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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000327/FIN-2010004-03 | 2010Q3 | Green | H.14 | NRC identified | Failure to Maintain Thermal Power Less Than Licensed Limit | The inspectors identified a Green non-cited violation of Unit 2 TS 6.8, Procedures and Programs, for the failure to take prompt action to maintain reactor thermal power less than the licensed power limit of 3455 megawatts thermal (MWt) in response to a transient caused by the loss of a condensate booster pump, as required by station procedures. The licensee entered this issue into their corrective action program as PER 259098. The licensee is currently evaluating for planned corrective actions. The finding was determined to be greater than minor because it was similar to example 8.b. of IMC 0612 Appendix E. Additionally, it was associated with the Human Performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective relative to the fuel cladding barrier since operation above the licensed power limit reduces analyzed margins to fuel cladding damage. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since only the fuel cladding barrier was affected. The cause of this finding was determined to have a cross-cutting aspect of Conservative Assumptions and Safe Actions in the area of Human Performance associated with the Decision Making component. The decision to take no operator action in response to the thermal power transient reflected a non-conservative assumption that average thermal power could be allowed to exceed the licensed limit without operator action while the feedwater control system responded to the transient associated with the condensate pump failure (H.1(b)). (Section 4OA3.3 |
| 05000327/FIN-2010004-04 | 2010Q3 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements that meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. Unit 2 TS 6.8.1.a required, in part, that written procedures be established, implemented, and maintained covering the activities specified in Appendix A, Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors, of Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operations), Revision 2, dated February 1978. RG 1.33 Appendix A Section 9.a required that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, on April 5, 2010, written procedures appropriate to the circumstances were not established which adequately prescribed the performance of maintenance that could affect the performance of safetyrelated equipment. Specifically, the maintenance instructions for reassembly of the SFP pump C-S backup breaker failed to include instructions for proper reassembly, which resulted in the breaker being installed in the 2A1 shutdown board and restored to service without arc chutes, causing the shutdown board to be inoperable for greater than its TS allowed outage time. The licensee entered the issue into the corrective action program as PERs 228519 and 228818. The finding was determined to have very low safety significance (Green) because there was no actual loss of safety system function, and there was no significant increase in the likelihood of a fire | |
| 05000327/FIN-2010005-01 | 2010Q4 | Green | NRC identified | Failure to Collect Reactor Coolant Pump Oil Leakage | The inspectors identified a Green non-cited violation of 10 CFR 50 Appendix R, Section III.O, Oil collection system for reactor coolant pump, for the licensees failure to ensure the capability of the reactor coolant pump (RCP) oil collection system to collect and drain all RCP oil leakage. System configuration and procedural deficiencies resulted in the inability of the oil collection system to collect and drain all RCP oil leakage. Approximately 2-3 gallons of oil leakage were identified on the containment floor following Unit 1 shutdown for a refueling outage. The licensee entered this issue into their corrective action program as PERs 270216, 278689, and 284244. Corrective actions included revision to applicable plant procedures to prevent the condition from occurring, as well as plans to evaluate a design change to modify the system configuration. The finding was determined to be greater than minor because it was associated with the protection against external factors attribute of the initiating events cornerstone, and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety function during shutdown as well as power operations. Specifically, the likelihood of a fire in the containment building was elevated due to the failure to maintain combustible material (RCP oil) within the boundaries of the oil collection system. Using IMC 0609 Appendix F, Fire Protection Significance Determination Process, the inspectors assumed that the condition represented a low degradation of the fire protection program element of fire prevention through control of combustible materials. Therefore, the finding was determined to be of very low safety significance (Green). No cross-cutting aspect was identified. The issue was not reflective of current licensee performance, since both the bowl drain line configuration (last modified in 1993) and the seal standpipe filling procedure (in place since at least 2000) had been in place for a number of years | |
| 05000327/FIN-2010005-02 | 2010Q4 | Green | H.4 H.5 | NRC identified | Failure to implement Technical Specification requirements to vent ECCS piping | The inspectors identified a Green non-cited violation of Technical Specification 6.8.1(c), Procedures and Programs, for the failure to establish surveillance test procedures to verify that ECCS piping systems were full of water by venting accessible piping high points on the suction side of the ECCS pumps as required by Surveillance Requirement (SR) 4.5.2.b.1. The licensee has entered this issue into their corrective action program as service request 291511. The finding was determined to be greater than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform surveillance tests on the ECCS system reduced the assurance that the system could respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since it was not a design or qualification deficiency, it did not represent the loss of a system safety function or the loss of any equipment trains, and is not potentially risk significant due to seismic, flooding or severe weather initiating events. Because site interdepartmental communication, coordination, and cooperation were not sufficient to identify the impact of changes to ECCS surveillance requirements on existing surveillance test procedures, the cross cutting aspect in the work control component of the human performance area applies to this finding (H.3(b)) |
