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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000389/FIN-2018003-01 | 2018Q3 | Saint Lucie | Failure to meet the Transient Combustible Requirements Specified by NFPA 805 | The 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. |
| 05000327/FIN-2018003-01 | 2018Q3 | Sequoyah | Licensee-Identified Violation | 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. |
| 05000369/FIN-2018003-01 | 2018Q3 | Mcguire McGuire | Failure to Adequately Document the Basis for a Change to the Emergency Plan | The inspectors identified a SL IV NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(3), for changes made to the McGuire Nuclear Station (MNS) Radiological Emergency Plan (E-Plan) that failed to demonstrate the changes would not reduce the effectiveness of the E-Plan. Specifically, the licensee did not provide an adequate analysis to determine that the removal of specific procedure references was not a reduction in effectiveness of the MNS E-Plan |
| 05000269/FIN-2018002-02 | 2018Q2 | Oconee | Failure to Coordinate a No-later-than Arrival Time for the Shipment of a Category 2 Quantity of Radioactive Material | The inspectors identified aSeverity Level IV NCV of 10 CFR 37.75(b) when the licensee failed to coordinate a no-later-than arrival time for a Category 2 shipment of radioactive material. Specifically, the licensee failed to recognize that a package of primary resin contained a Category 2 quantity of Cobalt-60 prior to shipment, and therefore failed to arrange a no-later-than arrival time with the receiving licensee. |
| 05000287/FIN-2018002-01 | 2018Q2 | Oconee | Failure to Perform ISI General Visual Examinations of Containment Moisture Barrier Associated with Containment Liner Leak Chase Test Connection Piping | The inspectors identified a Green NCV of 10 CFR Part 50.55a, Codes and Standards, involving the licensees failure to properly apply Subsection IWE, of ASME Section XI, for conducting general visual examinations of the leak chase test connection piping at the concrete floor interface which provides a moisture barrier to the containment liner seam welds. |
| 05000321/FIN-2018001-02 | 2018Q1 | Hatch | Licensee-Identified Violation | 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: Hatch Nuclear Plant Technical Specification (TS) 5.7.2 states in part, areas with radiation levels greater than 1000 mRem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in 1 hour measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry.Contrary to the above, February 6, 2018, the licensee identified dose rates of 72 Rem/hr on contact, and 3.9 Rem/hr at 30 cm on the U-1 bottom head drain valve located in the 127 foot elevation of the Subpile room, in the Unit 1 Drywell. For approximately 4 hours, the entrance to the room was not locked or continuously guarded to prevent unauthorized entry as required by TS 5.7.2. Significance/Severity: The finding was of very low safety significance (Green) because it was not an as low as reasonably achievable (ALARA) planning issue, there was no overexposure nor potential for an overexposure, and the licensees ability to assess dose was not compromised.Corrective Action Reference(s):The licensee identified and documented the failure to control access to the Lock High Radiation Area (LHRA) in Condition Report 10458608. |
| 05000321/FIN-2018001-01 | 2018Q1 | Hatch | Failure to comply with Type B shipping container Certificate of Compliance (CoC) requirements. | An NRC Identified Green NCV of 10 Code of Federal Regulations (CFR)71.17, General license: NRC-approved package, was identified for the licensees failure to comply with the Type B shipping container Certificate of Compliance (CoC) requirements. 10 CFR 71.17(c)(2)states, in part, that a holder of a General license to utilize an NRC-approved package shall comply with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part. Specifically, on several occasions the licensee placed in transit Type B containers which did not pass the CoC leak test requirement(s). |
| 05000324/FIN-2018001-01 | 2018Q1 | Brunswick | Inadequate Instruction to Perform Inspections on Emergency Ventilation Dampers | A self-revealing Green NCV of TS 5.4.1a, Procedures, was identified when the licensee failed to properly provide adequate work instructions associated with the control room emergency damper inspections. Specifically, the licensee disconnected the damper air supply line without adequate work instruction guidance, which caused a loss of Control Building Heating, Ventilation and Air Conditioning (HVAC) and Control Room Emergency Ventilation (CREV) Systems resulting in a safety system functional failure. |
| 05000269/FIN-2017004-01 | 2017Q4 | Oconee | Failure to Identify Sensitive Equipment During Modification Results in Loss of Safety Function | A self-revealing Green non-cited violation (NCV) of Oconee Nuclear Station Technical Specification (TS), Section 5.4, Procedures, was identified for the licensees failure to identify sensitive equipment in a work area that warranted implementation of compensatory measures as required by station procedure AD-EG-ALL-1180, Engineering Change (EC) Walkdowns. During the design and planning phase of a station modification, the licensee failed to identify sensitive components located in the subject work area and subsequently failed to implement adequate protective measures as defined in station procedures to prevent plant impacts during modification installation. The licensee entered this issue into their corrective action program (CAP) as nuclear condition report (NCR)02131608 and implemented corrective actions to identify other positionable components required for emergency power source operability that would require the use of protective measures, as defined by AD-OP-ALL-0204, Plant Status Control, in order to prevent inadvertent operation. The licensee created a formal Engineering department communication which included lessons learned from the event and familiarization with the EC walkdown checklist. The signs on the governor actuator cabinets were also revised to emphasize the sensitive nature of the equipment. The licensees failure to properly identify sensitive equipment and implement compensatory measures to prevent plant impacts as required by station procedure AD-EG-ALL-1180 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the loss of the emergency AC power path function for 11 hours and 31 minutes. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. Inspection Manual Chapter 0609, Appendix A required a detailed risk evaluation because the finding represented a loss of system and/or function. A regional senior reactor analyst (SRA) performed the detailed risk evaluation using SAPHIRE Version 8.1.6 and a modified Version 8.50 of the SPAR Model for Oconee. The SRA developed two change sets to model the total exposure time for the finding. The first simulated a common cause failure of both Keowee units with an exposure time of 7 hours. The second simulated the failure of both Keowee units while the standby buses were energized by the Lee Station for 5 hours. The result was less than 1E-6 for each Oconee unit, which would be a finding of very low significance (Green). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the area of human performance, in that the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5) |
| 05000287/FIN-2017004-02 | 2017Q4 | Oconee | Failure to Properly Risk Screen Work Within Two Feet of a Single Point Vulnerability Component | A self-revealing Green NCV of Oconee Nuclear Station TS, Section 5.4, Procedures, was identified for the licensees failure to identify and properly risk screen work within 2 feet of a single point vulnerability (SPV) component in accordance with procedure AD-OP-ALL-0201, Protected Equipment. Specifically, the transmission and Oconee organizations failed to recognize that planned maintenance on a breaker in the 525 kilovolt (kV) switchyard was within 2 feet of an SPV component and, as a result, appropriate planning and oversight were not in place to prevent a plant trip during maintenance activities. The licensee entered this issue into their CAP as NCR 02138958. Corrective actions included revisions to station and transmission procedures to ensure inclusion of appropriate SPV program information, addition of the SY special emphasis code to all switchyard type work which require coordination of transmission resources, and the addition of the T1 trip/transient risk special emphasis code to all breaker failure relays in the 230 kV and 525 kV switchyard cabinets containing SPV components.The licensees failure to identify and properly risk screen the planned maintenance on PCB-57 as work within 2 feet of an SPV component in accordance with AD-OP-ALL-0201 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, human errors led to a Unit 3 main generator lockout, which resulted in a reactor trip. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. The inspectors determined the finding was of very low safety significance (Green) because the finding did not represent a transient initiator that caused both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (i.e. loss of condenser, loss of feedwater). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the human performance area, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5) |
