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05000266/FIN-2018003-032018Q3Point BeachLicensee-Identified ViolationViolation: Point Beach Nuclear Plant, Units 1 and 2, Renewed Operating License Condition 4.F, requires the licensee to implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, as specified in the license amendment requests and as approved in the safety evaluation report dated September 8, 2016. Section 2.4.3.2, of NFPA 805, states that the PSA (Probabilistic Safety Assessment) evaluation shall address the risk contribution associated with all potentially risk-significant fire scenarios.Contrary to the above, from February 14, 2017 through June 14, 2018, the licensees PSA failed to address the risk contribution associated with all potentially risk-significant scenarios. Specifically, the licensee improperly excluded the risk contribution from 27 electrical panels because they had incorrectly concluded that internal fires would not propagate outside the panel walls due to them being misclassified as well-sealed. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) 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). The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The finding did not screen to green using questions 1.5.1A and 1.5.1B and thus required a detailed risk evaluation. The Senior Reactor Analyst performed walkdowns of dominant fire sequences and conducted an onsite review of the licensee's fire calculation which confirmed that the increase in risk due to the finding was less than 1E6/year (Green).
05000456/FIN-2018003-022018Q3BraidwoodMinor ViolationAll Braidwood Station EDG governors were replaced during the late 1990s. During design testing, the licensee noted that the historical EDG frequency response had changed slightly due to installation of new electronic governors. Prior to these governor replacements, EDG frequency was always above 57 hertz (Hz) during load sequencing. However, with the newly installed electronic governors, 1A and 2A EDG frequency was observed to dip below the 57 Hz under frequency relay setpoint following start of the 1A and 2A motor-driven AF pumps. (Note that because the 1B and 2B AF pumps are diesel-driven, there is no corresponding impact on the 1B or 2B EDGs.) As a result, an external 2-second time delay, provided by an Agastat time delay relay, was incorporated into the under frequency trip logic for the 1A and 2A EDGs to provide an additional margin for frequency recovery following motor-driven AF pump load starts. The Braidwood governor modification was installed in 1998, with the external time delay added to the 1A and 2A EDGs as part of the design changes to prevent inadvertent actuations of the under frequency logic.During the licensees investigation into the issue discussed in the subject LER, it was identified that the external Agastat time delay was installed incorrectly on the 1A EDG. Specifically, the original trip logic wiring had not been properly removed, which permitted the actuation of the under frequency trip after the original 0.5 second internal time delay through the bypassing of the additional 2.0 second external time delay. The wiring error was introduced during the original modification installation in October 1998. Screening: The inspectors determined that the error was of minor safety significance. Absent the mechanical binding of the manual fuel trip lever and associated linkage, as discussed in NCV 05000456/201800301 in this report, the 1A EDG had performed reliably and satisfactorily during surveillance testing prior to the Unit 1 refueling outage testing in April of 2018. Additionally, the inspectors determined that the error, having occurred some 20 years ago, was not indicative of current licensee performance.Violation: This failure to comply with the requirements of 10 CFR Part 50, Appendix B, Criterion III , Design Control, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000266/FIN-2018003-022018Q3Point BeachLicensee-Identified ViolationViolation: Point Beach Nuclear Plant, Units 1 and 2, Renewed Operating License Condition 4.F, requires the licensee to implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, as specified in the license amendment requests and as approved in the safety evaluation report dated September 8, 2016.Section 1.5.1, Nuclear Safety Performance Criteria, of NFPA 805, stated in part, that fire protection features shall be capable of providing reasonable assurance that, in the event of a fire, the plant is not placed in an unrecoverable condition. To demonstrate this, the following performance criteria shall be met: (a) Reactivity Control; (b) Inventory and Pressure Control; (c) Decay Heat Removal; (d) Vital Auxiliaries; and (e) Process Monitoring.Section 1.5.1 (d), Vital Auxiliaries, of NFPA 805, stated that vital auxiliaries shall be capable of providing the necessary auxiliary support equipment and systems to assure that the systems required under (a), (b), (c), and (e) are capable of performing their required nuclear safety function. Contrary to the above, from March 16, 2018 through April 11, 2018, the licensee failed to ensure that vital auxiliaries were capable of providing the necessary auxiliary support equipment and systems to assure that the systems required under (a), (b), (c), and (e) are capable of performing their required nuclear safety function. Specifically, select 120 VAC instrument buses, needed as a vital auxiliary, would not have been energized during certain fire scenarios and compensatory measures were not implemented. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) 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). The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The finding did not screen to green using questions 1.5.1A and 1.5.1B and thus required a detailed risk evaluation. The Senior Reactor Analyst performed walkdowns of dominant fire sequences and conducted an onsite review of the licensee's fire calculation which confirmed that the increase in risk due to the finding was less than 1E6/year (Green).
05000456/FIN-2018003-012018Q3BraidwoodInadequate Detail in Maintenance Procedure for Emergency Diesel Generator 2-Year Inspection Contributed to 1A Emergency Diesel GeneratorFuel Rack BindingA self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to include adequate detail within their maintenance procedures to enable proper lubrication of the emergency diesel generator (EDG) fuel rack control linkage. Specifically, the preventative maintenance template for the fuel rack control linkage required that the manual fuel trip lever and associated linkage be lubricated every 2 years. However, the licensees implementing 2year maintenance procedure failed to include specific instructions to disassemble the lever assembly for lubrication. This lack of lubrication contributed to the mechanical binding of the emergency diesel generator fuel rack and failure of the 1A EDG during surveillance testing on April 22,2018.
05000266/FIN-2018003-012018Q3Point 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.
