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05000282/FIN-2018003-0330 September 2018 23:59:59Prairie IslandFailure to Promptly Identify Degradation of the 122 DDCLP FOST Vent PipingThe inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, as of November 28, 2017, for the licensees failure to promptly identify a condition adverse to quality associated with 122 DDCLP FOST vent piping.
05000306/FIN-2018003-0430 September 2018 23:59:59Prairie IslandFailure to Promptly Identify and Correct 21 125 VDC Battery Lid Conditions Adverse to QualityThe inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, as of February 15, 2018, for the licensees failure to promptly identify and correct conditions adverse to quality associated with the 21 125 VDC battery system.
05000282/FIN-2018003-0130 September 2018 23:59:59Prairie IslandFailure to Repair a D2 EDG Jacket Water Leak per the Leak Management ProcessThe inspectors identified a finding of very low safety significance (Green) as of July 18, 2018, for the licensees failure to repair a D2 EDG jacket water leak per the Leak Management Process.
05000282/FIN-2018003-0230 September 2018 23:59:59Prairie IslandFailure to Maintain a Preventative Maintenance Strategy for 12 and 22 Cooling Water Pump Diesel EnginesThe inspectors identified a finding of very low safety significance (Green) and associated NCV of Prairie Island Technical Specification 5.4.1, Procedures, as of August 9, 2018, for the licensees failure to maintain a preventative maintenance strategy for sacrificial zinc anode plugs on the jacket water system for the 12 and 22 cooling water pump diesel engines (DDCLPs).
05000266/FIN-2018003-0130 September 2018 23:59:59Point BeachFailure to Perform Evaluations to Ensure that the Fabrication of Dry Cask Storage Systems Meets the Requirements of the Loading Certificate of ComplianceAn NRC-identified Severity Level IV NCV of 10 CFR 72.212 was identified when the licensee failed to perform written evaluations to ensure that the dry cask storage systems met the fabrication requirements of the Certificate of Compliance (CoC) to which they were loaded.
05000282/FIN-2018011-0130 June 2018 23:59:59Prairie IslandFailure to Justify Load Combinations Used in Main Steam Piping Stress AnalysisInspectors identified a Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate provisions from specified quality standards for load combinations into piping analysis. Specifically, in the analysis for the Class I main steam piping, the licensee combined the seismic Operating Basis Earthquake and safety relief valve operating loads by Square Root of Sum of Squares. Prairie Island Updated Safety Analysis Report and the Engineering Manual for piping system stress analysis do not permit the Square Root of Sum of Squares method for combining these loads.
05000282/FIN-2018011-0230 June 2018 23:59:59Prairie IslandPotential Failure to Protect Class I Structures, Systems,and Components from Tornado Generated Missiles

Inspectors identified a number of structure, systems,and components (SSCs) that lacked protection from tornado generated missiles. The following SSCs were identified: Division 1 and Division 2 Emergency Diesel Generators (D1/D2 EDGs)engine exhaust, fuel oil day tank vents, and main fuel oil storage tanks vents; and Diesel Driven Cooling Water Pumps (DDCWPs) main fuel storage tank vents, day tank vents, engine exhausts, and rooms ventilation intake and exhaust equipment. In various cases susceptible SSCs for redundant equipment (e.g. fuel tank vents) were right next to or within a few feet of each other such that a single missle could affect both trains of the system

A review of the sites licensing bases, including the original FSAR, identified the D1/D2 EDGs and the DDCWPs as Class I, safety-related SSCs, which are required to be designed to withstand, without loss of capability, environmental phenomena including tornadoes and tornado generated missiles. Specifically, the current USAR Table 12.2-1, Classification Of Structures, Systems and Components, list both systems as Class I and has two notes of interest. Note 1 applies to the Diesel Generators and their associated (Main) Fuel Oil Storage Tank, which states, in part, The indicated Design Class I is applicable to D1/D2 Diesel Generators and associated(emphasis added) safety related components and systems. The second note is listed at the beginning of the Table, which states,in part,To determine detail design classifications and boundaries separating different design classes within the overall classification scheme listed here, refer to controlled drawings. A review of controlled drawings, including NF-39255-1, Flow Diagram Diesel Generators D1 & D2 Unit 1 & 2,Revision 85, and NF-39232, Flow Diagram Fuel & Diesel System Unit 1 & 2, Revision 86,showed the fuel oil vents for the main storage tanks, fuel oil vents for the day tanks,engine exhaust piping,mufflers, and silencers for the D1/D2 EDGs and DDCWPs were classified as safety-related Class I SSCs. A review of the current UFSAR identified the following sections of interest:The USAR Section 1.5.I, Overall Plant Requirements, Criterion 2 -Performance Standards, Answer, established in part The system and components designated Class I in Section 12, in conjunction with administrative controls and analysis, as applicable, are designed to withstand, without loss of capability to protect the public, the most severe environmental phenomena ever experienced at the site with appropriate margins included in the design for uncertainties in historical dataThe USAR Section 12.2.1.1.a, Classification of Structures and Components, defines Design Class I as Those structures and components including instruments and controls whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of substantial amounts of radioactivity, and those structures and components vital to safe shutdown and isolation of the reactor.The USAR Section 12.2.5.1.g.1, Protection for Class I Items, establishes, in part, that Class I items are protected against damage from: Missiles from different sources.These sources comprise: Tornado created missiles.The USAR Section 12.2.1.3.2.c., Tornado Loads, defines the design tornado driven missile as assumed equivalent to an airborne 4 x 12 x 120 plank travelling end-on at 300 mph, or a 4000 lbs automobile flying through the air at 50 mph and at not more than 25 feet above ground level.Based on the above, the inspectors were concerned the susceptible SSCs could lose the capability to perform their safety-related function if they were impacted by tornado generated missiles. For example, an impact to the fuel oil vents could crimp the vent path resulting in a vacuum inside the tanks that could collapse the tank and/or cause the associated fuel transfer pump to lose net positive suction head
The licensee provided a position paper proposing the susceptible SSCs identified by the inspectors were meeting their current licensing bases and no further actions were required. The inspectors disagreed, but decided to request support from the Office of Nuclear Reactor Regulation (NRR) to obtain clarification on the sites licensing bases related to tornado generated missiles. Planned Closure Action: The inspectors have requested NRR to provide clarification on the sites current licensing bases regarding tornado generated missiles required protection.Licensee Action: Licensee is considering doing a self-review of design and licensing basis of the fuel oil storage tank vent lines to understand and clarify design class of the lines
Corrective Action Reference:501000012997
05000244/FIN-2018403-0130 June 2018 23:59:59GinnaSecurity
05000266/FIN-2018002-0130 June 2018 23:59:59Point BeachPrimary Auxiliary Building Floor Plug Removal Creates Unanalyzed Flood PathA Green finding and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to ensure that applicable regulatory requirements and design basis, for structures, systems, and components, were translated into procedures. Specifically, the licensee failed to include the floor plugs on the 26 level of the primary auxiliary building as credited flood barriers in procedure NP 8.4.7, PBNP Flooding Program.
