Semantic search

Jump to navigation Jump to search
 Start dateSiteTitleDescription
05000313/FIN-2018003-0430 September 2018 23:59:59Arkansas NuclearFailure to Verify Safety-Related 4160 V Breaker Operability Following Maintenance ActivitiesThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to perform post-maintenance testing to demonstrate component operability for the train A safety-related 4160 V switchgear A-303 breaker that provides power to the swing service water pump B (P-4B) after the breaker was racked in. The breaker subsequently failed to close when attempting to start the pump.
05000317/FIN-2018410-0130 September 2018 23:59:59Calvert CliffsSecurity
05000528/FIN-2018008-0430 September 2018 23:59:59Palo VerdeMinor ViolationFailure to promptly identify and correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI. The team identified a backlog of conditions adverse to quality that the licensee had failed to timely correct. The oldest of these conditions was approximately 10 years old, with several hundred having been identified at least two operating cycles prior to the inspection. The team determined that the licensee was appropriately addressing degraded components that had an impact on safety or security, but was not always tracking or timely correcting nonconformances with its design bases in cases where these nonconformances had been assessed as not impacting safety-related functions. Further, the licensee was unable to initially determine the scope of its nonconformance backlog. The licensee documented this deficiency as Condition Reports 18-13549 and 18-14426. Screening: The performance deficiency was minor because if left uncorrected it would not have led to a more significant safety concern and it did not adversely affect any cornerstone objectives. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000313/FIN-2018003-0630 September 2018 23:59:59Arkansas NuclearReactor Power Transient Caused by the Turbine Bypass Valve Failing OpenThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly pre-plan maintenance for the replacement of air supply tubing for turbine bypass valve CV-6687, which resulted in the failure of the air tubing, causing valve CV-6687 to fail open, which led to a manual reactor trip and a subsequent loss of the main condenser.
05000313/FIN-2018011-0130 September 2018 23:59:59Arkansas NuclearFailure to Properly Size the Unit 1 Emergency Diesel Generator Room Ventilation SvstemsAn NRC identified Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," was identified for failure to properly size the Unit 1 emergency diesel generator room ventilation systems to be capable of removing the design heat load during the most limiting design conditions while maintaining redundancy of the exhaust fans.
05000528/FIN-2018008-0530 September 2018 23:59:59Palo VerdeMinor ViolationFailure to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality as required by 10 CFR 50, Appendix B, Criterion VI. The team identified that the CAP procedure directed the use of the Cause Analysis Manual in performing some cause evaluations. This cause evaluation process is an activity affecting quality required by 10 CFR 50, Appendix B and the licensees Quality Assurance Program. The licensee failed to control the Cause Analysis Manual in accordance with the Palo Verde Nuclear Generating Station Operations Quality Assurance Program Description, Revision 0, Section 2.6, Document Control. The licensee documented this violation in Condition Report 18-13996. Screening: The performance deficiency is minor because if left uncorrected it would not have led to a more significant safety concern and it did not adversely affect any cornerstone objectives. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion VI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000313/FIN-2018011-0230 September 2018 23:59:59Arkansas NuclearFailure of Both Arkansas Nuclear One Units to Establish Adequate Corrective Actions Resulting in Excessive Instances of Damaged and Broken Internals of the Emergency Feedwater Pum o Turbine Steam Admission Check Valves.An NRC identified Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for failure to establish an adequate corrective action program and the resulting inability to correct a deficient system design which resulted in damaged and broken internals of the check valves admitting steam to the emergency feedwater turbine.
05000336/FIN-2018011-0130 September 2018 23:59:59MillstoneReviews of Incoming Industry Operation Experience Not CompletedThe inspectors identified that Millstone could not demonstrate that incoming industry operational experience reports (ICES) since 2015 had been properly reviewed for applicability to Millstone and for those items that were applicable, were evaluated and corrective actions developed as necessary as required by program guidance. A population of over 1600 ICES reports were identified where it could not be determined if required reviews were complete. Because there are parallel processes which may have reviewed these items, additional review is necessary to determine whether this issue represents a performance deficiency that is of more than minor significance. Therefore, this item is characterized as an unresolved item (URI). The purpose of the operational experience program is to identify conditions adverse to quality (CAQs) found at other plants, evaluate whether the concern is applicable to either Millstone unit, and evaluate and develop corrective actions for those CAQs when necessary. The inspectors noted that a performance improvement report (PIR) is automatically created for the Dominion fleet whenever an OPEX report is received (regardless of its source). Once the corporate PIR is generated, each site is required to check a box that it was received and also disposition it. The PIR remains opened until each site has completed this action. Prior to 2015, the corporate Operating Experience Coordinator would perform an applicability review and assign the remaining items to the site for further evaluation. When the corporate organization was reorganized, the headquarters review of OPEX became mostly administrative and the individual sites were expected to fully disposition the report. Since 2015, more than 1600 OPEX records were discovered that required disposition for Millstone. These records were still open and no records exist to show whether reviews were completed. Therefore it is uncertain if all applicable ICES reports were reviewed. Planned Closure Actions: The NRC will conduct a problem identification and resolution annual sample using NRC IP 71152 once Dominion has notified the NRC that they have completed their review of the 1600 ICES reports. Licensee Actions: Dominion wrote Condition Report (CR) 1105042 to capture the issue, conducted an investigation, and developed a plan to review the 1600 ICES reports which have no documented reviews. Dominion anticipates this review will be completed by the end of the first quarter of 2019.Corrective Action Reference: CR 1105042NRC Tracking Number: 05000336 & 05000423/2018-011-01
05000313/FIN-2018405-0130 September 2018 23:59:59Arkansas NuclearSecurity
05000313/FIN-2018003-0230 September 2018 23:59:59Arkansas NuclearFailure to Implement Welding Standard Guidance and Examination ProceduresThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to implement welding standard guidance and examination procedure guidance during the installation of the high pressure injection system drain line containing drain valves MU-1066A and MU-1066B. The drain line weld developed a crack that caused a leak shortly after plant startup that was determined to have been caused by grinding during the welding process, which was not permitted by the welding standard.
