Licensee-identified

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 SiteQuarterSignificanceCornerstoneViolation ofDescriptionSystem
05000528/FIN-2018008-03Palo Verde2018Q3Severity level IVMitigating Systems10 CFR 50.73
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
Technical Specification
This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: 10 CFR 50.73(a)(2)(i)(B) requires, in part, that the holder of an operating license shall submit an licensee event report within 60 days of discovery of the event, which includes any operation or condition which was prohibited by technical specifications. Contrary to the above, the licensee failed to submit a licensee event report within 60 days of April 23, 2016, after discovering that the Unit 1 channel C excore was in a condition which was prohibited by technical specifications. The detector was found in a configuration without o-rings at two electrical connection interfaces. Condition Report 16-06735 documented the non-conforming condition, but was closed without performing a reportability review. Significance/Severity Level: This violation was considered as traditional enforcement because the failure to notify the NRC had the potential for impacting the NRCs ability to perform its regulatory function. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the failure to report the condition prohibited by technical specifications was determined to be a Severity Level IV violation. Corrective Action Reference(s): Condition Report 18-02569
05000266/FIN-2018003-02Point Beach2018Q3GreenInitiating Events
Mitigating Systems
10 CFR 50.48
License Condition - Fire Protection
Violation: 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).
05000255/FIN-2018411-01Palisades2018Q3GreenPhysical Protection10 CFR 73
05000390/FIN-2018003-05Watts Bar2018Q3GreenMitigating SystemsLicense Condition - Fire ProtectionThis violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Watts Bar Nuclear Plant (WBN) Unit 1 Operating License Number NPF-90, Condition 2.F, requires, in part, that TVA shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Fire Protection Report for the facility, as approved in Appendix FF Section 3.5 of Supplement 18 and Supplement 29 of the SER (NUREG-0847). The WBN Fire Protection Report was developed for WBN to ensure compliance with the requirements of this license condition. Fire Protection Report, Part II, is the Fire Protection Plan. The Fire Protection Plan, Section 14, Fire Protection Systems and Features Operating Requirements (ORs), Subsection 14.10, Fire Safe Shutdown Equipment, paragraph 14.10.4, requires a fire watch to be established in auxiliary building room 757-A10 within one hour of closing pressurizer block valve 1-FCV-68-332-B. Contrary to the above, on July 19, 2018, the licensee failed to establish a fire watch in auxiliary building room 757-A10 within one hour of closing pressurizer block valve 1-FCV-68-332-B.
05000263/FIN-2018003-01Monticello2018Q3GreenInitiating Events
Mitigating Systems
10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power PlantsThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants; which requires, in part, that equipment qualified by test must be preconditioned by natural or artificial aging to its end of life or a shorter designated life considering all significant types of degradation which can have an effect on equipment function. Contrary to the above, on June 2, 2018, the licensee determined that EQ evaluation 608000000032, of MO2034, MO2035, MO2075, and MO2076 (HPCI and RCIC Steam Line Isolation Valves) internal actuator cables, failed to consider the temperature rise due to the high temperature process fluid in the vicinity of the affected components when aging (preconditioning) them and the unaccounted temperature rise shortened the life of some components to the point that they were no longer EQ qualified to the end of planned life. The unaccounted for process fluid temperature increases were verified by the licensee when thermography of the associated valves was performed. The licensee performed a prompt operability determination, entered the issue into the corrective action program (CAP) as CAP 501000012766 and performed a thermal life analysis engineering evaluation. Long-term corrective actions include replacement of the internal actuator cables during the next refueling outage. 10 Significance/Severity Level: This finding was more than minor because the performance deficiency was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, HPCI and RCIC Steam Line Isolation Valves are designed to provide reactor coolant pressure boundary, required for a safe reactor shutdown following a Design Basis Accident or transient. The finding was of very low safety significance (Green) because it was a design or qualification deficiency, did not involve an actual loss of safety system, did not represent actual loss of a safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hrs. Corrective Action Reference: 501000012766
05000373/FIN-2018412-01LaSalle2018Q3GreenPhysical Protection10 CFR 73
05000266/FIN-2018010-01Point Beach2018Q3GreenNo Cornerstone10 CFR 50 Appendix B Criterion XII, Control of Measuring and Test EquipmentThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Polic Violation: Title 10 CFR 50, Part B, Criterion XII requires that measures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.Contrary to the above, the licensee failed to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality were properly controlled. Specifically, the licensee did not include all M&TE devices in their control tracking program, which could result in instruments not being evaluated if associated M&TE fails its post-calibration.Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The inspectors assessed the significance of the finding using SDP Appendix A and concluded the violation was of very low safety significance (Green).
05000315/FIN-2018003-02Cook2018Q3No CornerstoneCertificate of Compliance (CoC) 1014, Amendment 9, Design Feature, Section 3.9, Environmental Temperature Requirements, requires building ambient temperatures be less than 110 degrees Fahrenheit during canister processing based upon the thermal analysis in the Holtec HI-STORM Final Safety Analysis Report, Revision 13. The thermal model documented in the Final Safety Analysis Report, Revision 13, Section 4.5.1, HI-TRAC Thermal Model, states that heat is passively rejected to the ambient from the outer surface of the HI-TRAC transfer cask by natural convection and thermal radiation. However, at D.C. Cook, the licensee uses additional shielding materials for as low as reasonably achievable purposes that are in contact with and in the general area of the HI-TRAC. The licensee requested Holtec to perform a site-specific thermal analysis, HI2177676, Thermal Evaluation of Shielding Package around the HI-TRAC at DC Cook, to include the shielding material in the thermal model. The analysis contained inputs that were different than the design basis calculation inputs, which were previously incorporated into Design Feature Section 3.9 and Approved Contents Section 2.4. The licensee performed a 10 CFR 72.48 Screening and Evaluation 2018013902, which concluded that shielding could be used without prior NRC approval and subsequently issued 212CR0017, which revised the 72.212 Report. The licensee implemented administrative controls on building temperature and fuel assembly heat load limits based upon the site specific thermal analysis. This unresolved item is being opened to determine if: A) the licensee is in compliance with Design Feature, Section 3.9, Environmental Temperature Requirements; B) the Design Feature Section 3.9 and Approved Contents Section 2.4 are non-conservative at D.C. Cook; and C) the licensee is in compliance with 10 CFR 72.48. Planned Closure Actions: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Materials Safety and Safeguards. Corrective Action References: AR 20184056; AR 20186342; AR 20186642
05000282/FIN-2018411-01Prairie Island2018Q3GreenPhysical Protection10 CFR 73
05000390/FIN-2018003-06Watts Bar2018Q3Severity level IVMitigating SystemsTechnical SpecificationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Watts Bar Unit 1 TS 3.8.1, AC Sources - Operating, Condition A, requires, in part, that an inoperable required offsite circuit be restored to operable status within 72 hours. Contrary to the requirements of Technical Specification 3.8.1, a required offsite circuit was determined to be inoperable from May 27, 2017, to June 2, 2017.