| 05000348/FIN-2017004-04 | 2017Q4 | Farley | Licensee-Identified Violation | 10 CFR 50.55 (a)(b)(5)(i) required in part that licensees must apply the most recent version of ASME BPV Code cases listed in Regulatory Guide 1.147, Revision 17. Contrary to the above, the licensee failed to perform augmented re-examinations on a 30-day periodicity as required by ASME Code Case N-513-3. A through-wall pinhole leak on the Unit 2 Train A Service Water strainer backwash piping was documented in condition report (CR) 10234480 on June 10, 2016. The service water system provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. The backwash piping is safety-related ASME Section III, Class 3 piping. An Immediate Determination of Operability Evaluation (IDO) was performed declaring the strainer operable but degraded non-conforming (OBDN). The licensee followed the guidance of ASME Code Case N-513-3, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping, Section XI, Division 1. The code case requires that an additional five similar susceptible locations be identified and inspected to ensure that another flaw does not exist. In addition to the expanded scope, the code case requires that frequent periodic inspections of no more than30-day intervals shall be used to determine if the flaws are growing to an unacceptable size. An additional CR (10236417) was initiated on June 15, 2016, to request work orders for inspection of these five locations. A total of three examinations were performed on a 30-day periodicity, the last being completed on August 22, 2016. CR 10416364 was initiated on October 5, 2017, documenting that no re-examinations on a 30-day periodicity were performed on the original leak location and the five additional locations since August 22, 2016. The ultrasonic examination was completed on October 5, 2017, and the degraded backwash piping was removed and replaced with new piping by WO SNC795917 on October 28, 2017. This finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency, it did not represent a loss of system safety function of a single train for greater than its TS allowed outage time, and it did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events. This finding was entered into the licensees CAP as CR 10416364. |
| 05000348/FIN-2017004-03 | 2017Q4 | Farley | Failure to Follow Procedure Resulted in Inoperable TDAFW pump | A self -revealing NCV of Technical Specification (TS) 5.4.1.a, Procedures, was identified when the Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) uninterruptible power supplies (UPS) swapped to a bypass power source during maintenance on November 5, 2017. As a result, the TDAFW pump was rendered inoperable. Failure to follow licensee procedure FNP-1-EMP-1352.01, TDAFW UPS Battery Weekly Battery Inspection, Version 19, as written was a performance deficiency. The operability of the TDAFW pump UPS was restored after approximately 3 hours. The licensee entered this issue into their Corrective Action Program (CAP) as Condition Report (CR) 10427370.The finding was more than minor because it was associated with the equipment performance attribute of the mitigating system cornerstone and adversely affected that cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences since the TDAFW pump was rendered inoperable. The significance of this finding was evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for findings at Power, dated June 19, 2012. This finding was determined to be of very low safety significance (Green) because all of the mitigating systems screening questions were answered NO. The inspectors determined the finding had a cross-cutting aspect of Avoid Complacency in the Human Performance area because the individuals involved in this maintenance did not recognize or plan for the possibility of mistakes and appropriate error reduction tools were not implemented. (H.12) |
| 05000364/FIN-2017004-02 | 2017Q4 | Farley | Failure to maintainan operable Oil Collection System on RCP 2B | An NRC-identified NCV of 10 CFR 50.48(c) and National Fire Protection Association Standard 805 (NFPA 805), Section 3.3.12, was identified for the licensees failure to maintain the Unit 2 RCP 2B oil collection system in an operable condition to perform its design function. Specifically, the licensee failed to ensure that the RCP 2B OSPS oil lift system enclosure collected all oil leakage from all potential leakage sites, including the oil lift system. The licensees failure to maintain the Unit 2 RCP 2B oil collection system in an operable condition to perform its design function was a performance deficiency. The licensee initiated CR 10428611, and determined an oil leak was not active. Another CR was initiated (10446206) to inspect and, if needed, repair this area at the next available opportunity.The finding was more than minor because if left uncorrected, the performance deficiency would have the potential to become a more significant safety concern. Specifically, failing to ensure that the RCP 2B Oil Spillage Protection System oil lift system enclosure collected all oil leakage from all potential leakage sites, including the oil lift system,presented a degradation of a fire confinement component which has a fire prevention function of not allowing an oil leak to reach hot surfaces. The significance of this finding was evaluated using IMC 0609, Appendix F, "Fire Protection Significance Determination Process, dated September 20, 2013, because the performance deficiency affected fire protection defense-in-depth strategies involving fire confinement. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) because the exposed fire area contains no potential damage targets that are unique from those in the exposing fire area. The inspectors determined the finding had a cross-cutting aspect of Design Margins in the human performance area because the licensee did not maintain fire protection defense-in-depth by ensuring the Unit 2 RCP 2B oil collection system was in an operable condition to perform its design function. (H.6) |
| 05000364/FIN-2017004-01 | 2017Q4 | Farley | Failure to Evaluate Impacts on the 2C RCP Oil Collection System | A self-revealing finding was identified for the licensees failure to evaluate the impacts to the Unit 2 Reactor Coolant Pump (RCP) 2C oil collection system when a service water (SW) leak was identified on the Unit 2 RCP motor air coolers. As a result, a strategy was not implemented to prevent service water from collecting in the 2C RCP oil collection system drain tank which impacted its design function while the plant was in Mode 1. The licensees failure to evaluate the potential impacts to the Unit 2 RCP 2C oil collection system during the operability/functionality evaluation of the SW leak associated with RCP motor air coolers was a performance deficiency. The licensee initiated condition reports (CRs) 10420400 and 10422562 and replaced the 2C RCP motor and leaking air cooler.The finding was more than minor because it was associated with the protection against external factors (fires) and adversely affected the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to maintain adequate capacity in the RCP 2C Oil Spillage Protection System (OSPS) oil collection tank presented a degradation of a fire confinement component which has a fire prevention function of not allowing an oil leak to reach hot surfaces. The significance of this finding was evaluated using IMC 0609, Appendix F, "Fire Protection Significance Determination Process, dated September 20, 2013, because the performance deficiency affected fire protection defense-in-depth strategies involving fire confinement. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) because the exposed fire area contained no potential damage targets that are unique from those in the exposing fire area. The inspectors determined the finding had a cross-cutting aspect of Evaluation in the problem identification and resolution area because the licensee did not fully evaluate the impacts of the RCP motor air cooler SW leak on the Unit 2 RCP oil collection systems. (P.2) |
| 05000424/FIN-2017003-05 | 2017Q3 | Vogtle | Licensee-Identified Violation | 10 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Contrary to the above, on June 28, 2017, the licensee failed to evaluate radiological conditions in room 1- AB -C-94, Back flushable Filter Crud Tank Pump Room, following the tank being placed in recirculation by Operations. On July 2, 2017, during a routine survey of room 1- AB- C-94, general area dose rates in the area were found to be as high as 600 mrem/hr. On the previous survey, conducted on June 19, 2017, maximum dose rates were found to be as high as 60 mrem/hr. This finding was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety SDP, and was determined to be of very low safety significance (Green) because the finding is not related to ALARA dose planning, did not result in an overexposure or the substantial potential for overexposure, and the ability to assess dose was not compromised due to the use of appropriate personnel dosimetry. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This issue was entered into the licensees corrective action program as CR 10383067. |