05000255/FIN-2018003-0130 September 2018 23:59:59PalisadesWire Not Landed on Safety Injection Initiation Relay CircuitThe inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018
05000313/FIN-2018011-0330 September 2018 23:59:59Arkansas NuclearFailure to Evaluate the Effects and the Suitability of Components in Containment from a Main Steam Line Break.The team identified an unresolved item (URI) related to the containment environment that would result from a main steam line break. Specifically, for ANO Unit 1 the licensee did not analyze the containment temperature, or evaluate the suitability of components in containment for the effects of a main steam line break (MSLB) accident. The Final Safety Analysis Report states, in part, that "At the end of Cycle 19, the original once through steam generators (OTSGs) were replaced. In support of Cycle 20 operation, an evaluation of the containment pressure/temperature response with the replacement OTSGs for loss of coolant accidents (LOCA) and MSLB was performed. For the MLSB, the containment pressure response with the replacement OTSGs was bounded by the current analysis. The post-MSLB temperature response w ith the replacement OTSGs would be worse. Entergy Operations, Inc. has adopted NUREG-0458 into the AN0-1 licensing basis which recognizes that the post-MSLB atmosphere may become superheated, but the temperature spike is of such short duration that the thermal lag of any SSC inside containment will not increase significantly. Consequently, the initial temperature peak does not define operating limits on any system, structure, or component (SSC) and the long-term containment temperature (which is essentially the saturation temperature) dominates the temperature response of SSCs. Therefore, as long as the peak MSLB pressure is less than the peak pressure following a LOCA, the temperature response of SSCs will still be defined by the LOCA." The NRC issued several bulletins subsequent to the issuance of NUREG-0458. Specifically IEB-79-01, as supplemented, and NRC Order CLI 80-21 state, in part, that "The Guidelines leave open the question of what standard will be applied to replacement parts in operating plants. Unless there are sound reasons to the contrary, the 1974 standard in NUREG-0588 will apply. The Guidelines and NUREG-0588 apply progressively less strict standards to the older plants. The justification for this position was not articulated at the time the older plants were grandfathered from the provisions of Reg. Guide 1.89." The NRC issued a Safety Evaluation Report to ANO, which states, in part, "A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, Section 50.49 of 10 CFR 50, specifies the requirements of electrical equipment important to safety located in a harsh environment. In accordance with this rule, equipment for Arkansas Unit 1 may be qualified to the criteria specified in either the DOR Guidelines or NUREG-0588, except for replacement equipment. Replacement equipment installed subsequent to February 22, 1983 must be qualified in accordance with the provisions of 10 CFR 50.49, using the guidance of Regulatory Guide 1.89, unless there are sound reasons to the contrary." The NRC issued Information Notice 85-39 states, in part, that the "Qualification of some replacement equipment was based on previously allowed DOR guidelines that stated "equipment is considered qualified for main steam line break environmental conditions if it was qualified for a loss-of-coolant accident environment in plants with automatic spray systems not subject to disabling single component failures." This basis of qualification is not acceptable without additional justification for replacement equipment that was procured and installed after February 22, 1983." The replacement steam generators have several design differences compared to the original steam generators. Specifically, the replacement steam generators were designed with larger secondary volumes, more tubes, flow-restricting venturis, and different materials (Alloy 690 vs. Alloy 600). Because the replacement steam generators were installed in 2005 (after 10 CFR 50.49 became effective on February 22, 1983) all replacement equipment must be qualified using the guidance of NUREG-0588 or Regulatory Guide 1.89. In addition, as stated above the licensee did not analyze or quantify the containment temperature that would result from a MSLB, and instead compared the containment pressures and the mass/energy releases that would result from a MSLB using the superseded guidance of NUREG-0458. The NRC team identified that there are several parameters that could have changed with the replacement steam generators which could impact the containment response. Specifically, input parameters such as: sub-compartment analysis, net positive suction head analysis, containment volume, heat sinks, properties of materials, heat transfer coefficients, initial conditions, and possibly cooling water temperature may affect the containment temperature response.
