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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000366/FIN-2018003-01 | 2018Q3 | Green | H.1 | Self-revealing | Inoperability of 2A EDG Due to Inadequate Acceptance Criteria for Determining Cleaning Requirements of Emergency Diesel Generator Day Tanks | The inspectors documented a Green, self-revealing, non-cited violation of Unit 2 Technical Specification 5.4.1(a) for the licensees failure to incorporate preventative maintenance criteria for Emergency Diesel Generator (EDG) day tanks as recommended by Regulatory Guide (RG) 1.33, 9.a. Specifically, procedure 52SV-R43-001-0, Diesel, Alternator, and Accessories Inspection, Ver. 30.4, did not contain deterministic criteria in the visual inspection of the fuel filters to initiate the cleaning of the EDG day tanks and thus prevent EDG inoperability. The EDG day tanks had never been inspected and cleaned. |
| 05000321/FIN-2018002-01 | 2018Q2 | Licensee-identified | Enforcement Action (EA)-18-100: Unanalyzed Conditions for a Postulated Fire Discovered During NFPA 805 Transition | On April 3, 2017, the licensee submitted Licensee Event Report (LER) 05000321, 366/2017-001-00: Unanalyzed Conditions for a Postulated Fire Discovered During NFPA 805 Transition documenting the discovery of a condition of non-compliance with the sites fire protection program (FPP). In preparation for transiting the fire protection licensing basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), a weak-link and operator manual action analysis was completed for Information Notice 92-18 type hot shorts on motor operated valves (MOV). The licensees examination of their Appendix R Safe Shutdown Analysis identified circuit configurations in multiple fire areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. The licensee failed to protect MOV cables associated with the RHR and RCIC emergency cooling systems in fire areas 0024 (Main Control Room), 1203F (Unit 1 Reactor Building), 1205F (Unit 1 Reactor Building), and 2203F (Unit 2 Reactor Building). Specifically, the licensee failed to ensure that fire induced cable impacts cannot bypass the limit and torque switches and result in physical damage to the MOVs, thus preventing the MOVs from being operated from the Main Control Room, Remote Shutdown Panel, or locally. This condition could prevent operators from achieving and maintaining safe shutdown (SSD) of the plant in the case of a postulated fire. A licensee-identified non-compliance with 10 CFR Part 50, Appendix R, Section III.G.2, was identified for the licensees failure to protect one of the redundant trains of equipment needed to achieve post-fire SSD from fire damage. Specifically, the licensee failed to use one of the means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. The inspectors performed a detailed review of the information and documents related to the LER and discussed the condition with the licensee to assess the adequacy of the licensees compensatory measures and corrective actions. Corrective Action(s): Hourly fire watches and Fire Action Statements were initiated to address the postulated condition for the identified MOVs. Additionally, the licensee committed to completing physical plant modifications to the impacted MOVs during the next Unit 1 and Unit 2 plant refueling outages to rectify the issue of potential spurious operation of the associated MOVs associated with this LER. Corrective Action Reference(s): The licensee entered this issue into their Corrective Action Program (CAP) as condition reports (CRs) 10326399, 10326401, 10326402, 10326404, and 10326405. Enforcement: Violation: 10 CFR Part 50.48(b)(1) requires that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G. 10 CFR 50, Appendix R, Section III.G.2, states, in part, that where cables or equipment, that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided: (a) separation of cables and equipment by a fire barrier having a 3-hour rating, (b) separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards and with fire detectors and an automatic fire suppression system in the fire area, or (c) enclosure of cables and equipment in a fire barrier having a 1-hour rating and with fire detectors and an automatic fire suppression system in the fire area. Contrary to the above, the licensee failed to use one of the means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. Specifically from October 1974 to April 2017, the licensee had not met the requirements of 10 CFR Part 50.48(b) to identify and protect cabling of 51 Unit 1 and Unit 2 RHR and RCIC emergency cooling system MOVs in fire areas 0024 (Main Control Room), 1203F (Unit 1 Reactor Building), 1205F (Unit 1 Reactor Building), and 2203F (Unit 2 Reactor Building). On April 3, 2017, the licensee identified the failure to protect equipment that was required to mitigate fire events and determined that fire damage could cause mal-operation of the affected MOVs, potentially leading to fire induced cable impacts which bypass the limit and torque switches and result in physical damage to the MOVs, thus preventing the MOVs from being operated from the Main Control Room, Remote Shutdown Panel, or locally. A fire-induced failure could have caused the loss of the required Safe Shutdown components. Severity/Significance: Failure to protect one train of cables and equipment necessary to achieve post-fire SSD from fire damage for fire areas designated in the Fire Protection Program (FPP) as meeting Appendix R, Section III.G.2, was a performance deficiency. This finding was more than minor because it was associated with the reactor safety mitigating system cornerstone attribute of protection against external events (i.e., fire). Specifically, failure to protect safe shutdown cables and equipment from fire damage negatively affected the reactor safety mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this issue relates to fire protection and this non-compliance was identified as a part of the sites transition to NFPA 805, this issue is being dispositioned in accordance with Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) of the NRC Enforcement Policy. The significance of this licensee-identified non-compliance with 10 CFR Part 50, Appendix R, Section III.G.2, was determined by the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. The quantitative screening approach performed by a Region II Senior Risk Analyst resulted in a calculated delta core damage frequency (CDF) of less than 1E-04, which screens this noncompliance to less-than-red significance. Additionally, in order to verify that this noncompliance was not associated with a finding of high safety significance (Red), inspectors reviewed qualitative and quantitative risk analyses performed by the licensee. These risk evaluations took ignition source and target information from the ongoing HNP fire PRA to demonstrate that the significance of the non-compliances were less-thanthan 1E-4/year). The inspectors also performed walk-downs to verify key assumptions were applicable. Based on the ignition frequency of fire sources in the affected areas, inspectors determined that the significance of this non-compliance was less-than-red. The inspectors also noted that the values in the licensees quantitative analysis were conservative, in that they used screening values instead of more detailed values. This provided additional confidence that this non-compliance was not associated with a finding of high safety significance (Red). The inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), a Confirmatory Order (ML16223A467) which extended the period for discretion, and Inspection Manual Chapter 0305. On April 4, 2018 (ML18096A955), the licensee submitted a license amendment request to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c). The inspectors reached this conclusion due to the fact that this issue was licensee-identified and will be addressed during the licensees transition to NFPA 805, it was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance (Red). | ||
| 05000321/FIN-2018001-01 | 2018Q1 | Green | NRC identified | Failure to comply with Type B shipping container Certificate of Compliance (CoC) requirements. | An NRC Identified Green NCV of 10 Code of Federal Regulations (CFR)71.17, General license: NRC-approved package, was identified for the licensees failure to comply with the Type B shipping container Certificate of Compliance (CoC) requirements. 10 CFR 71.17(c)(2)states, in part, that a holder of a General license to utilize an NRC-approved package shall comply with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part. Specifically, on several occasions the licensee placed in transit Type B containers which did not pass the CoC leak test requirement(s). | |
| 05000321/FIN-2018001-02 | 2018Q1 | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Hatch Nuclear Plant Technical Specification (TS) 5.7.2 states in part, areas with radiation levels greater than 1000 mRem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry.Contrary to the above, February 6, 2018, the licensee identified dose rates of 72 Rem/hr on contact, and 3.9 Rem/hr at 30 cm on the U-1 bottom head drain valve located in the 127 foot elevation of the Subpile room, in the Unit 1 Drywell. For approximately 4 hours, the entrance to the room was not locked or continuously guarded to prevent unauthorized entry as required by TS 5.7.2. Significance/Severity: The finding was of very low safety significance (Green) because it was not an as low as reasonably achievable (ALARA) planning issue, there was no overexposure nor potential for an overexposure, and the licensees ability to assess dose was not compromised.Corrective Action Reference(s):The licensee identified and documented the failure to control access to the Lock High Radiation Area (LHRA) in Condition Report 10458608. | |
| 05000321/FIN-2017004-02 | 2017Q4 | Green | NRC identified | Lack of Requirement for Engineering Evaluation of Scaffolding Near Safety-related Piping | An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to ensure engineering evaluations were performed when scaffolding was constructed within 2 inches of safety-related piping. The failure to ensure procedure NMP-MA-010, Erecting, Modifying, and Disassembling Scaffolding, required engineering evaluations when scaffolding was constructed within 2 inches of safety-related piping was a performance deficiency. The violation was entered into the licensees corrective action program as CR 10420643.The performance deficiency was more-than-minor because the licensees procedure, as written, would never require an engineering evaluation of any safety-related piping based on the exceptions granted in the procedure. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. The inspectors determined that the finding did not have an associated cross-cutting aspect because the discrepancy was introduced during a transition to a fleet standardized procedure, which occurred more than three years ago and was therefore not reflective of current licensee performance. | |