05000266/FIN-2018002-0230 June 2018 23:59:59Point BeachUnanalyzed Condition for Tornado Generated MissilesOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015, (ML15111A269) and revised on February 7, 2017, (ML16355A286). The EGM applies specifically to a SSC that is determined to be inoperable for tornado generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Point Beach, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states, in part, that SSCs, which are essential to the prevention and mitigation of nuclear accidents, shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon, such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice 53239 as an unanalyzed condition and potential loss of safety function. Enforcement discretion was previously authorized and documented in Inspection Report 05000266/2018001 (ADAMS Accession Number ML18128A229). Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify AOP13C, Severe Weather Conditions procedure, to include actions for removing potential airborne hazards and damage assessments for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies, was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession Number ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory measures to address the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which, for Point Beach, were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC, such that discretion was no longer needed.On April 26, 2018, the licensee submitted a request to extend the enforcement discretion in letter titled Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15002 for Tornado-Generated Missile Protection Non-conformances Identified in Response to Regulatory Issues Summary 201506, Tornado Missile Protection. On May 21, 2018, the NRC approved this request and extended the enforcement discretion until June 10, 2020. The disposition of this enforcement discretion closes LER 201800100.
05000282/FIN-2018002-0130 June 2018 23:59:59Prairie IslandResults of ISFSICask Array Dose Calculation Not Incorporated into FSARPrairie Island ISFSI FSAR, as updated, Revision 18, Section A7A.7 evaluates off-site dose rates for an array of ISFSI casks. In this dose rate calculation, explicit modeling credit is given to the earthen berm that surrounds the Prairie Island ISFSI as discussed in Section A7A.7.1. The earthen berm provides radiation shielding for the ISFSI. This calculation allows the licensee to demonstrate, in part, compliance with Title 10 of the Code of Federal Regulations (CFR) 72.104(a) which requires, in part, that, During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ. Calculation TN40HT0502, TN40HT Far Field Shielding Calculations, Revision 0, was performed by the licensee in support of a License Amendment Request (LAR) to modify the Prairie Island ISFSI TN40 cask design (designated as TN40HT casks). The TN40HT LAR was submitted to the NRC by the licensee on March 28, 2008. This dose rate calculation does not credit the earthen berm and, in part, also allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a). The licensee also provided this calculation directly to the NRC in a February 29, 2012, letter in response to a Request for Supplemental Information (RSI) from the NRC associated with the license renewal application for the Prairie Island ISFSI. Although the results from calculation TN40HT0502 for a single cask was incorporated into the Prairie Island ISFSI FSAR, Revision 18, in Tables A7A.22 and A7A.61, the results from TN40HT0502 for an array of casks which, in part, allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a), has not been incorporated into the ISFSI FSAR, Revision 18.Title 10 CFR 72.70, Safety analysis report updating requires, in part, that (a) Each specific licensee for an ISFSI shall update periodically, as provided in paragraphs (b) and (c) of this section, the FSAR to assure that the information included in the report contains the latest information developed (b) Each update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last update to the FSAR under this section. The update shall include the effects of: (2) All safety analyses and evaluations performed by the licensee in support of approved license amendments.This Unresolved Item is being opened to determine whether or not the licensee is required to update the ISFSI FSAR with the results of calculation TN40HT0502 for an array of casks in accordance with 10 CFR 72.70.Planned Closure Action: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Material Safety and Safeguards to determine whether or not calculation TN40HT0502 is subject to the FSAR updating requirements of 10 CFR 72.70 for the Prairie Island ISFSI.
05000244/FIN-2018002-0130 June 2018 23:59:59GinnaIncorrect Scaling Factors in Reactor Vessel Level Monitoring System Instrumentation Uncertainty CalculationThe inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure that adequate design control measures existed to verify the adequacy of the Reactor Vessel Level Monitoring System (RVLMS) uncertainty calculation. Specifically, Exelon failed to identify errors in the RVLMS uncertainty calculation which resulted in a reasonable doubt of operability for the system after a temporary modification was implemented.
05000266/FIN-2018001-0131 March 2018 23:59:59Point BeachFailure to Evaluate and Characterize Fire Protection Pipe Wall DegradationThe inspectors identified a finding of very low significance, for the failure to follow procedure NP 7.7.22, Service Water and Fire Protection Inspection Program. Specifically, Section 4.10, Degraded Component Characterization and System Failure Analysis, step 4.10.1 states, in part, the extent of pipe wall degradation shall be characterized by volumetric non-destructive examination (NDE) for subsequent flaw evaluation. The licensee identified pipe corrosion on November 28, 2012, and failed to characterize it by volumetric NDE.
05000266/FIN-2018001-0231 March 2018 23:59:59Point BeachFailure to Evaluate Material Acceptability for a Safety-Related DoorstopA self-revealed Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III was identified when the licensee failed to evaluate the suitability of material prior to installation in the plant. Specifically, the licensee installed a doorstop, which was fabricated from a length of Unistrut, behind a safety-related door. The Unistrut was not suitable for the application and caused the door to become wedged open.
05000244/FIN-2018011-0131 March 2018 23:59:59GinnaPotential Preconditioning of Turbine Driven Auxiliary Feedwater Surveillance TestingThe NRC identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XI, Test Control, because Exelon established unevaluated preconditioning, with a reasonable doubt of whether the preconditioning was acceptable, prior to testing of the turbine driven auxiliary feedwater pump. This results in the loss of as-found conditions which challenge the capability of the test to assure that the turbine driven auxiliary feedwater pump will perform satisfactorily in service.
05000306/FIN-2018001-0131 March 2018 23:59:59Prairie IslandQuestions Regarding Corrective Action Program, Use of Operating Experience, and Qualification of the 21 125 VDC Battery due to Cell Lid CrackingThe inspectors identified an unresolved item regarding manual override of the auto-closure function of component cooling water system valves. Specifically, the inspectors noted that the system was not protected from tornado generated missiles when valves CV39153 & CV39154 are opened per procedure to support system alignments. The inspectors initially determined that further review was needed to determine if Technical Specifications are met if/when CV39153 & CV39154 are maintained open.Corrective Action Reference: AR 501000001642; 2017 50.59 Potential PD Evaluation 1133; 08/15/2017 Closure Basis: The inspectors reviewed the license basis documentation, procedures, and interviewed licensee personnel, and did not identify any licensee failure to meet a requirement or standard.
05000266/FIN-2018001-0331 March 2018 23:59:59Point BeachInadequate Basis for Deletion of TRM 3.4.3 Primary System Integrity RequirementsThe inspectors identified a Severity Level IV, Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, and an associated finding of very low safety significance (Green) for failure to provide a written evaluation, which provided the basis for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for why deletion of the nondestructive examination requirements in Technical Requirements Manual (TRM) 3.4.3 for Primary System Integrity did not require prior NRC approval.