05000458/FIN-2018003-01River Bend2018Q3Mitigating Systems10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
10 CFR 50 Appendix B Criterion III, Design Control
This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the design basis for those structures, systems, and components to which Appendix B applies is correctly translated into specifications, drawings, procedures, and instructions. The design basis for the control building air conditioning system, as specified in the updated safety analysis report, requires that the system be capable of performing its safety function in the event of a single failure in any component. Contrary to the above, the licensee failed to assure that the design basis was correctly translated into specifications for the control building air conditioning system. Specifically, while reviewing the control logic for the control building air conditioning system, the licensee discovered that the control logic was designed such that a single failure in a component in the control logic could have prevented the system from performing its specified safety function.
05000482/FIN-2018010-05Wolf Creek2018Q3GreenMitigating Systems10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
10 CFR 50 Appendix B Criterion XVI, Corrective Action
This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected.Contrary to the above, prior to 2015, the licensee failed to promptly identify and correct a repetitive deficiency or non-conformance. Specifically, the licensee had identified a leaking flange on the residual heat removal heat exchanger since 1997. Prior to 1997 a different data base had been used to record boric acid leakage, and the data was not available during the inspection.Over the years since plant startup, the licensee had been diligent in completing boric acid evaluations on the leaking residual heat removal heat exchanger flange, indicating minimal wastage of the flange closure studs and nuts that had been subjected to boric acid. Corrective actions included cleaning up the boric acid leakage, and checking or re-torqueing the closure nuts. These measures did not correct the problem of the leaking heat exchanger flange. In 2015 the licensee completed an in-depth engineering evaluation of the leaking flange, including discussions with the heat exchanger manufacturer. New corrective measures included changing the torque values on the closure studs and nuts. The licensee is still evaluating the results of the corrective actions taken to preclude further leakage.Residual Heat Removal
05000272/FIN-2018003-03Salem2018Q3GreenMitigating Systems10 CFR 50 Appendix B Criterion V, Instructions, Procedures, and DrawingsThis violation of very low safety significant was identified by PSEG, has been entered into PSEGs CAP, and is being treated as a Green NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires activities affecting quality shall be prescribed by procedures, and shall be accomplished in accordance with these procedures. PSEG procedure MA-AA-716-011, Work Execution and Closeout, Revision 17, step 4.13.5, required order operations to be completed after the preventive maintenance WO was taken Technically Complete, or TECOd. Contrary to the above, preventive maintenance WOs 30319825 and 30320738 were TECOd by mechanical maintenance, on March 2 and April 9, 2018, respectively, without completing all of the WO operations. Specifically, maintenance technicians performed the monthly thermography on the 22 chiller evaporator divider plate gasket and took the preventive maintenance work order TECO and did not perform MA-AA-716-011, step 4.13.5to complete operation 0020 by notifying engineering that the thermography results were available for review. Consequently, leakage past the divider plate gasket went undetected from March 2 to April 30, 2018, until quarterly compressor thermography detected crankcase temperature above the action level on April 30, 2018. Maintenance immediately notified Operations of the elevated compressor temperature, and the 22 chiller was declared inoperable and removed from service emergently on April 30, 2018. Subsequent disassembly and inspection revealed internal compressor damage and pieces of the evaporator divider plate gasket in the compressor filter housing. PSEG replaced the compressor and restored the 22 chiller to OPERABLE on May 4, 2018
05000456/FIN-2018411-01Braidwood2018Q3GreenPhysical Protection10 CFR 73
05000390/FIN-2018003-07Watts Bar2018Q3GreenMitigating SystemsTechnical SpecificationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Watts Bar Unit 1 TS LCO 3.8.7, Inverters-Operating, requires that two inverters in each of the four channels shall be operable. Contrary to the above, the licensee failed to ensure that two inverters in each of the four channels were operable. Specifically, from April 9, 2017 to January 10, 2018 inverter 1-II was inoperable due to an unqualified class 1E capacitor associated with the inverter.
05000348/FIN-2018003-02Farley2018Q3GreenNo Cornerstone10 CFR 50.48
License Condition - Fire Protection
This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Farley Unit 1 Operating License Condition 2.C.(4) and Unit 2 Operating License Condition 2.C.(6), Fire Protection, required in part that Plant Farley shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(c) and NFPA 805. NFPA 805 section 3.2.3 stated, in part, procedures to accomplish compensatory actions implemented when fire protection systems and other systems credited by the fire protection program and this standard cannot perform their intended function shall be established. Licensee procedure FNP-0-SOP-0.4, Fire Protection Operability and LCO Requirements section 4.0 establishes compensatory action when fire protection systems and other systems credited by the fire protection program cannot perform their intended functions. Contrary to the above, since January 16, 2018 through August 28, 2018, the licensee failed to establish compensatory measures (fire watches) as required by licensee procedure FNP-0-SOP-0.4 on thirteen occasions. The cause of the fire watch discrepancies were mainly because Farley Operations staff lacked an adequate understanding and ownership of the fire watch implementation process.
05000275/FIN-2018404-03Diablo Canyon2018Q3Severity level IVPhysical Protection10 CFR 73
05000266/FIN-2018003-03Point Beach2018Q3GreenInitiating Events
Mitigating Systems
10 CFR 50.48
License Condition - Fire Protection
Violation: 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).