| 05000424/FIN-2017003-04 | 2017Q3 | Vogtle | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion XI, Test Control stated, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. UFSAR Section 8.1.4.3.C.2 stated that the onsite electrical system was designed in accordance with IEEE 308 -1974, Criteria for Class 1E Power System at Nuclear Generating Stations. IEEE 308 -1974 Section 6.3 recommended periodic tests be performed at scheduled intervals to detect deterioration of equipment to demonstrate operability of the components that are not exercised during normal operation. Contrary to the above, the licensee did not establish adequate test control measures to assure that the protective function of all 1E lockout relays were periodically verified. Specifically, there was no preventative maintenance to test the 1E lockout relays for non- MSPI loads. This condition has existed since plant initial operation and was identified during a licensee Nuclear Oversight audit on July 13, 2017. The inspectors determined this finding was of very low safety significance (Green) because the inspectors found no documented history of in- service failures of 1E lockout relays rendering safety -related equipment inoperative. This issue was documented in the licensees corrective action program as CR 10381797. |
| 05000424/FIN-2017003-01 | 2017Q3 | Vogtle | Failure to Implement and Establish Appropriate Work Instructions Affecting Safety-Related Chiller | A Self -Revealing, Green, non- cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow measurement and test equipment (M&TE) in support of essential safety features (ESF) chilled water pumps in- service testing (IST). As a result, the Unit 1 A train safety -related chiller was inadvertently rendered inoperable when technicians isolated a flow transmitter associated with the chillers auto -start control logic when installing and removing M&TE in support of the IST. The licensee entered this issue into their corrective action program (CAP) under condition report (CR) 10390340 and corrective action report 270610 and planned to revise the procedure. Failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow M&TE in support of ESF chilled water pumps IST, which can affect ESF chiller performance, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD). The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because while the unit 1 A train ESF chiller was rendered inoperable, it did not represent a loss of function of the train for greater than its TS Allowed Outage Time. The finding was assigned a cross cutting aspect of Challenge the Unknown because questions and risks regarding the use of flow M&TE for the test were not properly evaluated and managed before proceeding. (H.11) |
| 05000425/FIN-2017003-03 | 2017Q3 | Vogtle | Failure to Maintain Cleanliness of Motor Operated Valve Limit Switch Compartment | A Self -Revealing , Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to perform an adequate cleanliness inspection of the Unit 2 nuclear service cooling water (NSCW) system pump no. 6 discharge motor -operated -valve (MOV) limit switch compartment, as required by the maintenance procedure. As result , the valve failed to operate when demanded and rendered the NSCW pump inoperable. The failure to perform an adequate cleanliness inspection of NSCW pump no. 6 discharge MOV limit switch compartment following preventive maintenance, as required by maintenance procedure NMP -ES- 017- 008, was a performance deficiency (PD). The licensee cleaned affected MOV sub -components, verified proper operation, and restored operability of the pump. This issue was entered into the licensees CAP as CR10399054 . The performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The finding was determined to be of very low safety significance (Green) because although the performance deficiency affected the qualification and operability of the NSCW pump, it did not represent a loss of function of an NSCW train for greater than its TS Allowed Outage Time . The finding was assigned a cross cutting aspect of Avoid Complacency, because maintenance technicians did not recognize the possibility of making mistakes when performing routine tasks of inspecting and manipulating grease containing components inside the limit switch compartment. (H.12) |
| 05000425/FIN-2017003-02 | 2017Q3 | Vogtle | Failure to Maintain ECCS Flow Balance and Check Valve Inservice Test Procedure | An NRC- Identified, Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to maintain a Unit 2 surveillance procedure that demonstrated satisfactory performance of the forward flow safety function of emergency core cooling system ( ECCS ) check valves. The licensee revised and performed the test to verify satisfactory valve performance. This issue was entered into the licensees CAP as CR10410794. The failure to maintain procedure 14721D -2 to ensure test conditions that adequately demonstrated satisfactory performance of ECCS check valves 2- 1205- U6 -001/00 2, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD ). The performance deficiency was more than minor because if left uncorrected, it could result in degradation of ECCS check valves to go undetected. The finding was associated with the mitigating system cornerstone. The finding was determined to be of very low safety significance (Green) because the performance deficiency did not result in a loss of operability or functionality of ECCS check valves. The finding was assigned a cross cutting aspect of Resources, because the licensee did not ensure that an ECCS surveillance procedure was adequate to support nuclear safety . (H.1) |
| 05000364/FIN-2017003-01 | 2017Q3 | Farley | Failure to perform adequate corrective maintenance on the 2B EDG | The NRC identified a non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a, Procedures, for the licensees failure to implement corrective maintenance work order instructions to identify and replace piping as necessary for a degraded threaded joint on the 2B emergency diesel generator (EDG) jacket water keep warm system piping. As a result, a leak occurred at this threaded pipe joint during surveillance testing which rendered the 2B EDG inoperable. The inspectors determined that the failure to follow work order instructions to replace degraded jacket water system piping during corrective maintenance on the 2B EDG on March 3, 2017, was a performance deficiency (PD). The finding was more than minor because it was associated with the equipment reliability attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding was evaluated using IMC 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. Initial screening by the resident inspectors using the Saphire Farley 1 & 2 SPAR Model resulted in a potentially greater-than-green significance. Therefore, a detailed risk analysis was performed by a regional senior reactor analyst (SRA). The NRC Farley SPAR model was used for internal events, seismic and tornado/high winds risk estimates and the licensees Farley fire probabilistic risk assessment model was used for fire risk estimation. The major analysis assumptions included: a 51-day exposure period, EDG 2B operation at nominal failure to run probability until 8 hours when EDG assumed to fail due to the PD, PD treated as having common cause failure to run potential, no recovery of the 2B EDG was assumed, and no credit for FLEX equipment was assumed. The operation of the EDG for 8 hours prior to failure and remaining mitigating equipment limited the risk. The dominant sequence was a station blackout sequence consisting of a site-wide weather-related loss of offsite power, successful reactor shutdown, random failure to run of the 1/2A and 1C EDGs, failure of the 2B EDG due to the performance deficiency, failure to manually operate the turbine driven auxiliary feedwater pump long term, and failure to recover offsite power or an EDG leading to loss of core heat removal and core damage. The detailed risk evaluation (DRE) determined that the increase in core damage frequency due to the PD was <1.0 E-6 per year, a Green finding of very low safety significance. The finding had a cross-cutting aspect of Conservative Bias in the Human Performance area, because the decision to leave the diesel in a degraded condition following maintenance on March 3, 2017 was neither conservative nor prudent when additional action could have been taken to adequately repair or evaluate the threaded pipe joint (H.14). |
| 05000369/FIN-2017002-01 | 2017Q2 | Mcguire McGuire | Inadequate Survey Results in Unposted HRA | Green . A self -revealing Green non- cited violation (NCV) of 10 CFR 20.1501(a)(2) was identified for the licensees failure to conduct an adequate area radiation survey in Room 619 of the auxiliary building (waste gas decay tank (WGDT) room). Specifically, on April 19, 2016 , a high radiation area (HRA) was identifi ed near WGDT A in the WGDT room when a worker entering the area received a dose rate alarm on his electronic dosimeter (ED) and follow -up surveys revealed dose rates as high as 110 mrem/hr at 30cm. Also, as a result of the licensees failure to perform a survey, the area was not barricaded and posted in accordance with plant Technical Specification (TS) 5.7.1, High Radiation Area. The licensee immediately barricaded and posted the area as an HRA, performed an apparent cause evaluation to determine additional long term actions and entered the issue into their corrective action program as Nuclear Condition Report (NCR) 02021742. The licensees failure to conduct an area radiation survey to evaluate the magnitude and extent of radiation levels near WGDT A was a performance deficiency. This finding was determined to be more than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, failure to identify, post and control HRAs could allow workers to enter HRAs without knowledge of the radiological conditions in the area and receive unintended occupational exposure. The finding was evaluated using Inspection Manual Chapter (IMC) 0609 Appendix C, Occupational Radiation Safety Significance Determination Process. The finding was not related to the a s low as reasonably achievable (ALARA) planning, did not involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross -cutting aspect of avoid complacency in the area of human performance because the possibility of significant dose rate changes in the WGDT room during startup was a latent issue for which the licensee failed to recognize and plan. (H.12) |