05000282/FIN-2018001-0231 March 2018 23:59:59Prairie IslandQuestions Regarding the Corrective Action and Aging Management Programs Following the Discovery of 122 DDCLP FOST Vent Piping DegradationOn November 28, 2017, the inspectors identified a small hole in the vent piping for the below-ground 122 DDCLP FOST (located outside and adjacent to the plant screenhouse). The station generated AR 501000005894 and the shift manager declared the supported 22 DDCL pump operable-but-degraded with a temporary procedure change to AB4, Flood as a compensatory measure and wrapping of the pipe to preclude foreign material intrusion. The site backed up the immediate operability determination with a POD, evaluated past operability (no issues identified) and, subsequently replaced the affected portion of the pipe to restore full qualification. The inspectors concluded that these short term actions were acceptable to address the issue, but identified several concerns regarding prior actions to address the vent pipe corrosion. On March 1, 2018, the inspectors were provided the final evaluations for AR 501000005894. After review, the inspectors were concerned that the evaluations did not perform a sufficient review of: whether the corrective action program properly dispositioned corrosion of the pipe when first identified in July of 2015;whether the corrective action program and aging management program (AMP) performed as required to correctly classify and correct and/or manage the corrosion aging mechanism; and whether the extent of cause/condition for the adjacent 121 DDCLP FOST vent pipe was properly addressed.The inspectors passed these concerns to individuals in the engineering and regulatory affairs departments, but the licensee then stated that the evaluations provided on March 1, were, in actuality, still in a revision/review phase. The licensee stated that the final evaluations would likely address the inspectors concerns. On March 22, the inspectors were provided the final evaluations, but it appeared that only minor changes were made and the inspectors concerns were not addressed. On March 28, the inspectors again voiced their concerns with the licensee and two new ARs (501000010169 and 501000010178) were created documenting the following:the AR written identifying corrosion of the piping in July of 2015 was not evaluated under the AMP, the condition was determined to be operable and fully qualified, and it was closed to a work request to re-coat the piping but was never performed.the AR written in April of 2016 again noted the corrosion, but was closed to a non-conservative evaluation, the issue was not evaluated under the AMP, operability was again assessed as operable and fully qualified, and a work request was issued to apply a coating (not completed until May of 2017) 12 the AMP engineer was not consulted in 2015 or 2016 to determine if/how the issue fit into the AMP requirements for increased monitoring, development of acceptance criteria, and final corrective actions.Planned Closure Actions: To resolve this item, the inspectors will review planned actions regarding the degraded 121 DDCLP FOST vent pipe, further extent of condition reviews, and review planned licensee condition and causal evaluations regarding programmatic and/or human performance aspects of the issue.Licensee Actions: At the end of the inspection period, the licensee began excavation activities to replace the 121 DDCLP FOST vent pipe and had apparent cause and extent of condition evaluations in progress.Corrective Action Program References: ARs 501000005894, 501000010169 and 501000010178.
05000266/FIN-2018001-0431 March 2018 23:59:59Point BeachEnforcement Action: EA18030: Unanalyzed Condition for Tornado Generated MissilesOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015, (ML15111A269) and revised on February 7, 2017, (ML16355A286). The EGM applies specifically to an SSC that is determined to be inoperable for tornado generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of 12 Point Beach, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states in part that SSCs which are essential to the prevention and mitigation of nuclear accidents shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53239 as an unanalyzed condition and potential loss of safety function. Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources - Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession No. ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15-002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which for Point Beach were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed
05000244/FIN-2018011-0231 March 2018 23:59:59GinnaFailure to Procedurally Verify Fuel Transfer Cart Results in Fuel Interference EventA self-revealing Green non-cited violation (NCV)of Technical Specification 5.4.1.a, Procedures, was identified for the failure of Exelon to operate refueling equipment in accordance with technical procedures in April and May of 2017, which resulted in a fuel interference event, damage to the rod cluster control assembly, and the need for a detailed inspection of a fuel assembly
05000266/FIN-2017405-0131 December 2017 23:59:59Point BeachSecurity
05000244/FIN-2017004-0131 December 2017 23:59:59GinnaInadequate Component Monitoring Relating to Online Risk Management and AssessmentThe inspectors identified a finding because Exelon personnel did not follow Procedure WC-AA-101-1006, On-Line Risk Management and Assessment, Revision 2 to sufficiently monitor components such that the latest information was used to evaluate plant risk. Specifically, on December 27, 2017, Exelon failed to sufficiently monitor the diesel driven air compressor, commensurate with its operating history, such that a failure would be assessed and updated in the current plant risk assessment. Exelon entered this issue into the corrective action program (CAP) for resolution as action request (AR) 0487519. Corrective actions included declaring the diesel driven air compressor non-functional, transitioned to Yellow online plant risk, and completed restoration of the C Instrument Air Compressor.This finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, this issue is similar to Example 7.f of IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, because the overall elevated plant risk placed the plant into a higher licensee-established risk category. The inspectors evaluated this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 2, Assessment of (risk management actions) RMAs, to analyze the finding and calculated the incremental large early release probability using PARAGON, Exelons risk assessment tool, and found the increase in incremental large early release probability was less than 1E-7. The inspectors determined that if this condition existed for the full duration of the maintenance period, the large early release probability would have been 2.22E-7. Because the increase in incremental large early release probability, was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Exelon did not recognize and plan for the possibility of mistakes, latent issues and inherent risk, even while expecting successful outcomes. Specifically, Exelon did not ensure a component used to manage and assess risk was monitored at a frequency commensurate with its past performance. (H.12)
05000266/FIN-2017004-0131 December 2017 23:59:59Point BeachFailure to Implement Required Provisions of NFPA 805A finding of very low safety significance and associated NCV of Point Beach Nuclear Plant Units 1 and 2, Renewed Operating License Condition 4.F (fire protection) was identified by the inspectors for the licensees failure to either de-energize chemical and volume control system (CVCS) valve 1(2)CV285 or implement applicable compensatory measures. Specifically, the licensees National Fire Protection Association (NFPA) Standard 805 license basis and their fire protection program credited 1(2)CV285 as being de-energized to prevent fire-induced spurious operation from causing a loss of reactor coolant inventory. Immediate corrective actions included opening the 1CV285 circuit breaker on April 12, 2017. The circuit breaker for 2CV285 had been previously opened by the Unit 2 CVCS operations checklist on April 6, 2017.The finding was determined to be more than minor because it was associated with the Initiating Events cornerstone attribute of Protection Against External Factors (Fire) and 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. Specifically, by failing to de-energize 1CV285 and 2CV285, both Units 1 and 2 were vulnerable to a fire-induced intersystem loss of coolant accident through their respective excess letdown lines. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b the inspectors determined the finding degraded the fire protection defense-in-depth strategies. Therefore, screening under IMC 0609, Appendix F, Fire Protection Significance Determination Process, was required and directed the inspectors to continue to Significance Determination Process (SDP) Phase 2 Quantitative Screening Approach in IMC 0609, Appendix F. The Regional Senior Reactor Analyst (SRA) performed a detailed risk evaluation, which concluded that the risk was of very low safety significance or Green. This finding has a cross-cutting aspect in the area of human performance, Change Management, because the licensee did not use a systematic process for evaluating and implementing change so that 3 nuclear safety remains the overriding priority. Specifically, the licensee updated their operations checklist to maintain the CV285 valves de-energized, but failed to implement the checklist on the day the station transitioned to their NFPA 805 licensing basis. (H.3)
05000266/FIN-2017002-0230 September 2017 23:59:59Point BeachFailure to Identify Non-Conforming Conditionsafter Receipt of Anchor Darling Double Disc Gate Valve Related Part 21 ReportThe inspectors identified a finding of very-low safety significance (Green), and an associated (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action,for the licensees failure to identify a condition adverse to quality. Specifically, after receiving and reviewing the Flowserve 10 CFR Part 21 report, the licensee misunderstood the information provided and failed to identify 36 safety-related valves that were nonconforming. Of these 36 valves, 14 were identified as being susceptible to pin failure based on their torque setting, 6 of which had open or close safety functions. The licensee captured the inspectors concern in the CAP as AR 02212531, and AR 02212915. In addition, the licensee performed operability evaluations that concluded the affected valves remained operable.The performance deficiency was more-than-minor because it was associated with the equipment performance attribute of the Mitigating System and Initiating Event cornerstones, and adversely affected the cornerstone individual objectives. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding screened as of very-low safety significance (Green) by answering No to the questions contained in Exhibit 1, and in accordance with Exhibit 2, it did not result in the loss of operability or functionality of mitigating systems. The team did not identify a cross-cutting aspect associated with this finding because the most significant cause for the error was not reflective of current performance. Specifically, the Part 21 report and associated review by the licensee occurred in February 2013.
