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05000339/FIN-2018011-012018Q2North AnnaFailure to ensure compliance with the Technical Specification (TS) 5.4.1.a requirement relevant to procedures for plant firesThe NRC identified a Green finding and associated non-cited violation (NCV) of the TS 5.4.1.a requirement to establish and maintain fire contingency action procedures based upon the licensees failure to effectively perform reviews during the revisions of the procedures in accordance with procedure VPAP-0502, Procedure Process Control. The failure led to undetected errors and was a performance deficiency that was determined to be more than minor because, if left uncorrected, it could potentially lead to a more significant safety concern during Appendix R fire events.
05000280/FIN-2018010-012018Q1SurryFailure to implement the 10 CFR, Part 50, Appendix R, III.G.3 requirements consistent with fire protection license condition 3I.The NRC identified a Green finding and associated non-cited violation (NCV) of the requirements consistent with license condition 3.I, Surry Units 1 and Unit 2. Specifically, the licensee failed to adequately protect fiberglass pipe that is susceptible to fire damage and required for safe shutdown. By not protecting the pipe, the licensee did not ensure the alternative shutdown methodology was implemented with the independence as defined by the 10 CFR 50 Appendix R section III.G.3 requirements.
05000395/FIN-2017002-012017Q2SummerFailure to Implement Corrective Actions to Restore Compliance for Previous NRC-identified Green NCV 05000395/2013003-03The inspectors identified a Green finding with a cited violation of Operating Licensee Condition 2.C.(18) for failure to ensure that conditions adverse to fire protection as noted in a previous NRC-identified Green NCV, 05000395/2013003-03, Failure to Adequately Design, Install and Maintain Oil Collection Devices for Reactor Coolant Pump Motors, were corrected. Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for (1) a failure to ensure an adequate design for the oil lift pump enclosure, and (2) a failure to have oil collection components for internally leaked oil dripping from the motor air discharge ductwork flange. The licensee entered the issue in their corrective action program as condition report CR-17-03962.The inspectors determined that the failure to implement corrective actions for the oil collection system to restore compliance was a performance deficiency (PD). The inspectors used IMC 0612 and determined that the PD was more than minor and therefore a finding because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. This finding has a credible impact on safety because the failure to adequately install, maintain and design the oil collection system presented a degradation of a fire confinement component which has a fire prevention function of not allowing an oil leak to reach hot surfaces. This finding had been evaluated and screened to a low safety significance (Green) and documented in the previous NRC-identified Green NCV, 05000395/2013003-03. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.3 of the NRC Enforcement Policy. The inspectors used IMC 0310 and determined this finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance and restore compliance (P.3).
05000395/FIN-2017002-022017Q2SummerFailure to Provide NRC Staff Complete and Accurate InformationThe inspectors identified a severity level (SL) IV NCV of 10 CFR 50.9(a), Completeness and accuracy of information, involving licensee document,RC-13-0142, dated October 14, 2013. This document was a response to a request for additional information involving a license amendment request (LAR) to adopt NFPA 805 and contained an approval request, L12, associated with oil misting from the reactor coolant pumps. The licensee entered this violation into their corrective action program as CR-17-03961. The inspectors determined that the licensees failure to provide complete and accurate information associated with approval request, L12, was a violation of 10 CFR 50.9(a). Because this violation of 10 CFR 50.9(a) impacted the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using traditional enforcement (TE). Since the TE violation is associated with a previous Green reactor oversight process violation, and the misinformation was identified after the NRC relied on it for issuing a previous operating license amendment, the TE violation was determined to be a SL IV, NCV, consistent with the language of the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information. This violation involved TE; therefore a cross-cutting aspect was not assigned.
05000395/FIN-2017001-022017Q1SummerLicensee-Identified ViolationV.C. Summer Operating License condition 2.c(18) states in part that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(c), National Fire Protection Association (NFPA) 805 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on January 10, 2017, the licensee failed to establish procedures, SAP -131A, Fire Protection Program Surveillances and Compensatory Measures, Rev 3., FPP - 015, Shift Inspection, Rev. 7, and FPP -025, Fire Containment, Rev. 6A, to ensure the fire doors listed in TR07800- 020, NFPA 805 Monitoring Program Phase 1: Scoping, Rev. 0, were appropriately identified for adequate licensee actions concerning surveillances and degraded conditions. NRC used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, Attachment 1, dated September 20, 2013, to perform a Phase 1 analysis and determined that the finding was of very low safety significance (Green) based on the response for Question 1.3.1A, in which the reactor was able to reach and maintain safe shutdown. The licensee has documented this problem i n their CAP as CR- 17-00150.
05000395/FIN-2017001-012017Q1SummerLicensee-Identified ViolationTS 3/4.5.2, ECCS Subsystems T avg > 350 o F states in part that two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with one OPERABLE centrifugal charging pump in Modes 1, 2 and 3. TS 1.18, OPERABLE OPERABILITY, definition states in part that a subsys tem shall be OPERABLE when all necessary auxiliary equipment that are required for the subsystem are capable of performing their related support functions. Contrary to this, from June 20, 2015, through July 20, 2016, safety -related subsystem chiller, XHX0001A, an auxiliary component supporting the A train charging pump (XPP0043A) was incapable of performing its safety function resulting in the inoperability of XPP0043A for greater than the allowed action times of TS 3/4.5.2. A review of IMC 0609, Appendix A, determined the finding was of very low safety significance (Green) 14 because the finding was not a design deficiency and it did not result in a loss of function. The licensee has documented this problem in their CAP as CR -15- 04395
05000395/FIN-2016004-032016Q4SummerFailure to Accomplish Procedure for Foreign Material Exclusion Control Involving Failure of a Safety-Related BreakerGreen. The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to accomplish safety-related (SR) station administrative procedure, SAP-0363, Foreign Material and Debris Control, Revision 8H, for foreign material exclusion (FME) control during a SR breaker refurbishment. A subsequent breaker failure occurred due to foreign material. The licensee immediately initiated corrective actions to repair the breaker, and the licensee entered condition report, CR-16-03099, in their CAP. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone by adversely affecting the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the equipment reliability attribute was impacted because foreign material rendered the SR breaker nonfunctional causing inoperability of the pressurizer backup group 2 heaters for greater than the Technical Specification limiting condition for operation. The inspectors used IMC 0609, Significant Determination Process, Attachment 4, dated October 7, 2016, and Appendix A Exhibit 2, dated July 1, 2012, and determined that the finding required a detailed risk evaluation. A regional senior risk analyst performed a bounding risk evaluation in accordance with NRC IMC 0609 Appendix A using the VC Summer SPAR model. The finding was modelled as a transient initiator with a loss of the B EDG as a surrogate for the group 2 pressurizer heaters for a 94 hour exposure interval. The dominant sequence was a transient initiator with a consequential loss of offsite power without recovery, failure of the A EDG without recovery leading to a station blackout and loss of core heat removal after battery depletion. The risk was mitigated by the available normal and group 1 pressurizer heaters. The bounding assessment determined that the performance deficiency represented an increase in core damage frequency of < 1.0 E-6/year, a GREEN finding of very low safety significance. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of Problem Identification and Resolution and the aspect of work management, H.5, because the licensee failed to ensure the planning and execution of the respective work order for breaker refurbishment followed SAP-0363 for FME control to support nuclear safety-related work.