| 05000327/FIN-2010005-03 | 2010Q4 | Green | NRC identified | Failure to Use Worst Case 6900 VAC Bus Voltage in Design Calculations | The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to assure that applicable regulatory requirements for undervoltage (degraded) voltage protection, including those prescribed in TS section 3.3.14, table 3.3.14-2, were correctly translated into design calculation, SQNETAPAC, AC Auxiliary Power System Analysis, Rev. 36, which evaluated transient motor starting voltages at the beginning of a design basis loss of coolant accident (LOCA). The licensee has entered this into their corrective action program as PER 297671 This finding is more than minor because it affects the Design Control attribute of the Mitigating Systems Cornerstone. It impacts the cornerstone objective of ensuring the availability, reliability, and operability of the 6900 VAC safety buses to perform its intended safety function during a design basis event. The potential availability, reliability, and operability of the 6900 VAC safety buses during a potential degraded voltage condition was impacted as the licensee calculation used a non conservative degraded voltage input, with respect to the values specified in TS, into their safetyrelated motor starting and running calculations. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the finding represented a design deficiency confirmed not to result in the loss of functionality of safety-related loads due to the availability of load tap changers (LTCs) that are installed to improve a degraded voltage condition. The inspectors reviewed the performance deficiency for cross-cutting aspects and determined that none were applicable since this performance deficiency was not indicative of current licensee performance as the design calculation discussed above was not recently performed | |
| 05000327/FIN-2010005-04 | 2010Q4 | Green | Licensee-identified | Licensee-Identified Violation | On October 15, 2010, a one foot gap was discovered in the structural steel above the Unit 1 keyway inside the locked boundary of the Very High Radiation Area. Access to the Unit 1 keyway is obtained by entering through a locked caged door which provides access to an opening to the keyway. The location of the gap created the potential for individual to bypass the cage door and gain unauthorized access to the opening by climbing on the structural steel. 10 CFR Part 20.1602 requires that in addition to the controls specified in 10 CFR Part 20.1601, Control of access to high radiation areas, the licensee shall institute additional measures to ensure that an individual is not able to gain unauthorized or inadvertent access to Very High Radiation Areas. Contrary to the above, the gap in the structural steel provided a method for an individual to gain unauthorized or inadvertent access to the opening. The licensee identified the gap during routine surveys and checks of LHRAs and initiated immediate corrective action to control the VHRA. The licensee removed the lock from the caged door and relocated it to the metal louver covers directly above the opening to the keyway. Additional chains were added to the covers to reinforce and secure the access point directly above the opening. Although this event involved the failure to appropriately control access to a Very High Radiation Area, this finding is of very low safety significance 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 PER 266769. The long term corrective actions include fabricating additional materials to secure the area | |
| 05000327/FIN-2010007-01 | 2010Q2 | Green | P.2 | NRC identified | Violation of 10 CFR 50, Appendix B, Criterion V for Failure to Follow Procedure for Vendor Contact Program | The team identified a Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to properly maintain the vendor contact program for safety-related components. The team identified 37 examples of vendor technical manuals where the associated vendor had not been contacted in over three years. Procedure SPP- 2.5, Vendor Manual Control, required contact to be made with the vendors of safety-related components every three years to ensure that technical manuals . and vendor documents contained the most current and applicable information consistent with the guidance of Generic Letter (GL) 90-03. The team identified 37 examples of vendor manuals and technical documents where the associated vendor had not been contacted in more than three years with several examples extending to almost six years. The licensee entered this issue into their corrective action program with actions to make contact with the vendors for all documents identified as having not been verified with the vendor in over the required three years. This finding was entered into the licensees corrective action program as problem evaluation reports (PERs) 224364 and 224975. As an immediate corrective action, the licensee is ensuring that the vendor manuals and documents associated with safety-related components are being verified as most current with the respective vendors. This finding is more than minor because it affected the Mitigating Systems Cornerstone objective of ensuring the availability and reliability of safety systems, is related to the attribute of Procedure Quality (i.e., Maintenance and Testing (Pre-Event) Procedures) and represented a programmatic break-down which if left uncorrected, could become a more significant safety concern. The team assessed this finding using the SDP and determined that the finding was of very low safety significance (Green) because the inspectors found no documented occurrences where the lack of vendor contact ultimately resulted in the inability of a safety-related component to perform the intended safety function and will be treated as an NCV. The inspectors determined that the thorough evaluation of problems such that the resolutions address problems and extent of conditions, as necessary was a significant cause if this performance deficiency. The plant experienced a reactor trip in 2009 which was determined to have been caused, in part, by a vendor manual associated with a feedwater regulating valve (FRV) not being updated. The FRVs are components with both safety-related and non-safety-related features. The extent of condition of the corrective actions associated with this failed to identify the programmatic breakdown of the TVA vendor contact program for safety-related components. This is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution (P.1.(c)). (Section 1R21.2.3) |