05000266/FIN-2017003-0130 September 2017 23:59:59Point BeachInappropriate Instructions for Testing Safety-Related Power SuppliesA finding of very low safety significance and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to have instructions of a type appropriate to the circumstances. Specifically, the instructions for testing a refurbished safety-related power supply did not contain acceptance criteria to ensure that the power supply voltage output did not exceed the maximum voltage requirements established by the vendor of the downstream level transmitter. Immediate corrective actions included evaluating the voltage output of operating power supplies to ensure the voltage at their associated transmitters was within vendor specifications. The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to lead to a more significant safety concern. Specifically, power supplies could have been placed back in service producing voltage levels at the downstream safety-related transmitters exceeding their vendor requirements. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on October 7, 2016. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered No to the screening questions. This finding has a cross-cutting aspect in the area of human performance, Design Margins, because the licensee did not ensure that design margins were carefully guarded. (H.6)
05000282/FIN-2017003-0130 September 2017 23:59:59Prairie IslandFailure to Ensure Correct Operation of Meteorological TowerA finding of very-low safety significance, and an associated NCV of Technical Specification (TS) 5.4.1 was identified by the NRC inspectors for the failure to implement and maintain procedures to ensure adequate operation of a meteorological tower. The licensee entered this issue into their Corrective Action Program (CAP) as CAP 501000001091, dated July 27, 2017. The licensee had initiated efforts to assess and remove unnecessary vegetation growth. The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, Issue Screening, because the finding impacted the Plant Facilities/Equipment and Instrumentation Attribute of the Public Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, existing meteorological tower procedures did not include the assessment and subsequent removal of trees that could impair the correct operation of sensors located at the 10 meter elevation of the tower. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008. The violation was of very-low safety significance (Green) because: it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or Title 10 of the Code of Federal Regulations (CFR), Part 20.1301(e) criteria. The inspectors concluded that the most significant contributing cause of the performance deficiency involved the Resolution cross cutting component in the area of problem identification and resolution because this issue was previously entered into the licensees CAP in 2015 and closed with no action taken. (P.3)
05000266/FIN-2017003-0230 September 2017 23:59:59Point BeachService Water Cable Support FailureA finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the failure to promptly identify and correct degraded structural supports 3 for safety-related cables, a condition adverse to quality. Specifically, the licensee failed to repair or replace degraded service water pump cable supports after they identified the degraded supports in 2011. The licensee was in the process of scheduling the cable support repairs at the end of the inspection period. The inspectors determined that the continued non-compliance does not present an immediate safety concern because, given the weight pressing onto the cables, the insulation should remain intact. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Reliability and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure of the service water motor cable support allowed the structural beam to drop and metal cable clamps to impinge on the insulation of the 480 volt safety-related cables. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on October 7, 2016. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered No to the screening questions. This finding has a cross-cutting aspect in the area of human performance, Conservative Bias, because the licensee did not use decision making-practices that emphasize prudent choices overt those that are simply allowed. (H.14)
05000266/FIN-2017007-0130 September 2017 23:59:59Point BeachFailure to Correct a Condition Adverse to Quality Associated with a Seismic Interaction of the Motor-Driven Auxiliary Feedwater PipingThe NRC identified a finding of very-low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to correct a Condition Adverse to Quality (CAQ) associated with a seismic piping interaction affecting the Motor Driven Auxiliary Feedwater (MDAFW) system. Specifically, the licensee identified a flange clearance to the Unit 1 MDAFW suction piping was nonconforming and captured it in the Corrective Action Program (CAP) as Action Request (AR) 01684524. However, the licensee closed the AR without correcting the CAQ. The licensee captured the inspectors concern in the CAP as AR 02212810 and performed an evaluation that reasonably concluded the MDAFW remained operable.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability determination which concluded the stresses resulting from the seismic interaction would reasonably be bounded by the applicable stress operability limits. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance because the performance deficiency occurred more than 3 years ago. Specifically, the licensee closed AR 01684524 without correcting this CAQ on September 20, 2011.
05000266/FIN-2017002-0130 June 2017 23:59:59Point BeachFailure to Evaluate Operating ExperienceGreen . A finding of very low safety significance was self -revealed f or the failure to follow program description PI AA 102, Operating Experience Program, Revision 3. Specifically, the licensee failed to evaluate operating experience that applied to Point Beach that identified the potential for cable connectors to disconnect due to machine vibration. PI AA 102, Section 5, Instructions, Step 5.1(3), Screening Operating Experience Items, states, If the initial screening indicates potential applicability to a NextEra Energy nuclear plant, program (including corporate administered programs), policy, process, or procedure; then an evaluation is conducted. Subsequently, a disconnected magnetic speed sensor cable on the G 04 emergency diesel generator caused a failure during a surveillance run attempt. The licensees short -term corrective actions included reconnecting the G 04 emergency diesel generator ( EDG ) magnetic speed senor cable and installing lock -wire to prevent the connector from unintentionally disconnecting. The licensees long- term corrective actions included changing their maintenance procedures to check connector tightness on the diesels periodically. The inspectors determined that the failure to evaluate the external operating experience was contrary to licensee program descript ion PI AA 102 and was a performance deficiency. The finding was determined to be more than minor because the failure to evaluate operating experience was associated with the Mitigating Systems cornerstone attribute of Equipment Reliability and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, to this finding. The inspectors answered Yes to question A within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix G , Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014 . The in spectors referenced Exhibit 3Mitigating Systems Screening Questions. The finding screened as of very low safety significance (Green) because the inspectors answered No to the screening questions. The inspectors did not identify a cross -cutting aspect. The cause of the finding occurred in 2012 and was not reflective of present performance.
05000282/FIN-2017002-0230 June 2017 23:59:59Prairie IslandFailure to Implement the Emergency PlanA self-revealed finding, and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54 (q)(2), and 10 CFR 50.47 (b)(5) was identified on August 13, 2016, after a Notice of Unusual Event (NOUE) was declared due to reactor coolant system leakage greater than 25 gpm, the Shift Emergency Communicator (SEC) did not notify the States, Locals, and Tribal Community within 15 minutes of the classification.The inspectors reviewed IMC 0612, Appendix B, and determined that the finding was more than minor because it adversely affected the Emergency Response Performance attribute of the EP cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Since the finding involved a failure to implement emergency preparedness requirements, the inspectors reviewed IMC 0609, Appendix B, Attachment 1, and determined that this was a finding of very-low significance (Green) because it involved the failure to notify the offsite response organizations as required in the Emergency Plan after the classification of an NOUE. The cause of this finding involved the cross-cutting area of human performance, with the aspect of procedure use and adherence because the SEC did not appropriately follow the notification procedure. (H.8)
05000282/FIN-2017201-0130 June 2017 23:59:59Prairie IslandSecurity
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05000282/FIN-2017002-0330 June 2017 23:59:59Prairie IslandFailure to Make an 8Hour Report Required by05000306/201700203 10 CFR 50.72(b)(3)(ii)(B)The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.72(b)(3)(ii)(B) due to the licensees failure on March 20, 2017, to report an unanalyzed condition within eight hours of discovery. Specifically, removing the lower latch assembly of a transom above Door 225, a steam exclusion barrier, during maintenance resulted in the inoperability of the Units 1 and 2 safeguards batteries and Auxiliary Feed Water (AFW) systems, and Unit 1 safeguards bus as determined by CAP 1549724.The inspectors determined that the failure to submit a report required by 10 CFR 50.72 for the unanalyzed condition described above was a performance deficiency. The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the information that 10 CFR 50.72 reporting serves. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement process. The inspectors determined that this issue was a SL IV violation based on Example 6.9.d.9 in the NRC Enforcement Policy. Example 6.9.d.9 specifically states, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. Because the issue has been evaluated under the Traditional Enforcement process, there was no cross-cutting aspect associated with this violation.