05000266/FIN-2018003-042018Q3Point BeachLicensee-Identified ViolationViolation: Title 10 CFR 72.150 states The licensee . . . shall prescribe activities affecting quality by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall require that these instructions, procedures, and drawings be followed. Contrary to the above, on June 5 and 6, 2018, the licensee failed to follow procedures for an activity affecting quality. Specifically, during a dry run in the primary auxiliary building (PAB) over the spent fuel pool (SFP), the pin lock for the pin which engages the Point Beach pool lift yoke to the PAB overhead crane was not correctly engaged when lifting the transfer cask (TC) out of the pool. After the TC was set down in the decon area, the lift yoke was then left unattended over the SFP over spent fuel. This is not in accordance with procedure RP 17 Part 4, Revision 26, Step 5.1.4 for engagement of the pin lock, and not in accordance with procedure MAAA2121000, Revision 16, Step 4.5.3 for leaving a load suspended and unattended.Severity Level: The inspector determined the violation was more than minor, as informed by Inspection Manual Chapter (IMC) 0612 Appendix E, Example 4.k., in that there was a credible load drop scenario that could impact safety-related equipment. In accordance with Section 2.2 of the Enforcement Policy and IMC 0612, Appendix B, Issue Screening, Independent Spent Fuel Storage Installations are not subject to the Significance Determination Process and are not subject to the Reactor Oversight Process, so violations identified at ISFSIs are assessed using traditional enforcement. Consistent with the guidance in Section 1.2.6.D of the Enforcement Manual, if a violation does not fit an example in the Enforcement Policy Violation Examples, it should be assigned a severity level: (1) commensurate with its safety significance; and (2) informed by similar violations addressed in the Violation Examples. The inspector found no similar violations in the violation examples. This violation was determined to be a Severity Level IV in that there was no load drop, and that the weight of any load on the pin would contribute to opposing any potential movement of the pin.
05000266/FIN-2018002-022018Q2Point 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.
05000266/FIN-2018002-012018Q2Point 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-2018001-052018Q1Point BeachLicensee-Identified ViolationViolation: Technical Specification (TS) 3.0.4 states in part that entry into a MODE or other specified condition in the Applicability of a limiting condition for operation (LCO) shall only be made when the LCOs Surveillances have been met... TS 3.7.5 Auxiliary Feedwater (AFW) Limiting Condition SR 3.7.5.1 required in part Verify each AFW manual, power operated, and automatic valve in each water path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. Contrary to the above, at 1500 on October 29, 2017, Unit 1 entered MODE 3 and the licensee failed to verify that AFW (System required for MODE 3) turbine driven (TD) AFW steam supply valves 1MS235 and 1MS237 were in the correct (open) position. These valves were in fact shut rendering the TDAFW pump inoperable until the licensee identified this error and opened these valves at 1610 on October 29, 2017(reference; Licensee Event Report 05000266/201700200, Operation or Condition Prohibited by Technical Specifications). Significance/Severity: This licensee identified finding, affected the Mitigating Systems Cornerstone and was screened in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, issued June 19, 2012. Because of the short duration (~1 hour) that the TDAFW pump was not operable, the inspectors determined that this finding is of very low safety significance (Green) because: the performance deficiency was not a design or qualification issue; it did not represent a loss of the system function; the train was neither inoperable for greater than its allowed outage time nor was it inoperable for greater than 24 hours; and was not part of an external event mitigating system. Corrective Action Reference: AR 02233500 Made Mode Change With Inoperable TDAFW
05000266/FIN-2018001-042018Q1Point 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
05000266/FIN-2018001-032018Q1Point 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.
05000266/FIN-2018001-022018Q1Point 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.
05000266/FIN-2018001-012018Q1Point 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.
05000282/FIN-2017004-012017Q4Prairie IslandLicensee-Identified ViolationTechnical Specification 5.7.1 states, High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates. Technical Specification 5.7.1, further requires in part, that each entryway to such an areashall be barricaded and conspicuously posted as a high radiation area.Contrary to the above, on October 19, 2017, a licensee system engineer identified during the performance of a maintenance and engineering inspection that a chain that functioned as the barricade for the 22 reactor coolant pump vault, a posted high radiation area, was not installed. The licensee documented this issue in CAP 501000004026. The inspectors determined that this issue was of very-low safety significance (Green) after reviewing IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that this finding was not an ALARA Planning or Work Control issue; was not an overexposure; was not a substantial potential for overexposure; and the ability to assess dose was not compromised.
05000282/FIN-2017004-022017Q4Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.54(q)(2) requires, in part, that a holder of a nuclear power reactor operating license shall follow and maintain the effectiveness of an emergency plan that meets the requirements in Title 10 CFR Part 50, Appendix E and the planning standards of Title 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4) requires, in part, that the onsite emergency response plans for nuclear power reactors must meet the following standard: a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.Contrary to the above, between November 22, 2000 and September 22, 2017, the licensee failed to maintain the effectiveness of an emergency plan that met the requirements of the planning standards of 10 CFR 50.47(b). Specifically, on September 22, 2017, the licensee identified that prior assessments of NRC Information Notice 9745, Supplement 1, Environmental Qualification Deficiency for Cables and Containment Penetration Pigtails, and a subsequent industry-initiated study to determine signal errors for Prairie Islands Unit 1 & 2 containment high range radiation monitors 1R48, 1R49, 2R48 & 2R49 (used in the licensees emergency classification and action level scheme) that impacted operability of the monitors, failed to restore capability to classify EALs during certain design basis accidents.The violation was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring capability of implementing adequate measures to protect the health and safety of the public in 33 the event of a radiological emergency. The inspectors referenced IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.41 and Figure 5.41. The finding was determined to be of very low safety significance (Green) because timely and accurate EAL classification capability for an event at the General Emergency level was unaffected due to redundant and diverse indications.In response, the licensee entered the issue into the CAP as CAP 501000001861, declared the containment high range radiation monitors inoperable per TS 3.3.3, Event Monitoring Instrumentation, implemented Emergency Plan interim measures to make the emergency response organization aware of the issue, performed an extent-of-condition review, and submitted a letter to the U.S. NRC within 14 days as required by TS. Final corrective actions included the addition of a note to the Prairie Island EAL matrix to acknowledge the potential for TIC errors for the containment high range radiation monitors during the first 5 minutes for post-loss of coolant accident (LOCA) or main steam line break events inside containment.