05000336/FIN-2018403-0130 September 2018 23:59:59MillstoneSecurity
05000530/FIN-2018003-0130 September 2018 23:59:59Palo VerdeFailure to Maintain Command and Control During a Feedwater Control Valve MalfunctionWhile reviewing the licensee response to a Unit 3 feedwater pump trip, reactor cutback, reactor trip, and main steam isolation system actuation on June 27, 2018, the inspectors identified that the licensee did not meet the command and control standards outlined in station Procedure 40DP-9OP02 Conduct of Operations, Revision 72. Specifically, senior reactor operators in the control room did not effectively coordinate manual main feedwater output adjustments in the control room or operator actions in the field in response to an apparent valve failure with the activities of non-licensed operators locally evaluating the equipment condition in the field. These uncoordinated actions resulted in a significant plant transient
05000313/FIN-2018003-0330 September 2018 23:59:59Arkansas NuclearFailure to Provide Complete and Accurate Information in a License Amendment Request to Change Emergency Action Level RequirementsThe inspectors identified a Severity Level IV non-cited violation because the licensee provided inaccurate information to the NRC in a license amendment request for an emergency action level scheme change. Specifically, the licensee provided information about the availability of the postaccident sampling system building radiation monitor and the Unit 1 level instrumentation that was material to the licensing decision, but not accurate. The NRC approved an emergency action level scheme change on November 9, 2012 (ADAMS Accession No. ML12269A455) to allow Arkansas Nuclear One to adopt the Nuclear Energy Institute (NEI) 99-01, Revision 5, scheme. Subsequently, the licensee identified that two of their current emergency action level thresholds could not be implemented in accordance with their emergency classification procedure: On May 26, 2017, Condition Report CR-ANO-2-2017-03161 documented that postaccident sampling system building radiation monitor 2RX-9840 should be removed from all regulatory commitments because the postaccident sampling system had been removed from service, and its building would not be monitored for radiological releases. Radiation monitor 2RX-9840 was being used as a means to evaluate emergency action levels AU1, AA1, AS1, and AG1. In addition, it was used in the loss/potential loss of containment (CNB6) for fission product emergency action levels. The condition report noted that requirements for the postaccident sampling system had been removed from Arkansas Nuclear One licenses in August 2000 and the licensee had abandoned the systems valves (March 2003, EC-ANO-1779), removed power from the postaccident sampling system ventilation system (January 2004), and made radiation monitor 2RX-9840 nonfunctional (May 2008, Condition Report CR-ANO-2-2008-01439 and Work Order 150817). On March 15, 2018, Condition Report CR-ANO-C-2018-01121 documented that the Unit 1 level instrumentation set point used in emergency action level CA1 was below the indicating range of the instrument. The emergency action level indicated that a loss of Unit 1s reactor vessel inventory was shown by an indicated level less than 368 feet, 0 inches. Therefore, the lowest level indicated on the instrument would be higher than the level used in making the emergency classification decision. The inspectors reviewed the licensees license amendment request, dated December 1, 2011 (ADAMS Accession No. ML113350317), Proposed Emergency Action Levels Using NEI 99-01, Revision 5, Scheme, and the licensees response to a request for additional information dated July 9, 2012, (ADAMS Accession No. ML12192A090) to determine whether the conditions identified in the corrective action program existed at the time the licensee requested the license amendment and whether the request correctly described the instruments. The inspectors identified: The December 1, 2011, submittal incorrectly indicated that radiation monitor 2RX-9840 was a viable means of classifying emergency action levels AU1, AA1, AS1, and AG1, as well as providing input for the evaluation of fission product barrier emergency action levels. In the response to NRCs request for additional information (RAI) dated July 9, 2012, the licensee provided additional details about the super particulate iodine noble gas (SPING) radiation monitors used in this application. Response to Question 3 associated with emergency action levels AA1, AS1, and AG1 stated: Each SPING is associated with a particular ventilation pathway and provides continuous monitoring of air discharged via the respective release pathway. The license reviewer concluded that all of the SPING monitors included in the license amendment request were operable and continuously monitoring the specified release pathways, thereby being capable of measuring the radiation levels described in the proposed emergency action levels. 17 The December 1, 2011, submittal indicated that loss of Unit 1 reactor vessel inventory for emergency action level CA1 was a vessel level less than 368 feet, 0 inches. This issue was NRC-identified because when the licensee identified the emergency action level errors, they took action to correct the errors, but failed to address the failure to ensure that technical information provided to the NRC in support of the license amendment request was complete and accurate in all material respects. Corrective Actions: To correct the Unit 1 reactor vessel level emergency action level threshold error, the licensee issued communications regarding correct application of the emergency action level on March 15, 2018, followed by implementation of a change to Procedure OP-1903.010, Emergency Action Level Classification, Revision 56, dated June 26, 2018, with the corrected level. The use of radiation monitor 2RX-9840 is being removed from the emergency action levels as part of an emergency action level scheme change submitted to the NRC on March 29, 2018 (ADAMS Accession No. ML18088B412 and ML18094A155). In the interim, the licensee issued communications to emergency director-qualified staff members to ensure they are aware of the error, how to address it if implementing emergency action levels, and to inform them of the corrective actions in progress. Additionally, the licensee issued Condition Report CR-ANO-C-2018-03597, dated September 13, 2018, for the incomplete and inaccurate emergency action level submission examples to address the completeness and accuracy issues identified by the inspectors.
05000313/FIN-2018003-0130 September 2018 23:59:59Arkansas NuclearFailure to Translate the Design Requirements into Instructions for Refueling Emergency Diesel GeneratorsThe inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate current design into instructions for Unit 1 and Unit 2 diesel fuel oil transfer system. Specifically, the licensee failed to translate the current diesel fuel oil transfer system design into instructions to refuel Unit 1 and Unit 2 safety-related fuel bunkers, T-57 and 2T-57, if the non-safety bulk diesel fuel oil tank T-25 was unavailable following a design basis event (e.g., tornado, external flooding, or earthquake) for which it was not designed to withstand.
05000336/FIN-2018003-0130 September 2018 23:59:59MillstoneFailure to Assure that Safety-Related Service Water Piping Conformed to the Procurement DocumentsThe inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, when the licensee failed to identify that a replacement service water pipe spool (JGD-1-25) was not in conformance with the American National Standards Institute (ANSI) B31.1 code, a condition of the purchase order, and was installed in the plant.
05000528/FIN-2018008-0130 September 2018 23:59:59Palo VerdeInadequate Corrective Actions For Missing Control Room Hand-Switch Operator KnobThe team reviewed a Green, NRC identified, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct the failures of multiple control room hand-switch operator knobs.
05000313/FIN-2018003-0530 September 2018 23:59:59Arkansas NuclearFailure to Maintain Main Feedwater Pump B Discharge Pressure in Band Caused a Reactor TripThe inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to implement Procedure OP-1102.002, Plant Startup, Revision 106. Specifically, control room operators failed to maintain main feedwater pump discharge pressure in the required band to control flow to the steam generators during a plant startup. As a result, the only operating main feedwater pump tripped on high discharge pressure, causing an automatic reactor trip.
