whatsoever, instead allowing the corrective actions associated with those issues to be placed in the plants backlog of unscheduled work.
05000346/FIN-2011004-042011Q3Davis BesseInadequate Weld Records for CRDM HousingsA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, were identified by the inspectors for the licensees failure to establish adequate measures (e.g., perform a review of radiographic (RT) film weld records) to ensure material procured from a contractor (replacement control rod drive mechanism (CRDM) housings) met the American Society of Mechanical Engineers (ASME) Code. Consequently, two replacement CRDM housings were procured with RT film weld records that did not conform to the ASME Code-required film density ranges. As a corrective action, the licensee returned the affected CRDM housings to a vendor facility for completion of new RT film records prior to installation on the replacement vessel head. The violation was entered into the licensees corrective action program (CAP) as condition report (CR) 2011-00750. The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of Equipment Performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Absent NRC identification, the failure to complete an adequate RT examination of welds on two CRDM housings could have allowed unacceptable weld flaws to be placed in service. Specifically, weld flaws such as cracks, can reduce the CRDM housing integrity, and place the reactor coolant system (RCS) at an increased risk for through-wall leakage and/or failure. Because this finding was identified prior to placing the CRDM housings into service, the inspectors answered No to the Significance Determination Process Phase 1 screening question: Assuming worst case degradation, would the finding result in exceeding the Technical Specification (TS) limit for any RCS leakage or could the finding have likely affected other mitigation systems resulting in a total loss of their safety function assuming the worst case degradation? Therefore, the finding screened as having very low safety significance. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee staff failed to ensure adequate supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Absent NRC intervention, the failure to establish adequate measures to ensure material procured from a contractor (replacement CRDM housings) met the ASME Code would have allowed welds on two housings with non-conforming RT records to be placed into service.
05000346/FIN-2011004-062011Q3Davis BesseLicensee-Identified ViolationTS 5.4.1(a) requires the licensee to establish, implement, and maintain applicable written procedures for the safety-related systems and activities recommended in RG 1.33, Revision 2, Appendix A. Section 9.a, Procedures for Performing Maintenance, of RG 1.33, Revision 2, Appendix A, further states, in part, that: Maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to this requirement, on September 2, 2011, licensee personnel failed to properly rig and lift a new safety-related battery charger (DBC1PN) into position. Specifically, the personnel conducting the rigging activity switched from a four-point lift configuration to a two-point lift configuration when one of the lifting bolts atop the battery charger cabinet was inadvertently sheared off. This lifting configuration change was performed with an approved lift plan that contained inadequate technical/engineering guidance. When the-component was subsequently lifted, unbalanced forces resulting from the two-point lifting configuration caused several welds on the cabinet to crack, rendering the cabinet seismically unqualified. The objective of the Mitigating Systems Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A key attribute of this objective is human performance, and specifically, procedure use and adherence. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. The licensees failure to use technically adequate written procedures or instructions for the rigging and lifting configuration resulted in damage to safety-related battery charger DBC1PN that rendered it seismically unqualified and added significant time to it being inoperable. The licensee had entered this issue into their CAP as CRs 2011-02288 and 2011-02290. Corrective actions planned by the licensee include either weld repairs to the cabinet to restore its seismic qualification or replacement of the entire battery charger, and a re-examination of lifting and rigging practices.
05000341/FIN-2011003-022011Q2FermiSpent Fuel Cask Lay-down Areas Did Not Meet Seismic Category I RequirementsA finding of very low safety significance and associated NCV of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for failure to provide adequate design control measures for the reactor building radial girders, reactor building concrete floor slab and beam structures, spent fuel pool structure, and spent fuel cask leveling plate which were used to support the spent fuel cask placement. Specifically, the inspectors identified four examples where the licensee failed to perform adequate evaluations of the reactor building radial girders, reactor building concrete floor slab and beam structures, spent fuel pool structure, seismic restrain for multiple purpose canister cask transfer configurations, and spent fuel cask leveling plate in accordance with Seismic Category I requirements as defined in the Updated Final Safety Analysis Report, Section 3.8.4.5.1. The licensee documented the violation examples in condition assessment resolution documents (CARDs) 10-21097, 10-21205, 10-21943, 10-22955, 10-25226, 11-22993, and 11-25507. The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency could lead to a more significant safety concern. The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 -- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered yes to the question; is the finding a design qualification deficiency confirmed not to result in loss of operability or functionality in the Mitigating Systems column based on the licensee revising design calculations and initiating modifications where necessary to demonstrate compliance. The inspectors concluded the finding was of very low safety significance (Green). The inspectors identified a Human Performance, Work Practices, management and supervisory oversight (H.4.c) cross-cutting aspect associated with this finding. Specifically, the licensee failed to have adequate oversight of design calculations and documentation for establishing structural adequacy of the reactor building concrete floor slab, spent fuel pool structure and the spent fuel cask leveling plate used to support spent fuel cask placement.