| 05000327/FIN-2017002-01 | 2017Q2 | Sequoyah | Licensee-Identified Violation | Unit 1 and Unit 2 technical specifications LCO 3.7.10 required that if both trains of CREVS become inoperable than LCO 3.0.3 shall be immediately entered. Additionally, LCO 3.0.3 requires both units to be placed in Mode 3 within seven hours if the condition was not rectified. Contrary to the above, on August 10, with both trains of CREVS rendered inoperable, both units remained in Mode 1 for a period of approximately 24 hours. The finding was entered into the licensees CAP as CR 1201905. This finding was assessed using NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, and was determined to be of very low safety significance (Green) due to the finding only representing a degradation of the radiological barrier function provided for the control room. |
| 05000327/FIN-2017002-02 | 2017Q2 | Sequoyah | Licensee-Identified Violation | Unit 1 and Unit 2 facility technical specifications LCO 3.6.10 required two operable EGTS systems in Modes 1 through 4. Contrary to the above, on August 2, 2016,during a system review, plant engineers noted a design flaw that could have resulted in one train of EGTS being rendered inoperable since initial plant operation. This problem was entered into the licensees CAP as CR 1198440 and CR 1200028. The TVA probabilistic risk assessment model does not consider the EGTS in core damage and large early release frequencies. The EGTS system is designed to maintain the shield building at a negative pressure and filter any leakage past the steel liner during a design basis event. With the EGTS inoperable, dose would still remain below 10 CFR 100 limits. The finding was screened using IMC 0609, Appendix A At Power Operation, and was determined to be of very low safety significance (Green). According to Exhibit 3, an issue related to degradation of the radiological barrier function of the reactor building is considered to be of very low safety significance. |
| 05000364/FIN-2017002-03 | 2017Q2 | Farley | Failure to perform adequate corrective maintenance on the 2B EDG | To Be Determined (TBD). The NRC identified an apparent violation (AV) of Technical Specification (TS) 5.4.1.a, Procedures, for the licensees failure to implement corrective maintenance work order instructions to identify and replace a degraded jacket water fitting on the 2B emergency diesel generator (EDG) jacket water keep warm system piping. As a result, a leak occurred on the 2B EDG jacket water piping system during surveillance testing which rendered the EDG inoperable. 3 The inspectors determined that the failure to follow work order instructions to replace degraded jacket water system piping during corrective maintenance on the 2B DG on March 3, 2017, was a performance deficiency. The finding was more than minor because it was associated with the equipment reliability a ttribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding was evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. Initial screening by the resident inspectors using the Sapphire Farley 1 & 2 SPAR Model resulted in a potentially greater-than-green significance. Therefore, a detailed risk analysis will be performed by a regional senior reactor analyst (SRA). The inspectors determined the finding had a cross-cutting aspect of Conservative Bias in the Human Performance area, because the decision to leave the diesel in a degraded condition following maintenance was neither conservative nor prudent when additional action could have been taken to adequately repair or evaluate the piping connection (H.14). |
| 05000324/FIN-2017001-01 | 2017Q1 | Brunswick | Nonfunctional Sprinklers in the Service Water Building Without Compensatory Measures | Green . An NRC- identified Green non -cited violation ( NCV ) of License Condition 2.B.(6), Fire Protection Program, was identified for the licensees failure to implement compensatory measures for nonfunctional sprinklers. Specifically, from January 11, 2017, until January 14, 2017, fire sprinklers were impaired when scaffold ing was built over the service water (SW) system discharge valves without the proper fire protection evaluation and compensatory measures , as required by licensee procedure 0PLP -01.2, Fire Protection System Operability, Action, and Surveillance Requirements . The licensees corrective actions included declaring the sprinklers nonfunctional, and implementing an hourly fire watch and backup suppression until the scaffold could be removed. This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 2091795. The inspectors determined that the licensees failure to implement compensatory measures for nonfunctional sprinklers , in accordance with procedure 0PLP -01.2, was a performance deficiency. The finding was more than minor because it was associated with the Protection against E xternal Events attribute (i.e. fire) of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this resulted in nonfunctional sprinklers in a safety -related area without compensatory measures. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because the finding affected the fixed fire protection system capability. Using IMC 0609, Appendix F, Attachment 1, Fire Protection S DP Phase 1 Worksheet, dated September 20, 2013, the finding was assigned to the Fixed Fire Protection System category because the nonfunctional sprinklers affected the automatic fire suppression system. Proceeding to Task 1.3.1 of IMC 0609, Appendix F, Attachment 1, the inspectors determined the finding was of very low safety significance (Green), because with the sprinklers nonfunctional, the reactor was able to reach and maintain safe shutdown. The finding has a cross -cutting aspect in the area of human performance associated with the field presence attribute because leaders did not observe, coach, and reinforce standards and expectations regarding scaffolding . Deviations from standards and expectations for building scaffolding near fire protection sprinklers were not corrected promptly. (H.2) |
| 05000324/FIN-2017001-02 | 2017Q1 | Brunswick | Failure to Control a Temporary Fire Ignition Source Near the Unit 2 Standby Liquid Control Pump Motor and Cables | Green . An NRC- identified Green NCV of License Condition 2.B.(6), Fire Protection Program, was identified for the licensees failure to adequately control fire ignition sources in the Unit 2 standby liquid control (SLC) pump ar ea in accordance with licensee procedure AD -EG -ALL -1523, Temporary Ignition Source Control. Specifically, between January 7, 2017, and January 13, 2017, a temporary electric portable heater was energized 2 feet from an SLC pump motor without continuously attending the temporary ignition source or obtai ning a continuous fire watch. The licensees c orrective actions included turning off the heater and removing it from near the SLC pumps. This issue was entered into the licensees CAP as NCR 2091736. The inspectors determined that the licensees failure to control fire ignition sources in accordance with licensee procedure AD -EG -ALL -1523, was a performance deficiency. The finding was more than minor because it was associated with the Protection Against External Events attribute (i.e. fire) of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the temporary ignition source could have affected a nearby safety -related SLC pump motor and cables, which provide a shutdown mitigation function. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Proc ess, dated September 20, 2013. Using IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, dated September 20, 2013, the findi ng was assigned to the Fire Prevention and Administrative Controls category because the portable heater is part of the plants combustible materials control program. Proceeding to Task 1.3.1 of IMC 0609, Appendix F, Attachment 1, the inspectors determined the finding was of very low safety significance (Green), because even if one train of SLC had been inoperable, the reactor was able to reach and maintain safe shutdown. This finding had a cross cutting aspect in the area of human performance associated wi th the teamwork aspect because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions. (H.4) |