05000389/FIN-2018003-0130 September 2018 23:59:59Saint LucieFailure to meet the Transient Combustible Requirements Specified by NFPA 805The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.48(c), National Fire Protection Standard NFPA 805, requirements. Specifically, the licensee failed to comply with transient combustible control requirements in high risk fire zones as required by NFPA 805 and implemented by licensee procedure ADM-19.03, Transient Combustible Control.
05000382/FIN-2018002-0130 June 2018 23:59:59WaterfordFailure to Ensure Appropriate Chemistry Controls on the Component Cooling Water Heat ExchangersThe inspectors reviewed a self-revealed, Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred because the licensee did not prescribe procedures for preventing fouling of the component cooling water heat exchangers that were appropriate to the circumstances. Specifically, the licensee did not require in its instructions for adding biocide to the auxiliary component cooling water system that additions be coupled with running the associated auxiliary component cooling water pump or other means of ensuring that the biocide would be sufficiently circulated through the system. As a result, on February 8, 2018, component cooling water heat exchanger B failed a performance test and therefore would not maintain required design basis temperatures under all accident conditions due to biological fouling.
05000255/FIN-2018415-0130 June 2018 23:59:59PalisadesSecurity
05000423/FIN-2018010-0530 June 2018 23:59:59MillstoneInadequate Test Control of ECCS Valve InterlocksThe team identified a finding of very low safety significance (Green) involving an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Specifically, Dominion did not ensure that all testing required to demonstrate that emergency core cooling system (ECCS) valve interlock circuits would perform satisfactorily was being performed. The team determined that certain interlocks associated with ECCS valve 3SIL*MV8804A control circuit were not properly tested to demonstrate that the valve would not open if interlocks had not been met or would open, when required, with minimum interlock requirements met during design basis accidents.
05000336/FIN-2018010-0130 June 2018 23:59:59MillstoneOver-Duty Breakers on Safety-Related Buses on Unit 2The team identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control. Specifically, Dominion incorrectly concluded that the 480V safety-related breakers were conforming to the plants licensing basis following their identification that the calculated short circuit fault current exceeded the breaker rating. Dominions evaluation failed to take into consideration that non-class 1E loads fed from safety-related buses must be isolated from the class 1E system by fully qualified safety-related isolation devices (breakers). Dominions design basis requires that a circuit fault on the non-class 1E side of the isolation device shall not cause the loss of the associated safety-related system
05000313/FIN-2018002-0130 June 2018 23:59:59Arkansas NuclearFailure to Implement Procedural Guidance to Close Spent Fuel Pool Cooler Outlet Crosstie ValveThe inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One (ANO) Unit 1 Technical Specification (TS) 5.4.1.a for the licensees failure to implement Procedure OP-1102.015, Filling and Draining the Fuel Transfer Canal, Revision 44. Specifically, operators failed to close spent fuel pool cooler outlet valve SF-9 while lining up to fill the fuel transfer canal (FTC) from the borated water storage tank (BWST). As a result, the licensee drained approximately 2600 gallons from the SFP to the FTC.
05000255/FIN-2018011-0230 June 2018 23:59:59PalisadesFailure to Set Action Levels to Ensure that the Assumptions in the Engineering Analysis Remain Valid

The inspectors reviewed a sample of equipment located in the fire areas selectedfor inspection to determine if the licensee had established a proper method of monitoring that equipment as required by NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required that, A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee utilized Procedure EN-DC-357, NFPA 805 Monitoring Program, Revision 2,to ensure that, the assumptions in the NFPA 805 engineering analyses remain valid by executing an effective and ongoing monitoring program.The inspectors selected the high pressure air compressor (C-6B) and high pressure safety injection pump (P-66B), both of which were located in the West Safeguards Room. The licensee considered these components to be high-safety significant (HSS) structures, systems, or components (SSCs). The licensee chose to monitor the unavailability of these components utilizing the Maintenance Rule (10 CFR 50.65).The licensee set the Maintenance Rule allowable unavailability action level threshold for the high pressure air compressorat 5E-2 (5percent)whereas they assumed in their fire PRA an unavailability of 9.86E-3 (approximately 1percent). For the high pressure safety injection pump the licensee set the Maintenance Rule allowable unavailability at 1.5E-2 (1.5percent) whereas they assumed in their fire PRA an unavailability of 6.32E-3 (approximately 0.6percent). The inspectors believed that by relying on the less conservative action level thresholds in the Maintenance Rule the licensee failed to ensure that the assumptions in the engineering analysis (fire PRA) remained valid.The licensee stated in Procedure EN-DC-357, Section 1.0, Purpose, that, The NFPA 805 Monitoring Program ensures that the assumptions in the NFPA 805 engineering analyses remain valid by executing aneffective and ongoing monitoring program. Under Section 3.0, Definitions, the licensee defined, Action Level Threshold, as, When establishing the action level threshold for reliability and availability, the action level should be no lower than the Fire Probabilistic Safety Analysis (also called fire PRA) assumptions. The licensee stated in Section 5.3.3(c) that, If HSS SSCs have been identified in using the Maintenance Rule guidelines, the associated SSC specific performance criteria may be established as in the Maintenance Rule, provided the criteria are consistent with the Fire Probabilistic safety Analysisassumptions... The inspectors believed that Procedure EN-DC-357 required the licensee set the action level thresholds no lower than the fire PRA assumptions. Procedure section 5.3.4(b)(1) required that HSS equipment that is not sufficiently tracked in the Maintenance Rule be added to the NFPA 805 Monitoring Database. The licensee did not add the high pressure air compressor and the high pressure safety injection pump into the NFPA 805 Monitoring Database. In the SER 2015-2-27 dated February 27, 2015, in which the staff approved the licensee NFPA805 License Amendment Request, the staff noted that the licensee will develop an NFPA 805 Monitoring Program consistent with Frequently Asked Question (FAQ)10-0059. The staff also noted that the stated development of the Monitoring Program would include a review of existing surveillance, inspection, testing, compensatory measures, and oversight