05000341/FIN-2011003-032011Q2FermiMethodology Used in the HI-STORM/HI-TRAC Stack-up and Evaluation May Be InadequateA URI was identified by the inspectors regarding regulatory requirements and acceptable analytical methods to demonstrate seismic adequacy during vertical transfer of the MPC from the HI-TRAC to the HI-STORM during a postulated design basis earthquake event. Specifically, the inspectors identified a number of concerns pertaining to the licensees calculation performed to demonstrate that a free-standing configuration during vertical transfer of the MPC will not tip-over or excessively slide during a postulated design basis seismic event. Calculation SS-0003-2, Volume I, DCD 1, Rx/Aux. Bldg-Final Load Verif. Phase 2, First Floor Truck Bay Area (HI 2084016, Slab S-1, Beams 1B12, 1B2), Revision 0, evaluated the adequacy of a free-standing structural configuration during vertical transfer of the MPC from the HI-TRAC to the HI-STORM. The transfer includes the HI-TRAC placed on top of the HI-STORM with a mating device interposed between the two. All three components are placed on top of a trolley (low profile transporter) that can move along rails on the floor of the reactor building. A seismic analysis of the configuration was performed by the licensee using time history seismic input into the Visual Nastran computer code. The analysis model evaluated multiple freestanding bodies responding to the input seismic motion with friction at various contact surfaces acting as resisting forces. The inspectors identified a number of concerns regarding the calculation. The inspectors concerns regarding regulatory requirements and acceptable analytical methods were discussed with the Division of Spent Fuel Storage and Transportation staff and the licensee. In response to inspector concerns, the licensee decided to abandon the plan to use a freestanding stack-up configuration and instead, at this time, provide physical restraint of the structural configuration during MPC transfer operations. The inspectors did not make a determination of a performance deficiency or significance of these concerns by the end of the inspection. The licensee documented the inspectors concerns in CARD 10-22717. In addition, the inspectors identified the licensee is planning to deploy additional configurations of free-standing, non-stacked casks during ISFSI loading operations within the reactor building, using similar analytical methods. This issue will be a URI pending further review of the calculation by the inspectors after the Division of Spent Fuel Storage and Transportation provides inspection and regulatory guidance pertaining to seismic analysis of unrestrained structures and components. In addition, the inspectors will assess the acceptability of other free-standing, non-stacked casks, as needed.
05000341/FIN-2011003-012011Q2FermiEntry to a High Radiation Area on the Wrong Radiation Work PermitA finding of very low safety significance (Green) was self revealed when two radiation workers entered a high-radiation area without proper authorization. This issue was an NCV of licensee Technical Specification 5.4.1, Procedures. Specifically, radiation workers failed to adhere to a radiation work permit that limited access in the radiologically restricted area to radiation areas. This issue was placed in the licensees corrective action program as CARD 10-29820. The finding was more than minor because the individuals entered into a high radiation area on the wrong RWP, which is similar to the example in IMC 0612, Appendix E, Example 6.H, that states entry to a high radiation area is, not minor if: The individual was not authorized to enter a high radiation area. In addition it is associated with the human performance attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure adequate protection from exposure to radiation. The finding was determined to be of very low safety significance (Green) because it did not involve the as-low-as-reasonably-achievable program, did not involve an over exposure, did not involve a substantial potential for an over exposure, and did not compromise the ability to assess dose. The finding was not associated with a cross-cutting aspect as no aspects listed in IMC 0310 were characteristic of the finding.
05000341/FIN-2011002-012011Q1FermiFailure to Fully Evaluate the Failure of H2 O2 Sampling Pump Trips During CalibrationThe inspectors identified a finding and associated NCV of Technical Specification 5.4.1 for failure follow their Conduct Manual MES-43, Instrument Calibration Specification Sheets (ICSS), as established in Regulatory Guide 1.33, Appendix A.10, to ensure proper verification and calibration of the H2 O2 sample pump trip switch had been done during the annual preventative maintenance (PM) calibration. Specifically the engineering organization did not verify the actual setpoint until the inspector requested the calculations, then the licensee determined that the setpoint was out of tolerance. The licensee entered this into their corrective action program (CAP) as CARD 11-23023. The licensee completed the re-calibration of the flow switch. The inspectors determined that the failure to have a proper calibration of the switch was within their ability to foresee and correct, since the licensee failed to perform an evaluation when it was identified that the pump could trip at a flow setpoint in their normal band of operation established in procedures. Therefore the issue was a performance deficiency. This finding impacted the Mitigating System Cornerstone. The inspectors determined this finding was more than minor because, if left uncorrected, the early loss of the H2O2 sampling pump could have lead to a more significant safety concern and it was similar to the more than minor example of IMC 0612 Appendix E, 4.c. The flow switch for the H2 O2 sampling pump was outside of the acceptable range and would trip early causing a loss of the H2O2 monitoring system. This could complicate the verification of mitigating system equipment in a timely manner during plant events. The finding was determined to be of very low safety significance, Green, using IMC 0609, Significance Determination Process, Attachment 0609.04, Table 4a as all Mitigating System Cornerstone answers were no. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action, because Fermi 2 personnel proceeded in the face of uncertainty or unexpected circumstances by continuing with the calibration procedure and equipment use even though the pump tripped repeatedly at a setpoint value which the procedure established as acceptable, without performing an engineering evaluation that either determined the cause or provided conclusive justification for continued operation.
