Licensee-identified
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| Site | Quarter | Significance | Cornerstone | Violation of | Description | System | |
|---|---|---|---|---|---|---|---|
| 05000373/FIN-2018412-01 | LaSalle | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000282/FIN-2018411-02 | Prairie Island | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000255/FIN-2018411-02 | Palisades | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000456/FIN-2018411-01 | Braidwood | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000282/FIN-2018411-01 | Prairie Island | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000255/FIN-2018411-01 | Palisades | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000313/FIN-2018405-02 | Arkansas Nuclear | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000275/FIN-2018404-03 | Diablo Canyon | 2018Q3 | Severity level IV | Physical Protection | 10 CFR 73 | ||
| 05000275/FIN-2018404-02 | Diablo Canyon | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000354/FIN-2018403-02 | Hope Creek | 2018Q3 | Green | Physical Protection | 10 CFR 73 | ||
| 05000263/FIN-2018012-03 | Monticello | 2018Q3 | Green | Initiating 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 | |
| 05000482/FIN-2018010-05 | Wolf Creek | 2018Q3 | Green | Mitigating Systems | 10 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 |
| 05000266/FIN-2018010-01 | Point Beach | 2018Q3 | Green | No Cornerstone | 10 CFR 50 Appendix B Criterion XII, Control of Measuring and Test Equipment | 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 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). | |
| 05000528/FIN-2018008-03 | Palo Verde | 2018Q3 | Severity level IV | Mitigating Systems | 10 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 | |
| 05000528/FIN-2018008-02 | Palo Verde | 2018Q3 | Green | Mitigating Systems | 10 CFR 50 Appendix B 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 | Main Steam Line |
| 05000390/FIN-2018003-07 | Watts Bar | 2018Q3 | Green | Mitigating Systems | Technical Specification | 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: 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. | |
| 05000390/FIN-2018003-06 | Watts Bar | 2018Q3 | Severity level IV | Mitigating Systems | Technical Specification | 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: 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. | |
| 05000390/FIN-2018003-05 | Watts Bar | 2018Q3 | Green | Mitigating Systems | License Condition - Fire Protection | This 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. | |
| 05000266/FIN-2018003-04 | Point Beach | 2018Q3 | Green | No Cornerstone | 10 CFR 72 | Violation: 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. | |
| 05000397/FIN-2018003-04 | Columbia | 2018Q3 | Green | Occupational Radiation Safety | 10 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. | |
| 05000266/FIN-2018003-03 | Point Beach | 2018Q3 | Green | Initiating 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). | |
| 05000272/FIN-2018003-03 | Salem | 2018Q3 | Green | Mitigating Systems | 10 CFR 50 Appendix B Criterion V | This 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 | |
| 05000483/FIN-2018003-02 | Callaway | 2018Q3 | Green | Mitigating Systems | Technical Specification - Procedures | This 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 |
| 05000315/FIN-2018003-02 | Cook | 2018Q3 | No Cornerstone | Certificate 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 | |||
| 05000266/FIN-2018003-02 | Point Beach | 2018Q3 | Green | Initiating 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). | Decay Heat Removal |
| 05000289/FIN-2018003-02 | Three Mile Island | 2018Q3 | Severity level Minor | No Cornerstone | 10 CFR 50 Appendix B 10 CFR 50 Appendix B Criterion XVI | 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. | |
| 05000348/FIN-2018003-02 | Farley | 2018Q3 | Green | No Cornerstone | 10 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-2018003-02 | Browns Ferry | 2018Q3 | Green | No Cornerstone | 10 CFR 50.48, Fire Protection | 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 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. | |
| 05000237/FIN-2018003-02 | Dresden | 2018Q3 | Severity level IV | No Cornerstone | 10 CFR 50.59, Changes, Tests and Experiments | Violation: 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. | Low Pressure Coolant Injection |
| 05000338/FIN-2018003-01 | North Anna | 2018Q3 | Green | Mitigating Systems | Technical Specification | 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 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 | |
| 05000458/FIN-2018003-01 | River Bend | 2018Q3 | 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 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. | ||
| 05000263/FIN-2018003-01 | Monticello | 2018Q3 | Green | Initiating Events Mitigating Systems | 10 CFR 50.49 | 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. 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 | |
| 05000395/FIN-2018003-01 | Summer | 2018Q3 | Green | Mitigating Systems | 10 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. | |
| 05000327/FIN-2018003-01 | Sequoyah | 2018Q3 | Green | Mitigating Systems | 10 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. | |
| 05000331/FIN-2018003-01 | Duane Arnold | 2018Q3 | Green | 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 | 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 |
| 05000424/FIN-2018003-01 | Vogtle | 2018Q3 | Green | Emergency Preparedness | 10 CFR 50.47 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 |
| 05000263/FIN-2018002-01 | Monticello | 2018Q2 | Green | Emergency Preparedness Mitigating Systems | 10 CFR 50.72 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness 10 CFR 50.47 10 CFR 50 Appendix E 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. | |
| 05000369/FIN-2018002-01 | McGuire | 2018Q2 | Severity level IV | Mitigating Systems | 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 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. | |
| 05000293/FIN-2018002-06 | Pilgrim | 2018Q2 | Severity level Minor | No Cornerstone | 10 CFR 50.9, Completeness and Accuracy of Information 10 CFR 50.73, Licensee Event Report System | 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. | |
| 05000219/FIN-2018410-04 | Oyster Creek | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000454/FIN-2018410-01 | Byron | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000261/FIN-2018410-01 | Robinson | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000458/FIN-2018406-03 | River Bend | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000272/FIN-2018403-02 | Salem | 2018Q2 | Green | Physical Protection | 10 CFR 73 | ||
| 05000390/FIN-2018050-01 | Watts Bar | 2018Q2 | Green | Initiating Events Mitigating Systems | 10 CFR 50 Appendix B Criterion III, Design Control | This 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 | Emergency Core Cooling System |
| 05000315/FIN-2018002-02 | Cook | 2018Q2 | Green | Initiating Events Mitigating Systems Pr Safety | 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 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 | |
| 05000259/FIN-2018002-03 | Browns Ferry | 2018Q2 | TBD | Mitigating Systems | 10 CFR 50.48, Fire Protection | An Apparent Violation (AV) of 10 CFR 50.48(c)(3)(ii) was identified for the failure to perform a required analysis using the methodology in Chapter 2 of NFPA 805 for the RHRSW piping as a result of a postulated fire scenario. | Residual Heat Removal Service Water |
| 05000315/FIN-2018002-03 | Cook | 2018Q2 | Green | Emergency 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 | |
| 05000373/FIN-2018002-03 | LaSalle | 2018Q2 | Green | Initiating Events | 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: 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 Safety Relief Valve Main Steam |
| 05000382/FIN-2018002-03 | Waterford | 2018Q2 | Green | Mitigating Systems | Technical Specification | 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: 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 | |
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