| 05000321/FIN-2017004-03 | 2017Q4 | Green | NRC identified | A violation of Technical Specification (TS) 3.4.3 was identified because two of eleven safety relief valves were found to be outside the tolerance allowed by TS Surveillance Requirement (SR) 3.4.3.1 for the opening set-point pressure. | Description: During the February 2017 Unit 2 refueling outage, all eleven 3-stage safety relief valves (SRVs) were removed and replaced. The SRVs were Target Rock model 0867F, a 3-stage valve design which was in its first use on Unit 2. This design was adopted as a corrective action to address corrosion bonding experienced by 2-stage SRV model 7687F valves which were previously in use at Hatch. "As-found" testing results indicated two of the eleven SRVs had experienced a setpoint drift during the previous operating cycle which resulted in their failure to meet the Technical Specification (TS) opening setpoint pressure as required by TS Surveillance Requirement (SR) 3.4.3.1. The SRV pilot valves were disassembled and inspected to determine the reason for the drift. The licensee determined that the abutment gap closed pre-maturely most likely due to loose manufacturing tolerances. For the 3-stage design, the pilot disc seating stresses should increase proportionally as reactor pressure increases to where a mechanical gap within the valve stem mechanism, referred to as the abutment gap, is closed. Additional pressure increases will cause the valve stem mechanism to reduce the disc seat pressure until the valve eventually opens. This same cause was previously identified in 2016 (CAR 264544) after two of eleven SRVs removed from Unit 1 also experienced setpoint drift. Because the Unit 2 valves were already installed when the cause was initially identified, there was no opportunity for the licensee to take corrective actions for the valves that are the subject of this LER. Additionally, there were no symptoms available to operators or maintenance personnel to indicate the potential for the set point drift prior to post-service testing. As a corrective action, when the eleven valves were removed for post-service testing, the licensee installed eleven refurbished pilot valves that underwent the corrective actions identified by CAR 264544 which included the vendors usage of revised tolerances.Enforcement: Hatch Unit 2 TS limiting condition for operation 3.4.3, Safety/Relief Valves, required 10 of 11 SRVs be operable in MODES 1, 2 and 3. With two or more SRVs inoperable, the required TS action must be taken by the applicable completion time. Contrary to the above, Unit 2 operated from the initiation of the degraded condition until February 6, 2017, with two SRVs inoperable. The inspectors concluded that the violation would normally be characterized as a Severity Level IV violation because it was of very low safety significance (Green). However, the NRC is exercising enforcement discretion (EA-18-006) in accordance with Section 3.10 of the Enforcement Policy because the violation was not associated with a licensee performance deficiency. This issue was documented in the licensees corrective action program as CR 10382586. | |
| 05000321/FIN-2017502-01 | 2017Q4 | Green | Licensee-identified | Licensee-Identified Violation | The following licensee-identified violation of NRC requirements was determined to be of very low safety significance or Severity Level IV and meet the NRC Enforcement Policy criteria for being dispositioned as a Non-Cited Violation. Because it had the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with Inspection Manual Chapter 0612, Appendix B. This finding was also determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involved the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan was reduced. Title 10 of the Code of Federal Regulations, Part 50.54(q) states, in part, that a licensee may make changes to emergency plans without prior NRC approval only if the changes do not reduce the effectiveness of the plans and the plans, as changed, continue to meet the standards of 50.47(b) and the requirements of Appendix E. Proposed changes that reduce the effectiveness of the approved emergency plans may not be implemented without application to and approval by the NRC. Contrary to the above, on multiple occasions between 2008 and 2014, the licensee implemented changes to their Radiological Emergency Plan and Emergency Action Levels (EALs) which reduced the effectiveness of the Plan. Specifically, the licensee deleted and/or changed EAL threshold values, all of which would have resulted in a change that reduced the effectiveness of the approved Emergency Plan and was implemented without application to and approval by the NRC. Because the violation was entered into the licensees corrective action program as Condition Report 10421212, it is being treated as a Green non-cited licensee-identified SL IV violation consistent with Section 2.3.2 of the Enforcement Policy. | |
| 05000366/FIN-2017004-01 | 2017Q4 | Green | H.9 | NRC identified | Continuous Fire Watch or Compensatory Measures Not Established per FHA | An NRC-identified non-cited violation (NCV) of Unit 2 License condition 2.C.(3)(a) Fire Protection was identified when on October 17, 2017, the licensee failed to establish a continuous fire watch or alternative compensatory measures required by Hatchs Fire Hazards Analysis (FHA), Appendix B, while the carbon dioxide fire protection system was nonfunctional during a routine maintenance outage for the 2C emergency diesel generator. Failure to establish a continuous fire watch or alternative compensatory actions as required by Hatchs Fire Hazards Analysis, Appendix B, when the low pressure carbon dioxide storage system became inoperable on October 17, 2017, was a performance deficiency. The licensee restored compliance on October 25, 2017, when the double fire door was shut, restoring functionality of the carbon dioxide system. The licensee entered this issue into the corrective action program as Condition Report (CR) 10423361.This performance deficiency was more-than-minor because the failure to establish a continuous fire watch or alternative compensatory measures adversely affected the reliability of the carbon dioxide system and/or compensatory measures. The finding screened to green because the alternate train of safe shutdown remained operable. The inspectors determined this performance deficiency had a cross cutting aspect in the Human Performance Area Training attribute because of the observed weakness in the application of FHA applicability statements. (H.9) |
| 05000321/FIN-2017009-01 | 2017Q4 | Green | H.14 | Self-revealing | Failure to Implement Corrective Actions to Preclude Repetition of Reactor Scrams Due to IRM Spiking | A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to establish measures to assure that corrective action was taken to preclude repetition of a significant condition adverse to quality (SCAQ). Specifically, the licensee failed to implement corrective actions that would have increased the reliability of the intermediate range monitoring (IRM) system and prevent repetitive reactor scrams for similar reasons. The licensee entered this issue in the CAP as CR 10356172.The failure to ensure corrective actions were taken for a SCAQ to preclude repetition of reactor scrams due to IRM spiking caused by external electronic noise was a performance deficiency. This performance deficiency is more-than-minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective because it resulted in a reactor scram, which upsets plant stability and challenges critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609 Appendix A, the inspectors determined that this finding is of very low safety significance (Green) because it did not involve the loss of mitigation equipment per Exhibit 1.B Transient Initiators. The inspectors determined that the finding had a cross-cutting aspect of Conservative Bias (H.14) within the cross-cutting area of Human Performance because the licensee failed to use decision making-practices that emphasize prudent choices over those that are simply allowable. |
| 05000321/FIN-2017003-01 | 2017Q3 | Green | NRC identified | Installation of Non-Conforming RPS Equipment | An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control was identified for failure to translate regulatory requirements and the design basis of the scram discharge volume (SDV) thermal probes into the System Evaluation Document, which resulted in the installation of a nonsafety-related terminal board in the reactor protection system (RPS). As an immediate corrective action the licensee installed fully qualified equipment. The failure to classify reactor protection system components as safety-related in accordance with design documents was a performance deficiency. The violation was entered into the licensee's corrective action program as CR 10344772.The performance deficiency was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the ensured reliability of the RPS system wasadversely affected because the installed components were not qualified for the application. The team used IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0612, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green), because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability. The inspectors determined that this finding did not have an associated cross-cutting aspect because this finding did not occur within the previous three years and is not reflective of current licensee performance. | |