05000395/FIN-2016004-052016Q4SummerLicensee-Identified ViolationV.C. Summer Operating License condition 2.c(18) states in part that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(c), National Fire Protection Association (NFPA) 805 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on September 14, 2016, the licensee failed to implement the requirements of procedure, Fire Protection Procedure, FPP-025, Fire Containment, Rev. 4, to ensure that fire door and SPB, DRIB/107, did not remain open in excess of 1 hour. NRC IMC 0609.04 and NRC IMC 0609 Appendix A screening determined that the finding represented a loss of the short term heat removal safety function within the Mitigating Systems Cornerstone and required a detailed risk evaluation. A bounding analysis was performed by a regional SRA using the VC Summer SPAR model. The finding was modelled as a Steam Line Break Outside Containment (SLBOC) initiating event assessment. A 3 hours and 22 minutes bounding exposure was utilized. No recovery was assumed. Equipment impacts from potential HELBs were determined using the results from Gothic Analyses performed to assess the temperature, pressure and relative humidity increases in mild environment spaces in the Intermediate Building due to the various HELB boundary breaches associated with the finding. For non-Loss of Offsite Power (LOOP) conditions, HELB impacts in mild areas were minimal. For LOOP conditions, the HELB was assumed to impact the chilled water system such that one train of safety related equipment was assumed failed as a bounding impact. The dominant sequence was a SLBOC initiator with a failure to isolate the break and a failure of high pressure injection impacted by loss of chilled water leading to loss of core heat removal. The risk evaluation determined that the finding represented a risk increase of < 1.0E-6/year, a GREEN finding of very low safety significance. The licensee has documented this problem in their CAP as CR-16-04703
05000395/FIN-2016004-062016Q4SummerLicensee-Identified Violation10 CFR 50, Appendix B, Criterion V, requires in part that activities affecting quality shall be accomplished by documented instructions of a type appropriate to the circumstances. Contrary to this, on August 4, 2009, the licensee failed to accomplish documented work instructions contained in ECR50585 to install floor drain orifices to act as SPBs to protect SR components from the effect of a HELB. NRC IMC 0609.04 and NRC IMC 0609 Appendix A screening determined that the finding represented a loss of the short term heat removal safety function within the Mitigating Systems Cornerstone and required a detailed risk evaluation. A bounding analysis was performed by a regional SRA using the VC Summer SPAR model. The finding was modelled as a Steam Line Break Outside Containment (SLBOC) initiating event assessment. A forty five day bounding exposure was utilized. No recovery was assumed. Equipment impacts from potential HELBs were determined using the results from Gothic Analyses performed to assess the temperature, pressure and relative humidity increases in mild environment spaces in the Intermediate Building due to the various HELB boundary breaches associated with the finding. For non-Loss of Offsite Power (LOOP) conditions, HELB impacts in mild areas were minimal. For LOOP conditions, the HELB was assumed to impact the chilled water system such that one train of safety related equipment was assumed failed as a bounding impact. The dominant sequence was a SLBOC initiator with a failure to isolate the break and a failure of high pressure injection impacted by loss of chilled water leading to loss of core heat removal. The risk evaluation determined that the finding represented a risk increase of < 1.0E-6/year, a GREEN finding of very low safety significance. The licensee has documented this problem in their CAP as CR-16-04716.
05000395/FIN-2016004-072016Q4SummerLicensee-Identified ViolationTS 6.8.1, Procedures and Programs, requires in part that written procedures be maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33, Rev. 2, Section 9, Procedures for Performing Maintenance. Contrary to this, on October 4, 2016, the licensee determined they had failed to establish station procedures to adequately remove ventilation access covers for fire damper inspections during plant modes that would not present a HELB challenge to those components qualified for mild EQ conditions. NRC IMC 0609.04 and NRC IMC 0609 Appendix A screening determined that the finding represented a loss of the short term heat removal safety function within the Mitigating Systems Cornerstone and required a detailed risk evaluation. A bounding analysis was performed by a regional SRA with the following major analysis assumptions: a fifty one minute exposure period, a HELB frequency of 7.7E-3/year from the NRC VC Summer SPAR model, and a conditional core damage probability given a HELB of 1.0. The risk increase due to the finding was <1.0E-6/year, a GREEN finding of very low safety significance. The risk was mitigated by the short exposure. The licensee has documented this problem in their CAP as CR-16-05696.
05000425/FIN-2016004-012016Q4VogtleFailure to Implement Maintenance Procedure for Electrical Grayboot Connectors(Green). A self-revealing non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly install shims when assembling electrical connectors on Unit 2 main steam isolation valve (MSIV) HV-3026B, in accordance with maintenance procedure 25709-C, Instructions for EGS Grayboot Connection Kit Installation, Ver. 21.1. The licensee replaced the affected connectors and entered the issue in their corrective action program under condition reports (CR) 10279411, and 10268507, and technical evaluations (TE) 970299, 968149, and 970300, to evaluate and develop additional training for maintenance technicians, enhance the maintenance procedure, and conduct extent of condition. The performance deficiency (PD) was more-than-minor, because it adversely effected the Initiating Events cornerstone objective when Unit 2 received an automatic reactor trip and safety injection on March 14, 2015. Also, if left uncorrected, the PD would result in moisture intrusion and degradation of MSIV connectors and potentially lead to a more significant safety concern. The finding was determined to be Green, because the PD did not result in a loss of mitigation equipment used to transition the reactor to a stable shutdown condition. The finding was assigned a cross cutting aspect of Procedure Adherence, because maintenance technicians failed to adhere to procedural guidance in Attachment 1 of 25709-C for installing the connector shims. (H.8)
05000395/FIN-2016004-012016Q4SummerFailure to Establish Procedures for Corrective Actions to Address Conditions Adverse to Fire ProtectionGreen. The inspectors identified a Green, non-cited violation (NCV) of the V.C. Summer Nuclear Station Operating License, Condition 2.C (18), Fire Protection Program, for the failure to establish procedures requiring corrective action for conditions, including significant and repetitive, adverse to fire protection. The licensee immediately notified the corrective action program (CAP) supervisor and entered the problem into their CAP as condition report CR-16-05270. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the performance deficiency (PD) was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. Specifically, the failure to establish corrective action program requirements specific to fire protection with appropriate definitions for significant and repetitive would result in corrective actions not commensurate with the significance of the adverse condition. The inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, Attachment 1, dated September 20, 2013, to perform a Phase 1 analysis and determined that the reactor oversight process (ROP) finding was of very low safety significance (Green) based on the response for Question 1.3.1A, in which the reactor was able to reach and maintain safe shutdown. While the licensee does not have the required corrective actions defined, they have generally addressed conditions adverse to fire protection within the existing corrective action program. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of human performance and the aspect of resources, H.1, because the licensee leadership failed to ensure that adequate procedures were in place to address significant and repetitive conditions adverse to fire protection.