| 05000324/FIN-2017001-03 | 2017Q1 | Brunswick | Failure to Install Flood Barrier Seals Around the EDG 2 Four -Day Fuel Oil Tank Vents | Green . An NRC- identified Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the failure of the licensee to install flood barrier seals around the emergency diesel generator (EDG) 2, four -day fuel oil tank v ent as described in engineering change (EC) 400606. This result ed in a nonfunctional flood barrier into the EDG 2 four -day tank room. As an immediate corrective action, the licensee grouted the opening to prevent water intrusion into the EDG 2 four -day f uel oil tank room. The licensee entered this issue into the CAP as NCR 2093563. The inspectors determined the failure of the licensee to control the design of the installation of the new EDG 2 four -day fuel oil tank vent was a performance deficiency. Th e finding is more than minor because it is associated with the protection against external factors (i.e., flood hazard) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability , and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., c ore damage). Specifically, the licensee failed to install flood barrier seals around the EDG 2 four -day fuel oil tank vent designed to mitigate a flood of the EDG 2 four -day fuel oil tank room. Using IMC 0609, Appendix A, issued June 9, 2012, The SDP for Findings At -Power, the inspectors determined the finding screened to Exhibit 4, External Events Screening Questions, since the finding involved the loss of equipment specifically designed to mitigate a flood. The inspectors determined the finding screened to Green since if the flood barrier is assumed to be completely failed, it 4 would not result in the inoperability or degradation of EDG 2, and would not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. The finding has a cross -cutting aspect in the area of human performance associated with the design margins attribute because the licensee failed to maintain equipment within design margins and failed to change margins through a systematic and rigorous process. Specifically, the licensee changed the installation of the EDG 2 fuel oil tank roof vent without ensuring flood protection during the modification. (H.6) |
| 05000324/FIN-2017001-04 | 2017Q1 | Brunswick | Failure to Enter the Technical Specification for an Inoperable 1D Control Room Air Conditioning Unit | Green . An NRC- identified Green NCV of Technical Specification (TS) 3.7.4, Control Room Air Conditioning (AC) System, was identified for the failure to declare the 1D control room AC unit inoperable. Specifically, on December 1, 2016, the licensee failed to declare the 1D control room AC unit inoperable due to extensi ve corrosion on the support channels . As a result, the 1D control room AC unit was inoperable from December 1, 2016, until the next time it was inspected on January 30, 2017, and exceeded the TS allowed outage time. As corrective actions, the licensee replaced the supports of the 1D and 2D control room AC units and inspected the 2E control room AC unit for corrosion. The licensee entered this issue into the CAP as NCRs 2113799 and 2113800. The inspectors determined the licensees failure to declare the 1D control room AC unit inoperable on December 1, 2016, and enter TS 3.7.4 was a performance deficiency. The finding was more than minor because it was associated with the structures, systems, and components ( SSC ) att ribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, this resulted in the 1D control room AC unit being inoperable from December 1, 2016, to January 30, 2017. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At -Power, the inspectors det ermined the finding was of very low safety significance (Green) because the finding did not only represent a degradation of the radiological barrier function for the control room and the finding did not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere. This finding had a cross cutting aspect in the area of problem identification and resolution associated with the resolution aspect because the licensee failed to take effective corrective actions to addres s issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not correct the degradation of the 1D control room AC unit until the unit was inoperable. (P.3) |
| 05000324/FIN-2017001-05 | 2017Q1 | Brunswick | Licensee-Identified Violation | TS limiting condition for operation (LCO) 3.3.3.1, Condition F, Post Accident Monitoring (PAM) Instrumentation, states in part , with the DWHRRMs inoperable, a Special Report shall be submitted to the Commission within the next 14 days. Contrary to the above, the licensee failed to identify the inoperability of the DWHRRMs after the NRC Information Notice 97 -45 Supplement 1 was issued. In particular, the DWHRRMs signals cables are susceptible to thermally induced currents which can degrade the accuracy of DWHRRMs . The required action of LCO 3.3.3.1 , action F, was not perf ormed from 1998 until December 5, 2016. Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, inspectors determined that this violation was of very low safety significance (Green) because the finding is related to Emergenc y Preparedness Requirements that are not associated with a planning standard function (e.g.,10 CFR 50.54(q), 10 CFR 50.54(t), and requirements in Appendix E to 10 CFR Part 50 that do not support a planning standard function ). Other parameters could be used to validate the indications from the DWHRRMs . The corrective action is to restore the monitors to operable. This issue was documented in the licensees CAP as NCR 2066681 |
| 05000366/FIN-2017001-01 | 2017Q1 | Hatch | Failure to Identify Abnormal Condition on 2C EDG Cross Drive Assembly | Green . A self -revealing non- cited violation (NCV) of Hatch Unit 2 Technical Specification 5.4.1 was identified when technicians performing maintenance on the 2C emergency diesel generator observed pitting on the lower crank component gears and did not initiate a condition report as required by procedure 52SV -R43 -001- 0, Diesel, Alternator, and Accessories Inspection. The licensees failure to initiate a condition report, as required by 52SV -R43 -001- 0 Diesel, Alternator, and Accessories Inspection, for the pitting observed on the lower crank component gears was a performance deficiency. The violation of regulatory requirement occurred on or about November 2015 until the licensee replaced the 2C EDG cross drive assembly and restored compliance on August 25, 2016. The violation was entered into the licensees corrective action program as CR 10263236. The performance deficiency was more than minor because if left uncorrected, the failure to evaluate gear pitting would allow progression of a degradation mechanism to the point of EDG inoperability. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At -Power, dated June 19, 2012. Because all four questions in Section A of Exhibit 2, Mitigating Systems Screening Questions, were answered no, the finding screened as Green. The inspectors determined that this finding had a cross -cutting aspect in the Resources aspect of the human performance area, because the licensee did not ensure adequate procedural guidance to recognize the difference between normal and destructive pitting. (H .1) |
| 05000321/FIN-2017001-02 | 2017Q1 | Hatch | Licensee-Identified Violation | Unit 2 Technical Specification 3.6.1.3 requires each PCIV be operable in Mode 1. With one PCIV inoperable, the affected penetration flow path must be isolated by use of at least one closed and de -activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. Contrary to the above, on November 6, 2016 at 21:51 operators tagged valve 2E41F111, a PCIV, open with the breaker off. Subsequently, a licensed operator performing a main control room board walk down noted the PCIV was inoperable and, on November 8 at 0151, operators closed and de -activated an automatic valve in the line to rest ore compliance. Inspectors screened the finding in accordance with IMC 609 Appendix A The Significance Determination Process (SDP) for Findings at -Power. The finding screened as very low safety significance (Green) because the questions in Appendix A E xhibit 3 for reactor containment were answered no. This issue was documented in the licensees corrective action program as CR 10295889. (Section 4OA3.2) |
| 05000321/FIN-2017001-03 | 2017Q1 | Hatch | Licensee-Identified Violation | Technical Specification 5.7.1 requires, in part, entrances into areas in which the intensity of r adiation is > 100 mrem/hr but < 1000 mrem/hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, to be controlled by requiring issuance of a Radiation Work Permit (RWP). Contrary to this, On September 9, 2016, two in dividuals entered a High Radiation Area in the Unit 2 SE Diagonal 87' elevation to calibrate an RHR Service water transmitter without the proper briefing or RWP. The individuals were briefed and permitted to enter the HPCI Room area instead of this area. This finding was of very low safety significance (Green) because there was no substantial potential for overexposure and the licensees ability to assess dose was not compromised. The immediate corrective actions were documented in CR 10271667. The long term corrective actions include continuing training suc h that all craft personnel are exposed to the remediation scenario. (Section 2RS1 ) |