| 05000321/FIN-2017002-01 | 2017Q2 | Green | Self-revealing | Hardened grease prevents 1RHRSW pump breaker operation | Green. A self-revealing, Green, non-cited violation (NCV) of Hatch Unit 1 Technical Specification 5.4 Procedures, was identified when procedures to rejuvenate grease in the 1C' residual heat removal service water (RHRSW) Pump breaker were not implemented resulting in failure of the pump to start. The violation was entered into the licensees corrective action program as condition report (CR) 10263236 and the breaker was replaced to restore compliance. Failure to rejuvenate the lubricating grease on 4kv DHPVR breakers in accordance with vendor guidance was a performance deficiency. Specifically, the hardened grease prevented the 1C RHRSW pump breaker from closing resulting in the inoperability of the 1C RHRSW pump. The performance deficiency was associated with the Mitigating Systems cornerstone and was more than minor because it adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At-Power, dated June 19, 2012. Because all four questions in Section A of Exhibit 2, Mitigating Systems Screening Questions, were answered no, the finding screened as Green. The inspectors determined that this finding did not have an associated cross cutting aspect because this finding is not reflective of current licensee performance. | |
| 05000366/FIN-2017002-02 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | Performance of Operations with Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment | The inspectors reviewed this LER for potential performance deficiencies and/or violations of regulatory requirements. In February 2017, during the Unit 2 refueling outage, operations with the potential to drain the reactor vessel (OPDRV) activities were performed while in Mode 5 (Refueling Mode) contrary to Technical Specification (TS) 3.6.4.1. Enforcement Guidance Memorandum (EGM) 11-003, Revision 3, provided required interim actions which were incorporated into procedure 31GO-OPS-025-0 Operations with the Potential to Drain the Reactor Vessel. This procedure was used during the OPDRV activities for the Unit 2 refueling outage. LER 05000366/2017-001- 00 is closed. Description: The inspectors reviewed the plants implementation of Enforcement Guidance Memorandum 11-003 during maintenance activities which had the potential to drain the reactor vessel during the Unit 2 refueling outage. The activities were: Local power range monitors removal and replacement February 10, 2017; Control rod drive insert / recouple activity February 11, 2017; and Hydraulic Control Unit Venting February 12-13, 2017. 15 These activities took place without secondary containment being operable. Inspectors verified compliance with the guidelines of Enforcement Guidance Memorandum 11-003 prior to and during these activities. This condition was documented in the licensees corrective action program as CR 10329405, 10329857, 10330152, and 10330153. Enforcement: Unit 2 TS 3.6.4.1 required, in part, that activities that had the potential to drain the reactor vessel be conducted only with secondary containment operable. Contrary to that requirement, the licensee conducted activities that could cause the reactor vessel to drain while secondary containment was inoperable. The NRC is exercising enforcement discretion (Enforcement Action (EA)-17-124) in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy because the violation was identified during the discretion period described in Enforcement Guidance Memorandum 11-003. Therefore, the NRC will not issue enforcement action for this violation, subject to the license amendment request which was submitted on April 20, 2017. | |
| 05000321/FIN-2017002-03 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | Noncompliance for Providing Inadequate Procedural Guidance for Post-Fire Safe Shutdown | Introduction: The inspectors identified a noncompliance with Hatch Technical Specification 5.4.1.a for the licensees failure to provide adequate procedural guidance in post-fire safe shutdown abnormal operating procedure of Abnormal Operating Procedure (AOP) 34AB-X43-001-1, Fire Procedure. Specifically AOP 34AB-X43-001-1 directs operators to perform manual actions that may not be adequate to reopen a credited valve that has spuriously closed. Description: During the transition to NFPA 805, the licensee identified multiple instances of cables for equipment required to achieve SSD not meeting the separation requirements of the current licensing basis. The licensee determined that this condition existed for FA 1105, East Cableway Foyer. It was discovered that cables were identified in the current Safe Shutdown Analysis Report (SSAR) for HPCI Steam Supply Isolation motor operated valve 1E41-F002 . These cables were dispositioned by taking an Operator Manual Action (OMA) to open links BB-49 and BB-57 in panel 1H11-P622. Further evaluation showed that the OMA would prevent the valve from spuriously clos ing, but it would not re-open the valve after a spurious closure, due to the power supply for this valve being unavailable due to fire impacts. The licensee determined that these conditions were caused by methodology weaknesses in the sites fire safe shutdown analysis. Upon discovery, the licensee implemented compensatory measures in the form of posting a roving fire watch in fire areas of concern, and revised the affected procedure. 19 Analysis of the Problem Failure to adequately implement the requirements contained in 10 CFR Part 50.48(b)(1), and Hatch Renewed Operating License Condition 2.C.(3) and 2.C.(3)(a), for Units 1 and 2 was a performance deficiency. This finding was more than minor because it was associated with the reactor safety mitigating system cornerstone attribute of protection against external events (i.e., fire). Because this issue relates to fire protection and this non-compliance was identified as a part of the sites transition to NFPA 805, this issue is being dispositioned in accordance with Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) of the NRC Enforcement Policy. In order to verify that this non-compliance was not associated with a finding of high safety significance (Red), inspectors revi ewed qualitative and quantitative risk analyses performed by the licensee. These risk evaluations took ignition source and target information from the ongoing Hatch fire PRA to demonstrate that the significance of the non-compliances were less-than-Red (i.e. core damage frequency (CDF) less than 1E-4/year). Inspectors determined that cables associated with some of the VFDRs were not located in the zone of influence (ZOI) of any credible ignition source. For cables that were located in the ZOI of a credible ignition source, inspectors were able to perform a calculation to determine the change in conditional core damage probability (CCDP), based on the postulated fire-affected equipment not being available. Based on these screenings, inspectors determined that the significance of this non-compliance was less- than-Red. A bounding risk assessment was performed by a regional SRA which included the review of the licensee and inspectors risk evaluations and confirmed the CDF risk increase due to this condition was less than 1E-4, and therefore less than RED. The inspectors determined that no cross-cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. Enforcement of the Problem 10 CFR Part 50.48(b)(1) requires that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Sections III.G, III.J, and III.O. Section III.G.2 requires, in part, that where cables and equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located in the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided: o separation of cables and equipment by a fire barrier having a 3-hour rating; or o separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. Fire detection and automatic fire suppression shall be installed in the fire area; or o enclosure of cables and equipment of one redundant train in a fire barrier having a 1-hour fire rating. Fire detection and automatic suppression shall be installed in the fire area. 20 Section III.G.3 requires, in part, that alternative shutdown capability be provided where the protection of systems whose function is required for how shutdown does not satisfy the requirement of Section III.G.2. Additionally, Hatch Technical Specifications 5.4.1.a, Procedures for Unit 1 states that written procedures shall be established, implemented, and maintained covering activities listed in NRC Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Item 6.v of Appendix A lists Plant Fires as an activity that requires written procedures. Contrary to the above, the licensee failed to meet the requirements of its documented fire protection program since initial plant licensing, in that: The licensee did not meet the requirements of 10 CFR Part 50, Appendix R, Section III.G.2 in that the licensee did not ensure that one of the redundant trains was free of fire damage by providing one of the following means stated in Section III.G.2. The licensee did not ensure alternative shutdown capability be available for 2 fire areas where the guidelines for ensuring one redundant train for safe shutdown be free of fire damage, as required by 10 CFR Part 50, Appendix R, Section III.G.3. The licensee failed to provide adequate procedural guidance to ensure fire safe shutdown due to a fire in FA 1105. CRs generated for these issues are listed in the Documents Reviewed section. Because the licensee committed to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c), t he NRC is exercising enforcement and reactor oversight process (ROP) discretion (EA-17-120) for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, this issue was identified and will be addressed during the licensees transition to NFPA 805, it was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance (Red) | |
| 05000321/FIN-2017002-04 | 2017Q2 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.6.4.1 requires secondary containment be operable in Mode 1 and during movements of irradiated fuel assemblies in the secondary containment. Contrary to the above, on February 8 at 1035, with Unit 1 operating at 100 percent RTP and Unit 2 conducting refueling operations, secondary containment was made inoperable when Unit 2 reactor building containment was breached for a scheduled refueling outage and a configuration control error on the Unit 2 standby gas treatment system provided a uncontrolled opening into the secondary containment for the Unit 1 reactor building and the common refueling floor. A temporary blind flange had been incorrectly installed on the upstream side vice downstream side of the Unit 2 standby gas treatment inlet isolation valve when the valve had been removed from the system for testing. This configuration rendered secondary containment for the Unit 1 reactor building and the common refueling floor inoperable. A senior reactor operator performing a plant tour noted the incorrect flange configuration and at 2017 on February 17, the blind flange was moved to the downstream side of the Unit 2 standby gas treatment inlet isolation valve to restore compliance. Inspectors screened the finding in accordance with IMC 609 Appendix A The Significance Determination Process (SDP) for Findings at-Power. The finding screened as very low safety significance (Green) because the questions in Appendix A Exhibit 3 for Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building, were answered no. This issue was documented in the licensees corrective action program as CR 10332592. | |