05000424/FIN-2018003-01Vogtle2018Q3GreenEmergency Preparedness10 CFR 50.47, Emergency Plans
10 CFR 50.54
10 CFR 50.47(b)(4)
10 CFR 50.54(q)
This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 CFR Part 50.54(q)(2), required, in part, the licensee shall follow and maintain the effectiveness of its emergency plan that meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4), required, in part, a standard emergency classification and action level scheme, the bases of which include facility and system effluent parameters, is in use by the nuclear facility licensee. Contrary to the above, from January 30, 2018 to July 20, 2018, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, Units 1 and 2 procedure 19200, F-O Critical Safety Function Status Tree, version 1.0, specified over-conservative reactor coolant system (RCS) temperature values for determining a critical safety function RED Path on RCS Integrity used to evaluate emergency classification FA1 (Alert), potential loss of RCS barrier, in response to a rapid RCS cooldown event.Reactor Coolant System
05000255/FIN-2018411-02Palisades2018Q3GreenPhysical Protection10 CFR 73
05000528/FIN-2018008-02Palo Verde2018Q3GreenMitigating Systems10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
10 CFR 50 Appendix B Criterion V, Instructions, Procedures, and Drawings
This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V requires, 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 May 24, 2007, the licensee failed to perform the installation of the Unit 1, channel C excore nuclear instrument preamplifier connection, an activity affecting quality, in accordance with these instructions, procedures, or drawings. The licensee determined that a human performance error occurred during the performance of the 2007 work order which explicitly stated that the o-rings were required for environmental qualification. As a result, the excore detector would not have performed its safety function during a design basis main steam line break. Significance/Severity Level: The team determined this finding was of very low safety significance (Green) because a minimum of two excore detector channels always remained available to trip the reactor during a main steam line break. Redundant channels were not affected and were available to perform the required safety function to trip the reactor. Corrective Action Reference(s): Condition Report 18-12217
05000289/FIN-2018003-02Three Mile Island2018Q3Severity level MinorNo Cornerstone10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
10 CFR 50 Appendix B Criterion XVI, Corrective Action
This violation of minor significance was identified by the licensee and has been entered the licensee corrective action program and is being treated as a minor violation, consistent with the NRC Enforcement Policy. During TMIs 2015 refueling outage (T1R21) NRC and the licensee identified issues regarding reactor building pre-staging of materials were documented in NRC inspection report 05-289/2017008 (ADAMS Accession Number ML17191A697). Exelon evaluated and documented corrective actions in ACE report 2578255 which included an action to conduct an effectiveness review of those corrective actions. On October 18, 2017, after refueling outage T1R22, Exelon completed this effectiveness review. Exelon concluded that the implemented corrective actions were ineffective based on an adverse trend of licensee-identified reactor building pre-staging issues during the T1R22 refueling outage preparations. Exelon documented the results of the effectiveness review under assignment 21 of ACE 2578255 and the adverse trend in issue report 4051608. Primarily, direct oversight by Exelon staff during all phases of pre-staging, as approved by the management review committee, was not implemented and resulted in improper storage of materials in the reactor building during pre-staging activities. The improper storage was identified by Exelon during end-of-day walkdowns, from September 11 thru September 14, 2017, and documented in the corrective action program. All other corrective actions from ACE 2578255 were properly implemented. Screening: Exelons failure to implement the approved corrective actions is a performance deficiency. The inspector evaluated the significance in accordance with IMC 0612, Appendix B, Issue Screening. The inspector determined that this issue was of minor safety significance because non-compliant material configurations in the reactor building were corrected before being left unattended at the end of shift and that the corrective actions determined by ACE 2578255, except for direct Exelon supervision during pre-staging activities, were adequately implemented. Enforcement: Exelon identified this violation and documented the issue in report assignments 2578255-21 and 4051608-02. Exelon has initiated actions to include direct Exelon supervision to the current pre-staging corrective actions (AR 4051608-03) and will conduct an effectiveness review of pre-staging activities after the next outage (AR 2578255-22). This failure to comply with 10 CFR Part 50 Appendix B Criterion XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000313/FIN-2018405-02Arkansas Nuclear2018Q3GreenPhysical Protection10 CFR 73
05000397/FIN-2018003-04Columbia2018Q3GreenOccupational Radiation Safety10 CFR 20
10 CFR 20.1902
This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 CFR 20.1902(a) requires the licensee to post each radiation area with a conspicuous sign bearing the radiation symbol and the words "CAUTION, RADIATION AREA."Contrary to the above, from November 9, 2017 to November 13, 2017, the licensee failed to post a radiation area with a conspicuous sign bearing the radiation symbol and the words "CAUTION, RADIATION AREA."The licensee moved two resin liners with high dose rates into the turbine building truck bay. Once the resin liners were in the turbine building truck bay, a high radiation area boundary was posted around them. However, the dose rates outside the truck bay doors were not verified. On November 13, 2017, the licensee, while conducting routine area surveys, identified an unposted radiation area outside the turbine building truck bay doors, which resulted from the resin liners inside of the truck bay area. The licensee secured the radiation area and adequately posted it, as required.
05000327/FIN-2018003-01Sequoyah2018Q3GreenMitigating Systems10 CFR 50 Appendix R, Appendix R to Part 50-Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979
License Condition
This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Sequoyah Unit 1 Operating License Condition 2.C(16) and Sequoyah Unit 2 Operating License Condition 2.C(13) require in part that TVA shall implement and maintain in effect all provisions of the approved fire protection program. The Sequoyah fire protection report describes how the licensee complies with applicable sections of 10 CFR 50, Appendix R, including Section III.L.1 which states in part that alternative or dedicated shutdown capability provided for a specific fire area shall be able to achieve cold shutdown conditions within 72 hours and maintain cold shutdown conditions thereafter. Contrary to the above, since implementation of the Sequoyah Fire Protection Program, the licensee failed to maintain all aspects of the approved program. Specifically, in August 2018, the licensee discovered that the sites ability to achieve cold shutdown conditions within 72 hours would be challenged due to an inadequate evaluation of the RHR pumps functionality during certain Appendix R fire scenarios.
05000282/FIN-2018411-02Prairie Island2018Q3GreenPhysical Protection10 CFR 73
05000333/FIN-2018002-01FitzPatrick2018Q2GreenNo Cornerstone10 CFR 71
10 CFR 71.5, Transportation of Licensed Material
49 CFR
This violation of very low safety significance was identified by Exelon and has been entered into Exelons CAP and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy Violation: 10 CFR 71.5 requires that licensees who transport licensed material comply with the applicable requirements of the Department of Transportation (49 CFR). 49 CFR 172.202(a)(1) and (a)(2) require that the shipping description on the shipping paper include the proper shipping name and identification number for the material. 49 CFR 172.302(a) requires that shipments in bulk packages be marked with the identification number. Contrary to the above, on July 12, 2016, the shipping description on the shipping paper for shipment JAF-2016-1613 from FitzPatrick to Tennessee did not include the proper shipping name and identification number for the material. Exelon identified the error during a subsequent review of the shipping paperwork. Significance/Severity Level: No examples of transportation issues are presented in IMC 0612, Appendix E (Examples of Minor Issues). IMC 0609, Appendix D, Section VII.C.e.1 lists examples of Green findings that include documentation deficiencies including failure to properly document compliance with 49 CFR requirements such as shipping papers. Corrective Action Reference: Exelon placed this issue into its CAP as CR-JAF-2016-02857. Corrective actions included providing a corrected shipping paper to the facility in Tennessee that had received the package.
05000272/FIN-2018403-02Salem2018Q2GreenPhysical Protection10 CFR 73
05000373/FIN-2018002-03LaSalle2018Q2GreenInitiating EventsTechnical SpecificationThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification LCO 3.4.4 (applicable for Modes 1, 2 and 3) states: The safety function of 12 safety relief valves (S/RVs) shall be OPERABLE, and Action Statement A states that One or more required S/RVs inoperableA.1 be in mode 3 in 12 hours and A.2 be in Mode 4 in 36 hours. Technical Specification SR 3.4.4.1 states that Verify the safety function lift setpoints of the required S/RVs are as follows
Number of S/RVs Setpoint (psig
2 1205 36.