05000282/FIN-2017004-032017Q4Prairie IslandLicensee-Identified ViolationPrairie Island TS LCO 3.0.3 requires, in part, that when an LCO is not met and an associated ACTION is not provided, action shall be initiated within 1 hour to place the unit in MODE 3 within 7 hours.Contrary to the above, at 1556 hours on May 4, 2016, the licensee failed to place Unit 2 in MODE 3 within 7 hours due to no associated ACTION provided within TS 3.6.5, Containment Spray and Cooling Systems for two containment cooling trains not OPERABLE. Specifically, between May 4 and May 5, 2016, operators failed to recognize that with the ongoing unplanned inoperability of the 122 control room chiller, and the subsequent unplanned inoperability of the A train #23 CFCU, the 122 control room chiller was a required support system for the B train #22 and #24 CFCUs. Therefore, with both of the Unit 2 CFCU trains inoperable, LCO 3.0.3 was required to be entered to place Unit 2 in Mode 3 within 7 hours. Because the supported system TS applicability was not recognized, LCO 3.0.3 was not entered as required and both trains of Unit 2 CFCUs were inoperable for approximately 35 hours.Because the inspectors answered No to questions B.1 and B.2 under Exhibit 3, Barrier Integrity Screening Questions of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the finding screened as very low safety significance (Green). The issue was entered into the licensees CAP as CAP 501000002726. Corrective actions included re-assessing shared system LCOs between Units 1 and 2, revising the LCO tracking database, implementing new standards for LCO 3.0.6 applications, and revisions to the Safety Function Determination Program.
05000282/FIN-2017004-042017Q4Prairie IslandLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to the above, on October 17, 2017, with Unit 2 in Mode 5, Cold Shutdown, the licensee failed to accomplish procedure 2C12.2, Purification and Chemical Addition Unit 2; Revision 34. Specifically, control room operators signed off steps as completed without validating that the procedure actions were performed in the field. These procedure steps that intended to close letdown valves and open purification valves, resulted in unintended transfer of primary coolant from the RCS to the chemical and volume control system hold-up tank instead of back to the RCS. In turn, this resulted in a reduction in RCS inventorywith reactor vessel level at approximately 1 foot below the flange (reduced inventory operations). Due to operators quickly recognizing a lack of letdown flow as discussed during a pre-job brief, the purification evolution was halted and actions were taken to restore reactor vessel level.Because the inspectors answered No to questions B.2 and B.3 under Exhibit 2, Initiating Events Screening Questions of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). Specifically, the loss of inventory event was self-limiting such that the leakage would have stopped before impacting the operating method of decay heat removal (shutdown cooling via RHR in this case). The issue was entered into the licensees CAP as CAP 501000003923. Corrective actions included an operations department human performance clock reset to share the lessons learned from the event.
05000266/FIN-2017004-012017Q4Point 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-2017003-022017Q3Point 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-2017003-012017Q3Point 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)
05000266/FIN-2017002-012017Q2Point 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.
05000266/FIN-2017001-012017Q1Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 5.5.14, Safety Function Determination Program (SFDP), due to the failure to detect a loss of safety function and ensure appropriate actions were taken during maintenance activities conducted during performance of WO 40513133 for troubleshooting the check source drive mechanism for RE235, control room noble gas monitor, on January 18, 2017. In addition to the troubleshooting activities in WO 40513133, the licensee concurrently performed preventative maintenance on W14A, F16 control 32 room charcoal filter fan, and W13B2, control room recirculation fan. Due to these activities, the licensee implemented procedure NP 10.3.8, Safety Function Determination Program, to ensure that a loss of safety function was detected and the appropriate actions were taken for the equipment out of service associated with the CREFS. Specifically, NP 10.3.8, step 4.2.2 stated, Perform Loss of Safety Function Evaluation. Contrary to NP 10.3.8, step 4.2.2, an adequate loss of safety function evaluation was not performed for the CREFS system based on the equipment that was out of service. As a result of the inadequate loss of safety function evaluation, the licensee did not perform the Required Actions of TS limiting condition for operation (LCO) 3.7.9, Control Room Emergency Filtration System (CREFS), Condition C. The inadequate loss of safety function evaluation was identified when an operator wrote an action request that questioned condition of the CREFS during maintenance activities on January 18, 2017. TS 5.5.14, Safety Function Determination Program, required, in part, that if a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. Contrary to the above, on January 18, 2017, the licensee did not enter the appropriate Conditions and Required Actions of the LCO in which a loss of safety function existed. Specifically, the licensee did not adequately implement procedure NP 10.3.8, step 4.2.2, which resulted in the licensee not performing the Required Actions of TS LCO 3.7.9, Condition C. The licensee entered this issue into the CAP as AR 02183341. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors answered Yes to Question 1 in Exhibit 3, Section C, Control Room Auxiliary, Reactor, or Spent Fuel Pool Building. This resulted in the finding screening as Green.
05000266/FIN-2016004-012016Q4Point 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.
05000266/FIN-2016004-022016Q4Point BeachLicensee-Identified ViolationThe following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. The licensee identified a finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to the failure to properly implement instructions in Work Order (WO) 40461957 for the replacement of the power range nuclear instrument (NI) 1N-43 gain potentiometer vernier dial. Specifically, step 5 of the WO stated, Replace the gain pot vernier with the preset spare. Prevent movement of the potentiometer shaft as much as possible. Contrary to the WO instructions, the technician performing the work believed it was necessary to dial the gain potentiometer to zero before replacing the dial and in doing so caused the 1N-43 NI high flux trip function to become inoperable. This was identified when the control room operators observed the indicated NI power reading for 1N-43 decrease to 82 percent and questioned the technician performing the work about the observed power change. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, on December 2, 2016, the licensee did not accomplish activities affecting quality in accordance with the documented instructions. Specifically, the licensee did not follow step 5 of the work instructions in WO 40461957, causing the NI high flux trip function for 1N43 to become inoperable. The licensee entered this issue into the CAP as AR 02172378. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors answered no to all questions in Exhibit 2, Section C, Reactivity Control Systems. This resulted in the finding screening as Green.