05000395/FIN-2016004-042016Q4SummerFailure to Update FSAR with a New Design Function for the Equipment and Floor Drain SystemSL IV. The NRC identified a severity level IV (SL IV) non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.71(e) for the licensees failure to update the final safety analysis report (FSAR) with the latest information developed, regarding the design functions of the equipment and floor drain system. Specifically, the licensee failed to update the FSAR to reflect the high-energy line break (HELB) steam propagation barrier (SPB) function of the floor drain system following installation of new floor drain orifices used as SPBs. The licensee entered this issue into their corrective action program as CR-16-06003. The inspectors treated the noncompliance with 10 CFR 50.71(e) as traditional enforcement because not having an updated FSAR hinders the licensees ability to perform adequate 10 CFR 50.59 evaluations and impacts the NRCs ability to perform its regulatory function such as license amendment reviews and inspections. This was determined to be a SL-IV violation of 10 CFR 50.71(e) because it was similar to the NRC Enforcement Policy, Section 6.1.d.3, SL IV example of, a licensee fails to update the FSAR as required by 10 CFR 50.71(e) but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000395/FIN-2016004-022016Q4SummerFailure to Promptly Identify and Correct a Condition Adverse to Quality for B Emergency Diesel Generator Exhaust Manifold Weld IndicationsGreen. A self-revealing, Green NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the failure to promptly identify and correct a condition adverse to quality (CAQ) involving welded joint indications in the B emergency diesel generator (EDG) exhaust manifold. The licensee immediately removed the EDG from service to perform repairs, and the issue was entered into the licensees CAP as Condition Report CR-16-05421. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor and therefore a finding, because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and the respective attribute of equipment performance. Specifically, the B EDG was declared operable but degraded or nonconforming due to a circumferential weld failure and resulting separation of an exhaust manifold joint causing a small reduction in EDG power. The inspectors used IMC 0609, Significant Determination Process, Attachment 4, dated October 7, 2016, and Appendix A Exhibit 2, dated July 1, 2012, and determined the finding was of very low safety significance or Green because the finding was not a design deficiency or loss of function. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of problem identification and resolution and the aspect of resolution, P.3, because the licensee failed to take effective corrective actions commensurate with an issues safety significance in that they failed to promptly identify and correct a CAQ involving welded joint indications in the B EDG exhaust manifold.
05000424/FIN-2016003-012016Q3VogtleFailure to Properly Implement Fire Door InspectionsAn NRC-identified Green non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.d, Procedures, was identified for the licensees failure to correctly verify fire door gaps at the strike plate area and between meeting edges of double swinging metal doors were within acceptable limits. The licensee initiated hourly roving fire watches for these fire doors and took corrective maintenance action to restore affected fire doors within limits. The licensee documented this condition in condition reports 10254221 and 10252774. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Hazards (i.e. fire) and adversely affected the cornerstone objective in that door gaps outside the required limits compromised the doors fire rating qualification. The finding was determined to be of very low safety significance (i.e. Green) because either the combustible loading on both sides of each door was representative of a fire duration of less than 1.5 hours or each door maintained at least a 1-hour fire endurance rating. The finding had a cross-cutting aspect of Training in the Human Performance area because the licensee did not ensure there was adequate training to properly inspect station fire doors (H.9).
05000424/FIN-2016003-022016Q3VogtleLicensee-Identified ViolationTitle 10 CFR Part 50.54(q)(2) required, in part, that a licensee shall follow and maintain the effectiveness of its emergency plan that meets the planning standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4) required, in part, that a standard emergency classification and action level scheme is in use by the nuclear facility licensee. Contrary to these requirements, since 2008, emergency action level EAL HA1 #5 was not translated from the emergency plan to implementing procedure NMP-EP-110 GL03 (formally 91001-C) during the 2008 revision of the emergency plan. The licensee entered the issue into their corrective action program as CR 10251396. The inspectors determined that the finding was of very low safety significance (Green) because the finding constituted an ineffective EAL rather than a failed risk-significant planning standard.
05000395/FIN-2016003-012016Q3SummerFailure to Meet HRA Entry Requirements (Two Examples)The inspectors identified two examples of a Green, self-revealing, non-cited violation (NCV) of Technical Specification (TS) 6.12.1, High Radiation Area. TS 6.12.1 requires that entries into high radiation areas (HRAs) be controlled with issuance of a radiation work permit (RWP) and that individuals entering these areas be made knowledgeable of the dose rates. Contrary to that, on two separate occasions, workers made entries into HRAs without being issued an appropriate RWP and without being knowledgeable of area dose rates. Specifically, on March 28, 2016, a worker tagging a pump on the auxiliary building (AB) 400-01 slab entered a HRA without the required radiological briefing and appropriate RWP. Also, on April 18, 2016, a worker performing dry cask welding operations in the fuel handling building entered a HRA without the required radiological briefing and appropriate RWP. The licensee entered these events into their corrective action program as condition reports CR-16-01528 and CR-16-01863. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure and the ability to assess dose was not compromised. Therefore, the finding was determined to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Avoid Complacency (H.12) because in both examples there were repostings, radiation areas were upgraded to HRAs due to changing radiological conditions, and prior to entry the workers failed to stop and get updated conditions and to adhere to the postings.
05000395/FIN-2016003-022016Q3SummerFailure to Prescribe Work Instructions for a Temporary Repair on a Safety-Related ComponentThe inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to prescribe instructions for a temporary repair of the safety-related C component cooling water (CCW) pump outboard bearing. The licensee entered condition report, CR-16-04576, in their corrective action program for appropriate response. The inspectors determined that the failure to prescribe documented work instructions of a type appropriate to the circumstances for the temporary repair of the C CCW pump outboard bearing was a performance deficiency (PD). The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone by adversely affecting the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design control attribute was impacted because not prescribing instructions that follow vendor instructions for temporary repairs on the safety-related pump resulted in improper repairs causing reasonable doubt in operability. The inspectors evaluated the finding in accordance with IMC 0609, Significant Determination Process, Attachment 4 and Appendix A, and determined that the finding was of very low safety significance, Green, because it did not represent an actual loss of a safety-related train since the C CCW pump was operable but degraded. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of Human Performance and the aspect of resources, H.1, because the licensee failed to ensure instructions were adequate and available to support nuclear safety-related work.
05000395/FIN-2016003-032016Q3SummerLicensee-Identified ViolationV.C. Summer Operating License condition 2.c(18) states in part that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(c), National Fire Protection Association (NFPA) 805 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on September 14, 2016, the licensee failed to implement the requirements of procedure, Fire Protection Procedure, FPP-025, Fire Containment, Rev. 4, to ensure that fire door and SPB, DRAB/514, remained operable/functional. The inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and performed a Phase 1 analysis to determine the finding was of very low significance or Green. The fire confinement program element was not of low degradation, the non-suppression probability was 0.1, the fire frequencies related to the affected fire zones AB01.21.02 and FH01.04 were 2.79E-3 and 3.98E-4 respectfully, and the duration of the component inoperability was approximately 1 hour or 0.000114, which resulted in screening check frequency of 3.63E-8 that was less than the screening criteria of 1E-6. The licensee has documented this problem in their CAP as CR-16-05073.