05000341/FIN-2011002-022011Q1FermiDesign Control Measures Failed to Ensure Adequacy of the Design Relating to the Reactor Building Crane Support Structure and Reactor Building SuperstructureA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to ensure the adequacy of the design for the reactor building crane support structure and reactor building superstructure. Specifically, the inspectors identified six representative examples where the licensee failed to perform adequate design calculations resulting in the design not being in conformance with Seismic Category I requirements as defined in Updated Final Safety Analysis Report (UFSAR) Sections 3.8.4.3.1 and 3.8.4.5.1 and referenced codes. The licensee documented the corrective actions in CARDs 10-22393, 10-22958, 10-22979, 10-23882, 10-24166, 10-26278 and 10-26691. The licensee also performed a re-analysis of the reactor building crane support structure and reactor building superstructure to address the deficiencies, and determined the structure to be operable but nonconforming and initiated modifications. The inspectors determined the licensees failure to meet design requirements for Seismic Category I compliance for the reactor building crane support structure and reactor building superstructure was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, compliance with Seismic Category I requirements for the reactor building crane support structure and superstructure was to demonstrate safe handling of heavy loads over the reactor core, the spent fuel pool, or safety-related components. Also, the performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as having very low safety significance (Green) because it was a design deficiency that did not to result in a loss of functionality/operability. The inspectors did not identify a cross-cutting aspect associated with this finding because the concern was related to calculations from the 1980s and 1990s and thus was not necessarily indicative of current licensee performance (Section 1R20).
05000341/FIN-2011002-032011Q1FermiReactor Scram due to Loss of VacuumA finding of very low safety significance (Green) for failure to evaluate and incorporate the operating experience received from the Boiling Water Reactors Owners Group (BWROG) Off-Gas committee was self-revealed when Fermi 2 experienced a reactor scram due to degraded condenser vacuum on October 24, 2010. The cause of the loss of vacuum was the failure of No. 3 steam jet air ejector (SJAE) steam supply to nozzle gasket, which caused steam erosion of the seating surface and loss of capacity. The licensee repaired the air ejector. The inspectors determined this finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The issue resulted in a scram. This finding was determined to be of very low safety significance, Green, because, while it did contribute to the likelihood of a reactor trip, it did not contribute to the likelihood that mitigating equipment would not be available. This finding was not cross-cutting because the licensee received the operating experience input over 3 years ago and was not necessarily indicative of current licensee performance. Finally, no violation of NRC requirements was identified since the SJAEs and the off-gas system are nonsafety-related.
05000341/FIN-2011002-042011Q1FermiLicensee-Identified ViolationDuring Refueling Outage 14 in 2010 the licensee performed a walkdown to inventory and validate the debris source term inside the primary containment. The licensee identified drywell insulation in recirculation piping areas that was not the appropriate material for use in that area of the drywell. The types of insulation in the areas potentially affected by a loss-of-coolant accident were evaluated during the ECCS strainer replacement project and associated EDPs in 1998. The min-K insulation was to be replaced with a more acceptable type (NUKON) during the implementation of the EDPs. Ten of the locations containing the min-K insulation were identified following the outage. The licensee indentified the insulation for replacement at the next available drywell opening. The replacement occurred during the January 2011 planned outage. Appendix B, Criteria III of 10 CFR 50, Design Control, states in part The design control measures shall provide for verifying or checking the adequacy of design.... Contrary to the above, from 1998 and prior to January 2011,the licensee did not maintain accurate design information and failed to identify these sections of insulation for replacement in 1998. Because the amount of insulation was small and within the existing available design margin for the ECCS suction strainer debris source term (DC-5979), the strainer was still operable; and the finding was determined to be of very low safety significance. The licensee entered the issue into their corrective action program as CARD 10-32197, Confirmation of insulation type installed in localized recirulation piping areas within the drywell. The finding was determined to be a licensee-identified NCV of 10 CFR 50 Appendix B, Criteria III.
05000305/FIN-2011002-012011Q1KewauneeMisapplication of Code Acceptance Criteria for Weld FlawsA finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors on March 3, 2011, for the licensees failure to establish a procedure that incorporated the American Society of Mechanical Engineers Code acceptance criteria for evaluation of flaws detected during ultrasonic examinations. Consequently, the licensee applied incorrect acceptance criteria to the flaws identified during ultrasonic examination of a weld on the chemical and volume control system seal water injection filter 1A housing. Licensee corrective actions included: evaluation of weld flaws to ensure they met applicable Code criteria and revision of a site procedure to incorporate appropriate Code acceptance criteria. The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Absent NRC identification, the failure to provide Code acceptance criteria could have allowed components with unacceptable cracks to be returned to service. Cracks in components returned to service would place safety-related piping systems at increased risk for through-wall leakage and/or failure. The licensee promptly corrected this issue before components with unacceptable flaws were returned to service. The inspectors answered No to the Significance Determination Process Phase I screening question, Assuming worst case degradation, would the finding result in exceeding the Technical Specification (TS) limit for any reactor coolant system leakage or could the finding have likely affected other mitigation systems resulting in a total loss of their safety function assuming the worst case degradation. Therefore, this finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not effectively implement human error prevention techniques. Specifically, the lack of procedure acceptance criteria was caused by inadequate peer checking during the licensees review and approval of the procedure for evaluation of non-destructive examination data (H.4(a)).
