Licensee-identified

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 SiteQuarterSignificanceCornerstoneViolation ofDescriptionSystem
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.
05000259/FIN-2018003-02Browns Ferry2018Q3GreenNo Cornerstone10 CFR 50.48This 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: Violation: 10 CFR 50.48(c)(3)(ii) required, in part, the licensee to complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan. NFPA 805 Chapter 2, section 2.4.2.2.1, Circuits Required in Nuclear Safety Functions required, in part, that circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation or that result in the mal-operation of the equipment identified. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.
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).
05000275/FIN-2018404-02Diablo Canyon2018Q3GreenPhysical Protection10 CFR 73
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
05000331/FIN-2018003-01Duane Arnold2018Q3GreenMitigating 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
A violation of very low safety significance (Green)was identified by the licensee and has been entered into the corrective action program. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. System Design Specification APEDA61019, Pressure Integrity of Piping and Equipment Pressure Parts Data Sheet, required in the applicable castings section T1.3.3.b, all accessible surfaces including machine surfaces shall be examined by either the magnetic particle or liquid penetrant method in either the furnished or finished condition. Contrary to the above, in October 2016, measures were not established to assure that applicable design basis requirements as defined in 10 CFR 50.2 were translated into work instructions repairing the B inboard main steam isolation valve, CV 4415, during RFO 25. Specifically, instructions to perform a NDE of machined surfaces following the valve repair were not included in the work package. As a result, the non-destructive examination was not performed prior to placing the valve into service.Main Steam Isolation Valve
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
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.
05000483/FIN-2018003-02Callaway2018Q3GreenMitigating SystemsTechnical Specification - ProceduresThis 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.Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 6 of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for combating emergencies and other significant events. The licensee established Emergency Operating Procedure (EOP) ES-0.2, Natural Circulation Cooldown, Revision 9, in part, to meet the regulatory requirement. Figure 1 of ES-0.2 allowed cooldown rates that exceeded the values used in the license basis for radiological consequence analyses and exceeded the values used in the design of the nitrogen accumulators for atmospheric steam dumps and turbine-driven auxiliary feedwater system actuations. This issue was discussed in Licensee Event Report 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown Contrary to the above, from April 29, 2008 through May 7, 2018, the licensee failed to maintain procedures for combating emergencies and other significant events. Specifically, the licensee failed to maintain EOPs for natural circulation cooldown. This performance deficiency resulted in atmospheric steam dumps and turbine-driven auxiliary feedwater systems being rendered inoperable due to depletion of the safety-related actuation nitrogen.Auxiliary Feedwater
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.
05000395/FIN-2018003-01Summer2018Q3GreenMitigating Systems10 CFR 50.48
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. V.C. Summer Operating License condition 2.c(18) states in part that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(c), National Fire Protection Association (NFPA) 805 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to the above, on January 10, 2017, the licensee failed to implement an established procedure, FPP-025, Fire Containment, Revision 6D, to ensure fire door DRAB/319 closes and latches on its own power.
05000282/FIN-2018411-02Prairie Island2018Q3GreenPhysical Protection10 CFR 73
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).
05000266/FIN-2018003-04Point Beach2018Q3GreenNo Cornerstone10 CFR 72Violation: 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.