05000395/FIN-2016002-012016Q2SummerFailure to Adequately Manage Risk of Maintenance Activities Following Risk Model UpdatesThe inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, involving the licensees failure to develop and implement specific risk management actions (RMAs) for a yellow risk condition associated with solid state protection system (SSPS) surveillance testing. The issue was entered into the licensees corrective action program (CAP) as condition report (CR)-16-02504. The inspectors identified a performance deficiency (PD) for the failure to manage the increase in risk associated with A train SSPS surveillance testing which was indicative of the lack of programmatic requirements for assessing and managing risk subsequent to equipment out of service (EOOS) model updates. The inspectors reviewed inspector manual chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because (1) it was associated with the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences, and (2) if left uncorrected the PD would have the potential to lead to a more significant safety concern. Specifically, the failure to manage the increase in risk jeopardizes the availability of remaining safety systems to combat the consequences of an initiating event. The inspectors reviewed IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined that the finding was of very low safety significance, Green, because the incremental core damage probability (ICDP) for the SSPS surveillance test was less than 1E-6. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined that this finding had a cross-cutting aspect in the area of Work Management (H.5), because the licensee did not develop specific RMAs for a yellow risk condition which was indicative of the lack of programmatic requirements for assessing and managing risk subsequent to EOOS model updates. (Section 1R13)
05000395/FIN-2016001-022016Q1SummerLicensee-Identified ViolationTS 3/4.5.2, ECCS Subsystems Tavg > 350 oF states in part that two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with one OPERABLE centrifugal charging pump in Modes 1, 2 and 3. TS 1.18, OPERABLE OPERABILITY, definition states in part that a subsystem shall be OPERABLE when all necessary auxiliary equipment that are required for the subsystem are capable of performing their related support functions. Contrary to this, from July 24, 2015, through September 17, 2015, safety-related subsystem chiller, XHX0001A, an auxiliary component supporting the A train charging pump (XPP0043A) was incapable of performing its safety function resulting in the inoperability of XPP0043A for greater than the allowed action times of TS 3/4.5.2. A review of IMC0609, Appendix A, determined the finding was of very low safety significance (Green) because the finding was not a design deficiency and it did not result in a loss of function. The licensee has documented this problem in their CAP as CR-15-04395.
05000395/FIN-2016001-012016Q1SummerFailure to Implement Adequate Administrative Controls Following a Departure from NFPA 80-1973 and Provide NRC Staff Complete and Accurate InformationThe inspectors identified a Severity Level IV, non-cited violation (NCV) of 10 CFR 50.9(a), Completeness and accuracy of information, and an associated Green non-cited violation of V.C. Summer, Operating License Condition 2.C.(18) for a NFPA 80-1973 code deviation that was not discussed in the licensees NFPA 805 license amendment request (LAR), and would adversely affect the ability to achieve and maintain safe shutdown in the event of fire. The associated engineering evaluation relied on inadequate administrative controls to ensure the associated replacement doors in the intermediate building, DRIB/105A&B, were kept closed as a basis for not following NFPA 80-1973 which required the fire doors be self-closing. The licensee entered the violations into their corrective action program as condition reports CR-15-04027 and CR-16-00242 respectively. The inspectors identified a reactor oversight process (ROP) performance deficiency (PD) for the failure to provide adequate administrative controls to allow departure from NFPA 80-1973 requirements, which resulted in replacement of a self-closing fire door with two non-self-closing fire doors, DRIB/105A&B, that adversely affected the ability to achieve and maintain safe shutdown in the event of fire since they were found open on multiple occasions. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the ROP PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors, such as fire, and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, Attachment 1, dated September 20, 2013, to perform a Phase 1 analysis and determined that the ROP finding was of very low safety significance (Green) based on the response for Question 1.4.3.A in which the combustible loading on both sides of DRIB/105A&B was less than 120,000 BTU/ft2. Furthermore, the inspectors determined that the associated fire zone area (IB 7) with multiple equipment trains used a pre-action sprinkler system and automatic fire detection. The inspectors also determined that the licensees failure to include the departure from NFPA 80-1973 in their NFPA 805 license amendment request was a violation of 10 CFR 50.9(a). Because this violation of 10 CFR 50.9(a) had the potential to impact the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using traditional enforcement (TE). Since the TE violation is associated with a Green ROP violation, and the misinformation was identified after the NRC relied on it for issuing a previous operating license amendment, the TE violation was determined to be a Severity Level IV violation, consistent with the language of the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 14, 2014, and determined the cause of this finding involved the cross-cutting area of problem identification and resolution, P.3, because the licensee failed to ensure that adequate administrative controls were in place after the fire doors were found open multiple times.
05000266/FIN-2015004-032015Q4Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 5.4.1, Procedures for the failure to maintain the emergency operating procedures (EOPs). The licensees TS 5.4.1 required, in part, that written procedures shall be maintained including the EOPs required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. During design reviews, the licensee discovered that following a 2012 calculation update, the licensee inconsistently applied pre and post-modification uncertainties that had resulted from a 2010 modification associated with the sensitivity and calibration of both units Subcooling Margin Monitors. Ultimately the calculative errors resulted in 19 EOP Subcooling setpoints being incorrectly calculated. These Subcooling setpoints are used throughout the licensees EOPs network to provide operators with discrete indications for key EOP decision making. Contrary to the above, from April 12, 2012 through November 5, 2015, the licensees EOP network of procedures for both Unit 1 and 2, contained the incorrect setpoints for decision points with respect to subcooling. The licensee entered this issue into the CAP as AR 02089011 and AR 02099152. The inspectors consulted the Region III Senior Reactor Analysts and determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 1, 2012 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors determined that the issue was a design or qualification deficiency confirmed not to result in a loss of operability; therefore, answered yes to question A.1 in Exhibit 2, Section A, Mitigating SSCs and Functionality. This resulted in the finding screening as Green.
05000301/FIN-2015004-022015Q4Point BeachInadequate Evaluation of Non-Conforming Auxiliary Feedwater System Pipe DefectsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to maintain a Unit 2 auxiliary feedwater system (AFW) pipe segment containing linear defects in accordance with the design and material specifications. As a corrective action, the licensee performed light filing to remove the defects from this pipe segment. The licensee entered the failure to maintain the AFW pipe segment in accordance with the design into the corrective action program (CAP) as action request (AR) 02084077, and was evaluating additional corrective actions. This finding was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to maintain the Unit 2 AFW pipe segment containing linear defects in accordance with the design and material specifications could result in an increase in the possibility of pipe leakage or failure. In addition, the failure to maintain the AFW pipe segment containing linear defects in accordance with the design and material specification adversely affected the Mitigating System Cornerstone attribute of Equipment Performance because it could result in failure of AFW piping which would reduce the availability and reliability of the this mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors answered Yes to screening question A.1 of Exhibit 2. Although this finding adversely affected the design or qualification of the AFW pipe segments, the finding screened as very low safety significance (Green), because it did not result in the loss of operability or functionality of the affected pipe segment. This finding has a cross-cutting aspect in the Teamwork (H.4) component of the human performance cross-cutting area. Specifically, the licensees Projects Team responsible for the AFW modifications did not effectively communicate and coordinate with the licensees Programs Engineering Group for resolution of the AFW pipe nonconforming conditions to ensure nuclear safety was maintained.