05000305/FIN-2011002-022011Q1KewauneeInadequate Work Instructions Results in Potential Orange PathA finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to implement procedures for shutdown operations involving shutdown operations safety assessments. Specifically, OU-KW-201, Shutdown Safety Assessment Checklist, step 3.3.1, stated, in part, that a shutdown safety assessment was required to be completed in accordance with the procedure for core cooling; however, the inspectors noted that the February 28, 2011, 6:00 p.m. analysis credited the safety injection system feed and bleed as an available alternate decay heat removal system when the system was not available as described in Section 5.3.2, Available/Availability, for work scheduled at that time on the emergency core cooling system (ECCS) sump. The licensee initiated condition report CR415539, and at the end of the inspection period, the licensee was performing a causal evaluation to determine the causes of the event and develop corrective actions. On February 28, as a remedial corrective action prior to the start of work, additional steps to the work instructions were added to ensure the equipment would meet the intended function, operators were designated to perform the local manual operations and a pre-job brief was conducted that provided training for using the equipment in the given situation. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre-event) and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the availability of the ECCS sump was integral to ensuring that the plant was not in an orange risk path for the evolutions completed on February 28. The inspectors screened the finding as of very low safety significance (Green) because the finding did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re-established and, therefore, did not require a Significance Determination Process phase 2 or phase 3 analysis. The finding has a cross-cutting aspect in the areas of human performance, work control, because the licensee failed to plan the work activities by incorporating the need for planned contingencies and compensatory actions to ensure the ECCS sump was available to ensure an orange risk path for core cooling was not entered (H.3(a)).
05000305/FIN-2011002-032011Q1KewauneeUnintended Voiding of the Reactor Vessel Closure HeadA finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to establish, implement, and maintain procedures for shutdown operations involving the draining of reactor coolant system (RCS) inventory. Specifically, on March 21, 2011, during a pressurizer draindown evolution, licensed operators unknowingly created a gas void in the reactor vessel closure head (RVCH) that displaced water to a level near the RVCH flange. Subsequent evaluation determined that the procedure for draining the RCS did not contain adequate guidance to ensure that an unacceptable void in the RVCH was not present prior to or formed during operations draindown activities. The licensee subsequently entered the issue into its corrective action program as CR418537 and performed a remedial corrective action of removing the gas void that accumulated in the RVCH. At the end of the inspection period, the licensee was performing an apparent cause evaluation to determine the causes of the event and develop additional corrective actions. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of operating procedure quality and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the formation of the gas void in the RVCH displaced RCS inventory and could have challenged the ability to remove decay heat in the event of a loss of shutdown cooling. The Region III senior reactor analyst determined that this issue is best characterized as a finding of very low safety significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because operations personnel did not follow or implement the guidance contained in plant procedures. Specifically, procedure OP-KW-AOP-RC-002 prescribed actions to take if a gas void formed in the RVCH that resulted in RVLIS level readings less than 88 percent, which had occurred several hours prior to the start of a pressurizer draining evolution (H.4(b)).
05000305/FIN-2011002-042011Q1KewauneePartial Loss of Offsite Power Caused by Less Than Adequate Interface and Oversight of Switchyard Modification WorkA finding of very low safety-significance was self-revealed for the failure to adequately control relay testing for switchyard breaker installations under Design Change WO KW100691871. Specifically, on March 10, 2011, Dominion Electrical Transmission technicians deviated from standard work practices to test a relay via an internal corporate server, which caused a partial loss of offsite power to the plant through the loss of the main auxiliary transformer backfeed to safety-related bus 6. Licensee corrective actions included a human performance and safety stand-down for substation personnel on the day of the event, the development of a mitigating strategy that outlined expectations and implemented increased direct supervision on critical tasks, and the development of a formal memo describing expectations related to the restricted use of the server for performing remote testing of control functions. The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, had a different breaker been inappropriately tripped, the station could have experienced a total loss of offsite power. The inspectors concluded that the finding could be evaluated using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. Specifically, the inspectors qualitatively evaluated the finding by applying the spent fuel pool questions in the Fuel Barrier column of Table 4a, Attachment 4. The inspectors answered No to all three questions and determined that the finding was of very low safety significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because supervisory and management oversight of work activities, including contractors, was not implemented for this evolution (H.4(c)).
05000305/FIN-2011002-052011Q1KewauneeTechnical Support Center Diesel Fails To LoadOn March 10, 2011, the licensee inadvertently opened a switchyard breaker that was providing power to various non-safeguards busses, as well as, bus 6, a 4160-volt safeguards bus. The TSC diesel automatically started as expected, however, the output breaker failed to close and power bus 46, as designed. The licensees troubleshooting and investigation determined that the TSC output breaker failed to close because of a failed breaker latching relay. The licensee replaced the relay and restored the TSC diesel to functional status. The licensees apparent cause evaluation was still in progress at the conclusion of the inspection period, and the inspectors did not have enough information to determine if a performance deficiency existed. Pending further review and inspection, this issue was considered a URI (URI 05000305/2011002-05, Technical Support Center Diesel Fails To Load).