05000266/FIN-2016003-012016Q3Point BeachLicensee-Identified ViolationThe licensee identified a non-cited violation of very low safety significance of 10 CFR 72.150,Instructions, Procedures, and Drawings. Title 10 CFR 72.150, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established PBF5101, Fuel/Insert/Component Movement Authorization, Revision 17, as the implementing procedure for dry fuel storage fuel loading, an activity affecting quality. Procedure PBF5101 contains instructions for fuel handlers to move specific fuel assemblies from specific spent fuel pool locations into specific dry shielded canisters (DSCs) and DSC locations. Contrary to the above, on July 18, 2016, the licensee failed to follow PBF5101. Specifically, the licensee was utilizing a PBF5101 labeled for DSC25 during the loading of DSC24. This resulted in three fuel assemblies being incorrectly loaded into DSC24. The license entered the issue into its corrective action program under AR 02144237, dated July 18, 2016, and initiated actions to perform an apparent causal evaluation. The inspectors identified that DSC24 and DSC25 have identical design characteristics and therefore there was no actual safety significance to this event. Consistent with the guidance in Section 2.2 of the NRC Enforcement Policy, ISFSIs are not subject to the Significance Determination Process and, thus, traditional enforcement will be used for this issue. However, the inspectors determined that the violation significance could be informed by the significance determination process as no similar violations existed in the enforcement policy violations examples. The inspectors determined that the violation could be evaluated using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 3, Barrier Integrity Screening Questions. This resulted in the violation screening as Severity Level IV.
05000266/FIN-2016002-012016Q2Point 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-2016002-042016Q2Point 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-2016002-052016Q2Point 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.
05000266/FIN-2016002-062016Q2Point 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.
05000266/FIN-2016002-022016Q2Point 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-2016002-032016Q2Point 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.
05000301/FIN-2016001-012016Q1Point BeachFailure to Follow Electrical Safety Procedures Results in Plant TransientA finding of very low safety significance was self-revealed for the licensees failure to follow electrical safety procedures when hanging danger tags on electrical components with exposed conductors. Specifically, danger tags were attached directly to the exposed energized portion of switchgear test switches, which exposed employees to an electrical hazard and contributed to the lockout of the 2X-01 main transformers and the subsequent Unit 2 plant transient. The licensees corrective actions included a change to tagging procedures to include specific direction for tagging knife switches. The proposed changes included a prohibition for hanging tags on metal parts of the switches, and installing robust operational barriers using tags plus devices when danger tags are to be utilized. The inspectors determined that the finding was more than minor because it was associated with the human performance attribute of the initiating events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to use insulated tools on exposed electrical equipment greater than 50 volts presented an electrical injury hazard and actually resulted in a plant transient for Unit 2, which included lifting of a pressurizer power-operated relief valve (PORV), loss of forced reactor coolant system (RCS) flow, and actuation of the auxiliary feedwater (AFW) system. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, because Unit 2 was in mode 3 at the time of the event. Additionally, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012 applied. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Transient Initiators screening question. This finding has a cross-cutting aspect of Resources (H.1), in the area of Human Performance for failing to ensure that personnel, equipment procedures and other resources were available and adequate to support nuclear safety. Specifically, the licensee failed to ensure that employees had all necessary tools, direction, and supervision to support successful work performance.
05000282/FIN-2015004-032015Q4Prairie IslandLicensee-Identified ViolationTitle 10 CFR 50.48(b)(2) requires, in part, that all nuclear power plants licensed to operate before January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections III.G, III.J, and III.O. Appendix R, Section III.G.3 of 10 CFR Part 50, requires, in part, that alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section. In addition, fire detection and a fixed fire suppression system shall be installed in the area, room, or zone under consideration. Contrary to the above, on April 19, 2015, the licensee failed to ensure that alternative or dedicated shutdown capability and its associated circuits were independent of cables in the area. Specifically, procedure F5 Appendix B, Control Room Evacuation (Fire), Revision 31, did not contain actions to isolate the RCP breaker circuits to prevent restarting due to a fire induced loss of remote trip and loss of RCP seal cooling water that could lead to an increased rate of seal degradation and a small break loss of coolant accident. These actions were required to achieve and maintain safe shutdown in the event of a fire that resulted in functional loss and/or evacuation of the control/relay and cable spreading rooms. Section 9.1 of the NRC Enforcement Policy allows the NRC to exercise enforcement discretion for certain fire protection related non compliances identified as a result of a licensees transition to the new risk informed, performance based fire protection approach included in 10 CFR 50.48(c), and for certain existing non compliances that reasonably may be resolved by compliance with 10 CFR 50.48(c) as long as certain criteria are met. This risk informed, performance based approach is referred to as NFPA 805, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. The licensee is in transition to NFPA 805 and therefore the licensee-identified violation was evaluated in accordance with the criteria established by Section 9.1(a) of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) for a licensee in NFPA 805 transition. The inspectors determined that for this violation: (1) the licensee would have identified the violation during the scheduled transition to 10 CFR 50.48(c); (2) the licensee had established adequate compensatory measures within a reasonable time frame following identification and would correct the violation as a result of completing the NFPA 805 transition; (3) the violation was not likely to have been previously identified by routine licensee efforts; and (4) the violation was not willful. The finding also met additional criteria established in section 12.01.b of IMC 0305, Operating Assessment Program. In addition, in order for the NRC to consider granting enforcement discretion the violation must not be associated with a finding of high safety significance (i.e., Red). The licensee performed risk evaluation V.SPA.15.012, Revision 3, dated December 18, 2015, and determined that this issue was not associated with a finding of high safety significance. A region III senior reactor analyst (SRA) reviewed the evaluation and concluded that the result was reasonable and that the finding was less than Red and eligible for enforcement discretion. The dominant core damage sequence from the licensees evaluation involved an electrical cabinet fire in the relay room involving the cables that could cause spurious operation of the RCPs and that would lead to alternate shutdown. The licensee identified several conservative assumptions in the analysis. The SRA agreed that some were conservative, notably that any fire affecting the cables in the relay room that could cause a spurious start of an RCP would also result in a loss of all seal cooling due to fire damage. The SRA used IMC 0609, Appendix F, Fire Protection Significance Determination Process, to review the results of the licensees evaluation. The relay room is similar to a cable spreading room with electrical cabinets. The fire frequency for this room in Appendix F is 6E3/yr. The probability of non-suppression was estimated to be 2E2 and the spurious operation probability was assumed to be 0.6. The product of these values (7.2E5/yr) represents a bounding relay room fire scenario delta core damage frequency (CDF) for this finding. Since the bounding result is consistent with the licensees conclusion, the SRA determined that the delta core damage frequency for the finding was less than 1E4/yr, which is less than Red. In addition, the licensee entered this issue into their corrective action program as CAP 01475242. As a result, the inspectors concluded that the violation met all four criteria established by Section 9.1(a) and that the NRC was exercising enforcement discretion to not cite this violation in accordance with the Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues.