05000266/FIN-2015004-012015Q4Point BeachFailure to Follow Fire Protection Program Requirements for Care, Use and Maintenance of Fire HoseThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of license condition 4.F for the licensees failure to have procedures or instructions to prevent firefighting booster hoses from being kinked and/or twisted on hose reels. Specifically, booster hoses were installed on hose reels in both units containments and in the turbine building (TB), which were twisted and kinked. The licensees corrective actions included rewinding hoses in the Unit 2 containment, four hoses in the TB, and creating compensatory measures for hose reels for the Unit 1 containment. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee failed to ensure that activities such as inspection, testing, and maintenance of fire protection systems were prescribed and accomplished in accordance with documented instructions, procedures, and drawings. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1A, because the inspectors determined that the impact of a fire would be limited to one train/division of equipment for the affected fire areas and at least one credited safe shutdown path would be unaffected. This finding has a cross-cutting aspect of Training (H.9), in the area of human performance, because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce, and instill nuclear safety values. Specifically, the inspectors determined that operations personnel were not adequately trained to recognize deficiencies associated with firefighting equipment standards, such as kinked and twisted hoses on hose reels, and subsequently failed to initiate actions to remedy such conditions.
05000395/FIN-2015004-042015Q4SummerLicensee-Identified ViolationV.C. Summer Nuclear Station TS 6.8.1 states in part that procedures shall be implemented for the Fire Protection Program. Contrary to the above, on March 3, 2015, April 7, 2015, October 10, 2015, and October 11, 2015, the licensee failed to implement Fire Protection Program (FPP) procedure, FPP-25, Fire Containment, Revision 4H, in that required fire protection permits were not obtained while Appendix R fire doors were left open without being manned. Specifically, Appendix R fire door DRSW/302 was found blocked open by a trash can on March 3, 2015, Appendix R fire door DRSW/203 was found blocked open by multiple cords on April 7, 2015, and Appendix R fire door DRIB/314 was found open on October 10, 2015 and again on October 11, 2015. The inspectors used IMC 0609, Appendix F, Attachment 1, to determine that the finding was of very low safety significance (Green) because smoke or heat detection was present in all adjacent fire areas. Further, since plant personnel would be alerted in the event of a fire and the doors could then be closed, equipment required for safe shutdown would not be impacted. Fire door DRSW/302 was closed upon discovery, while fire permits and fire watches were required for DRSW/203 and DRIB/314 to support ongoing plant maintenance. The CRs for doors DRSW/302, DRSW/203 and DRIB/314 being found open are CR-15-01015, CR-15-01546 and CR-15-04950 respectively.
05000395/FIN-2015004-032015Q4SummerLicensee-Identified Violation10 CFR 50, Appendix B, Criterion XVI states in part that conditions adverse to quality shall be promptly identified and corrected. Contrary to this, on August 4, 2015, the licensee discovered degradation of the B SW intake screen allowing the introduction of fish into the B SW pump bay. The licensee had initiated WO0601588 in 2006 to repair/rebuild the screen but failed to correct. A review of IMC0609, Appendix A, determined the finding was of very low safety significance (Green) because the finding was not a design deficiency and it did not result in a loss of function. The licensee has documented this problem in their CAP as CR-15-03574.
05000395/FIN-2015004-022015Q4SummerFailure to Accomplish Procedure for Diagnostic Testing Resulting in Valve FailuresThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which requires in part that activities affecting quality shall be accomplished in accordance with procedures. Specifically, the licensee failed to accomplish preventative maintenance diagnostic testing in accordance with their station administrative program procedure, SAP-160, Motor Operated Valve Program, Revision 1, to identify degradation of a torque switch that led to two failures of stroke time testing of A train reactor building spray (SP) sump isolation valve, XVG03005A-SP. This also resulted in a loss of safety function involving reactor building spray. The licensee entered the problem into their corrective action program as condition report, CR-15-00541. The inspectors identified a performance deficiency (PD) for the failure to accomplish the requirements of SAP-160 leading to two failures of XVG03005A-SP. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor because it adversely impacted the barrier integrity cornerstone objective to provide reasonable assurance that the reactor building or containment protects the public from radionuclide releases caused by accidents or events and the related attribute of structures, systems and components (SSC) performance. Specifically, the licensee failed to perform preventative maintenance diagnostic testing required by SAP-160 to identify degradation of a torque switch for XVG03005A-SP. The inspectors used IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, dated July 1, 2012, and IMC 0609, Appendix H, Containment Integrity Significance Determination Process, dated May 6, 2004, and determined the finding was of very low safety significance or Green, because the finding did not represent a significant impact to Large Early Release Failure. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of problem identification and resolution and the aspect of evaluation, P.2, because the licensee failed to thoroughly evaluate the failures of XVG03005A-SP to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000395/FIN-2015004-012015Q4SummerDeparture from NFPA 80-1973 for Replacement Fire DoorsAn unresolved item (URI) was identified by the inspectors during the walkdown of the Intermediate building fire area involving an engineering justification for a departure from NFPA 80-1973 as required by the Fire Protection Program for replacement fire doors DRIB/105A and DRIB/105B located in the intermediate building. The inspectors identified an issue of concern regarding replacement of fire door DRIB/105 with a single door jamb containing two fire doors DRIB/105A and DRIB/105B. These replacement doors were installed in a back to back configuration to provide a pressure barrier function in addition to the fire barrier function but were not self-closing as required by NFPA 80-1973. The licensee subsequently initiated CR-15-04027 to evaluate this issue of concern. Pending completion of additional evaluations needed to determine the existence of any related performance deficiencies, this is identified as URI 05000395/2015004-01, Departure from NFPA 80-1973 for Replacement Fire Doors.
05000266/FIN-2015003-042015Q3Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 3.8.9; Distribution SystemsOperating, Condition A, which required the licensee to immediately declare associated supported features inoperable for the 4.16 kV safeguards busses. Failure to implement this action subsequently required the licensee to place both units in mode 5 within 36 hours. Contrary to the above, the licensee discovered that numerous occasions existed over the past three years where safetyrelated 4.16kV switchgear associated with B Train EDGs was inoperable due to the inoperability of the W-185A and W-185B, 1A-06 and 2A-06 Switchgear room fans, which were required support systems for the EDGs and associated switchgear. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, and determined that the finding required a detailed risk evaluation which was performed by Region III SRAs. The SRAs gathered data from licensee GOTHIC model calculations, licensee engineering evaluations associated with the POR of the condition and the NRCs Standard Plant Analysis Risk model. Based on the SSCs being available for their respective 24-hour mission time(s), the SRAs determined that the increase in CDF for this issue was negligible and the delta risk is of very low safety significance (i.e., Green). The licensee reported this condition in LER 2015-004-00, which was closed in Section 4OA3 of this report. The licensees corrective actions included improving administrative and procedural controls for removing these fans from service and used lessons learned from this condition to implement corrective actions to improve procedural guidance for similar activities where ventilation systems may cause support system inoperabilities.