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
05000338/FIN-2018003-01North Anna2018Q3GreenMitigating 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.a of the Enforcement Policy. Violation: TS 5.4.1.a, requires in part, that written procedures shall be established per Revision 2 of Regulatory Guide 1.33, Appendix A, of which part 9.a requires written procedures and documented instructions appropriate to the circumstances for performing maintenance that can affect the performance of safety related equipment. Contrary to the above, on June 12, 2018, the licensee failed to adequately establish a procedure appropriate to the circumstances during maintenance on the safety-related main control chillers. Specifically, licensee mechanical preventative maintenance procedure, 0-MPM-0806-02, Inspection of Control Room Chillers, Revision 0, did not provide a proper method to adequately monitor the Freon level in main control room chillers. Consequently, the licensee discovered a low Freon level condition on main control room chiller 1-HV-3-4B, which rendered the chiller inoperable. Significance: The inspectors reviewed Exhibit 2 Mitigating Systems Screening Questions of IMC 0609 Appendix A, The Significance Determination Process (SDP) for findings at Power and determined this finding was of very low safety significance, Green, because there was no design deficiency, it did not represent a loss of system or function, and did not represent an actual loss of function for greater than its TS allowed outage time. Corrective Action Reference: CR109958
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
05000263/FIN-2018012-03Monticello2018Q3GreenInitiating Events
Mitigating Systems
10 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 CAP. Therefore, this finding being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.Enforcement:Violation: 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. Updated Final Safety Analysis Report, Appendix I,Evaluation of High Energy Line Breaks Outside Containment,Table I.5-2, Table of System Effects,Revision 36P, listed the Division II emergency power system as available during HELBs outside containment. Contrary to the above, on July 29, 1974, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically,the Division II emergency power system would not be available during a HELB outside containment.Procedure B.09.07-05, Operations Manual Section 4.16 kV Station Auxiliary, Revision 53,had actions that required entry into the lower 4kV area to permit repowering Division II emergency power systems but this area would be inaccessible during the event. Significance: The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences.Specifically, the performance deficiency resulted in a condition were the Division II emergency power system would not be available during HELBs outside containment. The inspectors assessed the significance of the finding using the SDP in accordance with IMC 0609, 11 Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 2, Mitigating System Screening Questions,and concluded the violation was of very-low safety or security significance (Green)because the licensee reasonably demonstrated an alternate strategy was available to timely reach and maintain cold shutdown conditions. Corrective Action References: CAP501000011837, CAP 50100001593
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.
05000237/FIN-2018003-02Dresden2018Q3Severity level IVNo Cornerstone10 CFR 50.59, Changes, Tests and ExperimentsViolation: Dresden Technical Requirements Manual (TRM) Control Program (Appendix G of TRM), Section 1.5, Program Implementation, requires that proposed changes to the TRM are screened and reviewed under the 10 CFR 50.59 process in accordance with plant specific procedures. Contrary to the above, in October 2017 Dresden station approved and implemented an extension to the surveillance frequency of DIS 150020, Division I & II Low Pressure Coolant Injection (LPCI) Pumps Suction and Injection Valves Circuitry Logic System Functional Test, on Unit 2 per the Surveillance Frequency Control Program (SFCP) without the required 50.59 review.
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 VThis 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
05000354/FIN-2018403-02Hope Creek2018Q3GreenPhysical Protection10 CFR 73
05000456/FIN-2018411-01Braidwood2018Q3GreenPhysical Protection10 CFR 73
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
05000259/FIN-2018010-01Browns Ferry2018Q2GreenMitigating Systems10 CFR 50.48
License Condition
The Browns Ferry Nuclear Plant, Unit 3, Renewed Facility Operating License, DPR-68, License condition 2.C(7) required, in part, that TVA Browns Ferry Nuclear Plant shall 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)... Specifically, 10 CFR 50.48(c)incorporated by reference National Fire Protection Association Standard 805 (NFPA 805), and NFPA 805 section 2.4.2.2.2, Other Required Circuits, required in part, Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. Contrary to the above, since June 22, 2016, when the NFPA 805 requirements went into effect, the licensee did not implement and maintain in effect all provisions of the approved fire protection program, because the licensee did not correctly evaluate circuits that share common power supply for their impact on their ability to achieve nuclear safety performance criteria in accordance with NFPA 805.Significance: The team evaluated the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, for Mitigating Systems, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued May 2, 2018, and determined the finding to be of very low