| 05000327/FIN-2010007-02 | 2010Q2 | NRC identified | Degraded Voltage Relay Issue | The team indentified an Unresolved Item (URI) regarding calculations that supported the degraded voltage protection scheme. The calculations that analyzed the Class 1E 6900 VAC and 480 VAC motor loads took credit for using administrative controls for limiting the minimum 161kV offsite power supply bus voltage and credited performance of the non-safety-related automatic load tap changers on the CSSTs to limit the minimum voltage on the Class 1E 6900 VAC and 480 VAC buses. The calculations did not evaluate the Class 1E 6900 VAC and 480 VAC motor loads at the worst case possible low voltages which could drop as low as the bottom end of the acceptable tolerance band of the degraded voltage relays. Offsite power is normally provided to the Class 1E 6900 VAC buses from the 161kV offsite power system through the CSSTs. The CSSTs have automatic load tap changers which are designed to maintain approximately 6900 VAC on the Class 1E buses through a dynamic range of 161kV offsite power supply voltages. The Class 1E 480 VAC buses are then powered from fixed-tap 6900/480VAC transformers powered from the respective Class 1E 6900 VAC buses. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Appendix 8-A, Branch Technical Position PSB-1: Adequacy of Station Electric Distribution system Voltages, Rev. 2(07/1981) is part of the licensing basis for the Sequoyah Nuclear Plant. This document states, in part, that the selection of under-voltage and time-delay setpoints shall be determined from an analysis of the voltage requirements of the Class 1E loads at all onsite distribution levels. Calculation SQNETAPAC, AC Auxiliary Power System Analysis, Rev. 36 evaluated transient motor starting volatges at the beginning of a design basis loss of coolant accident (LOCA) and was based on the voltages where the minimum 161kv offsite power supply bus voltage was limited by taking credit for administrative controls rather than assuming a worst-case 161kV offsite power supply volatge drop which would still allow voltage recovery to the degraded voltage relay reset setpoint (minus septoint tolerance) before the expiration of the degraded voltage relay nominal 9.5 second time delay and thereby leave the Class 1E 6900 VAC buses connected to the offsite power supply. In addition, calculations for motor starting during steady-state conditions credited voltage improvement based on performance of the non-safety related CSST automatic load tap changers instead of being based on worst-case conditions. This issue is unresolved pending further inspection to determine (1) the actual worst-case voltage required to be analyzed on the Class 1E 6900 VAC and 480 VAC buses for safety-related loads in accordance with the facility licensing basis; and (2) the impact of not using the worst-case bus voltage afforded by the degraded voltage protection scheme in safety-related 6900 VAC and 480 VAC motor starting studies. (URI 05000327, 328/2010007-01, Worst Case 6900 VAC Bus Voltage in Design Calculations | ||
| 05000327/FIN-2010404-01 | 2010Q4 | Green | NRC identified | Security | ||
| 05000327/FIN-2010404-02 | 2010Q4 | Green | Licensee-identified | Security | ||
| 05000327/FIN-2011002-01 | 2011Q1 | Green | H.5 | Self-revealing | Reactor Trip due to Unplugged Steam Dump Load Reject Controller | A self-revealing finding was identified for the licensees failure to perform adequate post-maintenance testing, as specified by procedures SPP-8.3, Post- Modification Testing, revision 10, and NPG-SPP-06.3, Pre-/Post-Maintenance Testing, revision 0, in conjunction with a work order which implemented a plant modification on Unit 1 and included the relocation of the steam dump load reject controller. This resulted in a manual trip of Unit 1 following a turbine trip from 26 percent rated thermal power due to the steam dump load reject controller power supply not being properly connected. The licensee entered this issue into their corrective action program as PERs 285349. The licensee implemented corrective actions to include a revision to post-modification testing procedures to require an additional post maintenance testing (PMT) review for large/complex modifications, as well as revision to applicable maintenance procedures to require verification for plugin type connections. The finding was determined to be greater than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, this finding resulted in a reactor trip. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because although it did contribute to the likelihood of a reactor trip, it did not contribute to the likelihood that mitigating systems will not be available. The cause of this finding was determined to have a cross-cutting aspect of Work Planning, in the area of Human Performance associated with the Work Control component. The work planning processes failed to identify the need to include steps to verify the operational status of the controller following completion of the activity, considering the physical conditions and requirements associated with relocating the device. (H.3(a)). |