8processes for adequacy. The staff concluded in SER 2015-2-27 that since the final values for availability and reliability, as well as the performance criteria for the SSCs being monitored, have not been established for the Monitoring Program as of the date of this SER, completion of the licensee's NFPA 805 Monitoring Program is an implementation item. Furthermore, the staff concluded that there is reasonable assurance that the licensee will develop a Monitoring Program that meets the requirements specified in Sections 2.6.1, 2.6.2, and 2.6.3 of NFPA 805Section 2.6 of NFPA 805 stated in part that, Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee interpreted this statement to mean that utilizing the existing Maintenance Rule unavailability values is consistent with its commitment in SER 2015-2-27 and would allow the site to appropriately monitor the availability and reliability of fire protection systems and features. The licensee also performed sensitivity studies on the differences in the unavailability values of fire protection systems and features between the Maintenance Rule criteria and the fire PRA values and determined that they were not risk-significant. The inspectors questioned the appropriateness of the licensees interpretation of assumptions as described in Section 2.6 of NFPA 805 above. The inspectors believed that the licensee should monitor the unavailability of fire protection systems and features utilizing the same values as thosedocumented in the fire PRA associated with the NFPA 805 License Amendment Request. The licensee further stated that they were waiting for guidance from the NRCs Office of Nuclear Reactor Regulation and the industry who were working on revising guidance in FAQ10-0059, NFPA 805 Monitoring, to determine if they needed to change their approach. That guidance document was in the process of being revised during the inspection. The inspectors needed to determine if the licensees approach to monitoring the availability and reliability of the fire protection systems and features using the Maintenance Rule monitoring values in order to ensure that the assumptions in the engineering analysis remained valid was an acceptable approach.Planned Closure Action(s): The inspectors will await clarification from the Office of Nuclear Reactor Regulation in order to determine if a performance deficiency exists.Licensee Action(s): The licensee plans to follow the resolution of FAQ 10-0059, Revision 6, and take the appropriate corrective actions based on the guidance provided in that FAQ.
05000528/FIN-2018002-0130 June 2018 23:59:59Palo VerdeFailure to Re-baseline Valve Stroke Times as Required by ASME OM CodeThe inspectors identified a Green, non-cited violation of Palo Verde Technical Specification 5.5.8, Inservice Testing Program, which requires inservice testing of ASME Code Class 1, 2, and 3 components in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). On October 22, 2017, the licensee failed to establish new stroke time reference values for Unit 1 safety injection (SI) valve 660 following maintenance which could affect the valves performance
05000336/FIN-2018010-0430 June 2018 23:59:59MillstoneFlood Seals Not Installed in Unit 2 A EDG and Auxiliary Building PenetrationsThe team identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XIV, Corrective Actions. Dominion identified a condition adverse to quality but did not correct the condition. Specifically, Dominion performed evaluations and walk downs in 2012 and 2016 to validate that all necessary flood seals for design basis and beyond design basis flood events had been properly installed. Dominion determined that they could not verify 50 wall penetrations had seals installed and entered the deficiency into the corrective action program. The team noted that an electrical conduit that passed through a Unit 2 A emergency diesel generator (EDG) building exterior wall, located below the design basis flood height, was one of the penetrations in question. During the inspection, following NRC questions, Dominion removed the electrical conduit cover plate and confirmed that a seal was not installed.
05000530/FIN-2018002-0330 June 2018 23:59:59Palo VerdeFailure to Assess the Operability of a Degraded or Nonconforming Structure, System, or ComponentThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to evaluate conditions adverse to quality for impacts on the operability of the essential spray ponds.
05000382/FIN-2018002-0230 June 2018 23:59:59Waterford10 CFR 50.59 Evaluation Associated with Emergency Feedwater Logic ModificationThe licensee changed the emergency feedwater logic, as described in the Updated Final Safety Analysis Report (UFSAR), Section 7.3.1.1.6, from flow control mode to level control mode during a safety injection actuation signal. To accomplish this change, the licensee had to modify the following logic system signals and setpoints: steam generator critical level, steam generator lo level, steam generator lo-lo level, safety injection actuation, control board manual control, and the steam generator lo-lo level annunciator. The NRC team questioned whether the emergency feedwater modification required additional information to determine if the 10 CFR 50.59 evaluation was adequate, or if NRC approval was needed for the change. Specifically, the NRC team questioned if the emergency feedwater logic change: used a method of evaluation other than what was described in the UFSAR (e.g. the use of the TRANFLOW program) or would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a system important to safety. Specifically, because the emergency feedwater logic change introduced the potential to overcool the reactor, and substituted a previous automatic action for manual operator action, the NRC team questioned if the change and associated 50.59 evaluation addressed these concerns. Planned Closure Actions: The NRC and the licensee are working to gather more information related to the Final Safety Analysis Report-described methods for steam generator analyses and if the change resulted in a more-than-minimal increase in risk. Specifically, the licensee plans to provide an analysis that demonstrates the emergency feedwater logic change would not result in a more than minimal increase in the likelihood of an overcooling accident. Licensee Actions: The licensee has implemented a compensatory measure to take manual control of the emergency feedwater system during a safety injection signal such that an overcooling event will be prevented. Corrective Action References: CR-WF3-2017-06067, CR-WF3-2017-05882, CR-WF3-2017-05173
05000255/FIN-2018011-0130 June 2018 23:59:59PalisadesFailure to Maintain Adequate Fire Protection System Functional Test ProcedureThe inspectors identified a finding of very-low safety significance and associated violation of Technical Specification 5.4.1, Procedures,for the licensees failure to maintain fire protection system functional test procedure. Specifically, the licensee failed to maintain Procedure RO-52, Fire Suppression Water System Functional Test and Fire Pump Capacity Test, by failing to include appropriate acceptance criteria in the procedure to demonstrate fire protection system functionality.