3 1195 35.
2 1185 35.
4 1175 35.
2 1150 34.
Contrary to the above, during portions of previous Unit 1 and 2 operating cycles from 2012 through January of 2017, two main steam S/RVs did not meet these lift pressure setpoint requirements. Specifically S/RV 2B21F013C lifted at 1131 psig instead of from 1139.8 to 1210.2 psig and S/RV 2B21F013L lifted at 1130 psig instead of from 1159.2 to 1230.8 psig (reference: Licensee Event Report 05000374/201700400; 01, Two Main Safety Relief Valves Failed Inservice Lift Inspection Pressure Test.
Significance/Severity: This licensee identified finding affected the Initiating Events Cornerstone and was screened in accordance with Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power. The two affected SRVs lifted low outside of their setpoint band, which was conservative with respect to maintaining the reactor coolant system overpressure protection safety function of these valves. Therefore, the inspectors determined that this finding is of very low safety significance (Green) because after a reasonable assessment of degradation, the finding would not have resulted in exceeding the reactor coolant system leak rate for a small LOCA and did not affect other systems used to mitigate a loss-of-coolant accident. Corrective Action Reference: AR 3974669
Reactor Coolant System
05000263/FIN-2018002-01Monticello2018Q2GreenEmergency Preparedness
Mitigating Systems
10 CFR 50.72
10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness
10 CFR 50.47, Emergency Plans
10 CFR 50 Appendix E, Appendix E to Part 50-Emergency Planning and Preparedness for Production and Utilization Facilities
10 CFR 50.54
10 CFR 50.54(q)
10 CFR 50.47(b)(8)
This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section2.3.2 of the Enforcement Policy.Enforcement: Violation: Title 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements of 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(8) requires, in part, that a licensee must provide and maintain adequate emergency facilities and equipment to support the emergency response plan.Contrary to the above requirements, on March 23, 2018, the licensee identified the site failed to maintain the effectiveness of the emergency plan by not providing and/or maintaining equipment capable of measuring the Immediately Dangerous to Life and Health (IDLH) concentrations for several toxic chemicals as required to properly classify an Alert Emergency Action Level (EAL). Specifically, while performing an emergency equipment inventory, the licensee identified that detector tubes (Draeger tubes) available to measure chlorine gas concentrations were not capable of measuring the IDLH concentration of 10 ppm required to identify the threshold level for classifying an Alert EAL (HA 3.1) since the measurement range of the available sample tubes was 50500 ppm.The inability to properly classify the Alert EAL represented a Loss of Emergency Assessment Capability and resulted in the licensees submission of Event Notification Report # 53298 in accordance with the requirements of 10 CFR 50.72(b)(3)(xiii). An immediate extent of condition review performed by the licensee identified additional deficiencies in adequate sampling methods for determining IDLH concentrations for Butadiene, Ethylene Dichloride, and Gasoline. Additionally, the licensee identified that in April 2015 there was missed opportunity to correct this deficiency when an Emergency Preparedness (EP) Coordinator, performing a Control Room Emergency Equipment Inventory, identified the need to order and replace the existing chlorine detector tubes. The EP Coordinator added the incorrect detector tubes to the existing inventory form without validating the tubes detection range and accuracy to ensure it was capable of detecting the IDLH threshold concentration level of 10 ppm.Upon identification of the issue, the licensee implemented compensatory measures for determining the EAL classification and entered the issue into the corrective action program (CR 501000009876). On May 08, 2018, the licensee implemented the sites new EAL classification procedure that was developed using NEI 9901, Revision 6, which does not require atmospheric sampling (use of detection tubes) for classification of EAL HA 3.1.Significance/Severity Level: Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.81, the inspectors determined this finding was 10 of very low safety significance (Green) because a significant amount of equipment necessary to implement the E-plan was not available or functional to the extent that any key ERO member could not perform his/her assigned functions, in the absence of compensatory measures (Degraded Planning Standard), specifically the ability to accurately classify the Alert EAL. Determining the finding significance using IMC 0609, Appendix B, Table 5.41, results in the same finding significance (very low significance) since the performance deficiency would have rendered an EAL initiating condition ineffective such that the Alert would have been declared in a degraded manner.Corrective Action Reference: 501000009876, CR Toxic Gas Detector Tube.
05000275/FIN-2018008-04Diablo Canyon2018Q2GreenMitigating Systems10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
10 CFR 50 Appendix B Criterion III, Design Control
This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program. This is being treated as a non-cited violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy. 7 Violation: Title 10 CFR Part 50, Appendix B, Criterion III, requires that measures shall include provisions to assure that appropriate quality standards are specified and included in design documents, and that deviations from such standards are controlled. Contrary to the above, from approximately February 2004 until August 2017, the licensee did not assure that appropriate quality standards were specified and included in design documents, and that deviations from such standards were controlled. Specifically, the licensee had classified the seat o-ring used in Crosby and Lonergan pressure relief valves (e.g., RV-354 and RV-355) servicing safety-related back-up air/nitrogen applications as non-safety related when they should have been classified as safety-related. Consequently, the o-rings were procured as commercial grade (non-safety related), not dedicated as safety-related and installed in safety-related equipment. Significance/Severity Level: This violation was more than minor because it had the potential to lead to a more significant safety concern if left uncorrected. Specifically, the use of non-qualified seat o-rings had the potential to cause excessive leakage past the seat, adversely affecting the fixed air/nitrogen volume required to operate safety-related equipment during a loss of normal air/nitrogen. Using IMC 0609, Appendix A, dated June 19, 2012, the team determined that this violation was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system or component, and operability was maintained. Corrective Action Reference(s): SAPNs 50935776 and 50970247
05000454/FIN-2018002-02Byron2018Q2GreenInitiating Events
Mitigating Systems
10 CFR 50 Appendix B Criterion V, Instructions, Procedures, and DrawingsA violation of very low safety significance was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, 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. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.Licensee procedure ERAA321, Administrative Requirements for Inservice Testing, stated in Step 4.10.5, that acceptance criteria are established using the reference values and the applicable ASME (American Society of Mechanical Engineers) Code. Paragraph ISTA3160, Test and Examination Procedures, of the ASME Operation and Maintenance of Nuclear Power Plants (OM) Code required in part that, Tests and examinations shall be performed in accordance with written procedures. The procedures shall contain the Owner-specified reference values and acceptance criteria. Paragraph ISTA9230, Inservice Test and Examination Results, of the ASME OM Code required, in part, that The results of tests and examinations shall be documented and shall include the following: comparison with allowable ranges of test and examination values, and analysis deviations and requirements for corrective action.Contrary to the above, from July 1, 2016, to May 30, 2018, the licensees procedures did not clearly document acceptance range, alert range, and required action values for the diesel oil (DO) transfer pump IST surveillance tests in accordance with the ASME OM Code. This resulted in several instances where the pump being tested did not meet IST criteria, but no action was taken. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedural Quality attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to clearly identify the acceptance criteria, alert range and required action ranges resulted an in organizational failure to declare the pumps inoperable and to perform required analysis of the equipments condition. The inspectors assessed the significance of the finding using SDP Appendix A and concluded the issue was of very low safety significance (i.e., Green).Corrective Action References: (1) AR 04142617, Acceptance Criteria Not Clearly Listed in DO Pump Procedures, and (2) AR 04142370, DO Pump Test Packages are Not Routed to the IST Coordinator.