| 05000327/FIN-2011002-02 | 2011Q1 | Green | NRC identified | \\\"Failure to Adequately Qualify Molded-Case Circuit Breakers to Safety-Related Application Through Commercial Grade Dedication\\\" | The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that appropriate quality standards were specified and included in design documents and that deviations from such standards were controlled. Specifically, the licensee failed to ensure that the molded case circuit breakers utilized in the station 120VAC vital instrument power boards were properly seismically qualified for their application. The licensee entered this issue into their corrective action program as PERs 264271, 266599, 286156, and 319161. Corrective actions included revision of applicable procedures to perform re-alignment of breakers in the vital instrument power boards. The finding was determined to be greater than minor because it was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that the 120VAC vital instrumentation board components had proper seismic qualification had the potential to affect the ability of safety-related equipment to perform its required function under design basis conditions. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it did not represent an actual loss of safety function. No cross-cutting aspect was identified, since the issue was determined to not reflect current licensee performance. | |
| 05000327/FIN-2011002-03 | 2011Q1 | Green | H.7 | Self-revealing | Loss of 480-V Motor Control Center Due To Inadequate Breaker Replacement Maintenance | A self-revealing non-cited violation of Unit 1 TS 6.8, Procedures & Programs, was identified for the licensees failure to provide adequate procedures for maintenance involving the replacement of a safety-related 480V breaker. This resulted in the normal feeder breaker for the safety related 1A2 reactor motor operated valve (MOV) board unexpectedly tripping open when energized following maintenance, causing a loss of power to the board. The licensee entered this issue into their corrective actions program as PER 320274. Licensee corrective actions included revising the applicable breaker maintenance procedure, and reinforcing expectations regarding peer checking and procedure use and adherence. The finding was determined to be greater than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loss of power to the 1A2 reactor MOV board resulted in the inoperability of its associated MOVs affecting two trains of AFW, one train of containment spray (CS), feedwater isolation valves, and containment isolation valves. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since it did not represent an actual loss of safety function of a single train for greater than the associated TS allowed outage time. The cause of this finding was determined to have a cross-cutting aspect in the area of Human Performance associated with the Resources component. The work package was not adequate to assure nuclear safety due to the complexity and ambiguity associated with the procedure step which involved the jumper installation requirement. (H.2(c)). |
| 05000327/FIN-2011002-04 | 2011Q1 | Green | H.7 | Self-revealing | Inadequate Maintenance Procedures Result in Inoperable Feedwater Regulating Valve | A self-revealing non-cited violation of 10 CFR 50 Appendix B Criterion V, Instructions, Procedures, and Drawings, was identified for the failure to provide adequate procedures for maintenance on the Unit 1 loop 3 feedwater regulating valve (FRV). The applicable procedures did not contain adequate guidance to ensure that the valves operability was not adversely affected during reassembly. As a result, the FRV was placed in a condition where it was unable to perform its function of main feedwater isolation as required by TS LCO 3.7.1.6. The licensee entered this issue into their corrective actions program as PERs 284451 and 314771. Corrective actions included revision to the applicable work procedure to ensure no inadvertent valve stem rotation during reassembly. The finding was determined to be greater than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, as a result of the maintenance activity, the FRV was placed in a condition where it was unable to perform its required function of main feedwater isolation. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since the finding did not represent an actual loss of safety function of a single train for greater than the associated TS allowed outage time. The cause of this finding was determined to have a cross-cutting aspect in the area of Human Performance associated with the Resources component, in that a complete work package which was adequate to assure nuclear safety was not provided for this maintenance activity. The work procedures did not include guidance to ensure that the operability of the FRV was not adversely affected. (H.2(c)). |
| 05000327/FIN-2011003-01 | 2011Q2 | Green | NRC identified | Failure to Perform Instrumentation Surveillance Testing within Required Frequency | The inspectors identified a non-cited violation of Units 1 and 2 TS Surveillance Requirement (SR) 4.0.2 for the licensees failure to perform SRs specified in Units 1 and 2 TS 3/4.3.1, Reactor Trip System Instrumentation, and 3/4.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, within the required surveillance frequencies. The inspectors identified eight examples over the last three years (five examples on Unit 1 and three examples on Unit 2) where the interval between tests of the automatic actuation logic and reactor trip breaker functions required by SRs 4.3.1.1.1 and 4.3.2.1.1 exceeded the maximum surveillance interval allowed by TS. The licensee entered this issue into their corrective action program as PER 369938. Corrective actions included ensuring that work control processes correctly implement the required surveillance intervals. The finding was determined to be greater than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, extending beyond the required maximum interval between TS surveillance tests affects the ability to confirm continued availability of TS equipment, and the ability to detect potential latent operability concerns in a timely manner. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since it did not represent an actual loss of safety function of a single train for greater than the associated TS allowed outage time. The inspectors did not identify that the cause of this finding was related to any of the cross-cutting aspects defined in IMC 0310, and therefore no cross-cutting aspect was assigned to this finding | |