05000341/FIN-2010005-022010Q4FermiStandby Liquid Control Test Tank OperabilityThe inspectors identified an unresolved item (URI) for past operability of the standby liquid control (SLC) system and for the adequacy of the procedures utilized to perform the periodic SLC pump test. Specifically, the inspectors identified that the demineralized water in the SLC test tank had not been drained following each periodic test of the SLC pumps as a legacy condition. Further, they noted the SLC test tank could affect the SLC system operability following a seismic event, if there was still demineralized water remaining in the test tank. The SLC system is designed to provide the capability of bringing the reactor to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC system is classified as a seismic category I system. The SLC testing subsystem, which is designed to periodically test the two SLC pumps, includes the SLC test tank, some valves and piping. The SLC test tank is described in the USFAR as seismic category II/I. Therefore, it is a non-safety-related component within a safety-related envelope; and while not required to maintain its operability, it must not impact the category I portions of the system. To test the SLC pumps, demineralized water is put into the test tank, the SLC tank remains isolated, and the SLC pumps are lined up and locally started to recirculate the demineralized water through the SLC testing subsystem. The inspectors questioned whether the demineralized water in the SLC test tank was drained following the periodic test. Further, they questioned whether the SLC test tank could affect the SLC system operability following a seismic event, if there was still demineralized water remaining in the test tank. The plant procedure (24.139.02) used to periodically test the SLC pumps, as required by TS 3.1.7, did not require draining of the SLC test tank following testing. The procedure did not incorporate the General Electric maintenance instruction guidance to drain the test tank following pump testing. As an interim measure, the SLC test tank was drained of demineralized water, and the SLC pump testing procedure was revised to include guidance to drain the SLC test tank following testing. The initial operability evaluation provided by engineering concluded that the mounting of the SLC test tank would remain in place, and it would not impact the adjacent safety-related equipment. However, there are several outstanding technical questions regarding the evaluation. Engineering will revise the evaluation of past operability. Then the inspectors will review the revised operability evaluation to determine final resolution of this issue. Because the licensee is performing an engineering analysis of the SLC test tank mounting, this issue will be carried as an unresolved item in this report (URI 05000341/2010005-02, Standby Liquid Control Test Tank Operability).
05000341/FIN-2010005-032010Q4FermiFailure of Condensate Filter Demineralizer D Main Drain ValveA self-revealed finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.5.4.c, Radiation Effluent Controls Program, were self-revealed for failure to monitor an effluent release path when the condensate filter demineralizer (CFD) D drain valve failed in the open position resulting in approximately 100,000 gallons of water being released into the radwaste and turbine buildings. An approximately 100-gallon mixture of the water and resin entered the plant sanitary waste system and traveled outside of the protected area as an unmonitored release. The design of the sanitary pipe that allowed crossing the power block boundary without a monitoring system was a performance deficiency. The licensee immediately stopped pumping sanitary waste and closed all facilities onsite until the system had been cleaned. The inspectors determined the finding was more than minor in accordance with IMC 0612, because the performance deficiency is associated with of the Plant Facilities/Equipment attribute of the Public Radiation Safety Cornerstone and the performance deficiency adversely affects the associated cornerstone objective. Specifically, the performance deficiency resulted in the unmonitored release of radioactive material to the public domain. The finding was assessed using the Effluent Release branch of the Public Radiation Safety SDP and was determined to be of very-low-safety significance, because the resultant dose impact to a member of the public from the radioactive release was less than the dose values in Appendix I to 10 CFR 50 and 10 CFR 20.1301 (e). Therefore, the finding is classified as Green. This finding has a cross-cutting aspect in the area of Human Performance, Decision-Making, Systematic Process. Specifically, the inspectors determined that design of the sanitary waste system was not properly evaluated and reviewed in a systematic process to meet the UFSAR requirements.
05000341/FIN-2010005-012010Q4FermiPlastic Face Shield Lost in the Reactor CavityA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified by the inspectors for the licensees failure to follow procedures and review the accident analysis in Updated Final Safety Analysis Report (UFSAR) Chapter 15 for the operability evaluation of a face shield lost in the reactor cavity, which could impact coolant flow to a fuel channel. Specifically, the licensee failed to follow Procedure MES27, Fermi 2 Engineering Support Conduct Manual, which requires evaluations needed to understand the potential consequences of the plant condition. As corrective action, the licensee revised their operability evaluation, EFA-B11-10-011, to include the needed information to address the accident analysis with potential flow channel blockage. The inspectors determined the finding was more than minor because it impacted the configuration control attribute of the Barrier Integrity Cornerstone in IMC 0612, Appendix B, Reactor Safety. The deficiency adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that the clad barrier would be effective as a barrier from releases during plant events, in that the deficient evaluation could challenge the clad integrity. The Finding was determined to be of very low safety significance, Green, because the licensee took action before reactor start-up to ensure additional evaluation was completed, and the issue affects the fuel barrier only, in accordance with Table 4a of IMC 0609.04. This finding has a cross-cutting aspect in the area of Human Performance, Resources, conservative assumptions, because the licensee failed to provide complete information in the operability determination that would allow Operations to fully understand the potential consequences of the issue. Specifically, the licensee judged that the condition remained bounded without defining the analyzed parameters, and the licensee failed to validate the underlying assumptions in the evaluation.
05000341/FIN-2010011-012010Q3FermiReactor Building Crane and Reactor Building Crane Support Structure/Superstructure Design Calculation Issue