05000261/FIN-2018410-01Robinson2018Q2GreenPhysical Protection10 CFR 73
05000315/FIN-2018002-03Cook2018Q2GreenEmergency Preparedness
Mitigating Systems
10 CFR 50.47(b)(8)This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy Violation: Title 10 of the Code of Federal Regulations (10 CFR) 50.47 b(8) requires that licensee emergency plans meet the standard of having adequate emergency facilities. The Cook Plant Emergency Plan states that the Technical Support Center (TSC) (an emergency facility) will be constructed to provide the same degree of radiological habitability as the Control Room under accident conditions. Contrary to the above, from January 24 to 30, 2018, the licensee failed to maintain the TSC as an adequate emergency facility, by installing a portable air conditioning unit in the Shift Managers office which compromised the ability of the TSC ventilation system to fulfill its function of providing the necessary radiological protection for the TSC. Specifically, the exhaust from the portable unit was routed to an existing ventilation duct of the TSC ventilation system, and a panel on one of the ventilation units was opened, exposing the TSC to the turbine building environment. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Facilities and Equipment attribute of the Emergency Preparedness cornerstone, whose objective is to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the significance of the finding usingSDP Appendix B and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20180952
05000369/FIN-2018002-01McGuire2018Q2Severity level IVMitigating SystemsTechnical SpecificationThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: NAC-Magnastor Certificate of Compliance 1031, Amendment 2, Technical Specifications SR 3.1.1.2 requires, in part, that the transportable storage canister (TSC) be backfilled with helium in the range of 0.694-0.802 g/liter prior to transport operations. Contrary to the above, on June 4, 2018, the licensee transported Magnastor cask 45 to the independent spent fuel storage installation pad with the TSC backfilled to approximately 0.85-0.89 g/liter due to the use of out of tolerance flow meters during backfilling operations. Significance/Severity Level: The inspectors determined that traditional enforcement is applicable for this NCV as it involved requirements pertaining to ISFSI operations and therefore the reactor oversight process is not applicable. The NCV was determined to be a Severity Level IV violation as it did not involve willfulness, was identified by the licensee, and was determined to be of minimal safety significance as the over fill of helium did not exceed any design parameters of the TSC during the transport operations.Corrective Action Reference: This issue was entered into the licensees corrective action program as NCR 2211048, Potentially Exceeding Magnastor Helium Density Upper Range.
05000324/FIN-2018011-03Brunswick2018Q2Mitigating SystemsThe licensee used 0.78 eV as the limiting activation energy for Rosemount transmitters. The activation energy was based upon an academic paper documenting experimental work performed for the early space program and first published in 1965. The paper cautioned the reader that the methods used were experimental and were not validated. A 0.5 eV activation energy for electronics was documented by the Electric Power Research Institute (EPRI) report NP-1558, which attributed it to electron migration of aluminum. Reports published by the Institute of Electrical and Electronics Engineers (IEEE) indicated that activation energies for various electronics and their failure modes could range from 0.5-0.66 eV. The licensee did not document an independent failure modes and effects analysis to justify the activation energy that they used. In addition, the licensee chose to use less limiting activation energies that were not proven to be justified. Finally, the licensee was unable to demonstrate acceptable margins for extrapolation confidence. The IEEE standard 323-1974, section 6.5.2, Mathematical Modeling, stated, the first step in the qualification by analysis is generally the construction of a valid mathematical model of the electric equipment to be qualified. The mathematical model shall be based upon established principles, verifiable test data, or operating experience data. The mathematical model shall be such that the performance of the electric equipment is a function of time and the pertinent environmental parameters. All environmental parameters listed in the equipment specification must be accounted for in the construction of the mathematical model unless it can be shown that the effects of the parameter of interest are dependent on the effects of the remaining environmental parameters. Planned Closure Actions: The team must determine whether the activation energy used for the transmitters was appropriate and, if not, whether the licensee had the responsibility to verify the information provided by their vendors and contractors. The region is discussing this issue with NRC headquarters to find a resolution to this issue.
05000315/FIN-2018002-02Cook2018Q2GreenInitiating Events
Mitigating Systems
Pr Safety
License ConditionThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: License conditions 2.C.(4) (Unit 1) and 2.C.(3)(o) (Unit 2) require implementation of the approved fire protection program. Per the Cook NFPA 805 Fire Protection Program Manual Sections 3.11.2 and 3.11.4, fire seals shall have at least a three hour fire rating. Contrary to the above, on February 6, 2018, the licensee identified multiple fire seals (many of which were between the control rooms and the cable spreading area underneath) that were degraded to the point that they could no longer meet the three hour rating requirement of Sections 3.11.2 and 3.11.4 of the Cook NFPA 805 Fire Protection Program Manual. Specifically, inadequate controls in the fire seal maintenance procedure and unclear guidance for Performance Verification department inspections led to a deterioration in seal quality. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone, whose objective is to ensure 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 usingSignificance Determination Process Appendix F and concluded the violation was of very low safety significance (Green).Corrective Action Reference: AR20181208
05000293/FIN-2018002-06Pilgrim2018Q2Severity level MinorNo Cornerstone10 CFR 50.73
10 CFR 50.9, Completeness and Accuracy of Information
This violation of minor significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a minor violation, consistent with the NRC Enforcement Policy. On June 22, 2015, Entergy submitted a licensee event report in accordance with 10 CFR 50.73 that contained information that was not complete or accurate in all material respects, contrary to the requirements in 10 CFR 50.9. Specifically, the licensee submitted Licensee Event Report 2015-004-00 to communicate the failure during testing of time delay Agastat relay 27A-B1X/TDDO intended to provide under-voltage protection for 480V emergency bus B6 by transferring power from bus B1 to bus B2. In the licensee event report, Entergy incorrectly documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. Upon identifying the issue, on March 8, 2016, Entergy submitted a revised licensee event report with the correct information. Enforcement: 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 22, 2015, Entergy provided information to the Commission that was not complete and accurate in all material respects. In the licensee event report, the licensee documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. However, bus B6 would actually have tripped from bus B1 and lost power completely. This information was material to the NRC because the NRC requires timely and accurate reporting of information related to events in order to evaluate the potential safety significance and required NRC response. Entergy identified the inaccuracy and entered the issue into its corrective action program (CR-PNP-2015-9762). On March 8, 2016, Entergy submitted a revision to the licensee event report (2015-004-01) that corrected the report. This failure to comply with 10 CFR 50.9 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes Licensee Event Report 05000293/2015-004-01.