| 05000327/FIN-2011003-02 | 2011Q2 | Green | Licensee-identified | Licensee-Identified Violation | The following violations of very low safety significance (Green) or Severity Level IV were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. TS 6.12.1 requires that entryways into HRAs with dose rates not exceeding 1 rem/hour at 30cm be barricaded. Contrary to this, on May 31, 2011, and again on June 16, 2011, HRA entryways into U2 Containment were not barricaded. In both examples, the accessible areas of U2 Containment contained dose rates >100mrem/hr at 30cm, but less than 1 rem/hr at 30cm. On May 31, without HP present, workers in the area repositioned the HRA swing gates to facilitate the installation of rail track into U2 Upper Containment. In the second example, on June 16, the HRA swing gate to U2 Lower Containment was propped open with scaffold leveling legs during demobilization. In both cases, the boundary re-positioning was such that the swing gates no longer provided adequate barriers to check the advance of an oncoming worker. These violations were discovered by HPTs performing their normal radiological control duties. Immediate corrective actions were taken upon discovery and documented in PERs 379547 and 390159. Although these events involved the failure to maintain proper controls for HRAs, this finding is of very low safety significance because there was no evidence of unauthorized worker entry into the affected areas, nor any unexpected radiation exposures to licensee personnel | |
| 05000327/FIN-2011004-01 | 2011Q3 | Severity level IV | NRC identified | Failure to Report System Actuation | The inspectors identified a non-cited violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to report an invalid system actuation. On May 5, 2011, a containment ventilation isolation (CVI) signal was inadvertently generated on Unit 2 while performing surveillance testing. This system actuation was not reported to the NRC as required by 10 CFR 50.73(a)(2)(iv) within 60 days of discovery of the event. This issue was entered into the licensees corrective action program as PER 417453, and was reported to the NRC as EN #47249 on September 8, 2011. This violation was determined to be applicable to traditional enforcement because of its potential to impact the ability of the NRC to perform its regulatory oversight function, and was therefore evaluated in accordance with the NRC Enforcement Policy. This issue was determined to be a Severity Level IV violation in accordance with Section 6.9.d.9 of the NRC Enforcement Policy. No cross-cutting aspect was assigned since traditional enforcement violations for which there are no associated ROP findings are not screened for cross-cutting aspects. | |
| 05000327/FIN-2011005-01 | 2011Q4 | Green | H.8 | Self-revealing | Failure to Follow Procedure for Loss of Power Diesel Generator Start Instrumentation Surveillance Testing | A self-revealing non-cited violation of Unit 2 Technical Specification (TS) 6.8.1.a was identified for the licensees failure to follow station procedures during the performance of a surveillance testing activity. While performing degraded voltage/load shed relay testing associated with the 2B 6.9kV shutdown board, the use of improper test equipment and the incorrect connection of test equipment resulted in a control power circuit fuse being blown, which caused inoperability of an emergency diesel generator and a motor driven auxiliary feedwater train. This issue was entered into the licensees corrective action program as Problem Evaluation Report (PER) 415324. The finding was determined to be greater than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to follow procedure steps resulted in inoperability of the 2B emergency diesel generator and the 2B motor driven train of auxiliary feedwater. Using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since it did not represent an actual loss of safety function of a single train for greater than the associated TS allowed outage time. The cause of this finding was determined to have a cross-cutting aspect in the area of Human Performance associated with the Work Practices component. The licensee failed to adequately implement human error prevention techniques such as self and peer checking (e.g. concurrent verification) while connecting test equipment. Additionally, maintenance personnel failed to question the use of test equipment which was different than what was specified in the procedure (i.e. proceeding in the face of uncertainty). |
| 05000327/FIN-2011005-02 | 2011Q4 | Green | H.14 | NRC identified | Failure to Maintain Thermal Power Less Than Licensed Limit | The inspectors identified a non-cited violation of Unit 1 Operating License DPR-77 Condition 2. (C).1 Maximum Power Level for the licensees failure to take prudent action to ensure that the licensed power limit was not exceeded during a pre-planned evolution which involved manual reactivity manipulations. Prompt action was not taken by operators to reduce power when reactor thermal power exceeded the licensed power limit during a control rod full out position reset activity. Additionally, prudent action to sufficiently reduce power prior to the activity to accommodate the power transient was not taken. This issue was entered into the licensees corrective action program as PER 437068. The finding was determined to be greater than minor because it was sufficiently similar to example 8.a. of IMC 0612 Appendix E, in that: 1) prudent action based on prior performance was not taken to reduce power prior to performing the evolution, and 2) operators did not promptly lower thermal power once the licensed limit was exceeded. Additionally, if left uncorrected the finding would have the potential to lead to a more significant safety concern, since operation above the licensed power limit has the potential to reduce analyzed margins to fuel cladding damage. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) under the barrier integrity cornerstone since only the fuel cladding barrier criterion was applicable. The cause of this finding was determined to have a cross-cutting aspect in the area of Human Performance associated with the Decision-Making component. Elements of non-conservative decision making which contributed to this performance deficiency included: (1) Prior to February 2011, station procedures required operators to monitor and maintain the 10-minute average of thermal power below the licensed limit. These requirements were revised and replaced with requirements to maintain the 1-hour average of thermal power below the licensed limit. This was a non-conservative decision made without due consideration of potential consequences or the need for supplemental guidance to maintain an appropriate and conservative approach to controlling thermal power under non-steady state conditions. (2) The decision was made to proceed with the rod withdrawal activity under the non-conservative assumption that there would be negligible reactivity effects, and without considering available data from previous performances. This did not reflect a philosophy of demonstrating that a proposed activity is safe in order to proceed rather than demonstrating that it is unsafe in order to disapprove the action. |
| 05000327/FIN-2011005-03 | 2011Q4 | Green | H.12 | Self-revealing | Nuclear Instrumentation System Channel Calibration Error | A self-revealing non-cited violation of Unit 2 TS 3.0.3 was identified for the licensees failure to place the unit in Mode 3 within seven hours when a Limiting Condition for Operation (LCO) was not met in Modes 1 and 2. The requirements of LCO 3.3.1, Reactor Trip System Instrumentation, associated with the power range neutron flux function in Modes 1 and 2 were not met for a period of approximately 24 hours. This was the result of an error made during the performance of a channel calibration activity, which caused one channel to be left in an inoperable condition. This issue was entered into the licensees corrective action program as PER 397142. The finding was determined to be greater than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inoperability of the N44 power range Nuclear Instrucment System (NIS) channel low range neutron flux trip function resulted in the failure to meet TS operability requirements associated with reactor trip system instrumentation. Using Inspection IMC 0609, Significance Determination Process, (SDP) Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be applicable to Phase 2 SDP screening since it represented the loss of a channel of a TS required function for greater than its TS allowed outage time. A Phase 2 analysis using Saphire 8 software with the Sequoyah SPAR model in the SDP mode was performed by a regional SRA. Using an exposure period of 1 day with a truncation value of 1E-13, a result of CDF << 1E-6, or very low safety significance (Green), was obtained. The cause of this finding was determined to have a cross-cutting aspect in the area of Human Performance associated with the Work Practices component. The licensee failed to adequately implement human error prevention techniques, such as self and peer checking, to ensure that the work activity was being performed on the correct component. |
| 05000327/FIN-2011005-04 | 2011Q4 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation. Title 10 CFR 50, Part 50.65(a)(4), states, in part, that before performing maintenance activities, licensee personnel shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, on November 16, 2011, before performing maintenance activities, licensee personnel did not adequately assess the increase in risk from the proposed maintenance activities. Specifically, the licensee failed to assess the increase in risk associated with Essential Raw Cooling Water (ERCW) pumps Q-A and R-A being unavailable for maintenance. This problem was entered into the licensees corrective action program as PER 465734. The finding was screened using Inspection Manual Chapter 0609, Appendix K Maintenance Risk Assessment and Risk Management Significance Determination Process, and was determined to be of very low safety significance. | |
| 05000327/FIN-2011006-01 | 2011Q1 | NRC identified | Use of Operator Manual Actions in Lieu of Protecting Cables and Equipment Required for POST-FIRE Safe Shutdown | The inspectors identified an unresolved item (URI) related to the licensees failure to protect cables and equipment required to achieve hot shutdown to ensure that one train was free of fire damage in accordance with the requirements of the approved fire protection program as described in SQN Operating License Conditions (OLCs) 2.C(16) and 2.C(13) for Units 1 and 2 respectively. The inspectors reviewed the licensees post-fire SSD methodology for a fire in FA FAA-054/Room A01 or FA FAA-081/Room A24. Post-fire SSD for either of these FAs would be achieved from the MCR using AOP-N.01, Plant Fires, and AOP-N.08, Appendix R Fire Safe Shutdown. During this review the inspectors noted that in lieu of protecting the cables of equipment identified by the licensee as being required for post-fire SSD (i.e., provide barriers and spatial separation with detection and suppression), the SQN SSD methodology (for the above two FAs and numerous other FAs) also credited the use of OMAs. These new OMAs were not submitted to the NRC for review and/or approval prior to incorporation into the plant AOPs. Sequoyah added these new OMAs to their FPP in 2002 without seeking prior NRC approval because they believed their current OLCs (approved by NRC SER dated August 12, 1997 and applicable to plants licensed after January 1, 1979) allowed them to make these changes without prior NRC approval if they could demonstrate that the changes had no adverse impact on SSD. Sequoyah Units 1 and 2 were both licensed to operate after January 1, 1979. The SQN evaluation (dated March 2002) which added the OMAs concluded that adding the OMAs did not adversely affect SSD because the licensee determined at that time that the OMAs were feasible. The inspectors concluded that the licensees methodology of allowing fire damage to occur (instead of protecting SSD cables and equipment) and relying on OMAs to achieve post-fire SSD would adversely affect the ability to achieve and maintain SSD in the event of a fire. The inspectors reviewed the SQN evaluation