05000336/FIN-2018410-0130 June 2018 23:59:59MillstoneSecurity
05000336/FIN-2018010-0330 June 2018 23:59:59MillstoneFailure to Correct Part 21 Power Supply DefectsThe team identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, Dominion did not accomplish repairs to safety-related power supplies in accordance with instructions and procedures. The team identified that actions taken by Dominion to address Part 21 Report #48863, Foxboro Power Supply Potential Failures due to Defective Tie Wraps and Holder, were performed without procedure or engineering evaluations and the work activities performed were not documented. Specifically, instrumentation and control technicians altered the safety-related power supplies without approved design documents, plant procedures, or work orders, and records of the completed activities were not available
05000528/FIN-2018002-0230 June 2018 23:59:59Palo VerdeFailure to Implement and Maintain Procedures Regarding Breathing Air QualityThe inspectors identified a Green, non-cited violation of 10 CFR 20.1703 for failing to implement and maintain written procedures to ensure that respiratory protection equipment (air compressors and bubble hood suites) supplied respirable air of grade D quality or better to radiation workers.
05000382/FIN-2018404-0130 June 2018 23:59:59WaterfordSecurity
05000423/FIN-2018010-0230 June 2018 23:59:59MillstoneOver-Duty Breakers on Safety-Related Bus 34C on Unit 3The team identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, Dominion did not adequately evaluate the results of the Unit 3 short circuit calculations for the 4.16 kV breakers. Dominions evaluation of the short circuit calculation results did not identify that the breakers were non-conforming to the licensing basis. The teams review of the calculation results found that the momentary and interrupting duty ratings of the 4kV safety-related breakers associated with Bus 34C were not within their short-circuit ratings when evaluated under design fault condition and, therefore, not in accordance with the licensing basis of the plant.
05000368/FIN-2018001-0231 March 2018 23:59:59Arkansas NuclearFailure to Preplan and Perform Service Water Pre-Screen MaintenanceThe inspectors reviewed a self-revealed,non-cited violation and associated finding of Arkansas Nuclear One, Unit 2, Technical Specification 6.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly preplan pre-screen cleaning maintenance, causing the trainB service water system to become inoperable
05000255/FIN-2018001-0231 March 2018 23:59:59PalisadesLicensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliances that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The EGM applies specifically to a structure, system, or component (SSC) that is determined to be inoperable for tornado-generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Palisades, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. On December 7, 2017, and as supplemented on January 18, 2018, Palisades submitted a request to the NRC to extend the enforcement discretion from June 10, 2018 to June 10, 2020 (ML17341A415 and ML18018A328, respectively). By letter dated February 16, 2018, the NRC granted the request to extend enforcement discretion until June 10, 2020 (ML18046A675). The EGM permitted NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provide additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within about 60 days of issue discovery. In accordance with the EGM, the comprehensive compensatory measures are toremain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Palisades was licensed prior to issuance of Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC). Specifically, GDC 2, Design Bases for Protection Against Natural Phenomena, and GDC 4, Environmental and Dynamic Effects Design Basis, discuss how SSCs important to safety shall be designed to protect against natural phenomena, such as tornadoes and shall be adequately protected against the dynamic effects of tornadoes, including protection against missiles. Palisades site-specific licensing bases compliance with GDC 2 and GDC 4 are described in the Updated Final Safety Analysis Report (UFSAR) Sections 5.1.2.2 and 5.1.2.4. Palisades protection of SSCs against tornado-generated missiles is also discussed in UFSAR Section 5.5, Missile Protection. On January 31, 2018, the licensee initiated condition report (CR) CRPLP201800556, which identified a nonconforming condition in the Palisades licensing basis. Specifically, the surge line from the component cooling water (CCW) surge tank to the CCW suction line was identified to be potentially vulnerable to a tornado missile through a doorway. The licensee previously identified a CCW system-related vulnerability on March 29, 2017. The March 29, 2017 CCW vulnerability and five additional vulnerabilities of other SSCs, which all received enforcement discretion, are documented in NRC Inspection Report 05000255/2017002 (ML17220A349). The licensee assessed this new vulnerability and concluded that previously established compensatory measures for the CCW system were adequate and no additional comprehensive compensatory actions were required. Therefore, the licensee declared the SSC operable, but nonconforming because no additional compensatory measures designed to reduce the likelihood of tornado-generated missile effects were required and the previously implemented compensatory measures were still in place. Corrective Action: The licensee documented the condition of the SSC in the CAP and documented the SSC as operable but nonconforming.Corrective Action Reference: CRPLP201800556 Enforcement: Violation: Enforcement discretion was applied to the required shutdown actions of the following Technical Specification (TS) Limiting Conditions for Operation (LCOs): TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); andTS 3.7.7, Component Cooling Water (CCW) System.Severity/Significance: The subject of this enforcement discretion associated with tornado missile protection deficiencies was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in EGM 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance (ML16355A286). 11 Basis for Discretion:The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory actions to resolve the nonconforming conditions within the required 60 days. These comprehensive measures were to remain in place until permanent repairs were completed, which for Palisades were required to be completed by June 10, 2020, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed.The disposition of this enforcement discretion closes LER 05000255/201700101, Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions.
05000528/FIN-2018001-0131 March 2018 23:59:59Palo VerdeInadequate Post Maintenance Test Instructions for Diesel Fuel Oil Transfer PumpThe inspectors reviewed a self-revealed, Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to prescribe appropriate work instructions for maintenance on the Unit 1 diesel fuel oil transfer pump A. Specifically, following power cable maintenance on November 9, 2017, the instructions for conducting a post-maintenance test for the transfer pump were inadequate to detect a high resistance connection in the associated motor control center.