05000315/FIN-2018002-04Cook2018Q2GreenMitigating Systems
Public Radiation Safety
10 CFR 20.1501This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: Title 10 Code of Federal Regulations; Part 20.1501(c) requires that the licensee shall ensure that instruments and equipment used for quantitative radiation measurements are calibrated periodically for the radiation measured. Contrary to the above, between November 2012 and May 2017 the licensee used the liquid scintillation counter for quantitative radiation measurements outside the range of equipment capability and the system calibration. The licensee analyzed the impact on the annual effluent reports and UFSAR limits between 1/8/2013 and 5/3/2017. The licensee entered the violation on the corrective action program. Licensee Identified Non-Cited Violation Significance/Severity Level: Green. The inspectors determined the performance deficiency was more than minor because it adversely affected the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors assessed the significance of the finding usingSDP Appendix D and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20174835
05000321/FIN-2018002-01Hatch2018Q2Initiating Events
Mitigating Systems
10 CFR 50.48
10 CFR 50 Appendix R, Appendix R to Part 50-Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979
Section III.G
On April 3, 2017, the licensee submitted Licensee Event Report (LER) 05000321, 366/2017-001-00: Unanalyzed Conditions for a Postulated Fire Discovered During NFPA 805 Transition documenting the discovery of a condition of non-compliance with the sites fire protection program (FPP). In preparation for transiting the fire protection licensing basis from 10 CFR 50.48(b) (Appendix R) to 10 CFR 50.48(c) (NFPA 805), a weak-link and operator manual action analysis was completed for Information Notice 92-18 type hot shorts on motor operated valves (MOV). The licensees examination of their Appendix R Safe Shutdown Analysis identified circuit configurations in multiple fire areas where an Appendix R postulated fire could impact the ability to achieve safe shutdown conditions. The licensee failed to protect MOV cables associated with the RHR and RCIC emergency cooling systems in fire areas 0024 (Main Control Room), 1203F (Unit 1 Reactor Building), 1205F (Unit 1 Reactor Building), and 2203F (Unit 2 Reactor Building). Specifically, the licensee failed to ensure that fire induced cable impacts cannot bypass the limit and torque switches and result in physical damage to the MOVs, thus preventing the MOVs from being operated from the Main Control Room, Remote Shutdown Panel, or locally. This condition could prevent operators from achieving and maintaining safe shutdown (SSD) of the plant in the case of a postulated fire. A licensee-identified non-compliance with 10 CFR Part 50, Appendix R, Section III.G.2, was identified for the licensees failure to protect one of the redundant trains of equipment needed to achieve post-fire SSD from fire damage. Specifically, the licensee failed to use one of the means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. The inspectors performed a detailed review of the information and documents related to the LER and discussed the condition with the licensee to assess the adequacy of the licensees compensatory measures and corrective actions. Corrective Action(s): Hourly fire watches and Fire Action Statements were initiated to address the postulated condition for the identified MOVs. Additionally, the licensee committed to completing physical plant modifications to the impacted MOVs during the next Unit 1 and Unit 2 plant refueling outages to rectify the issue of potential spurious operation of the associated MOVs associated with this LER. Corrective Action Reference(s): The licensee entered this issue into their Corrective Action Program (CAP) as condition reports (CRs) 10326399, 10326401, 10326402, 10326404, and 10326405. Enforcement: Violation: 10 CFR Part 50.48(b)(1) requires that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G. 10 CFR 50, Appendix R, Section III.G.2, states, in part, that where cables or equipment, that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided: (a) separation of cables and equipment by a fire barrier having a 3-hour rating, (b) separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards and with fire detectors and an automatic fire suppression system in the fire area, or (c) enclosure of cables and equipment in a fire barrier having a 1-hour rating and with fire detectors and an automatic fire suppression system in the fire area. Contrary to the above, the licensee failed to use one of the means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. Specifically from October 1974 to April 2017, the licensee had not met the requirements of 10 CFR Part 50.48(b) to identify and protect cabling of 51 Unit 1 and Unit 2 RHR and RCIC emergency cooling system MOVs in fire areas 0024 (Main Control Room), 1203F (Unit 1 Reactor Building), 1205F (Unit 1 Reactor Building), and 2203F (Unit 2 Reactor Building). On April 3, 2017, the licensee identified the failure to protect equipment that was required to mitigate fire events and determined that fire damage could cause mal-operation of the affected MOVs, potentially leading to fire induced cable impacts which bypass the limit and torque switches and result in physical damage to the MOVs, thus preventing the MOVs from being operated from the Main Control Room, Remote Shutdown Panel, or locally. A fire-induced failure could have caused the loss of the required Safe Shutdown components. Severity/Significance: Failure to protect one train of cables and equipment necessary to achieve post-fire SSD from fire damage for fire areas designated in the Fire Protection Program (FPP) as meeting Appendix R, Section III.G.2, was a performance deficiency. This finding was more than minor because it was associated with the reactor safety mitigating system cornerstone attribute of protection against external events (i.e., fire). Specifically, failure to protect safe shutdown cables and equipment from fire damage negatively affected the reactor safety mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this issue relates to fire protection and this non-compliance was identified as a part of the sites transition to NFPA 805, this issue is being dispositioned in accordance with Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) of the NRC Enforcement Policy. The significance of this licensee-identified non-compliance with 10 CFR Part 50, Appendix R, Section III.G.2, was determined by the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. The quantitative screening approach performed by a Region II Senior Risk Analyst resulted in a calculated delta core damage frequency (CDF) of less than 1E-04, which screens this noncompliance to less-than-red significance. Additionally, in order to verify that this noncompliance was not associated with a finding of high safety significance (Red), inspectors reviewed qualitative and quantitative risk analyses performed by the licensee. These risk evaluations took ignition source and target information from the ongoing HNP fire PRA to demonstrate that the significance of the non-compliances were less-thanthan 1E-4/year). The inspectors also performed walk-downs to verify key assumptions were applicable. Based on the ignition frequency of fire sources in the affected areas, inspectors determined that the significance of this non-compliance was less-than-red. The inspectors also noted that the values in the licensees quantitative analysis were conservative, in that they used screening values instead of more detailed values. This provided additional confidence that this non-compliance was not associated with a finding of high safety significance (Red). The inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), a Confirmatory Order (ML16223A467) which extended the period for discretion, and Inspection Manual Chapter 0305. On April 4, 2018 (ML18096A955), the licensee submitted a license amendment request to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c). The inspectors reached this conclusion due to the fact that this issue was licensee-identified and will be addressed during the licensees transition to NFPA 805, it was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance (Red).Primary containment
Remote shutdown
05000255/FIN-2018011-03Palisades2018Q2GreenMitigating Systems10 CFR 50.48
License Condition - Fire Protection
License condition 2.C(3)requires the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program that complies with Title 10of the Code of Federal Regulations(CFR), Part50.48(a) and 10 CFR 50.48(c), NFPA Standard NFPA 805, as approved in the Safety Evaluation Report (SER)dated February 27, 2015. Section 2.4.3.3 of NFPA 805 states, in part, that the Probabilistic Safety Assessment (PSA)(Probabilistic RiskAssessment (PRA))approach, methods, and data shall be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.Contrary to the above, from February 27, 2015, until May 14, 2018, the licensee failed to base the PSA (PRA) approach, methods, and data on the as-built and as-operated and maintained plant.Specifically, the licensees PSA (PRA) model/analysis credited the suppression system located in the cable spreading room to suppress a type 2 fire scenarios, whereas the actual room contained numerous obstructions by the stacked cable trays located near the ceiling that interfered with the water spray pattern discharged from the sprinklers from providing adequate water density pattern to suppress a fire in areas below the cable trays which contained electrical panels.Significance/Severity Level: The performance deficiency was determined to be more-than-minor, and therefore, a finding because the performance deficiency, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to correctly model/analyze the as-built condition of the suppression system located in the cable spreading room in the PRA could potentially affect the risk associated with a fire in the room and could result in inappropriately screening out the effects of otherchanges associated with the fire area.The finding was of very-low safety significance (Green). While there may be a change to the plants baseline risk as a result of this issue, this is a fire modeling issue only; no physical plant fire protection feature was altered by the fire PRA model. Therefore, there was no increase in actual core damage risk to the physical plant.