05000282/FIN-2017002-0130 June 2017 23:59:59Prairie IslandFailure to Properly Implement the Minor Maintenance Process During Door 225 Transom MaintenanceThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of TS 5.4.1.a, Procedures, associated with the licensees failure to properly implement Procedure FPWMMMP01, Minor Maintenance Process, Revision 5, while planning and performing maintenance on a steam exclusion barriertransom latch assembly. Specifically, on February 3, 2017, maintenance workers in coordination with the Fix-It-Now (FIN) Senior Reactor Operator (SRO) removed the lower latch assembly from a transom above Door 225 that rendered the steam exclusion barrier non-functional. Consequently, for an approximately five minute window during maintenance on the latch assembly, the 11 safeguards battery system was rendered inoperable with respect to a postulated turbine building High Energy Line Break (HELB) event. The licensee entered the issues into the Corrective Action Program (CAP) as CAPs 1548470 and 1549724.The inspectors determined that the licensees failure to properly implement procedure FPWMMMP01 as required by Technical Specification (TS) 5.4.1.a. was aperformance deficiency. The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating Systems Cornerstone attribute of Human Performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. Since the inspectors answered No to all questions within IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of Teamwork in the Human Performance cross-cutting area, and involved individuals and work groups not properly communicating and coordinating their activities within and across organizational boundaries to ensure nuclear safety was maintained. (H.4)
05000282/FIN-2017001-0131 March 2017 23:59:59Prairie IslandFailure to Evaluate Changes to NRC Approved MethodologySeverity Level IV/Green. The inspectors identified a Green finding and associated Severity Level IV Violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.59(d)(1), for the licensees failure to perform a written evaluation which provided the bases for t he determination that a change in the NRC approved Westinghouse methodology referenced in the Updated Safety Analysis Report (USAR) for evaluating the acceptability of reactor pressure vessel internals baffle former bolting distributions did not require a license amendment. This finding was entered into the licensees Correction Action Program ( CAP ) as CAP documents 1539487, Documentation Missing in 50.59 Screening 4443, dated October 26, 2016; 1552331, BFB Screen Referenced Eval for SER Limitation 4 No n-Existent, dated March 6, 2017; and 1552314, BFB Screening Lacks Documentation for SER Limitation 3, dated March 6, 2017. The licensee performed an operability determination and determined the baffle bolts were operable. The inspectors reviewed the operability determination and no performance deficiencies were identified in this determination. The inspectors determined that the licensees failure to perform a written evaluation, providing the bases for the determination that a change in the NRC approved Westinghouse methodology for evaluating the acceptability of baffle former bolting distributions did not require a license amendment, was a performance deficiency. This finding was also evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The performance deficiency was determined to be more -than -minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, compliance with the NRC approved methodology of WCAP 15029 PA ensured the baffle former assembly maintained its structural integrity, avoiding a failure or excessive deflection of the baffle plates, and hence the primary concern of ensuring the emergency core cooling system could continue to perform its design function of cooling the reactor core. The inspectors determined the finding could be evaluated using the Significance 3 Determination Process (SDP) in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At -Pow er, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone. The finding screened as having very- low safety significance (Green) because the emergency core cooling system maintained its operability , specifically with respect to performing its safety function of ensuring adequate core cooling. As such, the finding corresponded to a Severity Level IV Violation in accordance with Example 6.1.d.2 of the NRC Enforcement Policy. The inspectors did not identify a cross cutting aspect because the performance deficiency was from 2013, and hence the issue did not represent current performance
05000282/FIN-2016004-0331 December 2016 23:59:59Prairie IslandFailure to Adequately Calibrate an ElectrometerGreen. A finding of very low safety significance, and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 20.1501(c) was identified by the inspectors for the failure to adequately calibrate the electrometer utilized in the validation of a JL Shepherd Calibrator. Specifically on November 30, 2015, the licensee performed a validation of a JL Shepherd Calibrator to ensure its correct operation. The electrometer used was incorrectly calibrated. The electronics and the detectors were required to be calibrated as a set, and this was not performed. The licensee entered this issue into their CAP as CAP 1543432. The inspectors determined that the licensees failure to properly calibrate the electrometer was a PD. The PD was more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Occupational Radiation Safety Cornerstone attribute of Program and Process and affected the Cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors applied IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to this finding. Since the finding was not associated with as-low-as-reasonably-achievable (ALARA) planning or work controls, nor was there an overexposure or a substantial potential for an over exposure and the ability to assess dose was not compromised, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the PD was associated with the cross-cutting aspect of Challenge the Unknown in the Human Performance cross-cutting area, and involved the licensee not challenging an unauthorized substitution for part of the electrometer that was damaged during shipment. (H.11)
05000282/FIN-2016004-0131 December 2016 23:59:59Prairie IslandBaffle Former Bolting Acceptance CriteriaFrom October 17November 28, 2016, the inspectors conducted a review of the implementation of the licensees inservice inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS), risk-significant piping and components and containment systems. This inspection constituted one ISI sample (see Sections 1R08.1, 1R08.3 and 1R08.5 below), as defined in IP 71111.0805. .1 Piping Systems Inservice Inspection a. Inspection Scope The inspectors either observed or reviewed records of the following Non-Destructive Examinations (NDEs) mandated by the American Society of Mechanical Engineers (ASME), Section XI Code, to evaluate compliance with the ASME Code Section XI and Section V requirements, and if any indications and defects were detected, to determine if these were dispositioned in accordance with the ASME Code or an NRC-approved alternative requirement. Ultrasonic examination of tubesheet to shell for steam generator (SG) 11; Magnetic particle examination of an integral attachment support rod for SG 11; Visual examination of reactor vessel nuts and washers (1 through 16); and Unit 1 metallic containment liner visual examination in 2012. During non-destructive surface and volumetric examinations performed since the previous refueling outage, the licensee had not identified any recordable indications. Therefore, no NRC review was completed for this inspection procedure attribute. The inspectors either observed or reviewed the following pressure boundary welds completed for risk-significant systems since the beginning of the last refueling outage to determine if the licensee applied the preservice NDEs, and acceptance criteria required by the Construction Code and ASME Code, Section XI. Additionally, the inspectors reviewed the welding procedure specification and supporting weld procedure qualification records to determine if the weld procedures were qualified in accordance with the requirements of Construction Code and ASME Code Section IX. Unit 1 reactor coolant pump (RCP) seal replacements. b. Findings No findings were identified. .2 Reactor Pressure Vessel Upper Head Penetration Inspection Activities a. Inspection Scope The licensee did not perform any welded repairs to vessel head penetrations since the beginning of the preceding outage for Unit 1. Therefore, no NRC review was completed for this inspection procedure attribute. For the Unit 1 vessel head, no