05000266/FIN-2015003-032015Q3Point BeachFailure to Perform a Written Safety Evaluation for FSAR ChangesThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance for the licensees failure to perform a safety evaluation to demonstrate that the removal of statements from the FSAR did not require a license amendment. Specifically, the licensee failed to perform a safety evaluation to determine whether removing an FSAR statement, which defined the RHR pump cubicle design flood height as seven feet, could be performed without a license amendment. The licensee entered the deficiency in their CAP as Action Request (AR) 02069425 by which the licensee intends on re-evaluating the 1996 FSAR change. The inspectors determined that the finding was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, inappropriately removing the information from the FSAR allowed the licensee to decrease the design basis flood protection height of the RHR compartments and significantly reduced the available time to isolate the leaking RHR pump seal. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015003-022015Q3Point BeachPotential Failure of Multiple Safety-Related Trains During Flooding EventsThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensees failure to ensure that a non-Category I (seismic) component failure, that results in flooding, would not adversely affect safety-related equipment needed to get the plant to safe shutdown (SSD) or to limit the consequences of an accident. Specifically, the design of Point Beach did not ensure that the Residual Heat Removal (RHR) pumps would be protected from all credible non-Category I (seismic) system failures. The licensees corrective actions included an extensive internal flooding design review, which will result in an updated Final Safety Analysis Report (FSAR) with a more detailed description of the stations flooding licensing basis; modifications to multiple flood barriers to bring them into compliance with the licensees flooding licensing basis; installation of additional flood level alarms where necessary, and evaluation or modification of service water (SW) piping to properly qualify it as seismic. The inspectors determined that the finding was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design resulted in an unanalyzed condition and loss of safety function of the RHR system while the plants were in Modes 4, 5, and 6, when relying on the RHR system for decay heat removal. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors answered yes to question 2 of the screening questions because the finding represented a loss of safety function. Thus the inspectors consulted the Region III Senior Risk Analysts (SRAs) who performed a detailed risk evaluation and determined that the finding was of very low safety significance (Green). The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015003-012015Q3Point BeachIncomplete Functionality Assessment for Flooding in the Diesel Generator BuildingThe inspectors identified a finding of very low safety significance for the licensees failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 19. Specifically, when the licensee identified that internal flood sources in the diesel generator building (DGB) were larger than the drain capacity, they failed to identify all affected structures, systems, and components (SSCs). The DGB contains predominately Train B emergency power systems; however, the fuel oil transfer pumps for the Train A emergency diesel generators are located in the southeast corner of the building. The licensee failed to assess the effects of flooding on the Train A fuel oil transfer pumps. The licensees corrective actions included the creation of an adverse condition monitoring plan, which implemented an hourly flood watch in the DGB when the fire pump was manually started. The inspectors determined that the finding was more than minor, because if left uncorrected, it would potentially result in a more safety significant issue. Specifically, the failure to evaluate the effects of flooding on all SSCs resulted in inadequate compensatory measures. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. For the time period in question, May 17, 2015 to September 17, 2015, the inspectors reviewed the security door card reader reports and starting sump levels for the DGB and found that during times when the fire pumps were running, station personnel had toured the DGB at a frequency that would have identified flooding conditions before a loss of system function. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Evaluation (P.2), in the area of Problem Identification and Resolution (PI&R), for failing to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000395/FIN-2015003-022015Q3SummerLicensee-Identified ViolationTechnical Specification (TS) 6.8.4.e, Radioactive Effluent Controls Program, requires the control and assessment of radioactive effluents be performed per the methodologies in the ODCM. ODCM 1.2.1.1.b requires, when less than the minimum number of channels are operable on the Main Plant Vent-Exhaust System (RMA-0003), releases can continue provided continuous samples with auxiliary equipment are collected. Contrary to this requirement, on October 26, 2012, with RMA-0003 rendered inoperable due to a planned loss of power, releases continued for approximately 7 hours via this pathway without the collection of continuous samples with auxiliary sampling equipment. The license entered the event in the CAP as CR 12-04908. This finding was determined to be Green because it did not involve a substantial failure to implement the radioactive effluent release program or result in an effluent release of radioactive material that exceeded the dose values in Appendix I to 10 CFR Part 50 and/or 10 CFR 20.1301. The licensees determination that no detectable releases of radioactive material occurred while RMA-0003 was inoperable was reviewed by the inspectors.
05000395/FIN-2015003-012015Q3SummerFailure to Assess and Manage Risk Associated with Emergent WorkThe inspectors identified a non-cited violation of 10 CFR 50.65 (a)(4) which requires in part that the licensee assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, the licensee failed to assess and manage the increase in risk for emergent work on the B train service water (SW) pump motor breaker. The licensee entered the problem into their corrective action program as condition report (CR) 15-03194. The inspectors identified a performance deficiency (PD) for the failure to assess and manage the increase in risk for work activities associated the B SW pump motor breaker in accordance with 10 CFR 50.65 (a)(4). The inspectors reviewed IMC0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor because it adversely impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of equipment perfomance involving availability and reliability. Specifically, the failure to identify increases in operational risk and implement risk management actions adversely affected the availability and reliability of those systems relied upon to respond to plant events. The inspectors used IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined the finding was of very low safety significance or Green, because the Incremental Core Damage Probability Deficit for the timeframe the B SW pump was unavailable was less than 1E-6. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of human performance and the aspect of work management, H.5, because the licensee failed to assess and manage the risk commensurate with the emergent work involving the B SW pump motor.
05000395/FIN-2015002-012015Q2SummerFailure to Maintain Fire Door/Steam Propagation Barrier in Accordance With the Fire Protection Program ProcedureThe inspectors identified a non-cited violation of Technical Specifications (TS) 6.8.1.f, Fire Protection Program (FPP) procedures, which involved a failure to comply with the requirements of FPP-025, Fire Containment, Revision (Rev.) 4H, for maintaining the operability of a fire door and steam propagation barrier (SPB), DRAB/319. The licensee entered the problem into their corrective action program as condition report (CR) 15-00662. The inspectors identified a performance deficiency (PD) for the failure to maintain the fire door and SPB operable per the requirements of FPP-025. The inspectors reviewed inspector manual chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. In regards to the fire confinement function of DRAB/319, the inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and performed a Phase 1 analysis to determine the finding was of very low significance or Green. The fire confinement program element was not of low degradation, the non-suppression probability was 0.1, the fire frequencies related to the affected fire zones AB01.10 and FH01.01 were 3.31E-3 and 8.69E-5 respectfully, and the duration of the component inoperability was approximately 12 hours or 0.00137, which resulted in screening check frequency of 4.65E-7 that was less than the screening criteria of 1E-6. Additionally, the inspectors noted minimal fixed combustibles and ignition sources in the near vicinity of both sides of DRAB/319, and the fire detection instrumentation in both affected fire zones remained operable allowing an operator response in the event of a fire. In regards to the SPB function of DRAB/219, the inspectors used IMC 0609, Appendix A, SDP for Findings at- Power, dated June 19, 2012, and determined the finding was also of very low safety significance, or Green, because it was not a design deficiency or loss of system function impacting TS. The resulting increase of humidity above equipment qualification test limits of one train of reactor vessel level instrumentation system transmitters would likely not have resulted in a loss of function. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of human performance and the aspect of resources, H.1, because the licensee failed to ensure that the fire door closure mechanism was adequate to close the door for the protection of equipment important to safety.
05000298/FIN-2015001-012015Q1CooperInadequate Operations ProcedureThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the inadequate Operations Procedure 2.2.7, Condensate Storage and Transfer System, Revision 56. Specifically, the procedure did not require that the affected system, either the high pressure coolant injection system or the reactor core isolation cooling system, be declared inoperable when one or more of the high pressure coolant injection or reactor core isolation cooling test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, or RCIC-MOV-33, were moved off of their closed (passive safety function position) seats. The license entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-00274. The failure to establish and maintain a correct filling procedure required by Technical Specification 5.4.1.a. was a performance deficiency and resulted in the licensees failure to declare the high pressure coolant injection and reactor core isolation cooling systems inoperable when required to do so. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the high pressure coolant injection and reactor core isolation cooling systems were not declared inoperable when their test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, and RCIC-MOV-33, were taken off their normally closed (passive safety function position) seats. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction techniques. Specifically, licensee personnel fell into a pattern of acceptance regarding Procedure 2.2.7. This resulted in a failure to question the lack of an operability caution statement, even though there was other guidance in the inservice inspection program to that effect (H.12).