| 05000335/FIN-2017001-01 | 2017Q1 | Saint Lucie | Inadequate Procedure Results in Adding an Incorrect Lubrication Oil to the 1B CS Motor Inboard Bearing | An NRC-identified Green, non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for the licensees failure to establish, implement, and maintain written procedures covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensees failure to maintain a plant lubrication manual with correct lubrication oil specifications for the 1B containment spray (CS) pump motor resulted in adding unacceptably low viscosity lubrication oil to the inboard bearing of the 1B CS pump motor. Immediate corrective actions included restoring the 1B CS pump inboard bearing with the correct lubrication oil and placing the issue in the licensees corrective action program.The licensees failure to correctly specify the 1B CS pump motor inboard bearing lubrication requirements in licensee general maintenance procedure GMP-22 was a performance deficiency (PD). The PD was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedure resulted in adding the incorrect lubrication oil to the 1B CS pump motor bearing, causing the pump to be declared inoperable for approximately 56.5 hours. The finding screened to Green because the failure did not: (1) affect the design or qualification of the systems, structures and components, (2) represent an actual loss of function, and (3) represent an actual loss of function of at least a single train for greater than its TS allowed outage time. The finding involved the cross-cutting area of human performance, in the aspect of avoid complacency, in that, the individuals involved with the procedure revision did not implement appropriate error reduction tools to ensure the procedure was appropriately changed to reflect the new lubrication oil requirement (H.12). |
| 05000348/FIN-2016004-01 | 2016Q4 | Farley | Failure to Adequately Install an Oil Collection System on Reactor Coolant Pump Motors | An NRC-identified non-cited violation (NCV) of 10 CFR 50.48(c) and National Fire Protection Association Standard 805 (NFPA 805), Section 3.3.12, was identified for the licensees failure to comply with code requirements for design and installation of the Unit 1 Reactor Coolant Pump (RCP) oil collection system. The oil collection system did not include gaskets between the bolted joints on the RCP oil catch-basins, as required by the approved design for the Oil Spillage Protection System (OSPS). The licensees failure to install gaskets on the Unit 1 RCP oil collection systems was a performance deficiency. The licensee was informed of the inspector observation and initiated CR 10289565. Gasket material was installed on all three RCPs on October 23, 2016, as documented on WO SNC464660, SNC459614, and SNC406358. The performance deficiency was more than minor because if left uncorrected, the inadequate installation of the RCP oil collection system presented a degradation of a fire confinement function to prevent oil to leak onto hot surfaces. The significance of this finding was evaluated using IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because the performance deficiency affected fire protection defense-in-depth strategies involving fire confinement. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) because the exposed fire area contains no potential damage targets that are unique from those in the exposing fire area. The inspectors determined the finding had a cross-cutting aspect of Procedure Adherence in the human performance area because the vendor installing the oil catch-basins did not follow the RCP reassembly procedure which required gaskets between all bolted joints. (H.8) |
| 05000348/FIN-2016004-02 | 2016Q4 | Farley | Failure to Perform Adequate NTTF Flooding Walkdowns | An NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified because the licensee failed to identify and correct conditions adverse to quality associated with the flood protection design basis of the Unit 1 Auxiliary Building. Specifically, the licensee failed to identify missing condulet covers in electrical conduits that penetrate the Unit 1 auxiliary building below the flood protection design basis elevation of 154.5 feet (MSL). The inspectors determined that the failure to identify missing condulet covers in electrical conduits that penetrate the Unit 1 auxiliary building below the flood protection design basis elevation of 154.5 feet was a performance deficiency. The discovery of the missing condulet covers was captured in the licensees corrective action program with CR 10273516. The licensee implemented WO SNC815778 to replace missing condulet covers. Corrective actions to inspect the remaining below grade pipe trenches are being developed and scheduled. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events. Specifically, flood water could enter the Auxiliary Building Lower Equipment Room through unsealed electrical conduits and render the TDAFW Pump inoperable. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, issued June 19, 2012, the inspectors utilized Section B, External Event Mitigation Systems (Seismic/Fire/Flood/Severe Weather Degraded), and Exhibit 4 of Appendix A and determined the finding did not involve a total loss of any safety function, identified through a PRA, IPEEE, or similar analysis, that contributes to external event initiated core damage accident sequences (i.e., initiated by a seismic, flooding, or severe weather event). The two motor driven AFW pumps are also located in the lower equipment room but are protected behind watertight doors and can satisfy the AFW safety function. Therefore, the finding screened to Green. The inspectors determined the finding had a cross-cutting aspect of Procedures in the human performance area because the licensee missed two opportunities to follow the NEI 12-07 guidance to evaluate the adequacy of the flood protection features below the design basis flood protection elevation.(H.8) |
| 05000348/FIN-2016004-03 | 2016Q4 | Farley | Failure to Follow Procedure Resulted in Automatic Reactor Trip and Safety Injection | A self-revealing non-cited violation (NCV) of Technical Specification 5.4, Procedures, was identified on October 1, 2016, when the Unit 1 operations shift crew failed to comply with annunciator response procedure FNP-1-ARP-1.9, Ver. 50 for the JC4 annunciator. Conditions were met to trip the reactor, but the operations shift crew failed to do so. As a result, approximately 35 minutes later, MSIV 3369A closed which resulted in an automatic reactor trip and safety injection actuation. The failure of the operations shift crew to follow procedure FNP-1-ARP-1.9 was a performance deficiency (PD). This event was captured in the licensees corrective action program with condition report (CR) 10280729. The licensee established a root cause evaluation team, identified the root causes, and implemented corrective actions (CAR 266911). The PD was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone objective and adversely affected that objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically a manual reactor trip of Unit 1 as required by the ARP, would have prevented the automatic reactor trip and the automatic safety injection actuation. The significance of this finding was evaluated using IMC 0609, Appendix A, "The Significance Determination Process (SDP) for findings at Power, dated June 19, 2012. This finding was determined to be of very low safety significance (Green) because, while this issue resulted in a reactor trip, it did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined the finding had a cross-cutting aspect of Procedure Adherence in the Human Performance area, because the ARP was not followed and the operations crew did not trip the reactor as required by the procedure. (H.8) |
| 05000348/FIN-2016004-04 | 2016Q4 | Farley | Licensee-Identified Violation | 10 CFR 20.1501(a)(2) requires, in part, that licensees make surveys to evaluate the magnitude and extent of radiation levels and quantities of radioactive material. 10 CFR 20.1501(b) requires that the licensee shall ensure that instruments and equipment used for quantitative radiation measurements be calibrated periodically for the radiation measured. Contrary to this, on June 2, 2016, the licensee discovered that the surveillance procedure used to calibrate N1D21RE0001 and N2D21RE0001B (MCR Area Monitors) had been deleted and the monitors had not been calibrated for approximately six years. This condition was documented in CR 10231300. Upon re-calibration of N1D21RE0001, the low voltage power supply was found out of tolerance (CR 10256497), indicating that the radiation monitor might not have been able to perform its function of alerting MCR operators of changing radiological conditions. This condition was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety SDP, and determined to be of very low safety significance (Green) because the finding is not related to ALARA dose planning, did not result in an overexposure or the substantial potential for overexposure, and the ability to assess dose was not compromised due to the use of appropriate personnel dosimetry. |