05000382/FIN-2018001-0131 March 2018 23:59:59WaterfordFailure to Obtain NRC Staff Authorization Prior to Changing a Procedure that Impacts Implementation of Technical SpecificationsThe inspectors identified a Severity Level IV, non-cited violation of 10CFR50.59, Changes, Tests, and Experiments, Section (c)(1), for the licensees failure to submit and obtain authorization prior to implementation procedures described in the Final Safety Analysis Report
05000335/FIN-2018001-0131 March 2018 23:59:59Saint LucieImproper Evaluation of LCV-9005 position setpoints Leads to AFASOn November 19, 2013, during reactor startup activities, feedwater bypass valves, A (LCV-9005) and B (LCV-9006), were found to be operating at different throttle positions while maintaining their respective steam generator water levels. Valves LCV-9005 and 9006 were both originally installed in April 1978. LCV-9005 was replaced in 1994, with an equivalent valve, due to obsolescence. The original valve had a full open stroke length of 1.5 inches (in.), while the new equivalent valve had a full open stroke length of 2 in. to provide the same flow as the original valve. When installed, LCV-9005 was set up to limit its stroke length to 1.5 in., matching the replaced valve, and the associated drawings were never revised to show that the new valve had a full 2 in. open stroke length. In 2009, the distributed control system (DCS) was installed utilizing these drawings and was setup under the assumption that both valves, LCV-9005 and LCV-9006, were the same model valves and stroke lengths.The DCS system was designed to provide a signal to throttle the feedwater bypass valves following a reactor trip to 20 percent open to provide approximately 5 percent feed flow in order to recover steam generator water levels utilizing main feedwater. During Unit 2 startup activities in November 2013, the licensee noted a discrepancy in the valve positions for LCV-9006 and LCV-9005 when they were providing steam generator water level control. The licensee placed the issue in the corrective action program under Action Request (AR) 1921720 and determined that it was necessary to evaluate a revision of the LCV-9005 DCS setpoint, which was accomplished by an engineering condition evaluation under AR 1925428. The engineering condition evaluation was inadequate in that it failed to recognize the differences in the two different model valves, and therefore failed to provide adequate corrective actions to address performance issues associated with these differences.The final recommendation from AR 1925428 was that the current LCV-9005 setting did not impose any risk to the plant operation, as the 2A steam generator level had been within acceptable range with no control room alarm observed. Therefore, no setpoint change was required at that point.On October 26, 2017, following a Unit 2 trip, LCV-9005 was sent a digital DCS demand signal to be 20 percent open. Since the valve was locally set to have a maximum stroke of 1.5 in. instead of 2 in. open, the actual flow through the valve was less than 5 percent. This resulted in flow lower than needed to maintain 2A steam generator level, and caused level to lower, which eventually resulted in an actuation of the A train auxiliary feedwater actuation system (AFAS). Corrective Action(s):The licensee implemented corrective actions to: 1) properly set up LCV-9005 in order for it to have a full stroke length of 2 inches so that it could provide the required feedwater flow and, 2) update associated drawings to include correct stroke lengths.Corrective Action Reference(s): This issue was entered into the licensees CAP as AR 2232869
05000528/FIN-2018012-0131 March 2018 23:59:59Palo VerdeFailure to provide adequate guidance to personnel to assure degraded or deficient emergency lighting required for post-fire safe shutdown was corrected in a timely mannerThe licensee provides emergency lighting for access and egress paths, and to illuminate required safe shutdown components to safely shut down the reactor in case of fire requiring control room evacuation, in accordance with 10 CFR 50, Appendix R, Section III.J., Procedure 40DP-9ZZ16, Administrative Controls for Appendix R Equipment, Revision 13, Steps 4.5.1 and 4.6.1, limit the out-of-service time for Appendix R equipment and require restoring the equipment to service within 30 days. The licensee provides guidance to operations department watch standers to check for deficient lighting throughout the plant. Procedure 40DP-9OP20, Watch Standing Practices, Revision 48, Step 4.3.1.1 instructs watch standers to check for adequate or sufficient lighting, among other items during routine tours of the plant. On January 30, 2018, during the walkdown of the control room evacuation due to fire Procedure 40AO-9ZZ19, Control Room Fire, Revision 35, the inspectors identified multiple examples of Appendix R emergency light fixtures that were not functional. The licensee confirmed no condition report had documented these deficient emergency lights at that time. During operator rounds through the area, with offsite power available, the area had adequate and sufficient lighting from the normal AC lighting system. The guidance provided in Procedure 40DP-9OP20 was not adequate to instruct watch standers to identify when an Appendix R emergency light is not functional and to promptly identify the failure to ensure timely restoration of the light.Corrective Actions: The licensee documented the emergency light deficiencies in the corrective action program to initiate repairs, and also issued an Operations Communication Newsflash titled, Actions needed on Emergency (App R) Lighting, dated February 1, 2018. The Newsflash required all operations crews to review procedures for identifying emergency lighting deficiencies and to focus on the emergency light fixtures. The licensee has 30 days to repair and restore the emergency lights to operation. The licensee has a standing compensatory measure for emergency lighting requiring operators to obtain flashlights from the emergency storage locker during a control room evacuation due to fire. The licensee initiated a revision request to provide additional guidance in Procedure 40DP-9OP20, concerning the Appendix R emergency lighting.