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
05000293/FIN-2018002-04Pilgrim2018Q2GreenBarrier Integrity
No Cornerstone
10 CFR 50 Appendix B Criterion VThis 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 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, in part, that activities affecting quality shall be prescribed by documented instructions appropriate to the circumstances and shall be accomplished in accordance with the instructions. Contrary to the above, from January 1994 to June 2017, Entergy modified site surveillance procedure 8.M.3-18, Standby Gas Treatment System Exhaust Fan Logic Test and Instrument Calibration, without prescribing adequate documented instructions for the condition caused by the testing. Specifically, Entergy failed to identify that the procedurally prescribed lineup of the standby gas treatment system resulted in secondary containment being inoperable due to the large opening introduced into the system. Significance/Severity: The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance. Corrective Action Reference: CR-PNP-2017-11714 The disposition of this violation closes Licensee Event Reports 05000293/2017-013-00 and 05000293/2017-013-01.Secondary containment
Standby 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
05000277/FIN-2018010-03Peach Bottom2018Q2GreenEmergency PreparednessLicense Condition - Fire ProtectionThis violation of very low safety significance was identified by Exelon and has been entered into Exelons corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Peach Bottom Unit 2 and Unit 3 Renewed Facility Operating License Condition 2.C.(4) requires, in part, Exelon to implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report. Fire Protection Program, Peach Bottom Atomic Power Station, Units 2 and 3, is incorporated by reference into the Updated Final Safety Analysis Report, as discussed in Section 10.12, Fire Protection Program. Fire Protection Program Chapter 5.1, Methodology, assumes that two or more circuit failures resulting in spurious operation of two or more valves in series at a high/low pressure interface may occur due to a postulated fire in any given area.Fire Protection Program Chapter 6.2, Analysis of High/Low Pressure Interfaces, requires Exelon to address the situations for which the isolation valves at a given interface point consists of two electrically controlled valves in series and where the potential may exist for a single fire to cause damage to cables associated with both valves. 8 Contrary to above, as ofMarch 14, 2018, Exelon identified they failed to evaluate two motor-operated valves in series, MO-2-06-2663 and MO-2-06-038B for Unit 2, and MO-3-06-3663 and MO-3-06-038B for Unit 3, where the potential may exist for a single fire to cause damage to cables associated with both valves. Specifically, a postulated fire scenario could cause spurious opening of the valves, which may potentially result in a fire-induced loss of coolant accident through the high/low pressure system interface. Exelons evaluation identified the affected valves cables are routed through Fire Area 6N for the Unit 2 valves, and Fire Area 13N for the Unit 3 valves, and thus, a possibility exists for a single fire to cause damage to cables associated with both valves. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Significance/Severity: The inspectors performed a Phase 2 Significance Determination Process screening for this issue, in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected the post-fire safe shutdown category because of spurious operations of safe shutdown components. Based on a walkdown of Fire Areas 6N and 13N, the inspectors did not identify any credible fire ignition source scenarios that could affect both Unit 2 motor-operated valves or both Unit 3 motor operated valves. Therefore, based upon task number 2.3.5, the inspectors determined that this finding screened to very low safety significance (Green).Corrective Action References: IR 04115309, EC 623585, EC 623586
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
05000293/FIN-2018002-05Pilgrim2018Q2GreenNo Cornerstone10 CFR 50.72(b)(3)(v), Loss of Safety FunctionThis 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 NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR 50.72(b)(3)(v)(C) requires licensees to a notify the NRC within 8 hours any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above, Entergy did not make a required notification pursuant to 10 CFR 50.72(b)(3)(v)(C). Specifically, on June 20, 2017, secondary containment was declared inoperable due to simultaneous opening of both airlock doors, and Entergy did not make the required notification until June 22, 2017. Significance/Severity: This violation is being treated under the NRCs traditional enforcement process, for impeding the regulatory process, specifically Entergy did not make a required notification, as outlined in Inspection Manual Chapter 0612, Appendix B. The Reactor Oversight Processs 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. The severity of this violation was determined to be Severity Level IV, as outlined in Example 9 from Section 6.9.d. of the NRC Enforcement Policy. Corrective Action References: CR-PNP-2017-06380 and CR-PNP-2017-07015 The disposition of this finding closes Licensee Event Report 2017-011-00.Secondary containment
05000348/FIN-2018002-06Farley2018Q2GreenMitigating Systems10 CFR 50 Appendix B Criterion XIViolation: 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.
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.
05000458/FIN-2018406-03River Bend2018Q2GreenPhysical Protection10 CFR 73
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