which added the OMAs and concluded that the evaluation was not adequate to support the conclusion that adding the OMAs did not adversely affect post-fire SSD because the evaluation did not address defense-in-depth. During this inspection, SQN personnel stated that there were no plans at the time to implement modifications or submit the OMAs added in March 2002 for NRC review and approval. Subsequent to the onsite inspection, the licensee submitted additional information to the inspectors in support of their conclusion that the OMAs added in 2002 did not affect SSD. Additionally, the licensee initiated Problem Evaluation Report (PER) 324757 to address this issue. The inspectors are reviewing this information relative to the SQN licensing basis. This issue is identified as unresolved item (URI) 05000327, 328/2011006-01, Use of Operator Manual Actions in Lieu of Protecting Cables and Equipment Required for Post-Fire Safe Shutdown. This issue is unresolved pending further NRC review of this information and the Sequoyah fire protection licensing basis | ||
| 05000327/FIN-2011006-02 | 2011Q1 | Green | H.8 | NRC identified | Failure to Establish Compensatory Actions for Blocked Sprinklers | The inspectors identified a non-cited violation of Sequoyah Operating License Conditions 2.C.(16) and 2.C.(13) for Units 1 and 2, respectively, for failure to establish compensatory measures for an obstructed sprinkler system. Specifically, scaffolding installed in auxiliary building fire area FAA-054/Room A01 was in a configuration which obstructed sprinkler heads A198 and A208. The licensee entered this issue into the corrective action program as Problem Evaluation Report 321911 and implemented compensatory measures (fire watches) in accordance with the approved fire protection program. This finding was determined to be a performance deficiency because the licensee did not provide the required compensatory measures (fire watch) for sprinkler heads that were obstructed or blocked by scaffolding. This finding was more than minor because it affected the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire) and it affected fire protection defense-in-depth strategies involving suppression systems. Failure to provide compensatory measures for obstructed sprinkler heads A198 and A208 would have reduced the licensees ability to quickly extinguish a fire in the area. The inspectors evaluated this issue in accordance with Inspection Manual Chapter (IMC) 0609, Appendix F, Fire Protection Significance Determination Process, and determined that this finding was of very low safety significance (Green). The finding category of fixed fire protection systems was assigned, based upon that element of the plant fire protection program being impacted. The inspectors assigned a low degradation rating that reflected the severity of the observed deficiency. This rating was based upon meeting the criteria described in IMC 0609, Appendix F, Attachment 2 for fixed fire detection and suppression degradation. Specifically, less than 10% of the sprinkler heads were non-functional, there were functional heads within 10 feet of the combustibles of concern and the system was nominally code compliant. This finding was determined to have a cross cutting aspect in the Human Performance Area, Work Practices Component, because the licensee failed to define and effectively communicate expectations regarding procedural compliance and requirements for personnel to follow procedures |
| 05000327/FIN-2011006-03 | 2011Q1 | Green | NRC identified | Sprinkler System in Room 690.0-A1 of the Auxiliary Building Has Nfpa Code Deviation | The inspectors identified a non-cited violation of Operating License Conditions 2.C.(16) and 2.C.(13), for Units 1 and 2 respectively, for failure to install the automatic suppression system (sprinkler system) in the auxiliary building corridor 690 foot elevation, in accordance with applicable National Fire Protection Association (NFPA) Standard No. 13, Automatic Sprinkler Systems. Specifically, NFPA 13- 1975 required sprinklers to be installed within 12-inches of the ceiling. Portions of the auxiliary building sprinkler system were installed greater than 12-inches below the ceiling. As a result, the actuation of the fusible link type sprinklers would have been slower than originally intended after fire ignition. The licensee entered this issue into the corrective action program as Problem Evaluation Report 147467 and implemented compensatory measures (fire watches) in accordance with the approved fire protection program. This finding was determined to be a performance deficiency because the licensee did not locate the sprinkler heads according to the applicable Code of Record for the facility. The finding was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e. fire). It affected the objective of ensuring the reliability and capability of systems that respond to initiating events. The inspectors evaluated this issue in accordance with Inspection Manual Chapter (IMC) 0609, Appendix F, Fire Protection Significance Determination Process, and determined that this finding was of very low safety significance (Green). Pursuant to IMC 0609, Appendix F, the finding category was fixed fire protection systems. The finding was of very low safety significance because of the limited number of sprinkler heads improperly installed, the proximity of the existing sprinklers to combustibles of concern, and the suppression system was nominally code compliant for all other aspects of design and installation. The inspectors determined that there was no cross-cutting aspect associated with this finding because the condition has existed since initial plant licensing and was not reflective of present performance | |