05000390/FIN-2018002-02Watts Bar2018Q2Severity level IVNo CornerstoneTechnical SpecificationLER: 05000390, 391/2017-013-00, Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications, November 6, 2017. Violation: Watts Bar Unit 1 TS 3.7.12, Auxiliary Building Gas Treatment System (ABGTS), Condition A, requires that an inoperable ABGTS train to be restored to operable status within 7 days. Condition B of TS 3.7.12 requires the plant to be in Mode 3 within 6 hours and Mode 5 within 36 hours if one train of ABGTS is inoperable longer than 7 days. Contrary to the requirements of TS 3.7.12, ABGTS, train A was determined to be inoperable from July 7, 2017, at 2030 Eastern Daylight Time (EDT) to September 5, 2017, at 1645 EDT while the plant remained in Mode 1. Significance/Severity Level: This violation was characterized using traditional enforcement because the NRC determined that this violation was not reasonably foreseeable and preventable by the licensee and, therefore, is not a performance deficiency. The violation was assessed using Sections 2.2.4 and 6.1.d.1 of the NRCs Enforcement Policy and determined to be a SL IV violation. Corrective Action Reference(s): Condition Report (CR) 1335791Auxiliary Building Gas Treatment System
05000259/FIN-2018002-04Browns Ferry2018Q2GreenNo Cornerstone10 CFR 50.48LER 05000259, 260, 296/2018-003-00 identified a violation of 10 CFR 50.48(c)(4)(iii). This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR 50.48(c)(4)(iii) Fire Protection required, in part, that the licensee maintain fire protection defense in depth (post-fire safe shutdown capability). Contrary to the above, from October 28, 2015 until March 10, 2018, the C3 Emergency Equipment Cooling Water (EECW) pump did not have the Fire Protection Plan required backup control panel function. Significance/Severity: Using IMC 0609 Appendix F, the violation was screened to green following a risk analysis performed by the licensee that a NRC Senior Risk Analyst reviewed and agreed was correctly performed. Corrective Action Reference(s): CR 1394604Emergency Equipment Cooling Water
05000498/FIN-2018002-01South Texas2018Q2GreenInitiating Events
Mitigating Systems
Technical SpecificationThis violation of very low safety significance was identified by the licensee, has been entered into the licensees corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification 6.8.1.a requires that, Written procedures shall be established, implemented, and maintained covering the activities referenced below: The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Section 9.a, Procedures for Performing Maintenance, states, in part, that Maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. The licensee established Procedure COM-0001, Conduct of Maintenance, to guide maintenance craft on what to do if a condition or issue arises during a maintenance activity. Specifically, Section 1.4 Supervisor Responsibilities, states, in part, that, If we cannot find the problem with the component or piece of equipment, the issue must be raised to the Division Manager/General Supervisor BEFORE we close the work control document AND return the equipment to operations. Contrary to the above, on March 10, 2017, Unit 1 E1B undervoltage relay was found outside the technical specification acceptance criteria, and was retested until the relay it was back in tolerance and placed back into service (declared operable) instead of raising the issue up to the division manager for further evaluation. The issue was discussed with the electrical maintenance supervisor and the findings were documented in Condition Report 17-12616. The relay was declared operable and placed back into service. Subsequently, after review of the condition report, approximately 99 hours after the relay was declared inoperable, the relay was replaced, and the system declared operable. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the undervoltage relay was outside its tolerance and placed back into service without correcting the cause of being outside its tolerance. The inspectors assessed the significance of the finding using Exhibit 2, Mitigating Systems Screening Questions, of Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined this finding is not a deficiency affecting the design or qualification of a mitigating structure, system, and component that maintained its operability or functionality; the finding does not represent a loss of system and/or function; the finding does not represent an actual loss of function of at least a single train for greater than its Technical Specification-allowed outage time; and the finding does not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the inspectors determined the finding was of very low safety significance (Green). Corrective Action Reference: Condition Report 17-12616
05000390/FIN-2018050-01Watts Bar2018Q2GreenInitiating Events
Mitigating Systems
10 CFR 50 Appendix B Criterion III, Design ControlThis violation of very low safety significance (Green)was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a Non-CitedViolation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 of the Code of Federal Regulations(10 CFR) Part 50 (10 CFR 50), Appendix B, Criterion III, Design Control, requires the licensee to effectively implement design control measures for piping analysis calculations* associated with the Unit 1 and Unit 2 emergency core cooling systems (ECCS).Contrary to the above, since initial operation of Unit 1 in 1996 and Unit 2 in 2016, Tennessee Valley Authority failed to ensure the proper hydraulic time history was utilized in TVAs TPIPE special purpose computer program used to determine static and dynamic linear elastic analyses for the ECCS including the effects of pipe voiding. This resulted in non-conservative voiding design acceptance criteria for the RHR and SI systems of both units. This performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems 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 failure to utilize proper hydraulic time history in the licensees TPIPE computer model resulted in developing non-conservative voiding acceptance criteria that was used during operation that could challenge ECCS functionality. The finding was determined to be of very low safety significance since additional analysis determined with reasonable assurance that the systems remained operable but non-conforming and would have performed their safety function.Significance/Severity Level: Green. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the finding affected the design or qualification of mitigating systems; however, the mitigating systems maintained their operability. Corrective Action Reference:CR 1407257