The inspectors determined that an unresolved item (URI) existed concerning the design calculations that demonstrate the design adequacy of the reactor building crane, reactor building crane support structure and reactor building superstructure. The inspectors reviewed Calculation No. CN-23934, Reactor Building Crane for Enrico Fermi Atomic Power Plant No. 2 Bridge and Trolley Structural Calculations, dated March 8, 1973, Calculation No. 4.02.09, Reactor Building Superstructure Steel Girt and Column Framing Design, Volume I, Revision A, Calculation No. DC-6019, Assessment of the Interior Columns for the Reactor Building Steel Superstructure Including Crane Lifted Load, Volume IA, Revision 0, and Calculation No. 4.02.04, Superstructure Roof Framing Bracing System, Volume I, Revision A The inspectors were concerned that the reactor building crane and reactor building crane support structure/superstructure had been evaluated using acceptance criteria that were not contained within UFSAR Table 3.8-18, Loading Combinations for Steel Structures Elastic Design. Also, inspectors were concerned that the reactor building crane and reactor building crane support structure/superstructure had not been evaluated for the applied loads that were part of the design and licensing basis load cases/load combinations specified in UFSAR Table 3.8-18, Loading Combinations for Steel Structures Elastic Design. In response to the inspectors concern, the licensee initiated corrective action program documents related to the reactor building crane and reactor building crane support structure/superstructure. The licensee entered these items into their corrective action program (CAP) as the following condition assessment and resolution documents (CARD):