| 05000280/FIN-2016004-01 | 2016Q4 | Surry | Change of Surveillance Frequency Caused the Charging Service Water Header to Become Biologically Fouled | A self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI was identified because the surveillance procedure frequency used to flush the service water (SW) piping in Mechanical Equipment Room (MER)-3 and MER-4 was changed from two weeks to four weeks without sufficiently considering the effects of river conditions on biological growth and without getting management permission to change the periodicity. As a result of the periodicity change, the B charging (CH) and main control room (MCR) SW header became blocked with biological growth and was declared inoperable on September 22, 2016, during the performance of 0-OSP-VS-012, High Flow Flush of SW Strainers and Piping in MER 3 and MER 4. As immediate corrective action, the licensee cleaned the clogged SW strainer and completed the backflushing of the SW header. The SW flushing periodicity was restored to a two week frequency to be seasonally and risk assessed and reduced as heavy fouling season ends. This issue was documented in the licensees corrective action program (CAP) as CR 1048251. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the performance deficiency (PD) was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not affect the design or qualification of the charging pump service water pump system and it did not represent a loss of system safety function. This finding has a cross-cutting aspect in conservative bias aspect of the human performance area, H.14, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowed. |
| 05000281/FIN-2016004-02 | 2016Q4 | Surry | Inadequate Design Change Post Maintenance Testing Causes Water Intrusion into Station Service Transformer and a Reactor Trip | A self-revealing finding was identified because the test requirements section of the station service transformer (SST) design change (DC) was not comprehensive in that it did not test that the isolated phase bus ducting terminal boxes were constructed to prevent water intrusion into the boxes. This was discovered during a significant rainfall event partially caused by Hurricane Matthew, which filled up the A SST terminal box with water and eventually shorted the A phase of the main generator causing a Unit 2 main generator, main turbine, and subsequent reactor trip on October 9, 2016. As corrective action, sealant was applied to the SST terminal boxes on all seams and bolt holes; and weep holes with drain assemblies were installed on each box. This issue was documented in the licensees CAP as CR 1049987. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone, and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated October 7, 2016, the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because although the deficiency did cause a reactor trip, it did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the Operating Experience aspect of the Problem Identification and Resolution area, P.5, because the licensee did not evaluate and implement relevant external operating experience. |
| 05000269/FIN-2016004-01 | 2016Q4 | Oconee | Failure to Perform Appropriate Evaluation of Motor Operated Valve Actuator Output Capability | Green. The NRC identified a non-cited violation (NCV) of Title 10 Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly determine the bounding degraded voltage to be assumed in the determination of motor operated valve (MOV) actuator output capability. Specifically, the licensee did not use appropriate transient voltages as input into the evaluation of the capability of the MOVs that are required to reposition in response to an accident signal. In response, the licensee entered the issue into their corrective action program as nuclear condition report (NCR) 2056895 and planned to formally revise their calculations to reflect the current plant configuration. This performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees programmatic failure to use bounding terminal voltage values in the evaluation of their automatically actuated, safety-related MOVs did not ensure they would be capable of mitigating accidents when powered from sources other than the 230kV switchyard, thus resulting in doubt on their capability to perform their intended safety function. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the most recent transient analysis that was performed for the sources other than the 230kV switchyard was performed in 2012. |
| 05000269/FIN-2016004-02 | 2016Q4 | Oconee | Inappropriate Voltage Band in Lee Combustion Turbine Unit Operating Procedure | Green. The NRC identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to identify appropriate procedural updates that were needed to ensure the Lee combustion turbine (LCT) procedures were appropriate for the circumstances and maintained current. Specifically, the licensee did not include appropriate operational limitations in procedures associated with the LCTs. In response, the licensee generated NCR 2058763, verified the LCT automatic voltage regulator setpoint was, and had been, 13.8kV, and generated a corrective action to revise the affected procedures limits to 13.78kV, a value bounded by station analyses. This performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees failure to limit the operating voltage band of the LCTs to an amount that was demonstrated as acceptable by analysis resulted in doubt on their capability to provide power to safety-related equipment during an accident. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the update to the procedures occurred in January and October 2007, after replacement of the LCTs. |
| 05000321/FIN-2016003-02 | 2016Q3 | Hatch | Failure to Ensure Work Hours are Within Work Hour Limits | An NRC-identified non-cited violation (NCV) of 10 CFR Part 26, Fitness for Duty Programs, was identified when the licensee failed to ensure that personnel subject to work hour controls did not exceed 72 hours in a work week. The licensee entered this condition into their corrective action program as Condition Report 10214872 and restored compliance when the affected individuals received an adequate rest period. The failure to ensure that work hours for personnel subject to work hour controls were tracked in accordance with licensee procedures was a performance deficiency. The finding was more than minor because, if left uncorrected, the failure to appropriately implement work hour limitations for covered workers could adversely impact the conduct and oversight of work on safety significant components. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not result in an adverse impact to plant safety due to worker fatigue. The inspectors determined this performance deficiency had a cross-cutting aspect of Consistent Process in the Human Performance area because the licensee failed to assess which workers were subject to work hour limits. (H.13) |
| 05000321/FIN-2016003-01 | 2016Q3 | Hatch | Unit Downpower Caused by RFP Vent Line Failure | A self-revealing finding was identified when the licensee failed to install a reactor feed pump (RFP) vent line weld in accordance with plant procedures resulting in a failure that required an unplanned Unit 1 power reduction greater than 20%. Failure to install the correct weld thickness on the unit 1 B RFP vent line, as required by procedures, was a performance deficiency. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective in that an unplanned reactor power reduction was required from 100 percent to 60 percent RTP. The inspectors determined this finding was of very low safety significance (Green) because there was not a reactor trip or loss of mitigation equipment. The inspectors determined that this finding had a cross-cutting aspect in the Resolution aspect of the problem identification and resolution area, because the organization did not take effective corrective actions to address the previous weld configuration issue. (P.3) |
| 05000324/FIN-2016002-01 | 2016Q2 | Brunswick | Failure to Identity Broken Auto Start Control Relay on Emergency Diesel Generator 1 | An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because the licensee failed to promptly identify and correct a condition adverse to quality (CAQ) on emergency diesel generator (EDG) 1. Specifically, from February 7, 2016, until March 5, 2016, the licensee failed to promptly identify and correct a broken auto start control relay (ASCR) which resulted in reduced capacity of EDG 1 due to load oscillations and inoperability of EDG 1 due to oscillating between droop and isochronous mode. The oscillations could cause the EDG to not meet Technical Specification (TS) frequency and load requirements. The licensee replaced the ASCR and entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 2007720. The licensees failure to promptly identify and correct the broken ASCR, which resulted in reduced capacity and inoperability of EDG 1 due to load oscillations, was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct the malfunctioning ASCR resulted in reduced capacity of EDG 1 due to load oscillations, and could cause EDG 1 to not meet TS frequency and load requirements. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because it represented a loss of system and/or function, and the finding represented an actual loss of a function of a single train for greater than the TS allowed outage time. The regional Senior Reactor Analyst evaluated the finding and determined it to be Green. The risk was low because of the diverse sources of AC power available, and the long duration of some of the sequences allowed a greater potential for recovery of a failed AC power source. The dominant risk sequences contained common cause failure of the diesel generators, with the supplemental EDG aligned to the other unit, and non-recovery of offsite power or of an EDG. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the identification attribute because the licensee failed to implement a CAP with a low threshold for identifying issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to write a timely NCR and identify the load oscillations as a CAQ. (P.1) |