| 05000321/FIN-2017001-03 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.7.1 requires, in part, entrances into areas in which the intensity of r adiation is > 100 mrem/hr but < 1000 mrem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, to be controlled by requiring issuance of a Radiation Work Permit (RWP). Contrary to this, On September 9, 2016, two in dividuals entered a High Radiation Area in the Unit 2 SE Diagonal 87' elevation to calibrate an RHR Service water transmitter without the proper briefing or RWP. The individuals were briefed and permitted to enter the HPCI Room area instead of this area. This finding was of very low safety significance (Green) because there was no substantial potential for overexposure and the licensees ability to assess dose was not compromised. The immediate corrective actions were documented in CR 10271667. The long term corrective actions include continuing training suc h that all craft personnel are exposed to the remediation scenario. (Section 2RS1 ) | |
| 05000321/FIN-2017001-02 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | Unit 2 Technical Specification 3.6.1.3 requires each PCIV be operable in Mode 1. With one PCIV inoperable, the affected penetration flow path must be isolated by use of at least one closed and de -activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. Contrary to the above, on November 6, 2016 at 21:51 operators tagged valve 2E41F111, a PCIV, open with the breaker off. Subsequently, a licensed operator performing a main control room board walk down noted the PCIV was inoperable and, on November 8 at 0151, operators closed and de -activated an automatic valve in the line to rest ore compliance. Inspectors screened the finding in accordance with IMC 609 Appendix A The Significance Determination Process (SDP) for Findings at -Power. The finding screened as very low safety significance (Green) because the questions in Appendix A E xhibit 3 for reactor containment were answered no. This issue was documented in the licensees corrective action program as CR 10295889. (Section 4OA3.2) | |
| 05000366/FIN-2017001-01 | 2017Q1 | Green | H.1 | Self-revealing | Failure to Identify Abnormal Condition on 2C EDG Cross Drive Assembly | Green . A self -revealing non- cited violation (NCV) of Hatch Unit 2 Technical Specification 5.4.1 was identified when technicians performing maintenance on the 2C emergency diesel generator observed pitting on the lower crank component gears and did not initiate a condition report as required by procedure 52SV -R43 -001- 0, Diesel, Alternator, and Accessories Inspection. The licensees failure to initiate a condition report, as required by 52SV -R43 -001- 0 Diesel, Alternator, and Accessories Inspection, for the pitting observed on the lower crank component gears was a performance deficiency. The violation of regulatory requirement occurred on or about November 2015 until the licensee replaced the 2C EDG cross drive assembly and restored compliance on August 25, 2016. The violation was entered into the licensees corrective action program as CR 10263236. The performance deficiency was more than minor because if left uncorrected, the failure to evaluate gear pitting would allow progression of a degradation mechanism to the point of EDG inoperability. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At -Power, dated June 19, 2012. Because all four questions in Section A of Exhibit 2, Mitigating Systems Screening Questions, were answered no, the finding screened as Green. The inspectors determined that this finding had a cross -cutting aspect in the Resources aspect of the human performance area, because the licensee did not ensure adequate procedural guidance to recognize the difference between normal and destructive pitting. (H .1) |
| 05000321/FIN-2016004-01 | 2016Q4 | Green | P.1 | NRC identified | Failure to Establish Icing Controls on CAD Subsystem | An NRC-identified non-cited violation (NCV) of Hatch Unit 1 Technical Specification 5.4, Procedures, was identified when procedures did not include inspection criteria for ice buildup of the Unit 1 nitrogen storage tank piping. The licensees failure to establish controls to ensure that ice buildup on the Unit 1 Containment Atmospheric Dilution (CAD) subsystem piping did not exceed ten inches was a performance deficiency. The licensee entered the condition into their corrective action plan as CR10296584, and performed de-icing activities to remove the ice buildup. This performance deficiency was more than minor, because ice buildup on the CAD system may lead to CAD subsystem inoperability if left uncorrected. The finding screened as Green because the CAD subsystem remained operable. The inspectors determined that this finding had a cross-cutting aspect in the Initiation aspect of the problem identification and resolution area, because the licensee did not initiate a condition report upon initially identifying the issue. (P.1) |
| 05000321/FIN-2016404-01 | 2016Q3 | Green | NRC identified | Security | ||
| 05000321/FIN-2016003-02 | 2016Q3 | Green | H.13 | NRC identified | Failure to Ensure Work Hours are Within Work Hour Limits | An NRC-identified non-cited violation (NCV) of 10 CFR Part 26, Fitness for Duty Programs, was identified when the licensee failed to ensure that personnel subject to work hour controls did not exceed 72 hours in a work week. The licensee entered this condition into their corrective action program as Condition Report 10214872 and restored compliance when the affected individuals received an adequate rest period. The failure to ensure that work hours for personnel subject to work hour controls were tracked in accordance with licensee procedures was a performance deficiency. The finding was more than minor because, if left uncorrected, the failure to appropriately implement work hour limitations for covered workers could adversely impact the conduct and oversight of work on safety significant components. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not result in an adverse impact to plant safety due to worker fatigue. The inspectors determined this performance deficiency had a cross-cutting aspect of Consistent Process in the Human Performance area because the licensee failed to assess which workers were subject to work hour limits. (H.13) |
| 05000321/FIN-2016003-01 | 2016Q3 | Green | P.3 | Self-revealing | Unit Downpower Caused by RFP Vent Line Failure | A self-revealing finding was identified when the licensee failed to install a reactor feed pump (RFP) vent line weld in accordance with plant procedures resulting in a failure that required an unplanned Unit 1 power reduction greater than 20%. Failure to install the correct weld thickness on the unit 1 B RFP vent line, as required by procedures, was a performance deficiency. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective in that an unplanned reactor power reduction was required from 100 percent to 60 percent RTP. The inspectors determined this finding was of very low safety significance (Green) because there was not a reactor trip or loss of mitigation equipment. The inspectors determined that this finding had a cross-cutting aspect in the Resolution aspect of the problem identification and resolution area, because the organization did not take effective corrective actions to address the previous weld configuration issue. (P.3) |
| 05000321/FIN-2016002-02 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.9.4 requires control rod full-in position indication for each control rod be operable in Mode 5. With one or more control rod position indications inoperable, in vessel fuel movement must be suspended. Contrary to the above, on February 11, 2016, at 1156 the licensee initiated fuel move in the vessel with 20 control rod full-in position indications inoperable. On February 11, 2016, at 1320 the shift manager suspended moving fuel to restore compliance. Inspector screened the finding in accordance with IMC 609 Appendix G Shutdown Operations Significance Determination Process. The finding screened as very low safety significance (Green) because all the questions in Appendix G Attachment 1 were answered no. This issue was documented in the licensees corrective action program as CR 10181628. | |
| 05000321/FIN-2016007-01 | 2016Q2 | Green | P.2 | NRC identified | Failure to provide reasonable assurance that Appendix R time critical operator actions (TCOAs) can be completed in a timely manner | The NRC identified a Green non-cited violation (NCV) violation of Hatch Technical Specifications 5.4.1.d, Procedures, for Units 1 and 2, for not ensuring manual action feasibility for actions in fire area (FA) 0024. Specifically, the licensee failed to provide reasonable assurance that a credited manual action to ensure emergency power was both feasible and reliable in response to a fire event. The licensee plans to assess the issue and entered this violation into their Corrective Action Program (CAP) based upon CR10209664, CR10213119, & CR10212821. The licensees failure to provide reasonable assurance that Appendix R time critical operator actions (TCOAs) associated with fire events can be completed in a timely manner was a performance deficiency (PD). The PD was more than minor because if left uncorrected, it could to lead to a more significant safety concern. Specifically, the exclusion of TCOAs from a validation process could lead to plant or program changes that prohibit the completion of actions required to meet the licensing basis. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The deficiency was screened with IMC 0310, Aspects Within Cross Cutting Areas, to determine if any cross-cutting areas were applicable. The team concluded cross-cutting was applicable to the problem identification and resolution (PI&R) area, evaluation attribute due the licensees failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2). |