05000335/FIN-2018411-0131 March 2018 23:59:59Saint LucieSecurity
05000313/FIN-2018001-0131 March 2018 23:59:59Arkansas NuclearFailure to Establish Adequate Criteria for Flood Seal TestingThe inspectors identified a Green finding and associated non-cited violation of Unit1 Technical Specification 5.4.1.a and Unit 2 Technical Specification 6.4.1.a for the licensees failure to establish the criteria for ensuring the necessary conditions existed for a successful test of hatch flood seals. Specifically, Procedure OP 1402.240, Inspection of Watertight Hatches, Revision 1, did not contain adequate guidance to ensure that the auxiliary building was at a lower pressure than the turbine building such that puffing smoke on the turbine building side would allow a seal leak to be detectable.
05000335/FIN-2018450-0131 March 2018 23:59:59Saint LucieSecurity
05000317/FIN-2018001-0131 March 2018 23:59:59Calvert CliffsFailure to Conduct Adequate Radiation Surveys and Evaluate Potential Radiological HazardsAself-revealed Green non-cited violation(NCV)of Title 10 Code of Federal Regulations(10 CFR) 20.1501, Surveys and Monitoring: General, was identified when Exelon failed to perform adequate surveys of the 11 reactor coolant pump bay area following the aggregation of 25 high dose-rate in-core detectors in one area of the flooded refueling cavity, which is adjacent to the pump bay. Surveys were not performed as required after radiological conditions changed and radiological hazard mitigation measures, such as locking and controlling access in accordance with Exelon procedures, were not implemented, resulting in accessible dose-rates of up to 2,000 millirem per hour(mrem/hr)in the pump bay
05000255/FIN-2018001-0131 March 2018 23:59:59PalisadesFailure to Maintain an Appropriate Documented Work Instruction for Reassembly of Primary Makeup Tank Makeup Control Valve CV2008A self-revealed Green finding and an associated NCV of Technical Specification 5.4.1, Procedures, was identified for the licensees failure to have an adequate maintenance work instruction for the reassembly of Primary Makeup Tank Makeup Control Valve CV2008. Specifically, because a previous CV2008 maintenance activity failed to properly set the height of the CV2008 jam nuts, the valve guide key fell out of place and in December 2017, CV2008 was unable to be manually stroked during surveillance testing
05000335/FIN-2017004-0331 December 2017 23:59:59Saint LucieFailure to Identify and Correct a Condition Adverse to QualityThe NRC-identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for failure to identify and correct a condition adverse to quality. The licensee failed to identify that their procedures lacked actions to install control power jumpers that are required to defeat the reactor coolant systems (RCS) pressure interlocks for the shutdown cooling (SDC) suction line motor operated valves (MOVs) when aligning the plant for hot leg injection (HLI) and then correct the condition. Following the identification of this procedural vulnerability, the licensee fabricated control power jumpers and revised procedure 1-GME-100.03, Installation and Removal of Temporary Power Jumpers for MOV V3481, V3652, V3432 AND V3444, to provide direction for installation of power jumpers. In addition, the licensee performed a more detailed failure modes and effects analysis to ensure that the revised procedures accounted for all possible single failures. This issue has been entered into the licensees corrective action program (CAP) as CR 2217631.The PD was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone objective of ensuring the capability of the low pressure safety injection (LPSI) system to perform its required long term cooling safety function (HLI). The condition was evaluated by a Regional Senior Reactor Analyst and determined to have very low safety significance (Green) based on the low likelihood of a loss of coolant accident (LOCA) and low likelihood of electrical failures requiring jumpers to be installed. This issue and corrective actions were documented in the licensees CAP as Action Request (AR) 2217631. This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance.
05000382/FIN-2017008-0231 December 2017 23:59:59WaterfordFailure to Meet RG 1.9 Emergency Diesel Testing Requirements during Surveillance Test Results in Missed SurveillanceThe team identified a Green non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification Limiting Condition for Operation 3.8.1.1 for failure to maintain operability of two separate independent diesel generators. Specifically, on May 23, 2017, the licensee failed to verify that the train A emergency diesel generator energized all auto-connected shutdown loads through the load sequencer and operated for greater than or equal to five minutes in accordance with Technical Specification Surveillance Requirement 4.8.1.1.2.
05000255/FIN-2017007-0331 December 2017 23:59:59PalisadesContainment Dome Truss AnalysisThedome truss system was originally designed to support the containment liner plate and wet concrete during the construction of the containment dome (i.e., the liner plate initially acted as a form and the truss supported the form). After the concrete cured, the dome truss system was lowered away from the liner and was used to support the safety injection tanks(SITs) and CS system piping and their associated supports. The CS and SIT systemsare both safety-related which were required to be evaluated for seismic loads (self-weight and externally applied loads). The dome truss system would have alsobeen required to be evaluated for seismic loads. The UFSAR,Section 6.1,described the safety-related design function of the SITsystemwas to prevent fuel and cladding damage that could interfere with adequate emergency core cooling, and to limit the cladding-water reaction to less than approximately 1percentfor all break sizes in the primary system piping up to and including the double-ended rupture of the largest primary coolant pipe, for any break location, and for the applicable break time. Also,the SITsystem also functions to provide rapid injection of large quantities of borated water for added shutdown capability during rapid cooldown of the primary system caused by a rupture of a main steam line. UFSAR Section 6.2.1 described the safety-related design function of the CSsystem was to limit the containment building pressure rise and reduce the airborne radioactivity in containment by providing a means for spraying the containment atmosphere after occurrence of a LOCAor a main steam line break.The inspectors requested the design basis analysis of the dome truss system that considers the LOCA loading on the dome truss system as well as the seismic loading due to the applied design loads from the CS and SITsystem. During the time of the inspection, the licensee was unable to locate the dome truss analysis. In response to the inspectors concern, the licensee entered the issue into their CAP asCR 2017-05016, Dome Trusses, dated November 1, 2017. The licensee is investigating the containment dome truss analysis further with the vendor of the dome truss system.This issue is a URI pending additional inspector review of the design basis analysis for the containment dome truss system. (URI 05000255/2017007-03; Containment Dome Truss Analysis)