examination was required pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50.55a(g)(6)(ii)(D) for the current refueling outage. Therefore, no NRC review was completed for this inspection attribute. b. Findings No findings were identified. .3 Boric Acid Corrosion Control a. Inspection Scope The inspectors performed an independent walkdown of the RCS and related lines in the containment, which had received a recent licensee boric acid walkdown, and verified whether the licensees boric acid corrosion control visual examinations emphasized locations where boric acid leaks can cause degradation of safety significant components. The inspectors reviewed the following licensee evaluations of RCS components with boric acid deposits to determine if degraded components were documented in the CAP. The inspectors also evaluated corrective actions for any degraded RCS components to determine if they met the ASME Section XI Code. 11 RCP seal bowl. The inspectors reviewed the following corrective actions related to evidence of boric acid leakage to determine if the corrective actions completed were consistent with the requirements of the ASME Code Section XI and 10 CFR Part 50, Appendix B, Criterion XVI. CAP 1465567; 12 RCP Seal Leakage. b. Findings No findings were identified. .4 Steam Generator Tube Inspection Activities a. Inspection Scope The licensee did not perform in-situ pressure testing of SG tubes. Therefore, no NRC review was completed for this inspection attribute. For the Unit 1 SGs, no examination was required pursuant to the TSs during the current refueling outage. Therefore, no NRC review was completed for this inspection procedure attribute. b. Findings No findings were identified. .5 Identification and Resolution of Problems a. Inspection Scope The inspectors performed a review of ISI/SG-related problems entered into the licensees CAP, and conducted interviews with licensee staff to determine if: the licensee had established an appropriate threshold for identifying ISI/SG-related problems; the licensee had performed a root cause evaluation (if applicable) and taken appropriate corrective actions; and the licensee had evaluated operating experience and industry generic issues related to ISI and pressure boundary integrity. The inspectors performed these reviews to evaluate compliance with 10 CFR Part 50, Appendix B, Criterion XVI requirements. Documents reviewed are listed in the Attachment to this report. b. Findings (1) Baffle Former Bolting Analysis Acceptance Criteria Introduction: The inspectors identified an Unresolved Item (URI) concerning the analysis that demonstrated the design adequacy of the baffle former bolting under design and licensing basis loading conditions. Description: The inspectors reviewed WCAP 17586P, Determination of Acceptable Baffle-Barrel Bolting for Prairie Island Units 1 and 2, Revision 0; WCAP15030NPA, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions under Faulted Load Conditions, dated March 2, 1999; and Safety Evaluation by the Office of Nuclear Reactor Regulation of WCAP15029, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, dated November 10, 1998. The inspectors were concerned that the licensee had evaluated the baffle former bolting using acceptance criteria different than what was reviewed and approved by the Office of Nuclear Reactor Regulation. In WCAP15030NPA, Section 4.3.2 stated that the stress allowable for primary membrane and bending of irradiated bolt material is taken to 0.9 times Sy (yield stress of baffle bolt material) for the faulted load condition. The stress allowable used in WCAP 17586P was based on ASME, Section III, Appendix F, specifically: (minimum of (0.9 times Su) ultimate stress of baffle bolt material), maximum of (0.67 times Su, Sy + 1/3 (Su - Sy)). The inspectors also reviewed 10 CFR 50.59 Screening No. 4443, Determination of Acceptable Baffle-Barrel Bolting, dated January 24, 2013, to determine whether the licensee performed a 50.59 evaluation for the use of ASME, Section III, Appendix F acceptance criteria. However, the inspectors identified that the change for the use of ASME, Section III, Appendix F acceptance criteria in lieu of the acceptance criteria contained in Section 4.3.2 of WCAP15030NPA was not explicitly reviewed in 50.59 Screening No. 4443. In response to the inspectors concern, the licensee initiated CAP 1539487, Documentation Missing in 50.59 Screening 4443, dated October 26, 2016. This issue is an URI pending evaluation of these concerns by the licensee, subsequent inspector review, and discussion with the licensee and Office of Nuclear Reactor Regulation (URI 05000282/201600401; 05000306/201600401; Baffle Former Bolting Analysis Acceptance Criteria).
05000244/FIN-2016404-0131 December 2016 23:59:59GinnaSecurity
05000282/FIN-2016004-0231 December 2016 23:59:59Prairie IslandFailure to Properly Implement a Post-Maintenance Test Procedure during Safety Injection System Valve TestingGreen. A finding of very low safety significance was self-revealed, and an associated NCV of Technical Specification (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly implement surveillance procedure (SP) 1088B, Train B Safety Injection Quarterly Test, Revision 24, while performing a post-maintenance valve stroke test. Specifically, on November 14, 2016, while cycling a safety injection (SI) system pump suction valve, operators exposed the SI suction header to reactor coolant system (RCS) pressure, causing a relief valve to lift as designed, a subsequent unexpected RCS pressure drop below 240 pounds per square inch (psig), and requiring operators to trip both reactor coolant pumps (RCPs). The licensee entered the issue into the Corrective Action Program (CAP) as CAP 1541821. The inspectors determined that the licensees failure to properly implement procedure SP 1088B as required by TS 5.4.1.a was a performance deficiency (PD). The PD was determined to be more than minor and a finding in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Initiating Events Cornerstone attribute of Configuration Control and affected the associated Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, to this finding. Since the finding pertained to an event while the plant was shut down, the inspectors transitioned to IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Since the inspectors answered No to all questions within IMC 0609, Appendix G, Attachment 1, Exhibit 2, Initiating Events Screening Questions, the finding screened as very low safety significance (Green). The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the PD was associated with the cross-cutting aspect of Teamwork in the Human Performance cross-cutting area, and involved individuals and work groups not communicating and coordinating their activities within and across organizational boundaries to ensure nuclear safety was maintained. (H.4)
05000266/FIN-2016004-0131 December 2016 23:59:59Point BeachScaffolds Constructed Without Required Engineering ApprovalGreen: A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the licensees failure to follow step 4.1.3 (2) of procedure MAAA1001002, Scaffold Installation, Modification, and Removal Requests. Specifically, the licensee failed to obtain and document engineering approval for multiple scaffolds constructed in the cable spreading room that did not meet the separation criteria of Attachment 1 of MAAA1001002. The licensees short-term corrective actions included obtaining the appropriate engineering evaluations for the affected scaffolding and conducting a stand-down and information sharing with the scaffold builders to ensure they were aware of the importance of obtaining engineering approvals. The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, if the licensee continued to construct scaffolding without obtaining required engineering approvals, scaffolding could be constructed that was not seismically qualified and adversely affect the operability of surrounding structures, systems, and components (SSCs). The inspectors concluded this finding was associated with the Mitigating Systems cornerstone. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on October 7, 2016. Specifically, the inspectors used IMC 0609, Appendix A, SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered "No" to the screening questions. This finding has a cross-cutting aspect of Teamwork (H.4), in the area of Human Performance, for the failure of individuals and work groups to communicate and coordinate their activities across organizational boundaries to ensure nuclear safety is maintained. Specifically, the scaffold building team failed to communicate with the engineering organization to ensure the engineering evaluations were complete.