| 05000321/FIN-2016002-03 | 2016Q2 | Severity level Enforcement Discretion | NRC identified | Performance of Operations with Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment | In February 2016, during the Unit 1 refueling outage, operations with the potential to drain the reactor vessel (OPDRV) activities were performed while in Mode 5 (Refueling Mode) contrary to Technical Specification (TS) 3.6.4.1. These OPDRV activities were also performed during the Unit 2 Refueling Outage. Enforcement Guidance Memorandum (EGM) 11-003, Revision 3, provided required interim actions which were incorporated into procedure 31GO-OPS-025-0 Operations with the Potential to Drain the Reactor Vessel. This procedure was used during the OPDRV activities for the Unit 1 refueling outage. Enforcement: Unit 1 TS 3.6.4.1 required, in part, that activities that had the potential to drain the reactor vessel be conducted only with secondary containment operable. Contrary to that requirement, the licensee conducted activities that could cause the reactor vessel to drain while secondary was inoperable. The inspectors determined this was a Severity Level IV violation. The NRC is exercising enforcement discretion (Enforcement Action (EA)-16-158) in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy because the violation was identified during the discretion period described in Enforcement Guidance Memorandum 11-003. Therefore, the NRC will not issue enforcement action for this violation, subject to a timely license amendment request. | |
| 05000321/FIN-2016008-03 | 2016Q2 | Green | H.2 | NRC identified | Failure to Control Qualification of Purchased 1E Components in Accordance to IEEE 323- 1974 | The inspectors identified two examples of a non-cited violation of 10 CFR 50 Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, for failing to assure that vendors met the quality standards specified in procurement documents (IEEE 323-1974, IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations). The licensee entered this issue into the licensees corrective action program as CR10240023 and CR102399929. The licensee planned to ensure the adequate qualification of Class 1E components. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding to be of very low safety significance (Green) because the structure, system, or component maintained its operability or functionality. The finding was assigned a cross-cutting aspect of Field Presence (H.2), in the Human Performance area because senior managers did not ensure supervisory and management oversight of contractors. |
| 05000321/FIN-2016008-04 | 2016Q2 | NRC identified | Potential departure from protection system design basis | The inspectors identified a URI regarding the licensees compliance with the design bases for the Analog Transmitter Trip System (ATTS). On April 16, 2010, Unit 1 received an invalid emergency core cooling system (ECCS) LOCA signal on high drywell pressure due to a failed ATTS slave card that caused a voltage perturbation within a single panel. The licenses root cause identified that the ATTS panel design did not comply with channel independence (IEEE 279-1971 Section 4.6). A design change (SNC116736) separated the specific cards that caused the false trip. During the review of design change, the inspectors identified that the ATTS panel design for other cards also do not comply with channel independence. A single failure within the ATTS panels could result in an interaction that can cause a failure of two channels. The licensee has provided documentation related to the original submittal of the design to the NRC and the NRCs acceptance of the design. In addition, the UFSAR section 7.3.1.2, System Descriptions, for the ECCS stated, in part, Instrumentation and control are designed to establish that the following functions are met: A. Each instrument channel functions independently of all others. B. Sensing devices respond to process variables and provide channel trips at correct values. C. Sensors and associated instrument channels respond to both steady-state and transient changes in the process variable within specified accuracy and time limitations, and they provide channel trips at correct values even when affected by process variations that may extend grossly beyond the expected trip setpoint. D. Paralleled circuit elements can perform their intended function independently. E. Series circuit elements are free from shorts that can abrogate their function. F. Redundant instruments of logic channels are free from interconnecting shorts that could violate independence if a single malfunction should occur. There is not an immediate safety concern because no specific examples have been identified that would prevent an actuation from occurring. This URI is being opened to determine whether a performance deficiency exists and is identified as URI 05000321/2016008-04 and 05000366/2016008-04, Potential Departure from Protection System Design Basis. | ||
| 05000321/FIN-2016010-02 | 2016Q2 | Green | H.11 | NRC identified | Failure to Identify N2E Nozzle Weld Through-Wall Flaw | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to promptly identify a condition adverse to quality regarding a through-wall flaw in the safe end-to-nozzle weld of the reactor coolant system N2E nozzle. The licensee has since repaired the flaw, completed all required postrepair examinations, and entered this issue entered this into their corrective action program as CR 10247856. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors screened this finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012. Because after a reasonable assessment of degradation, the finding could neither result in exceeding the RCS leak rate for a small LOCA, nor likely affected other systems used to mitigate a LOCA resulting in a total loss of their function, the finding screened as Green. This finding has a cross-cutting aspect of Challenge the Unknown in the area of Human Performance (H.11) because upon discovery of a less robust configuration of the N2E nozzle overlay, the licensee failed to consider the implications on the flaw that had existed in that component since 1988. |
| 05000321/FIN-2016010-01 | 2016Q2 | Severity level III | NRC identified | Inaccurate Information Provided Regarding N2E Nozzle Weld Overlay | The NRC identified an AV of 10 CFR 50.9, Completeness and Accuracy of Information, for the licensees failure to provide data to the NRC that was accurate in all material aspects. Specifically, on two occasions (October 1995, May 2000), the licensee stated that weld 1B31-1RC-12BR-E-5 had been modified with a full-structural weld overlay (FSWOL), when in fact it had only been modified with a less robust design overlay (leak barrier). The NRC approved the licensees requests/proposed alternatives in part based on the inaccurate characterization of the welds. The licensee has since installed the FSWOL and entered the issue into the corrective action program as CR 10197850. The NRC is considering escalated enforcement on the basis that had the licensee provided accurate information, it would likely have caused the NRC to reconsider a regulatory position. | |
| 05000321/FIN-2016002-01 | 2016Q2 | Green | H.1 | Self-revealing | Failure to Implement Maintenance Procedure for Control Room Air Conditioning System | A self-revealing Green non-cited violation (NCV) of Hatch Unit 1 and Unit 2 Technical Specification 5.4, Procedures, was identified when the B main control room air conditioning condenser tripped on high discharge pressure due to an improperly adjusted water regulating valve. The licensee entered the condition into their corrective action program as CR 10217777 and adjusted the water regulating valve to the appropriate setpoint. Failure to adjust the water regulating valve in accordance with preventive maintenance procedure 52PM-Z41-002-1, Control Room Air Conditioning Maintenance, was a performance deficiency. The performance deficiency was more than minor because it associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective in that the failure resulted in the inoperability of the B main control room air conditioner. The finding screened as Green because the loss of component function did not significantly affect the function of the train or system. The inspectors determined that this finding had a cross-cutting aspect in the Resources aspect of the Human Performance area, because the licensee did not ensure that procedures were available and adequate to support nuclear safety (H.1). |
| 05000321/FIN-2016007-02 | 2016Q2 | Green | NRC identified | Passive Fire Protection | The NRC identified a Green NCV of Hatch Renewed Operating License Conditions (OLCs) 2.C.(3) and 2.C.(3)(a), for Units 1 and 2 respectively, because the licensee failed to adhere to branch technical position (BTP) Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1. Specifically, the licensee failed to implement the NFPA 80, Fire Doors and Windows, requirements to ensure fire confinement, thus affecting the defense in depth (DID) aspects. Description: During walkdowns of the chosen fire areas, the inspectors assessed whether the passive fire protection features adhered to the NFPA code commitments specified in the current licensing basis. Based upon the walkdown of the West DC Switchgear Room 2A (FZ 2018) and the adjacent access corridor (FZ 2014), the inspectors observed what they determined to be inadequate fire protection program implementation which would result in a degraded fire confinement ability between two fire zones. The code states, in part, that when doors are installed on only one face of a fire wall, heat responsive units shall be located on each side of the wall and interconnected so that the actuation of any one of them will permit the door to close. In this case, the heat responsive units were fusible metallic links designed to melt at a specific temperature and initiate door closure. The fusible links were not installed as required on both sides of the credited fire door between FZ 2018 and FZ 2014. Specifically, a link was only on one side of door 2L482C10. No link was installed on the side of the door in FZ 2014. The door in question was considered to be a Class A fire door and was designed to provide at least 3-hours of fire resistance