05000306/FIN-2015004-012015Q4Prairie IslandFailure to Meet ANSI N14.6 Section 5.3.1 RequirementsThe inspectors identified a finding of very low safety significance (Green), and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to incorporate the American National Standards Institute (ANSI) N14.61978, Section 5.3.1 required testing frequency for the reactor vessel head and reactor vessel internals lifting devices into the controlling preventive maintenance procedure. Compliance with the ANSI standard was documented in the Safety Evaluation Report (SER) for the licensees control of heavy loads. The licensee documented the issue in the corrective action program (CAP) as CAP 01497779 and performed testing on the reactor vessel head and internals lifting devices during the outage. The inspectors determined the licensees failure to comply with ANSI N14.61978, Section 5.3.1, for the continued use testing of special lifting devices was a performance deficiency (PD). The PD was determined to be more-than-minor and a finding because the PD was associated with the Initiating Events Cornerstone attribute of design control, and adversely affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, compliance with ANSI N14.61978, Section 5.3.1 ensured safe load handling of heavy loads over the reactor core, and/or over safety-related systems through established testing for the continued functionality of the special lifting devices. The failure to perform the required frequency of testing on special lifting devices could increase the likelihood of a load drop and could decrease the load handling reliability of the lifting device if the device were returned to service with potentially unacceptable flaws. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase I - Initial Screening and Characterization of Findings, Table 3. Since the finding was associated with shutdown conditions, the inspectors used Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined that none of the conditions constituting a loss of control were met, as described in Appendix G, Attachment 1, Phase I Operational Checklists for Both PWRs (Pressurized Water Reactors) and BWRs (Boiling Water Reactors), for this finding, and neither a Phase II nor a Phase III analysis was required. Therefore, the inspectors determined that this finding was of very low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Resources, for the licensees failure to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, the licensee staff evaluated NRC Information Notice (IN) 201412, Crane and Heavy Lift Issues Identified during NRC Inspections, in corrective action program (CAP) document 01457469. However, in CAP 01457469, the licensee concluded that issues identified in IN 201412 related to other licensees not performing testing in accordance with ANSI N14.6 requirements were not applicable to the licensee at the Prairie Island Nuclear Generating Plant. Therefore, the inspectors determined that there was a recent missed opportunity for the licensee to have reasonably identified that the current preventive maintenance procedure for special lifting devices was not in accordance with the ANSI N14.61978 requirements, as referenced in the SER.
05000282/FIN-2015004-022015Q4Prairie IslandFailure to Adequately Calibrate Liquid Effluent MonitorsThe inspectors identified a finding of very low safety significance (Green) and associated NCV of TS 5.5.1.a for the failure to comply with the Offsite Dose Calculation Manual (ODCM) for not using calibration sources that were traceable to the National Institute of Standards and Technology (NIST) or equivalent during the calibration of station effluent monitors. The licensee entered the issues into the CAP as CAPs 01490581 and 01500149. Immediate corrective actions included the re-calibration of impacted monitors and the performance of an extent of condition evaluation for other radiation monitor calibrations. The PD was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the plant facilities/equipment and instrumentation attribute of Public Radiation Safety and it adversely impacted the cornerstone objective of ensuring adequate protection of public health and safety due to failure to properly calibrate certain effluent monitors. Subsequent calibrations of the monitors determined that the monitor efficiency was previously overstated. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, but did not identify any similar examples. The finding was assessed using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated, February 12, 2008, and determined to be of very low safety significance (Green), because it was associated with the effluent release program but was not a failure to implement an effluent program, public dose did not exceed Appendix I criteria, and the limits in Title 10 CFR 20.1301(e) were not exceeded. A cross-cutting aspect was not assigned as this issue occurred numerous years ago. The station has since performed monitor calibrations with radioactive sources with known quality.