1. CARD 10-22393; NRC-ISFSI Issue Calculation Issues; dated March 19, 2010

2. CARD 10-22723; NRC ISFSI Issue Need to Revise Calculation CN-23934; dated March 29, 2010

3. CARD 10-22958; NRC ISFSI Issue Calculation 4.02.04; dated April 6, 2010

4. CARD 10-22979; NRC ISFSI Issue-Inspectors Questions About Calculations DC-6019 and 4.02.04; dated April 7, 2010

5. CARD 10-22979; NRC ISFSI Issue-Inspectors Questions About Calculations DC-6019 and 4.02.04; dated April 7, 2010

6. CARD 10-22981; NRC ISFSI Issue-Calculation CN-23934; dated April 7, 2010

7. CARD 10-23797; NRC ISFSI Issue-Revise Calculation 4.02.04; dated May 5, 2010

8. CARD 10-23882; NRC ISFSI Issue-RB5 Crane Calculation; dated May 7, 2010

9. CARD 10-24045; NRC ISFSI Issue-Calculation DC-6019 Load Combinations; dated May 14, 2010

10. CARD 10-24166; NRC ISFSI Issue-CMTRs used in calculation 4.02.09; dated May 19, 2010

11. CARD 10-26275; NRC ISFSI Issue-Clarification of Analysis in DC-6019; dated July 24, 2010

12. CARD 10-26276; NRC ISFSI Issue; dated July 24, 2010

13. CARD 10-26277; NRC ISFSI Issue-Licensing Basis for Allowable Axis Stresses in 4.02.04; dated July 24, 2010

14. CARD 10-26278; NRC ISFSI Issue-Stresses is DC-6019 and 4.02.09; dated July 24, 201

15. CARD 10-26279; NRC ISFSI Issue-Extraction of Loads Used in 4.02.04; dated July 24, 2010

16. CARD 10-26280; NRC ISFSI Issue-Acceptance Criteria in Calc. 4.02.09; dated July 24, 2010

17. CARD 10-26282; NRC ISFSI Issue-Use of CMTRs in EDP-34473; dated July 24, 2010

18. CARD 10-26415; NRC ISFSI Question on Design Basis for Response Spectra used in Design Calculation DC-6019; dated July 28, 2010

19. CARD 10-26545; NRC Question ISFSI-RB5 Structure Corner Column Buckling Calculation Compliance with AISC Code; dated July 30, 2010

20. CARD 10-26548; NRC Questions-ISFSI-RB5 Structure Corner Column Peak Critical Yield Analysis Not Evaluated as Plastic According to AISC requirements; dated July 30, 2010

21. CARD 10-26691; ISFSI NRC Issue - Crane Runway Girder Splice; dated August 4, 2010

This item is being held as an unresolved item (URI 05000341/2010011-01) pending evaluation of these concerns by the licensee and subsequent inspector review and discussion with the licensee. The licensee indicated that they would provide the inspectors multiple revised analyses addressing the issues relevant to the reactor building crane, reactor building crane support structure and reactor building superstructure on September 10, 2010

05000346/FIN-2010004-022010Q3Davis BesseCompliance with Spent Fuel Pool Storage Requirements

The inspectors indentified an unresolved item (URI) concerning an inconsistency with TS 3.7.16, which described acceptable fuel loading patterns within the spent fuel pool