05000244/FIN-2016003-0130 September 2016 23:59:59GinnaFailure to Perform Drills Required by the Site Emergency PlanThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) for Exelons failure to maintain an emergency plan that meets the requirements in Appendix E, Content of Emergency Plans, to Part 50 and the planning standards of 50.47(b). Specifically, Exelon did not perform a drive-in augmentation drill during the required 3-year cycle nor did they perform a health physics drill semi-annually as required by Ginnas Emergency Plan Implementing Procedure EP-AA-122-100, Drill and Exercise Planning and Scheduling. Immediate corrective actions included entering this issue into their corrective action program (CAP). This finding is more than minor because it is associated with the emergency response organization (ERO) readiness attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective to ensure that Exelon is capable of maintaining adequate measures to protect the health and safety of the public in the event of a radiological emergency. In accordance with IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Attachment 2, Failure to Comply Significance Logic, the inspectors determined that the performance deficiency affected planning standard 10 CFR 50.47(b)(14). The inspectors concluded that this performance deficiency matched an example on Table 5.14-1 Significance Examples 50.47(b)(14), for a Degraded Planning Standard Function. Specifically, two drills had not been conducted during a 2year (calendar) period in accordance with the emergency plan, thus constituting a degraded planning standard function which corresponds to a very low safety significance (Green) finding. The cause of the finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon did not schedule or plan for a drive-in augmentation drill or health physics drills in accordance with procedure EP-AA-122-100. (H.8)
05000266/FIN-2016002-0230 June 2016 23:59:59Point BeachSubmerged Safety-Related EDG Fuel Oil Transfer Pump CablesA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors, for the failure to maintain emergency diesel generator (EDG) fuel oil transfer pump safety-related cables in an environment for which they were designed. Specifically, the licensee allowed the safety-related cables to be submerged in water, which was outside of their design, in manhole Z066B. The licensees corrective actions included pumping the water out of the manholes, repairing the failed sump pump, level switch, and alarm circuit; and performing an engineering evaluation to quantify the level of degradation as a result of the submergence. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered "Yes" to the question does the SSC maintain its operability or functionality. Specifically, the submergence of the G01 and G02 EDG fuel oil transfer pump cables did not render the transfer pumps inoperable. This finding has a cross-cutting aspect Evaluation (P.2) in the area of problem identification and resolution, because the licensee did not thoroughly evaluate problems to ensure that resolutions address causes and extent of conditions, commensurate with their safety significance. Specifically the licensee failed to thoroughly investigate and prioritize the failure of the manhole alarm and pumping system according to the safety significance of the cables contained within the manholes which led to prolonged and unevaluated submergence of the cables.
05000266/FIN-2016002-0130 June 2016 23:59:59Point BeachFailure to Perform Required Fire Watches in Areas Containing Transient CombustiblesA finding of very low safety significance and associated NCV of license condition 4.F was identified by the inspectors for the licensees failure to conduct required fire watch inspections in accordance with the licensees Fire Protection Program requirements. Specifically, while conducting fire protection walkdowns of both units residual heat removal (RHR) pipeway and heat exchanger rooms, the inspectors discovered numerous transient combustible items in areas that the licensee had controlled using tamper seals on the entrances in lieu of physical entry. The licensees corrective actions included documenting and quantifying the removal of the items from the zones and additional actions to perform additional evaluation of the fire zones. The finding was determined to be more than minor because the failure to conduct the required fire watch inspections was associated with the Initiating Events cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage). Specifically, the failure to conduct the required fire watch inspections or meet the alternate measures specified by the licensees engineers, allowed unanalyzed transient combustibles and ignition sources to be present in fire zones that contained both trains of both units RHR pumps, heat exchangers and associated equipment. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue under the Phase 1 Screening Question 1.3.1A, and determined that determined that the finding was of very low safety-significance (Green), because the inspectors determined that the impact of a fire would not prevent either reactor from reaching and maintaining safe shutdown (hot). This finding has a cross-cutting aspect of Bases for Decisions (H.10), in the area of human performance, because the licensees leadership did not ensure that the bases for operational and organizational decisions are communicated in a timely manner. Specifically, the licensee did not periodically verify the understanding of the individuals assigned to fire watches, in particular, that the relief from physical entry and application of a tamper seal required a thorough tour of the zones following any entry into those fire zones.
05000266/FIN-2016002-0430 June 2016 23:59:59Point BeachViolation of Technical Specifications During Mode 4 Entry with LCO 3.6.6 Not MetA finding of very low safety significance and associated NCV of Technical Specification 3.0.4 was identified by the inspectors for the licensees failure to follow procedure OP 1A, Cold Shutdown to Hot Standby Unit 1 and checklist CL 2C, Mode 5 to Mode 4 Checklist. Specifically, the licensee entered Mode 4 from Mode 5 without meeting the requirements of LCO 3.0.4 for entering a Mode when an applicable LCO is not met. The licensee had not met LCO 3.6.6 because the control switches for two out of the required four containment accident recirculation fans were in their pullout position instead of the required automatic position. Corrective actions for this event included restoration of accident cooler fan control switches to automatic. Additional corrective actions included: performance of an apparent cause evaluation; changes to the licensees ORT 3 test procedures to restore accident fan cooler switches after completion of testing; updating OP 1A to include performance of a control room shift turnover checklist prior to changing modes; and planned enhancements to CL 2 series procedures to strengthen a note on the responsibility of the SRO when ensuring operability of LCOs. The inspectors determined that the finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow procedures OP 1A and CL 2C caused the licensee to unknowingly operate with multiple containment accident recirculation fans inoperable, which were required in Mode 4. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier Integrity Screening Questions, dated May 9, 2014. The inspectors answered no to the Containment Barrier Screening Questions and determined the finding had very low safety significance (Green). This finding has a cross-cutting aspect of Challenge the Unknown (H.11), in the area of Human Performance, for failing to stop when faced with uncertain conditions. Specifically, when the licensee assessed the illuminated Safeguards Equipment Locked Off alarm, during their control board walk down, they confirmed that the safety injection pump control switch was in pullout and was a reason for the alarm to actuate; however, they failed to confirm that other inputs to the alarm were also not valid.
05000266/FIN-2016002-0530 June 2016 23:59:59Point BeachFuel Assembly Move Sequence Planned IncorrectlyA finding of very low safety significance was identified by the inspectors, for the licensees failure to follow procedure REI 26.0, Fuel/Insert/Component Movement Planning. Specifically, the licensee failed to follow procedure REI 26.0, Step 5.5.7.b, which verified that the licensee would not place fuel assemblies with cooling times less than 295 days into spent fuel pool rack foot locations. The licensees corrective actions included completing additional spent fuel moves, which placed the spent fuel pool into an appropriate configuration. The inspectors determined that the finding was more than minor, because, if left uncorrected, it had the potential to become a more significant safety concern. Specifically, if the inspectors had not questioned the licensee about spent fuel pool rack foot locations, the spent fuel pool would have remained in an incorrect configuration. The inspectors concluded this finding was associated with the Barrier Integrity cornerstone. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix L, B.5.b Significance Determination Process, Table 2 Significance Characterization, The inspectors determined that the finding did not meet the criteria in Table 2 for a Greater-Than-Green significance; therefore, the finding was of very low safety significance (Green). This finding has a cross-cutting aspect of Avoid Complacency (H.12), in the area of Human Performance, for failing to implement appropriate error reduction tools. Specifically, the licensee became desensitized to overriding fuel placement constraints and failed to implement effective human performance tools to prevent the error.