05000266/FIN-2015004-032015Q4Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 5.4.1, Procedures for the failure to maintain the emergency operating procedures (EOPs). The licensees TS 5.4.1 required, in part, that written procedures shall be maintained including the EOPs required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. During design reviews, the licensee discovered that following a 2012 calculation update, the licensee inconsistently applied pre and post-modification uncertainties that had resulted from a 2010 modification associated with the sensitivity and calibration of both units Subcooling Margin Monitors. Ultimately the calculative errors resulted in 19 EOP Subcooling setpoints being incorrectly calculated. These Subcooling setpoints are used throughout the licensees EOPs network to provide operators with discrete indications for key EOP decision making. Contrary to the above, from April 12, 2012 through November 5, 2015, the licensees EOP network of procedures for both Unit 1 and 2, contained the incorrect setpoints for decision points with respect to subcooling. The licensee entered this issue into the CAP as AR 02089011 and AR 02099152. The inspectors consulted the Region III Senior Reactor Analysts and determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 1, 2012 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors determined that the issue was a design or qualification deficiency confirmed not to result in a loss of operability; therefore, answered yes to question A.1 in Exhibit 2, Section A, Mitigating SSCs and Functionality. This resulted in the finding screening as Green.
05000301/FIN-2015004-022015Q4Point BeachInadequate Evaluation of Non-Conforming Auxiliary Feedwater System Pipe DefectsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to maintain a Unit 2 auxiliary feedwater system (AFW) pipe segment containing linear defects in accordance with the design and material specifications. As a corrective action, the licensee performed light filing to remove the defects from this pipe segment. The licensee entered the failure to maintain the AFW pipe segment in accordance with the design into the corrective action program (CAP) as action request (AR) 02084077, and was evaluating additional corrective actions. This finding was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to maintain the Unit 2 AFW pipe segment containing linear defects in accordance with the design and material specifications could result in an increase in the possibility of pipe leakage or failure. In addition, the failure to maintain the AFW pipe segment containing linear defects in accordance with the design and material specification adversely affected the Mitigating System Cornerstone attribute of Equipment Performance because it could result in failure of AFW piping which would reduce the availability and reliability of the this mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors answered Yes to screening question A.1 of Exhibit 2. Although this finding adversely affected the design or qualification of the AFW pipe segments, the finding screened as very low safety significance (Green), because it did not result in the loss of operability or functionality of the affected pipe segment. This finding has a cross-cutting aspect in the Teamwork (H.4) component of the human performance cross-cutting area. Specifically, the licensees Projects Team responsible for the AFW modifications did not effectively communicate and coordinate with the licensees Programs Engineering Group for resolution of the AFW pipe nonconforming conditions to ensure nuclear safety was maintained.
05000266/FIN-2015004-012015Q4Point BeachFailure to Follow Fire Protection Program Requirements for Care, Use and Maintenance of Fire HoseThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of license condition 4.F for the licensees failure to have procedures or instructions to prevent firefighting booster hoses from being kinked and/or twisted on hose reels. Specifically, booster hoses were installed on hose reels in both units containments and in the turbine building (TB), which were twisted and kinked. The licensees corrective actions included rewinding hoses in the Unit 2 containment, four hoses in the TB, and creating compensatory measures for hose reels for the Unit 1 containment. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee failed to ensure that activities such as inspection, testing, and maintenance of fire protection systems were prescribed and accomplished in accordance with documented instructions, procedures, and drawings. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1A, because the inspectors determined that the impact of a fire would be limited to one train/division of equipment for the affected fire areas and at least one credited safe shutdown path would be unaffected. This finding has a cross-cutting aspect of Training (H.9), in the area of human performance, because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce, and instill nuclear safety values. Specifically, the inspectors determined that operations personnel were not adequately trained to recognize deficiencies associated with firefighting equipment standards, such as kinked and twisted hoses on hose reels, and subsequently failed to initiate actions to remedy such conditions.
05000266/FIN-2015003-012015Q3Point BeachIncomplete Functionality Assessment for Flooding in the Diesel Generator BuildingThe inspectors identified a finding of very low safety significance for the licensees failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 19. Specifically, when the licensee identified that internal flood sources in the diesel generator building (DGB) were larger than the drain capacity, they failed to identify all affected structures, systems, and components (SSCs). The DGB contains predominately Train B emergency power systems; however, the fuel oil transfer pumps for the Train A emergency diesel generators are located in the southeast corner of the building. The licensee failed to assess the effects of flooding on the Train A fuel oil transfer pumps. The licensees corrective actions included the creation of an adverse condition monitoring plan, which implemented an hourly flood watch in the DGB when the fire pump was manually started. The inspectors determined that the finding was more than minor, because if left uncorrected, it would potentially result in a more safety significant issue. Specifically, the failure to evaluate the effects of flooding on all SSCs resulted in inadequate compensatory measures. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. For the time period in question, May 17, 2015 to September 17, 2015, the inspectors reviewed the security door card reader reports and starting sump levels for the DGB and found that during times when the fire pumps were running, station personnel had toured the DGB at a frequency that would have identified flooding conditions before a loss of system function. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Evaluation (P.2), in the area of Problem Identification and Resolution (PI&R), for failing to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000266/FIN-2015003-022015Q3Point BeachPotential Failure of Multiple Safety-Related Trains During Flooding EventsThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensees failure to ensure that a non-Category I (seismic) component failure, that results in flooding, would not adversely affect safety-related equipment needed to get the plant to safe shutdown (SSD) or to limit the consequences of an accident. Specifically, the design of Point Beach did not ensure that the Residual Heat Removal (RHR) pumps would be protected from all credible non-Category I (seismic) system failures. The licensees corrective actions included an extensive internal flooding design review, which will result in an updated Final Safety Analysis Report (FSAR) with a more detailed description of the stations flooding licensing basis; modifications to multiple flood barriers to bring them into compliance with the licensees flooding licensing basis; installation of additional flood level alarms where necessary, and evaluation or modification of service water (SW) piping to properly qualify it as seismic. The inspectors determined that the finding was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design resulted in an unanalyzed condition and loss of safety function of the RHR system while the plants were in Modes 4, 5, and 6, when relying on the RHR system for decay heat removal. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors answered yes to question 2 of the screening questions because the finding represented a loss of safety function. Thus the inspectors consulted the Region III Senior Risk Analysts (SRAs) who performed a detailed risk evaluation and determined that the finding was of very low safety significance (Green). The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015003-042015Q3Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 3.8.9; Distribution SystemsOperating, Condition A, which required the licensee to immediately declare associated supported features inoperable for the 4.16 kV safeguards busses. Failure to implement this action subsequently required the licensee to place both units in mode 5 within 36 hours. Contrary to the above, the licensee discovered that numerous occasions existed over the past three years where safetyrelated 4.16kV switchgear associated with B Train EDGs was inoperable due to the inoperability of the W-185A and W-185B, 1A-06 and 2A-06 Switchgear room fans, which were required support systems for the EDGs and associated switchgear. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, and determined that the finding required a detailed risk evaluation which was performed by Region III SRAs. The SRAs gathered data from licensee GOTHIC model calculations, licensee engineering evaluations associated with the POR of the condition and the NRCs Standard Plant Analysis Risk model. Based on the SSCs being available for their respective 24-hour mission time(s), the SRAs determined that the increase in CDF for this issue was negligible and the delta risk is of very low safety significance (i.e., Green). The licensee reported this condition in LER 2015-004-00, which was closed in Section 4OA3 of this report. The licensees corrective actions included improving administrative and procedural controls for removing these fans from service and used lessons learned from this condition to implement corrective actions to improve procedural guidance for similar activities where ventilation systems may cause support system inoperabilities.