Limiting Condition for Operation (LCO) 3.7.16 describes the requirements for fuel assembly storage in the spent fuel pool. Spent fuel assemblies are categorized based on burn-up and initial enrichment, in accordance with TS figure 3.7.16-1. The approved loading patterns applicable to each fuel assembly category are specified in the Bases. On August 26 the inspectors reviewed CR 10-81824, which documented a potential non-compliance involving loading patterns in the Spent Fuel Pool rack modules. At the time of discovery of the issue on August 26, 2010, the Bases for LCO 3.7.16 stated that different loading patterns may be used in different rack modules, provided each rack module contains only one loading pattern. Contrary to this statement, the Davis-Besse spent fuel pool contained rack modules that used two different loading patterns. However, the spent fuel pool was configured in accordance with site procedures consistent with the criticality safety analysis. This analysis, and the use of two different loading patterns, was previously approved by the NRC in the safety evaluation report for license amendment 247

On August 27, 2010, the licensee submitted a change to the bases of LCO 3.7.16, which added a sentence stating, Two different loading patterns may be used in a single rack module, subject to certain additional restrictions. This sentence, which restored compliance with TS 3.7.16, was unintentionally removed from the Bases when Improved TSs were implemented at the plant on December 13, 2008. The inspectors continue to review the TS non-compliance and reporting requirements of this issue. Pending further review of the licensees evaluation of reportability, the issue is considered an unresolved item.

05000341/FIN-2010004-012010Q3FermiInadequate procedures to control the plant from the dedicated shutdown panelOn August 11 and 13, 2010, the inspectors walked down the dedicated shutdown system abnormal operating procedure 20.000.18. The inspectors identified an unresolved item (URI) for inadequacy of the procedures utilized to control the plant from the dedicated shutdown panel. Specifically, the inspectors identified that at the minimum staffing complement of non-licensed nuclear operators as defined in MOP03, Section 3.4.1, the outside rounds operator, who is the available nuclear operator designated to fulfill the position as the nuclear operator for dedicated shutdown, would not be able to complete the required actions to restore reactor pressure vessel makeup within 29 minutes, as required by the licensees feasibility study. At minimum staffing complement of non-licensed nuclear operators as defined in MOP03, Section 3.4.1, the outside rounds operator is the only available nuclear operator during a fire emergency that would require utilization of the dedicated shutdown panel for safe shutdown. The outside rounds operator could be at a location outside the protected area at the initiation of the abnormal event. The fire protection engineering evaluation (FPEE) 05-0012 assumes initial conditions based upon the safe shutdown nuclear operators starting from the main control room and proceeding with the non-supervisory reactor operator assigned to the dedicated shutdown panel and commencing implementation of the abnormal operating procedure 20.000.18. A delay of 7 minutes and 22 seconds was estimated by the licensee for the time it would take the outside rounds operator to proceed to the dedicated shutdown panel. Thus, the time to restore reactor pressure vessel makeup, which is defined as required by the FPEE within 29 minutes of initiation of the dedicated shutdown panel, would be estimated to be 33 minutes (rounded). It appears the Standard Operating Procedure 20.000.18 is inadequate in meeting the objective to restore reactor pressure vessel makeup. The licensee documented the issue in Card 10-27645. There is no current safety concern as an extra operator is on shift to cover the requirements. Further, procedures are being revised to ensure the operator is inside of the power block. This item is being held as an unresolved item (URI 05000341/2010004-01) pending evaluation of the timing and manning issues by the licensee.
05000341/FIN-2010003-012010Q2FermiFailure to Adequately Control Loose Materials near the SwitchyardA finding of very low safety significance was identified by the inspectors for the licensees failure to adequately control loose materials next to the 345kV switchyard. Specifically, the inspectors identified tarps next to the switchyard fence. Once this condition was identified, the licensee removed the material from the switchyard area. No violation of regulatory requirements occurred. The finding was greater than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the loose items could affect the proper operation of the switchyard during periods of high winds. This finding was determined to be of very low safety significance because the finding was not a loss of coolant accident initiator, did not increase the likelihood of a fire or a flood, and did not contribute to the likelihood that mitigating equipment relied upon during a loss of division 2 offsite power sources would not be available. The inspectors determined that the failure to ensure that procedure changes were incorporated in procedures following corrective actions from previous findings also affected the cross-cutting area of PI&R, Corrective Actions (P.1(d)).
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