| 05000324/FIN-2016002-02 | 2016Q2 | Brunswick | Failure to Verify or Check the Adequacy of Design of the EDG 3 Auto-Start Circuitry | A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to verify or check the adequacy of design of the EDG 3 emergency auto-start circuitry. Specifically, on October 24, 2011, the licensee failed to verify or check the adequacy of design of the fuse block holder modification to the EDG auto-start circuitry. This resulted in the fuse block holder connection becoming loose, a loss of continuity through the circuit, and the inoperability of EDG 3. The licensee replaced the fuse block holder, performed a continuity check, and plans to implement a design change to install continuity indication for continuous verification of continuity. The licensee entered this issue into the CAP as NCR 2007449. The licensees failure to verify or check the adequacy of design of the EDG 3 emergency auto-start circuitry fuse block holder modification was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This resulted in the fuse block holder connection becoming loose, a loss of continuity through the circuit, and the inoperability of EDG 3. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because it represented a loss of system and/or function, and the finding represented an actual loss of a function of a single train for greater than the TS allowed outage time. The regional SRA performed a detailed risk review for the finding. The finding was determined to be Green. The limited duration of the EDGs failure of the auto start, the ability to manually recover the EDG, and the availability of the other EDGs and of the supplemental EDG contributed to the low risk value. The dominant risk sequences were of low value, and were Station Blackout with failure to recover offsite power or the EDGs. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the identification attribute because the licensee failed to implement a CAP with a low threshold for identifying issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify EDG 3 was inoperable on February 7, 2016, when the indications were apparent. (P.1) |
| 05000321/FIN-2016010-02 | 2016Q2 | Hatch | Failure to Identify N2E Nozzle Weld Through-Wall Flaw | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to promptly identify a condition adverse to quality regarding a through-wall flaw in the safe end-to-nozzle weld of the reactor coolant system N2E nozzle. The licensee has since repaired the flaw, completed all required postrepair examinations, and entered this issue entered this into their corrective action program as CR 10247856. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors screened this finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012. Because after a reasonable assessment of degradation, the finding could neither result in exceeding the RCS leak rate for a small LOCA, nor likely affected other systems used to mitigate a LOCA resulting in a total loss of their function, the finding screened as Green. This finding has a cross-cutting aspect of Challenge the Unknown in the area of Human Performance (H.11) because upon discovery of a less robust configuration of the N2E nozzle overlay, the licensee failed to consider the implications on the flaw that had existed in that component since 1988. |
| 05000325/FIN-2016001-03 | 2016Q1 | Brunswick | Alert Declared Due to Electrical Fault Resulting in a Fire/Explosion | The inspectors are opening a URI to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, including enforcement action determinations, associated with the Alert declaration due to the electrical fault resulting in an explosion/fire in the Unit 1 BOP 4 kV switchgear bus area. The inspectors are opening a URI to review the root cause and determine if a performance deficiency exists. On February 7, 2015, operations personnel declared an Alert for Units 1 and 2, in accordance with Emergency Action Level HA 2.1 due to an explosion/fire in the Unit 1 BOP 4 kV switchgear bus area. A manual reactor SCRAM was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. The SAT experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. EDGs 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The licensee restored offsite power to the emergency buses and exited the NOUE. The licensee wrote NCR 1998726 to address this event. This issue is being tracked as a URI: URI 05000325/2016001-03; Alert Declared Due to Electrical Fault Resulting in a Fire/Explosion. |
| 05000324/FIN-2016001-02 | 2016Q1 | Brunswick | Failure to Identify and Correct a Condition Adverse to Quality Associated with the 2B NSW Pump Strainer | The inspectors identified a Green non-cited (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to identify and correct a condition adverse to quality associated with the 2B nuclear service water (NSW) pump strainer. Specifically, the licensee did not ensure the spacers/shims were filed down or seated appropriately, which resulted in the 2B NSW pump strainer shear key failures, and the unavailability of the 2B NSW pump on three separate occasions. As corrective actions, the licensee ensured the spacers/shims were filed down and seated appropriately for the 2B NSW pump strainer and changed the procedure to ensure these steps were performed in the future. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 1988423. The inspectors determined the licensees failure to ensure the 2B NSW pump strainer spacers/shims were filed down or seated appropriately was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this resulted in the failure of 2B NSW pump strainer shear key, and unavailability of the 2B NSW pump during repairs to the strainer. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the technical specification (TS) allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the evaluation attribute because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to evaluate the cause of the 2B NSW pump strainer shear pin failures to ensure the appropriate repair. (P.2) |
| 05000321/FIN-2016001-01 | 2016Q1 | Hatch | Reactor Coolant System N2E Weld Flaw | The inspectors identified an unresolved item associated with a flaw identified in the safe end-to-nozzle weld of the Reactor Coolant System N2E Nozzle. In July 2015, the licensee submitted a proposed alternative to ASME Code, HNP-ISI-ALT-15-01 (ML15183A354), to install a full-structural weld overlay on reactor coolant nozzle N2E (1B31-1RC-12-BR-E). This proposed alternative was approved by the NRC in December 2015 (ML15349A973). The licensee implemented this proposed alternative during the February 2016 refueling outage (1R27). After removing all but 1/16 of the existing overlay, the licensee performed a liquid penetrant examination and noted a pair of linear indications. Subsequently, the licensee determined that these indications were actually a single indication, and that it exceeded allowable size limitations according to ASME Code. Upon further review, the licensee realized that these indications were potentially the result of growth of an inner-diameter, surface-connected intergranular stress corrosion cracking (IGSCC) flaw found in 1988. The licensee has repaired the flaw, installed the full-structural weld overlay, and completed all required post-installation examinations. This is an unresolved item pending review of whether the licensee performed all required examinations of the N2E nozzle between 1988 and 2016, and whether the flaw exceeded minimum wall limitations at some point during prior operation. The issue will be tracked as URI 05000321/2016001-01, Reactor Coolant System N2E Weld Flaw. |
| 05000325/FIN-2016001-01 | 2016Q1 | Brunswick | ASME Section IX Weld Procedure Qualification | The inspectors identified an unresolved item (URI) associated with the qualification of the weld procedure specification (WPS) used for replacement of a portion of nuclear service water piping. Description: While conducting buried piping inspections in support of license renewal, the licensee identified pitting on the exterior wall of a portion of the Unit 1 nuclear service water supply header (1-SW-103-30-157). The licensee chose to address this by replacing the section of pipe (WO 12274010-08). The licensees repair/replacement plan for this activity identified that the requirements of ASME Section III, 1986 Edition, Subsection ND were applicable for the repair. By reference (ND-4320), several ASME Section IX Subsection QW requirements also applied. First, QW-200.2(f) allowed the use of multiple Procedure Qualification Records (PQRs) to produce a single WPS, provided that each essential variable is addressed by at least one PQR. Second, QW-403.8 and QW-404.30 established the requirements for two essential variables (base metal thickness qualified and filler metal thickness qualified, respectively) and referred to QW-451, which established the limits for both. The inspectors are opening a URI to review whether the licensees use of PQRs 1, 5, 193A and 193B to qualify the base metal and filler metal thickness ranges identified in WPS 01-1-04 and WPS 01-3-04 in accordance with Code was appropriate, and if a performance deficiency exists. The licensee wrote NCR 2009571 to address this issue. The issue is being tracked as a URI: URI 05000325/2016001-01, ASME Section IX Weld Procedure Qualification. |