between adjacent fire zones for a postulated fire event. Neither of these fire zones were protected with automatic suppression capability. In addition, there were existing exemptions in place for not meeting the 10 CFR Appendix R, III.G.2 requirements as referenced by the FHA. This further supported the need for ensuring the DID measures were adequate for fire confinement. In the second example, on October 15, 2015, the NRC resident inspector observed that an issue existed with the installation of an electro-thermal link designed to close the 2C EDG rolling fire door which separated FA 2407 from FA 0401. Specifically, it was noted that the device was installed in an improper configuration and that the electro-thermal link was mounted directly to the wall and secured using a nut and washer. In this configuration, the washer overlapped the seam of the electro-thermal link and hampered the links ability to separate and automatically close the door. The licensee declared the electro-thermal link non-functional and performed a functionality assessment. The assessment concluded that additional links designed to release the door in the event of a fire would have eventually fused, releasing the door. In addition, the gaseous carbon dioxide (CO2) fire suppression system protecting the 2C EDG would have retained the required CO2 gas concentration using the air control louvers, which were installed in series with the rolling fire door. The degraded electro-thermal link was corrected on November 10, 2015. Analysis: The licensees failure to ensure the DID aspects of the FPP were implemented consistent with the NFPA 80 requirement as specified by the current fire protection licensing basis was a PD. 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. Specifically, the lack of a required link above the fire door between the West DC Switchgear Room 2A and the adjacent access corridor fire zone and the improperly installed link between EDG Room 2C and the adjacent access corridor would have negatively impacted the expected response time of each of the fire doors to close. In addition, review of historical work orders and condition reports indicated problems with the air balance louvers coincident with the degraded ETL controlling the closure of the fire door would have impacted the likelihood of confining the CO2 gas at the required design concentration. In both these instances, the finding had a negative impact on the program DID aspects for the fire confinement category. In accordance with NRC IMC 0609, Significance Determination Process, Appendix F, the inspectors performed a Phase 1 analysis and determined the finding resulted in very low significance, Green, based on question 1.4.3-A since, in each case, the combustible loading on both sides of the barrier wall represented a fire duration less than 1.5 hours (i.e., less than 120,000 Btu/ft2). The team determined that no cross-cutting attributes were applicable based upon the issue being associated with meeting the original NFPA 80 design criteria at licensing. Enforcement: Hatch Operating License Condition (OLC) 2.C.(3) and 2.C.(3)(a), for Units 1 and 2 respectively, stated, in part, that Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report (UFSAR) for the facility. The E. I. Hatch UFSAR, Unit 2, Section 9.5, stated in part the plant fire protection system is described in the Edwin I. Hatch Nuclear Plant Units 1 and 2 Fire Hazards Analysis and Fire Protection Program (incorporated by reference into the FSAR). FHA Section 9.0, Appendix A Compliance Matrix, stated the licensee complied with the applicable sections of BTP APCSB 9.5-1. The General Guideline for Plant Protection section stated that the NFPA 80, Fire Doors and Windows was applicable for fire doors. Contrary to the above, the team identified two instances that were not consistent with the stated commitments. The licensee documented this issue with condition reports CR10085883, CR 10135493, CR 10144100, and CR10022283. The team reviewed the DID and the fire confinement provisions of NFPA 805 as referenced by Sections A.4.4.6.4 and 7.3.7. In addition, Section 5.11.3.1, which states in part that passive fire protection devices such as doors and dampers shall conform to the following NFPA standards, as applicable unless otherwise permitted by 5.11.3.2. Because Hatch committed to adopt NFPA 805 and the aforementioned issues meet the criteria as stated in the NRC Enforcement Policy (Policy), Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), the NRC will disposition the violations in accordance with the Policy and grant enforcement discretion. The NRC will also disposition the associated findings in accordance with Inspection Manual Chapter 0305, Section 11.05, Treatment of Items Associated with Enforcement Discretion. | |
| 05000321/FIN-2016008-02 | 2016Q2 | Green | P.2 | NRC identified | Failure to identify a condition adverse to quality for Masterpact 600V breakers | The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for failing to identify the applicability of US NRC Part 21 Report 2016-20-01 to the 1B emergency diesel generators (EDGs) motor control center (MCC 1B). The licensee entered this issue into the corrective action program for resolution as CR 10240007. For corrective actions, the licensee performed an immediate operability determination and established compensatory measures to reset the breaker linkage in the event that it malfunctions. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding to be of very low safety significance (Green) because the structure, system, or component maintained its operability or functionality. The finding was assigned a cross-cutting aspect of Evaluation (P.2), in the Problem Identification and Resolution area because the organization did not thoroughly evaluate the Masterpact breaker Part 21 to ensure that resolutions addressed the causes. |
| 05000321/FIN-2016008-01 | 2016Q2 | Green | H.9 | NRC identified | Failure to Adequately Qualify Modifications to Class 1E 4160V Buses | The inspectors identified a non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50 Appendix B, Criterion III, Design Control, for the failure to verify adequate design and qualification of Class 1E buses in accordance with Institute for Electronics and Electrical Engineering (IEEE) 279-1971, Standard Criteria for Protection Systems for Nuclear Power Generating Stations. The licensee entered this issue into the licensees corrective action program as CR10240030. The licensee planned to correct the issue prior to installing new transformers. The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding determined to be of very low safety significance (Green) because the system, structure, or component maintained its operability or functionality. The finding was assigned a cross-cutting aspect of Training (H.9), in the Human Performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce to adequately complete a modification of the Class 1E buses. |
| 05000321/FIN-2016001-01 | 2016Q1 | NRC identified | Reactor Coolant System N2E Weld Flaw | The inspectors identified an unresolved item associated with a flaw identified in the safe end-to-nozzle weld of the Reactor Coolant System N2E Nozzle. In July 2015, the licensee submitted a proposed alternative to ASME Code, HNP-ISI-ALT-15-01 (ML15183A354), to install a full-structural weld overlay on reactor coolant nozzle N2E (1B31-1RC-12-BR-E). This proposed alternative was approved by the NRC in December 2015 (ML15349A973). The licensee implemented this proposed alternative during the February 2016 refueling outage (1R27). After removing all but 1/16 of the existing overlay, the licensee performed a liquid penetrant examination and noted a pair of linear indications. Subsequently, the licensee determined that these indications were actually a single indication, and that it exceeded allowable size limitations according to ASME Code. Upon further review, the licensee realized that these indications were potentially the result of growth of an inner-diameter, surface-connected intergranular stress corrosion cracking (IGSCC) flaw found in 1988. The licensee has repaired the flaw, installed the full-structural weld overlay, and completed all required post-installation examinations. This is an unresolved item pending review of whether the licensee performed all required examinations of the N2E nozzle between 1988 and 2016, and whether the flaw exceeded minimum wall limitations at some point during prior operation. The issue will be tracked as URI 05000321/2016001-01, Reactor Coolant System N2E Weld Flaw. | ||
| 05000321/FIN-2015007-03 | 2015Q4 | Green | NRC identified | Failure to Assure that Class 1E Components were Qualified for Design Temperatures | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, for the licensees failure to ensure that adequate environmental test requirements were satisfied before relying on safety-related components to perform their intended safety functions. As an immediate corrective action, the licensee performed an operability evaluation and determined the components were operable. In addition, the licensee indicated that they planned to determine adequate corrective actions to restore full qualification of these commercial grade components, and entered this issue into their Corrective Action Program as Condition Report 10138133. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective, in that the licensee failed to verify the environmental qualification of safety-related components to ensure their performance up to the expected temperature of 150 degrees F. The finding was determined to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000366/FIN-2015007-05 | 2015Q4 | Green | NRC identified | Failure to Classify RCIC Sub-components as Safety-Related | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to classify components in accordance with Regulatory Guide 1.26 as specified by the Unit 2 Updated Final Safety Analysis Report, Section 3.2.2. As an immediate corrective action, the licensee performed an operability evaluation, and determined that the reactor core isolation cooling (RCIC) was operable. In addition, the licensee planned to reclassify the relief valve as safety-related, and entered this issue into their Corrective Action Program as Condition Reports 10132353, 10136685, and 10141965. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective, in that inadequate classification of the relief valves affected the reliability of safety-related function of the RCIC system. The finding was determined to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance (Section 1R21.2.b.5). | |