05000298/FIN-2014005-022014Q4CooperImplementation of Enforcement Guidance Memorandum 11-003, Revision 2, Causes Conditions Prohibited by Technical SpecificationsDuring Refueling Outage 28, Cooper Nuclear Station performed Operations with a Potential for Draining the Reactor Vessel (OPDRV) activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measure to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliances with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to: (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this Licensee Event Report for potential performance deficiencies and/or violations of regulatory requirements. The inspectors reviewed the stations implementation of the Enforcement Guidance Memorandum 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations with a potential for draining the reactor vessel activities were logged in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1 requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from October 3, 2014 to October 22, 2014, Cooper Nuclear Station performed operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. Specifically, the station conducted the following seven operations with a potential for draining the reactor vessel activities without an operable secondary containment: Draining reactor recirculation pump without the jet pump plugs fully installed Control rod drive maintenance Removal of jet pump plugs associated with reactor recirculation pump B maintenance Venting the control rod drives Defeating the scram function for two control rod drives and support IVVI inspections Reactor recirculation pump A seal maintenance Control rod drive freeze seal These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Condition Reports CR-CNS-2014-06293. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The Licensee Event Report is closed.
05000395/FIN-2014005-012014Q4SummerLicensee-Identified Violation10 CFR Part 50 Criterion XVI, Corrective Actions, requires in part that measures be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to this, the licensee failed to correct the valve packing friction and the degradation of the valve opening springs associated with the SW system outlet header component cooling loop B cross-connect Valve, XVG090627B-CC after the valve failed surveillance testing on October 30, 2012. In addition, the licensee failed to correct this condition adverse to quality during a mini-outage on March 30, 2013. This degraded condition resulted in the failure to stroke open of both valves XVG090627B-CC and XVG09627A-CC during the 2014 refueling outage and resulted in violations of Technical Specifications 4.0.5, Inservice Testing Surveillance, 3.7.3, Component Cooling Water System, and 3.7.4, Service Water System. The conditions adverse to quality were identified by the licensee and entered in the licensees corrective action program as CR- 13-00930, CR-14-01926, and CR-14-02282. The failure to correct a condition adverse to quality related to the service water (SW) system outlet header component cooling water (CCW) system loop cross-connect valves, XVG09627B-CC and XVG09627A-CC was a PD. Both valves XVG09627A-CC and XVG09627B-CC failed surveillance tests and both trains of the CCW system were simultaneously rendered past inoperable during the 2014 refueling outage. The PD was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609.04 and was determined to impact the Mitigating Systems Cornerstone. Significance Determination Process (SDP) screening performed in accordance with NRC IMC 0609, Appendix A, Exhibit 2, determined that the PD represented a loss of system function and required a Detailed Risk Evaluation. A Detailed SDP risk evaluation was performed by a regional SRA in accordance with the guidance of NRC IMC 0609, Appendix A, using the latest NRC Virgil C. Summer SPAR risk model. The major analysis assumptions included: Both valves considered simultaneously inoperable for a one year exposure period; Random or seismically induced pipe failure in combination with failure of the normal CCW supply, the Demineralized Water (DW) System, would be required to yield a demand for valves XVG09627A/B to open to fill the CCW system; Operator action was credited to identify and isolate leakage and restart the standby CCW train and re-align components per annunciator and abnormal operating procedures; EPRI generic pipe rupture frequency data was utilized. The dominant sequence was a seismic initiator which resulted in a loss of offsite power, loss of the DW system, and seismically induced leakage in the CCW system; failure of the Backup SW fill function due to the PD; failure of the operator to recover the CCW system resulting in a loss of seal cooling which led to a seal loss of coolant accident and core damage. The risk was mitigated by the low probability of the associated rupture initiators and the ability to manually operate the valves or start the alternate CCW train components. The Detailed SDP Risk Evaluation determined that the risk increase due to the PD was an increase in core damage frequency of <1 E-6/year, a GREEN finding of very low safety significance.
05000298/FIN-2014005-012014Q4CooperFailure to Follow Procedure for Post Maintenance TestingThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow Special Procedure GEH-TP-116, Procedure for the Operation and Maintenance of the REM*TAKE-2/D-100 Modified REM*TAKE 2, Revision 3, for postmaintenance testing following corrective maintenance. Specifically, the licensee did not follow post-maintenance testing requirements associated with the calibration of the bleeder valve for the REM*TAKE-2/D-100 tool following corrective maintenance to address water intrusion. This resulted in the bleeder valve being misadjusted and nullifying the fail-safe feature of the REM*TAKE-2/D-100 tool. With the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when the supplemental employee inadvertently pressed the disengage button. No reactor fuel was damaged as indicated by normal radiation levels and air samples on the refuel floor and reactor water coolant samples. The licensees immediate corrective actions for the event was to suspended all in-vessel maintenance activities and remove REM*Take-2/D-100 grapple from service and determined functionality of the tool. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-06809. The licensees failure to follow the post-maintenance testing requirements in Special Procedure GEH-TP-116 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the associated objective of maintaining functionality of fuel cladding. Specifically, with the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when a supplemental employee inadvertently pressed the disengage button. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 09, 2014, inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the fuel barrier because it: (1) does not increase the potential for failure of the freeze seal or if unmitigated have the potential to cause a disruption of residual heat removal/decay heat removal or a loss of inventory event; (2) does not involve two or more adjacent control rods with the potential to, or actually, add postive reactivity; and (3) does not degrade the ability to isolate a drain down or leakage path. The finding has a cross-cutting aspect in the area of human performance associated with the field presence component because the licensee failed to ensure supervisory and management oversight of work activities including contractors and supplemental personnel (H.2).
05000285/FIN-2014009-072014Q3Fort CalhounFailure to Accurately Model Flow Path for External Flood MitigationA non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to accurately model cell level control of river water during external flooding events. Specifically, the licensee failed to account for losses due to the physical obstructions of trash racks for inflowing river water, the decreased withdrawal rate of the raw water pumps due to fouling across the traveling screens, and a bounding in leakage rate for the sluice gates when the river level is at maximum level of 1014 mean sea level and the intake cell levels are at minimum level of 9769 . The licensee entered this issue into its corrective action program as Condition Report 2014-09155, performed an operability determination, and initiated action to update station calculations related to intake cell level control. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the finding would have the potential to lead to a more significant safety concern. Specifically, the failure to accurately model flow in and out of the cells could adversely affect the external flooding mitigation strategy beyond previously identified equipment capacities and operator actions. This finding was associated with the Mitigating Systems Cornerstone. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, in that the licensee failed to incorporate relevant internal operating experience related to previous NRC inspection into Calculation FC08081.