05000282/FIN-2016007-0130 June 2016 23:59:59Prairie IslandFailure to Ensure Breaker Main Contacts are Fully AlignedA finding of very low safety significance and associated non-cited violation of Technical Specification Section 5.4.1, Procedures, was identified by the inspectors for the licensees failure to ensure the 21 safeguards diesel exhaust fan main contact connectors were fully engaged and aligned as required per electrical maintenance procedures to ensure proper operation of the breaker. As part of their corrective actions, the licensee aligned and re-engaged the main contact connectors as necessary. In addition, the licensee ensured maintenance personnel were aware of the operating experience to prevent the same issue from occurring in the future. The violation was entered into the licensees corrective action program as Action Request 1525844. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone and the breaker failure led to the inoperability of the 21 safeguards diesel exhaust fan and impacted the availability of the 22 cooling water system diesel driven pump. This finding represented a loss of the 22 safeguards diesel cooling water pump function for longer than the Technical Specification allowed outage time of 7 days and therefore required a detailed risk evaluation. The regional senior reactor analyst performed a detailed risk evaluation of this finding using the Prairie Island Standardized Plant Analysis Risk Model revision 8.19 and determined the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because it was not indicative of current performance.
05000266/FIN-2016002-0630 June 2016 23:59:59Point BeachIncorrect Wiring Causes Transformer LockoutA finding of very low safety significance and associated NCVs of TS 3.8.1, AC Sources-Operating and TS 3.8.2, AC Sources-Shutdown, were self-revealed for the licensees failure to follow procedure RMP 90569B, 1X03, Protective Relay Calibration and Testing. Specifically, a wiring error in the 1X03 connection box, which occurred in 2013, caused the 1X03 transformers differential protection circuity to lockout the transformer at current levels below the design protection values. The licensees corrective actions included correcting the improper wiring in the 1X03 connection box and evaluating other work performed by the same vendor during that timeframe. The inspectors determined that the finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance and 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. Specifically, the lockout of 1X03 caused a loss of one of the licensees offsite power lines and also caused a loss of power to multiple station battery chargers placing Unit 2 into limiting condition for operation (LCO) 3.0.3. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012. The inspectors answered Yes to the Support System Initiators question; therefore, a Detailed Risk Evaluation was required. Based on the conclusions in the Detailed Risk Evaluation, the SRA determined that the finding was of very low safety significance (Green). This finding has a cross-cutting aspect of Avoid Complacency (H.12), in the area of Human Performance, for failing to implement appropriate error reduction tools. Specifically, the incorrectly performed procedure step, in RMP 9056-9B, clearly specified which terminal point to land the wires on, the terminal points were clearly labeled, and the step required a concurrent verification; however, even with those barriers in place, the task performers still landed the wires on the wrong location.
05000282/FIN-2016007-0230 June 2016 23:59:59Prairie IslandInadequate Operability DeterminationsA finding of very low safety significance with two examples and an associated non-cited violation of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish the requirements of procedure FPOPOL01, Operability/Functionality Determination, Revisions 14 and 15. Specifically, on two occasions, the licensee failed to properly evaluate potential operability concerns associated with the Unit 2 emergency diesel generator (EDG) day tanks and the Unit 2 train A cooling water (CL) system piping. The licensee entered the issues into the Corrective Action Program as Action Requests 1525842 and 1526070. The inspectors determined that the licensees failure to accomplish the requirements of procedure FPOPOL01, Operability/Functionality Determination, Revisions 14 and 15, to properly evaluate the operability issues associated with the Unit 2 EDG day tank fuel oil level and the Unit 2 CL system piping (both safety-related, mitigating systems) was a performance deficiency. The performance deficiency, with two examples, was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," it was associated with the Mitigating Systems Cornerstone attributes of Equipment Performance (for the Unit 2 EDGs) and Protection against External Factors (for the Unit 2 CL piping) and adversely affected the Cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events. The inspectors utilized IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding screened as very low safety significance (Green) since the inspectors answered Yes to Question 1 of Section A of Exhibit 2, Mitigating Systems Screening Questions. The inspectors concluded that this issue was cross-cutting in the area of Problem Identification and Resolution in the aspect of Evaluation. As defined in IMC 0310, Aspects Within the Cross-Cutting Areas, this aspect states, The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee had not thoroughly evaluated the operability issues associated with the Unit 2 EDG day tank levels and the Unit 2 CL piping structural integrity.
05000244/FIN-2016002-0130 June 2016 23:59:59GinnaIncorrect Emergency Action Level TableExelon identified that they had inadvertently made a change to the Ginna Emergency Plan. The NRC determined that this error is a preliminary White finding under the Reactor Oversight Process and a violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.54 (q)(2), Emergency Plans, because Exelon did not maintain the effectiveness of Ginnas Emergency Plan such that it met the requirements of Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, and the planning standards of 10 CFR 50.47(b). Specifically, Exelon implemented a revision to the emergency action level (EAL) table for the fission product barrier matrix that was incorrect with respect to the EAL threshold associated with potential loss of containment barrier. This could have resulted in an untimely declaration of a General Emergency or a failure to declare a Site Area Emergency during an actual event. Using IMC 0612, Appendix B, Issue Screening, the performance deficiency was determined to be more than minor because it impacted the procedure quality attribute of the Emergency Preparedness cornerstone and adversely affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, Exelons EAL table was revised without adequate technical reviews resulting in a discrepancy between the EAL table and the EAL technical basis. The EAL wording of Table F-1 containment barrier potential loss, block C.6 did not meet the minimum required operable equipment in all situations and could have resulted in a delayed General Emergency declaration or a failure to declare a Site Area Emergency. The inspectors utilized IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), to determine the significance of the performance deficiency. The performance deficiency is associated with the emergency classification system planning standard and is considered a risk-significant planning standard function. The inspectors were directed by the SDP to compare the performance deficiency with the examples in Section 5.4, 10 CFR 50.47 (b)(4), Emergency Classification System, to evaluate the significance of this performance deficiency. In accordance with Section 5.4, when an EAL has been rendered ineffective such that any General Emergency declaration would not be declared, but due to other EALs, an appropriate declaration would be made in a degraded manner or any Site Area Emergency would not be declared for a particular off-normal event, a degradation of risk-significant planning standard function (b)(4) is determined; and the finding is White. The finding has a cross-cutting aspect in the area of Human Performance, Change Management, because Exelon did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, Exelon did not maintain a clear focus on nuclear safety when implementing changes to the EALs resulting in a significant unintended consequence, the potential to make an untimely emergency declaration.
05000266/FIN-2016002-0330 June 2016 23:59:59Point BeachSuitability of Reactor Protection System and Engineered Safeguards System ComponentsDuring the review of the Reactor Protection System (RPS), the inspectors identified an Unresolved Item (URI) associated with components in both units RPS and engineered safeguards (ESF) system which contained components known to degrade with age, including electrolytic capacitors. In some cases, these components may have been installed as original plant equipment. During the inspectors review of system health reports associated with both Units 1 and 2 RPS, and ESF system as an extent of condition review, the inspectors identified a URI associated with components in hundreds of safety-related RPS and ESF printed circuit boards, power supplies, amplifiers, transmitters, and other related components that potentially exceeded their design criteria for the time period that the components were installed for which no evaluations existed. The inspectors determined that this was an issue of concern in which more information was needed to determine if the issue constituted one or more violations of NRC requirements. Specifically, the inspectors determined that subcomponents, including but not limited to electrolytic capacitors, were installed in both safety trains of both units RPS and ESF components, in some cases for over 40 years without any documented evaluation of age-related degradation mechanisms. The inspectors needed to evaluate the licensees operability determinations that resulted from this inspection activity, any engineering evaluations to provide justification for suitability with respect to design control, recovery plans, a review of the proposed preventative maintenance activities, current failure rates and drift trending, and any other information provided by the licensee that may provide a technically defensible basis for the continued operation. The issue is unresolved pending further NRC review of the licensees evaluation.