| 05000321/FIN-2015007-04 | 2015Q4 | Green | NRC identified | Failure to Verify Design Basis Timing Margins for Safety Related Motor Operated Valves | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate if transients in control power voltage could affect the design basis margins for the timing of safety-related motor operated valves (MOVs). The licensee planned to perform corrective actions to ensure that the safety analysis remains bounded, and entered this violation into their Corrective Action Program as Condition Report 10138053. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective, in that the failure to evaluate transients that effect the timing margins for NOVs affected the established reliability and capability of the valves. The finding was determined to be of very low safety significance (Green) because the deficiency did not result in actual loss of safety function. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance (Section 1R21.2.b.4). | |
| 05000366/FIN-2015007-01 | 2015Q4 | Green | NRC identified | Failure to Perform Adequate Circuit Breaker As-Found Testing | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion XI Test Control, for the failure to perform circuit breaker as-found electromechanical testing prior to inspecting, cleaning, and lubricating the mechanical components. The licensee planned to revise the test procedures to correct the deficiencies, and entered this violation into their Corrective Action Program as Condition Reports 10137545 and 10126677. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective, in that inadequate periodic testing to detect deterioration toward an unacceptable condition, the likelihood that these breakers could unpredictably fail when called upon increases with time in service. The finding was determined to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance | |
| 05000366/FIN-2015007-02 | 2015Q4 | Green | P.2 | NRC identified | Failure to Correct Nonconformances with Regulatory Guide 1.9-1971 | The NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct non-conformances with the acceptance limits established for the emergency diesel generator (EDG) test requirements. The licensee performed an operability evaluation, and determined the EDGs were operable based on successful completion of the required Technical Specification surveillance testing. In addition, the licensee planned to revise the EDG test procedures suitable for RG 1.9-1971 testing requirements, and entered this violation into their Corrective Action Program as Condition Report 10133018. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective, in that the failure to ensure that non-conformances with the acceptance limits were adequately incorporated into the EDG test procedures, which affected the reliability of the EDGs. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution (P.2) because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes, and extent of conditions, commensurate with their safety significance (PI.2) |
| 05000321/FIN-2015003-01 | 2015Q3 | Green | H.2 | NRC identified | Failure to Perform Adequate Surveillance on Fire Barriers and Penetration Seals | The NRC identified a non-cited violation (NCV) of Hatch Operating License Conditions (OLCs) 2.C.(3) and 2.C.(3)(a), for Units 1 and 2 respectively, for the licensee s failure to perform fire barrier penetration seal inspections in accordance with the requirements of Surveillance Requirement 2.1.1.c of Appendix B of the Fire Hazard Analysis (FHA). Specifically, the licensee failed to ensure that fire-rated penetrations and fire-rated barriers separating redundant safe-shutdown trains were adequate to keep a fire from spreading from one fire area to another. To restore compliance the licensee performed a 100 percent inspection of fire-rated penetrations to verify the material condition of the site s rated fire barrier penetrations. The licensee s failure to perform fire barrier penetration seal inspections was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e. fire), and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Based on the finding being of very low probability, the finding was determined to be of very-low safety significance (Green). The cause of the finding had a cross-cutting aspect in the area of Human Performance, field presence, because plant leadership did not reinforce standards and expectations, and did not ensure that deviations from standards and expectations were corrected promptly (H.2). Specifically, licensee oversight was not properly engaged to ensure that surveillances were performed adequately, and that deviations were addressed appropriately. |
| 05000321/FIN-2015003-03 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.4.3 requires 10 of 11 safety relief valves (SRVs) to be operable during Mode 1, 2, and 3. Contrary to the above, the licensee identified during bench testing that two safety relief valves failed to lift at the required technical specification setpoint, and therefore were inoperable in Mode 1, 2, and 3. Analysis showed that with the SRVs lifting at the as-found bench test setpoints, the SRVs still would have maintained reactor coolant system pressure below the TS safety limit requirements. The inspectors determined the violation was of very low safety significance (Green) because the SRVs maintained their functionality. This condition was documented in the licensees corrective action program as CR 10067922 | |
| 05000321/FIN-2015003-02 | 2015Q3 | Green | Self-revealing | 1A PSW Pump High Vibration Failure | A self-revealing, NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, Instructions, and Drawings, was identified when the licensee failed to provide instructions to ensure alignment of the 1A plant service water (PSW) pump column in the true vertical position. The failure to align the 1A PSW pump column resulted high stresses which caused the failure of the 1A PSW pump. To restore compliance, the licensee replaced the 1A PSW pump and revised the pump installation procedure to ensure the pump column is aligned in the true vertical position. Failure to provide instructions to ensure appropriate vertical alignment of the 1A PSW pump column was a performance deficiency. This performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective in that the misalignment of the pump column resulted in inoperability of the 1A PSW pump. A regional Senior Reactor Analyst (SRA) performed a detailed risk review of the finding. The SRA calculated the difference between the risk associated with loss of offsite power (LOOP) events with extended recovery times with the 1A pump available, and without the pump. Because of the low frequency of the seismic event, the finding was determined to be Green. The inspectors determined that this finding did not have an associated cross cutting aspect because this finding was not reflective of current licensee performance. | |
| 05000321/FIN-2015002-01 | 2015Q2 | Green | NRC identified | Failure to Maintain HELB Penetration | An NRC identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for failure to maintain reactor building residual heat removal (RHR) diagonal room penetrations in the designed configuration. The violation was entered into the licensees corrective action program as CR 10055943. The licensee issued work orders to seal the affected penetrations in accordance with design documents. The licensees failure to maintain the penetration seals in accordance with design drawings was a performance deficiency. The performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective in that the failure to maintain the design basis configuration compromised the capability of the RHR diagonal room wall to restrict a high pressure coolant injection (HPCI) high energy line break to the torus area. The finding was of very low safety significance (Green) because the loss of component function did not significantly affect the function of the train or system. The inspectors determined that the finding had a cross-cutting aspect of work management in the human performance area (H.5), because the licensees work process did not control work activities such that nuclear safety was the overriding priority. | |
| 05000366/FIN-2015002-02 | 2015Q2 | Green | Licensee-identified | Licensee-Identified Violation | On February 9, 2015, a violation of Unit 2 Technical Specification (TS) 3.6.1.3 was identified by the licensee. TS 3.6.1.3 requires primary containment isolation valves to be operable during Modes 1, 2, and 3. Contrary to this requirement, the A and C inboard main steam isolation valves (MSIVs) failed to close within the required isolation time during a technical specification surveillance test. Therefore, the A and C inboard MSIVs were inoperable when Unit 2 was in Modes 1, 2, and 3. The cause of the failure of the MSIVs to close within the required isolation time was excessive lubrication of the pistons and springs in the 2-way and 4-way valves within the pneumatic manifold assembly of the MSIV actuator. The excessive lubrication became tacky, causing a delay in the opening of the air supply and exhaust paths. Because the A and C outboard MSIVs closed within the required technical specification isolation time, the primary containment isolation safety function of the main steam lines was maintained. Therefore, this finding was determined to be of very low safety significance (Green). This condition was entered into the licensees corrective action program as CR 10036361. | |
| 05000321/FIN-2015002-03 | 2015Q2 | Severity level Enforcement Discretion | NRC identified | Performance of Operations with Potential to Drain the Reactor Vessel (OPDRV) in Mode 5 Without Secondary Containment | A violation of Unit 2 Technical Specification (TS) 3.6.4.1 was identified. Because the violation was identified during the discretion period described in Enforcement Guidance Memorandum 11-003, 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, subject to a timely license amendment request being submitted. | |