05000382/FIN-2018002-03Waterford2018Q2GreenMitigating SystemsTechnical SpecificationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification 3.6.3, Containment Isolation Valves, requires, in part, that when an isolation valve for containment penetrations associated with an open system are inoperable, the licensee must restore the inoperable valve(s) to operable status within 4 hours, isolate the affected penetration within 4 hours, or be in hot standby within the next 6 hours. Contrary to the above, between December 8, 2017, and December 11, 2017, with containment isolation valves inoperable, the licensee did not restore the inoperable valves to operable status within 4 hours, isolate the affected penetrations within 4 hours, or place the unit in hot standby within the next 6 hours. The licensee restored the valves to operable status on December 20, 2017, exceeding the Technical Specification 3.6.3 allowed outage time by approximately 70 hours. Significance/Severity Level: The finding was of very low safety significance (Green) because the containment isolation valves were maintained closed during the period and did not represent an actual open pathway in the physical integrity of the reactor containment. Corrective Action Reference: CR-WF3-2018-00983
05000348/FIN-2018002-06Farley2018Q2GreenMitigating Systems10 CFR 50 Appendix B Criterion XI, Test ControlViolation: 10 CFR 50, Appendix B, Criterion XI, Test Control, required in part, a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in all applicable design documents. Contrary to the above, the Unit 1 pressurizer power operated relief valve (PORV) PCV-445A was not set up properly for testing and the written test procedures did not incorporate the acceptance limits in all applicable design documents. Specifically, the open and closed limit switches were not set up properly which would result in shorter stroke times during testing per licensee procedure FNP-1-STP-45.11, Miscellaneous Cold Shutdown Valves Inservice Test. Additionally, licensee procedure FNP-1-STP-201.28, Pressurizer Power Operated Relief Valves Position Indication and Relay Logic Contact Verification Q1B31PCV0444B and Q1B31PCV0445A, Ver. 14, allowed a minimum stroke length of 0.5 inches while a vendor evaluation in Request for Engineering Review (RER) 941414 stated a minimum stroke length of 0.56 inches was required.
05000266/FIN-2018001-05Point Beach2018Q1GreenMitigating SystemsTechnical SpecificationViolation: 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 TDAFWAuxiliary Feedwater
05000263/FIN-2018001-02Monticello2018Q1GreenMitigating Systems10 CFR 50.59, Changes, Tests and Experiments
10 CFR 50.90, Application for Amendment of License, Construction Permit, Or Early Site Permit
Technical Specification
Violation: Title 10 CFR 50.59(d)(1) requires, in part, that the licensee maintain records of changes to the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c).These records must include a written evaluation which provides

the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to Paragraph (c)(2) of this section.Title 10 CFR 50.59(c)(2)(ii) requires that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the Final Safety Analysis Report (FSAR) (as updated).Technical Specification (TS) 3.3.1.1, Reactor Protection System (RPS) Instrumentation, states the RPS instrumentation for each function in Table 3.3.1.11 shall be operable. As specified in Table 3.3.1.11, Function 5, Main Steam Isolation Valve (MSIV) - Closure (8 channels) and Function 8, Turbine Stop Valve (TSV) Closure (4 channels) are required to be operable in Mode 1. TS 3.3.1.1, Condition C.1 states with one or more functions with RPS trip capability not maintained, to restore RPS trip capability in 1 hour and was applicable to both the MSIV and TSV RPS logic functional testing.Contrary to the above, on March 7, 2009 and July 11, 2009, the licensee failed to perform and maintain a written evaluation as required by 10 CFR 50.59(d)(1) to demonstrate a change to its facility did not require a license amendment. Specifically, the licensee incorrectly concluded in its 10 CFR 50.59 evaluation SCR080319, dated September 29, 2008, that no license amendment was required prior to implementing two surveillance test procedures; 0009 Turbine Stop Valve Closure Scram Test Procedure, Revision 16 on March 7, 2009 and; 0008 Main Steam Line Isolation Valve Closure Scram Test Procedure, Revision 20 on July 11, 2009. The test fixture was applied during quarterly surveillance testing through September 16, 2017.Implementation of procedures 0008 and 0009, respectively, resulted in the loss of RPS trip Function 5 (MSIV) and Function 8 (TSV) by bypassing more than the TS minimum allowed inputs per channel to maintain functionality, thereby violating the requirements of TS 3.3.1.1. Loss of these functions resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated) as specified by 10 CFR 50.59(c)(2)(ii).On November 14, 2017, the licensee generated CAP 501000005391 after conducting an operating experience evaluation of a similar event at another station concluding the event was applicable to the Monticello Plant. The surveillance procedures were immediately quarantined and subsequently revised on December 8, 2017 and December 11, 2017, to remove the use of the RPS test fixture.Significance/Severity Level:Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined this finding was of very low safety significance (Green) because it did not affect a single RPS trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown.The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV

The disposition of this violation closes LER 05000263/201700600.Corrective Action Reference: 501000005391
Reactor Protection System
Main Steam Isolation Valve
05000255/FIN-2018001-03Palisades2018Q1GreenMitigating SystemsTechnical SpecificationA violation of very low safety significance (Green) was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a Non-Cited Violation consistent with Section 2.3.2 of the Enforcement Policy. Enforcement:Violation: Technical Specification 3.7.6 requires that the combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T81) shall be greater or equal than 100,000 gallons. LCO 3.7.6, Condition A states that if the useable volume is not within this limit then A.1 Verify OPERABILITY of backup water supplies in 4 hours andA.2 Restore condensate volume to within limit in 7 days. Condition B states that if the Required Action and associated Completion Time is not met then B.1 Be in MODE 3 in 6 hours and B.2 Be in MODE 4 without reliance on steam generators for heat removal in 30 hours. Contrary to the above, on December 7, 2017 and March 3, 2016, the licensee failed to enter and comply with the actions required by LCO 3.7.6 Condition A and Condition B when Primary Makeup Tank Makeup Control Valve CV2008 could not be fully opened, resulting in a combined useable volume of the CST and T81 of less than 100,000 gallons.Significance/Severity Level: The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the CST and T81 volume were considered inoperable by the TS requirements, there was not a loss of safety function because credited backup water sources were available and operable.Therefore, the finding screened as Green.Corrective Action References: The licensee entered these issues into their CAP as CRPLP20175589, CRPLP20175554, CRPLP20175551, and CRPLP20161116Steam Generator
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