05000285/FIN-2014009-242014Q3Fort CalhounLicensee-Identified ViolationTitle 10 CFR 50.73(a)(1) requires, in part, that licensees shall submit a licensee event report for any event of the type described in this paragraph within 60 days after the discovery of the event. Contrary to the above, on February 5, 2012, November 15, 2011, and February 19, 2013, the licensee failed to submit a licensee event report for an event meeting the requirements for reporting specified in 10 CFR 50.73. Specifically, the licensee submitted Licensee Event Reports 2012-013, 2012-015 and 2013-001 greater than 60 days following discovery of a reportable event. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV, non-cited violation. The licensee entered this issue into their corrective action program as CR 2014-02792.
05000285/FIN-2014009-132014Q3Fort CalhounFailure to Perform Evaluation for Design ChangeA cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, was identified involving the failure to evaluate if a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the licensee did not evaluate a change that would permanently substitute a manual action for an automatic action to add water and nitrogen gas to the component cooling water surge tank. The licensee entered this issue into its corrective action program as Condition Report 2014-09080 and initiated action to evaluate the change to the component cooling water system. The NRC determined that the licensees failure to perform an evaluation prior to implementing a proposed change described in the Updated Safety Analysis Report was a violation of 10 CFR 50.59. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC evaluated the performance deficiency using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the team evaluated this finding using the significance determination process to assess its significance. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy this performance deficiency is being characterized as a Severity Level IV violation. The team determined that a cross-cutting aspect was not applicable to this finding because the issue was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-082014Q3Fort CalhounFailure to Report Loss of Environmental Qualification of Safety Related Limit Switches within Required Time LimitsA non-cited violation of 10 CFR 50.73(a)(1), Licensee Event Report System, was identified involving the failure to submit a required licensee event report. Specifically, the licensee failed to report within 60 days the discovery that NamcoTM Type EA 180 limit switches were not environmentally qualified as required due to inadequate maintenance procedures, a condition that resulted in operation prohibited by the plants technical specifications. The licensee restored compliance by submitting Licensee Event Report 05000285/2014-004 on June 20, 2014. The licensee entered this issue into its corrective action program as Condition Report 2014-08454. The NRC determined that the failure to submit a licensee event report within the time limits specified in regulations was a violation of 10 CFR 50.73. This violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV, non-cited violation. The NRC determined that a cross-cutting aspect was not applicable because the issue was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-042014Q3Fort CalhounFailure to Perform an Evaluation for a New Operator Manual Action to Refill Component Cooling Water System During Post- Accident ConditionsA non-cited violation of 10 CFR 50.59, Changes, Test, and Experiments, was identified involving the failure to evaluate if a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the licensee failed to evaluate if a change implemented under Engineering Change 59252 that credited the non-safety related demineralized water system as a make-up source to the component cooling water system during post-accident conditions represented an adverse change to the Updated Safety Analysis Report described design function. The licensee entered this deficiency into its corrective action program for resolution as Condition Report 2014-09151 and established action items to update Engineering Change 59252. The NRC determined that the licensees failure to perform an evaluation prior to implementing a proposed change described in the Updated Safety Analysis Report was a violation of 10 CFR 50.59. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the NRC evaluated the violation using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the NRC evaluated this finding using the significance determination process to assess its significance. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this performance deficiency is characterized as a Severity Level IV violation. The team determined that a cross-cutting aspect was not applicable because the issue involving the failure to perform an adequate 10 CFR 50.59 evaluation was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-022014Q3Fort CalhounMultiple Examples of Failure to Evaluate Operability of Degraded or Non-Conforming ConditionMultiple examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to follow Procedure OP-FC-108-115, Operability Determinations, Revision 0a. In each example, the team identified that the licensee failed to make an immediate determination of operability for a degraded or non-conforming condition or failed to make an immediate determination of operability based on a detailed examination of the deficiency. The licensee took immediate corrective actions to update the incomplete or inaccurate operability determinations and entered the collective failures to follow station operability procedures into their corrective action program as Condition Report 2014-09163. This performance deficiency was more than minor, and therefore a finding, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective of ensuring the reliability of systems that respond to initiating events. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to use decision-making practices that demonstrate that a proposed action is to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee made non-conservative decisions related to the impact of degraded or non-conforming conditions.
05000285/FIN-2014009-032014Q3Fort CalhounFailure to Adequately Perform an Operability Evaluation and a 50.59 EvaluationA non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to evaluate and implement adequate compensatory measures for a degraded condition associated with raw water pump AC-10C. Specifically, the licensees operability determination established a compensatory measure to place pump AC-10C in pull-to-lock, contrary to the system single failure analysis design criteria described in the Updated Safety Analysis Report. The licensee entered this issue into its corrective action program as Condition Reports 2014-09104 and 2014-08515 and performed an operability evaluation and associated 10 CFR 50.59 evaluation that used an acceptable compensatory measure to pump water from affected manholes prior to affecting the degraded power feeder cable for raw water pump AC-10C. The NRC evaluated this performance deficiency as both a reactor oversight process finding and a traditional enforcement violation. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of problem identification and resolution with an aspect of evaluation because the licensee failed to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2). In addition, because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function in that the failure to obtain a license amendment for a change that could result in a malfunction of a structure, system or component with a different result than previously evaluated in the Updated Safety Analysis Report is in violation of 10 CFR 50.59(c)(2)(vi), the NRC also evaluated the violation using traditional enforcement. Since this violation is associated with a Green reactor oversight process violation, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.
05000285/FIN-2014009-222014Q3Fort CalhounLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from initial construction until January 13, 2013, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to control the design inputs to ensure that piping in the chemical and volume control system would perform acceptably during a seismic event. This finding is of very low safety significance (Green) because a chemical and volume control system piping failure event is enveloped by the small break loss of coolant accident as described in Updated Safety Analysis Report Section 14.5.5. This issue was entered into the licensees corrective action program as CR 2013-01796.
05000285/FIN-2014009-212014Q3Fort CalhounFailure to Take Timely Corrective Actions for an Unsealed Raw Water System Control PanelA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take corrective actions to address a design deficiency affecting the control panel for raw water strainer AC-12B. Consequently, the panel experienced a water intrusion event on August 3, 2014, resulting in an unplanned inoperability of the raw water system. Following identification of this issue, the licensee implemented corrective actions to seal conduits leading to control panel AI-348 to prevent future water intrusion. The licensee entered this issue into its corrective action program as Condition Report 2014-09572. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of problem identification and resolution in that the licensee failed to adequately review and provide timely responses to past operating experience that demonstrated that panel AI-348 was susceptible to water intrusion.
05000285/FIN-2014009-202014Q3Fort CalhounFailure to Correct Conditions Adverse to Quality in the Diesel Generator Stating Air SystemA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective actions to address service life related degradation of the emergency diesel generator starting air system. As a result, diesel generator 1 failed to roll during planned surveillance testing due to a degraded diesel starting air valve. The licensee replaced the faulty starting air valve and implemented corrective actions to develop preventative maintenance strategies for the starting air system. The licensee entered this issue into the corrective action program as Condition Report 2014-09424. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings , Exhibit 3, Mitigating Systems Screening Questions, dated May 9, 2014, the finding was of very low safety significance (Green) because the finding does not represent a loss of system safety function and the finding does not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to recognize and plan for the possibility of latent issues, and inherent risk, even while expecting successful outcomes when determining the repair schedule for starting air valve SA-148.
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