| 05000366/FIN-2015001-02 | 2015Q1 | Green | H.8 | NRC identified | Failure to perform complete analysis of air samples | An NRC-Identified non-cited violation (NCV) of TS 5.4.1 was identified for the failure of the licensee to perform complete quantitative analysis of air samples using approved counting equipment as required by the licensees procedures. NMP-HP-301, Step 5.6, provides guidance for quantitative evaluation of air samples. On February 16, and 25, 2015, air samples for work activities in the Reactor Pressure Vessel head (RPV) and the Reactor Water Cleanup (RWCU) System heat exchanger were not quantitatively analyzed or evaluated for alpha activity even though the areas had been identified as having elevated alpha contamination levels. The licensee entered the issue into their corrective action program (CAP) as CR 10034556. The finding was more than minor because it was associated with the Occupational Radiation Safety Program attribute of exposure control and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from airborne radioactive material during routine civilian nuclear reactor operation. Failure to identify potentially significant contributors to internal dose could lead to unmonitored occupational exposures. The finding was determined to be of very low safety significance (Green) because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose related to As Low As Reasonably Achievable (ALARA) Planning and the ability to assess dose was not compromised during this instance. The cause of this finding was directly related to the cross-cutting aspect of following processes, procedures, and work instructions in the Procedure Adherence component of the Human Performance area. |
| 05000321/FIN-2015001-01 | 2015Q1 | Green | H.1 | NRC identified | Failure to perform adequate surveys of air samples for alpha activity | An NRC-Identified non-cited violation (NCV) of 10 CFR 20.1501(a) was identified for failure to perform an adequate survey. Air samples obtained in the reactor cavity and on the refuel floor during a contamination event indicating greater than 0.3 beta-gamma Derived Air Concentration (DAC) fraction level were not analyzed for alpha activity as required by the licensees procedures. Previous characterization of the area had determined the area to be an Alpha Level II area requiring additional assessment and evaluation of air samples. This violation was entered into the licensees CAP as CR 10033022. This finding is greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Program and Process (Monitoring and RP Controls) and adversely affected the cornerstone objective in that failure to identify potentially significant contributors to internal dose could lead to unmonitored occupational exposures. The finding was determined to be of very low safety significance (Green) because it was not related to As Low As Reasonably Achievable (ALARA) Planning and the ability to assess dose was not compromised during these instances. The cause of this finding was directly related to the cross-cutting aspect of leaders ensuing equipment, procedures, and other resources are available and adequate in the Resources component of the Human Performance area. |
| 05000321/FIN-2015001-04 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50.48(b)(1) required that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Sections III.G.2 or III.G.3. Contrary to the above, since November 1985, the licensee has not met the requirements of 10 CFR Part 50, Appendix R, Sections III.G.2 or III.G.3, in that the licensee failed to provided adequate protection of cables and equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions located in the same fire area by either (a) a 3-hour rated fire barrier; (b) 20 feet of spatial separation with detection and suppression installed in the fire area; or (c) a 1-hour rated fire barrier with detection and suppression installed in the fire area; or by providing alternative shutdown capability for the areas where adequate cable protection was not provided. This violation was determined to be of very low safety significance (Green) based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. This violation was documented in the licensees CAP as CRs 687178, 688543, 687173, and 692904 | |
| 05000321/FIN-2015001-06 | 2015Q1 | Severity level Enforcement Discretion | NRC identified | Unfused DC Ammeter Circuits Result in an Unanalyzed Condition | On April 28, 2014, the licensee submitted an LER documenting the discovery of a condition of non-compliance with the sites fire protection program (FPP). This condition could prevent operators from achieving and maintaining safe shutdown (SSD) of the plant, in the case of a postulated fire. The inspectors reviewed documents related to the LER and discussed the event with plant personnel to assess if the licensees compensatory measures and corrective actions were adequate. The licensee identified a non-compliance with Hatch Renewed License Conditions 2.C.(3) and 2.C.(3)(a), for Units 1 and 2. The licensee failed to provide short circuit protection for non-safety-related associated circuits which could result in a secondary fire in another fire area and adversely affect SSD capability. Description: During a review of industry operating experience (OE) related to unfused DC ammeter circuits the licensee determined that certain DC ammeter circuits lacked short circuit protection. A postulated fire in a fire area containing affected DC ammeter circuit cabling could result in concurrent shorts in the circuit. Due to the lack of short circuit protection, the resultant excessive current flow in the DC ammeter cable could result in a secondary fire in another fire area and adversely affect SSD equipment or cables for SSD equipment. Multiple fire areas in the Control Building were potentially affected. Section 9.6.2.4 of Appendix E of the licensees Fire Hazards Analysis (FHA) categorizes associated circuits of concern into 3 types. Type C associated circuits were defined as nonsafe shutdown circuits which shared a common enclosure with safe shutdown circuits and were not electrically protected by an automatic fault protection device or were not inherently self-protected because the circuit lacks sufficient energy to cause circuit damage. A subsequent paragraph in Section 9.2.6.4 stated that Type C associated circuits are electrically protected by automatic fault interrupting devices, do not carry sufficient energy to cause cable damage, and will not propagate fire into a common enclosure in another fire area. The licensees OE review determined that certain DC ammeter circuits were not provided with automatic fault interrupting devices, and thus, invalidates the SSD evaluation bases stated in Section 9.6.2.4 of the FHA. Upon discovery, the licensee implemented roving fire watches for the affected areas. Analysis: The licensees failure to provide short circuit protection for DC ammeter circuits is a performance deficiency. This finding is more than minor because it is associated with reactor safety mitigating system cornerstone attribute of Protection Against External Events (i.e., fire) and adversely affected the cornerstone objective in that not providing circuit protection could have affected the licensees SSD capability. Because this issue relates to fire protection, and this noncompliance was identified by the licensee as a part of the sites transition to NFPA 805, this issue is being dispositioned in accordance with Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) of the NRC Enforcement Policy. In order to verify that this non-compliance was not associated with a finding of high safety significance (Red), a bounding phase 3 SDP risk analysis was performed by a regional SRA using the guidance from NRC Inspection Manual Chapter 0609 Appendix F and NUREG/CR 6850 revision 0 and Supplement 1. The analysis used inputs from the licensees NFPA 805 project for ignition frequency and cable routing data. The major analysis assumptions were: a one year exposure period, two proper DC polarity hot shorts required to achieve the high current conditions for secondary fires, and all ignition sources for each affected fire zone assumed to damage the ammeter cables. Based on this bounding risk analysis, the regional SRA determined that this performance deficiency resulted in a CDF increase for each Hatch Unit 1 and 2 of less than 1E-4/year (i.e., less than Red). The licensee also performed a risk assessment using their Hatch fire probabilistic risk assessment model which also produced a result | |
| 05000321/FIN-2015001-05 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.7.2 requires areas with radiation levels greater than or equal to 1000 mrem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour measured at 1 meter from the radiation source or from any surface the radiation penetrates shall be provided with a locked or continuously guarded doors to prevent unauthorized entry. Contrary to this on 12/18/14, a RPT found the Unit 1 Recombiner Preheater B room door propped open and not posted as a LHRA. Follow-up surveys of the area identified maximum radiation levels of 1600 mR/hr at 12 inches from surface of the preheater. This finding was of very low safety significance (Green) because there was no substantial potential for overexposure and the licensees ability to assess dose was not compromised. This violation was documented in the licensees CAP as CAR 249078. | |
| 05000321/FIN-2015001-03 | 2015Q1 | Green | H.5 | Self-revealing | Failure to Identify Embedded Conduit prior to Core Drill Operations | A self-revealing non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Procedures, Instructions, and Drawings, was identified for failure to identify existing embedded conduit in the vicinity of prescribed core drills location. The violation was entered into the licensees corrective action program (CAP) as condition report (CR) 902506. Failure to provide adequate instructions in Design Change Package (DCP) SNC467474 to perform core drills in the Unit 2 control building to support conduit installations was a performance deficiency. This performance deficiency is more than minor because it affected the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective in that 2P41F316A was rendered incapable of performing its safety related function of closing in the event of an accident condition. The finding was screened as Green because the inoperability did not last longer than the technical specification (TS) allowed outage time. The inspectors determined the performance deficiency has a cross-cutting aspect of work management in the human performance area, because the licensees work process did not identify and manage the risk commensurate to the core drill work. |