05000266/FIN-2015007-012015Q3Point BeachFailure to Demonstrate the Functionality of a Credited Safe Shutdown ComponentThe inspectors identified a finding of very low safety significance and an associated NCV of license condition 4.F for the licensees failure to demonstrate the capabilities of systems needed to perform a design function for Appendix R cold shutdown. Specifically, none of the licensees tests, inspections, or maintenance activities demonstrated that CC-722A, the component cooling water pump suction cross-tie valve, was capable of being opened as required in AOP-10B, Safe to Cold Shutdown in Local Control. The licensee corrective actions included entering the issue into their CA program, declaring CC-722A non-functional, and commencing four-hour fire rounds. The inspectors determined the finding to be more than minor because the failure to demonstrate the capabilities of systems needed to perform a design function for Appendix R safe shutdown was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage). 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 affected the ability to reach and maintain safe shutdown, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1A, because the inspectors determined that the finding would not prevent the reactor from reaching and maintaining hot shutdown. This finding has a cross-cutting aspect of Resolution (P.3), in the area of problem identification and resolution, because the licensee did not take effective corrective actions to address the issue in a timely manner. Specifically, in 2007, the licensee identified that they had not been testing the valve as specified in their Fire Protection Evaluation Report and as of July 2015 had still not corrected it.
05000266/FIN-2015003-032015Q3Point BeachFailure to Perform a Written Safety Evaluation for FSAR ChangesThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance for the licensees failure to perform a safety evaluation to demonstrate that the removal of statements from the FSAR did not require a license amendment. Specifically, the licensee failed to perform a safety evaluation to determine whether removing an FSAR statement, which defined the RHR pump cubicle design flood height as seven feet, could be performed without a license amendment. The licensee entered the deficiency in their CAP as Action Request (AR) 02069425 by which the licensee intends on re-evaluating the 1996 FSAR change. The inspectors determined that the finding was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, inappropriately removing the information from the FSAR allowed the licensee to decrease the design basis flood protection height of the RHR compartments and significantly reduced the available time to isolate the leaking RHR pump seal. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000301/FIN-2015002-012015Q2Point BeachAuxiliary Feedwater Pump Trip Time Delay Relay Installed Past Evaluated Service LifeDuring the performance of the semi-annual full system lineup inspection sample, the inspectors identified an Unresolved Item (URI) associated with the Unit 2 turbine-driven AFW pump low suction pressure trip time delay relay potentially being past its design service life. During the inspectors review of corrective actions associated with an AFW full system alignment, the inspectors identified a URI associated with a time delay relay that potentially exceeded its design and evaluated maximum service life. The inspectors determined that this was an issue of concern in which more information was needed to determine whether a performance deficiency exists. Specifically, after the inspectors pointed out that the time delay relays were installed beyond the licensees documented evaluation for service life extension, the licensee further extended the relays service life based on historical plant testing. The inspectors needed to evaluate whether the information provided by the licensee was sufficient to provide a technically defensible basis for the additional service life extension. The issue is unresolved pending further agency review of the licensees evaluation (URI 05000301/201500201, Auxiliary Feedwater Pump Trip Time Delay Relay Installed Past Evaluated Service Life).
05000266/FIN-2015002-042015Q2Point BeachLicensee-Identified ViolationSection (b) of TS 5.7.1 requires, in part, that access toand activities ina high-radiation area be controlled by a radiation work permit or equivalent. Section (e) of TS 5.7.1 requires, in part, that entry into HRAs be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Contrary to the above, on April 14, 2015, an individual entered an HRA without being on a radiation work permit that allowed for HRA entry and was not made knowledgeable of the dose rates in the HRA. The licensee entered this issue into the CAP as AR 0204280. This violation is considered to be of very low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls; (2) there were no overexposures; (3) there was not a substantial potential for overexposures; and (4) the ability to assess dose was not compromised.
05000266/FIN-2015002-032015Q2Point BeachInadequate Measures to Control Spare Firing Card AssembliesA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, was self-revealed for the licensees failure to establish measures to ensure non-conforming tantalum electrolytic capacitors that were part of an assembly and that were beyond their recommended shelf-life would not be installed in safety-related equipment in the plant. The licensees corrective actions included repair of the D-107 battery charger, and updating maintenance and procurement requirements with component shelf-life information. The inspectors determined the finding was more than minor since the failure to ensure the quality of spare parts, if left uncorrected, could lead to a more significant safety concern. Specifically, the failure to control circuit boards which contained tantalum electrolytic capacitors that were beyond their shelf-life was self-revealed when the D-107 safety-related battery charger failed three days after the circuit boards were installed. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered "No" to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Change Management (H.3), in the area of Human Performance, for the licensees failure to use a systematic process for implementing changes so that nuclear safety remained the overriding priority.
05000266/FIN-2015002-022015Q2Point BeachFailure to Control Transient Combustibles During Service Water Pumphouse MaintenanceA finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1.h was identified by the inspectors for the failure to control transient combustible material in accordance with the licensees Fire Protection Program requirements. Specifically, the licensee installed a power cord in the north side of the service water pump room that was subsequently extended also into the south side of the service water pump room across a transient combustible exclusion boundary with no prior evaluation. The licensees corrective actions included immediately removing the power cord from the fire exclusion zone and standing-down the work group for a brief of the event and a review of the requirements for transient combustibles. The inspectors determined the finding was more than minor because the failure to identify the transient combustibles was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.B, because the inspectors assigned a Low degradation rating to the single cable that crossed through the exclusion zone. This finding has a cross-cutting aspect of Field Presence (H.2), in the area of human performance, because the licensees leadership did not ensure that oversight of work activities, including contractors and supplemental personnel was provided such that nuclear safety was supported.
05000266/FIN-2015001-042015Q1Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, due to the licensees failure to assure that all testing required to demonstrate that SSCs will perform satisfactorily in service is identified and performed in accordance with written test procedures. Specifically, the licensee unacceptably preconditioned the D107 battery charger by lifting and reseating the wire harness connector to the current limiter card prior to conducting required surveillance testing on the battery charger. Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, states, in part, that a test program shall be established to assure that all testing required to demonstrate that SSCs will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to this, from November 15, 2008 through on August 14, 2014, the licensee failed assure that all testing required to demonstrate that SSCs will perform satisfactorily in service is identified and performed in accordance with written test procedures. Specifically, procedure RMP 93596A, D105 Station Battery, D107 Battery Charger Maintenance and Surveillances, Revisions 08, which the licensee used to perform its 18-month TS surveillance, improperly sequenced the step to lift and reseat the current sensing and limiting card edge connector prior to performing the surveillance test. The current limiting function for the charger is necessary to prevent the charger input current from exceeding the supply breaker current setting and tripping the battery charger when its needed for accident mitigation. The licensee entered this issue into the CAP as AR 01993719. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors answered Yes to question number 3 and concluded a detailed risk evaluation was necessary. To evaluate this finding, the Senior Reactor Analysts assumed that the exposure time was one-year which is the maximum allowed by the SDP. The Point Beach Standardized Plant Analysis Risk model version 8.22 and Systems Analysis Programs for Hands-on Integrated Reliability Evaluations version 8.1.2 software was used to obtain a delta core damage frequency (CDF Internal) for internal events of 6.22E7/yr. The dominant core damage sequences involve a loss-of-offsite-power initiating event with a loss of reactor coolant pump seal cooling, a failure of rapid secondary depressurization, failure of the reactor coolant pump seals, failure of RCS cooldown (primary and secondary), and failure of high pressure recirculation. Since the total estimated change in core damage frequency was greater than 1.0E-7/yr, an evaluation was performed for external event delta risk contributions. The evaluation found that external event risk contribution was 3.09E7/yr, giving a total CDF of CDF Total = 6.22E7/yr + 3.09E7/yr = 9.31E7/yr. Large Early Release Frequency - Sequences important to Large Early Release Frequency include steam generator tube rupture events and inter-system loss-of-coolant-accident events. These were not the dominant core damage sequences for this finding. Based on the Detailed Risk Evaluation, the inspectors determined that the finding was of very low safety significance (Green).
05000266/FIN-2015001-032015Q1Point BeachLicensee-Identified ViolationThe licensee identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that design basis information was correctly translated into calibration procedures for both units over temperature delta temperature (OTDT) reactor trip instrumentation. On January 29, 2015, the licensee discovered that lead time-constants associated with the OTDT reactor trip function for all four channels on both units 1 and 2 were incorrect, and that the calibration procedures, 1/2ICP 04.001C, for this setting also contained the incorrect values. The licensee identified this issue in AR 02021827, which described the condition and stated that the incorrect lead time-constant values were not updated following EPUs for both units. Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from June 15, 2012 to January 29, 2015, the licensees calibration procedures 1/2ICP 04.001C did not contain the correct lead time-constant setting for the proper calibration for all channels of the TS required OTDT reactor trip instrumentation. The licensee entered this issue into the CAP as AR 02021827, realigned all eight of the affected channels (four on each unit), and initiated procedure changes to incorporate the correct values for these settings. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 1, 2012 and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated July 1, 2012. Because the licensee determined that operability was maintained, the inspectors answered "No" to all questions in Exhibit 2, Section A, Mitigating Structures, Systems, Components (SSCs), and Functionality. Therefore, the finding screened as very low safety significance (Green).
05000266/FIN-2015001-012015Q1Point BeachFailure to Process Vendor Technical InformationA finding of very low safety significance was identified by the inspectors for the failure to follow site procedure NP 7.2.13, Processing of Vendor Technical Information. Specifically, the licensee failed to process a vendor technical bulletin in accordance with NP 7.2.13. The technical bulletin provided relevant information related to the inspection, adjustment, and replacement of an electrical connector located in some of the licensees safety-related battery chargers. Procedure NP 7.2.13 ensured that relevant vendor correspondence received by the licensee was analyzed to identify specific actions needed to operate and maintain the plant safely. Licensee corrective actions included conducting a condition evaluation, which concluded that a lack of understanding of current vendor technical document process expectations may exist within key departments. The licensee plans to perform information sharing to increase awareness of expectations for processing vendor documents. The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more safety significant concern. Specifically, if a degraded connector was not identified and corrected during safety-related battery charger maintenance, the charger may fail to limit current and open the supply breaker to the battery charger. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Operating Experience (P.5), in the area of Problem Identification and Resolution, for the failure to systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner.
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