| 05000327/FIN-2011006-04 | 2011Q1 | Green | Licensee-identified | Licensee-Identified Violation | Sequoyah Fire Protection License Condition 2.C.(16) for Unit 1 requires that TVA implement and maintain in effect all provisions of the approved fire protection program referenced in the Sequoyah Nuclear Plants Final Safety Analysis Report (FSAR) as described in the Fire Protection Report (FPR). Part III of the FPR states that SQN must comply with 10 CFR 50 Appendix R, Section III.G. Section III.G.2 requires that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, on January 28, 2009, the licensee identified that cables associated with the Unit 1 6900-volt power supply in fire areas FAA 1 and 29 would not be free of fire damage to support safe shutdown (see Section 40A3). This violation is of very low safety significance (Green). This issue was determined to be of very low safety significance based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase II Quantitative Screening Approach. This violation was documented in the licensees CAP as PER 162189. | |
| 05000327/FIN-2011202-01 | 2011Q4 | Green | NRC identified | Security | ||
| 05000327/FIN-2011402-01 | 2011Q2 | Green | H.7 | NRC identified | Security | |
| 05000327/FIN-2011402-02 | 2011Q2 | Green | H.8 | NRC identified | Security | |
| 05000327/FIN-2012002-01 | 2012Q1 | Green | H.8 | NRC identified | Failure to Implement Procedures for Tornado Watch/Warning | The inspectors identified a non-cited violation of Units 1 & 2 Technical Specification of 6.8.1.a for the licensees failure to adequately implement procedure AOP-N.02, Tornado Watch/Warning, Revision 28. On March 2, 2012, the licensee entered AOP-N.02 due to a tornado watch/warning and failed to secure or remove loose material in the Switchyard/Transformer Yard as required by the procedure. This issue was entered into the licensees corrective action program as Problem Evaluation Report (PER) 515684. The performance deficiency was determined to be more than minor because it was associated with the protection against external factors attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to secure or remove potential missile hazards from the Switchyards/Transformer Yard increased the likelihood of a Unit trip and/or loss of offsite power event. The inspectors evaluated the significance of the finding using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The finding was determined to have a crosscutting aspect in the Work Practices component of the Human Performance cross-cutting area since the licensee failed to define and effectively communicate expectations regarding procedural compliance, and personnel failed to follow procedures. |
| 05000327/FIN-2012002-02 | 2012Q1 | Green | H.9 | NRC identified | Failure to Meet Fire Drill Training Requirements | The inspectors identified a non-cited violation of facility operating license DPR-77 condition 2.C.(16) and facility operating license DPR-79 condition 2.C.(13) for failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, Sequoyahs Fire Protection Report Part II, Section 9.3.b.2 Fire Drills requires a minimum of one drill per shift every calendar quarter, a minimum on one unannounced drill per shift per year, at least one drill per shift per year is performed on a backshift for each fire brigade, and fire brigade members including leaders shall participate in at least two drills per year. The inspectors identified multiple examples of the licensees failure to meet these requirements in calendar years 2010 and 2011. This issue was entered into the licensees corrective action program as Problem Evaluation Report (PER) 513378, 512736, and 527875. The performance deficiency was determined to be greater than minor because it was associated with the protection against external events 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 lack of adequate drill performance could negatively affect the fire brigades capability to combat a fire. Findings associated with performance of the fire brigade are not evaluated using IMC 0609, Attachment F, Fire Protection Significance Determination Process, and require NRC management review using Appendix M, Significance Determination Process Using Qualitative Criteria, as described in NRC Inspection Manual Chapter 0609.04, Table 3b, Phase 1 - Initial Screening and Characterization of Findings. Regional management concluded that the finding was of very low safety significance (Green) because it involved fire drill training rather than performance during an actual fire event. The finding was determined to have a crosscutting aspect in the Resources component of the Human Performance cross-cutting area since the licensee failed to ensure that personnel are available and adequately trained to assure nuclear safety. |
| 05000327/FIN-2012002-03 | 2012Q1 | Green | P.3 | NRC identified | Failure to Follow Corrective Action Program Procedures | The inspectors identified a finding for the licensees failure to meet the requirements of corrective action program procedure NPG-SPP-03.1.7, PER Actions, Revision 2. Specifically, the licensee failed to ensure that the corrective action plan and associated actions addressed the required action and schedule associated with PER 432510, which documented the need to address a condition involving water accumulation in manhole locations containing electrical cable runs. This issue was entered into the licensees corrective action program as Problem Evaluation Report (PER) 433761, 432510, and 505259. This issue was determined to be greater than minor because if left uncorrected, the issue could become a more significant safety concern. Specifically, the failure to take corrective action as required to address manhole water accumulation in a timely manner could result in the submergence of safety related cables and potentially adversely affect the ability of safely related equipment to perform required functions. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because although it did contribute to the likelihood that mitigating systems will not be available, no actual loss of safety functions occurred. The cause of this finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution associated with the corrective action program because the licensee failed to take appropriate corrective actions in a timely manner, commensurate with their safety significance and complexity. |
| 05000327/FIN-2012002-04 | 2012Q1 | Severity level IV | NRC identified | Inadequate 10 CFR 50.59 Evaluation for Implementation of Manual Actions to Cool RHR Suction Piping During a Mode 4 Loss of Coolant Accident | The inspectors identified a Severity Level IV (SL-IV) non-cited violation (NCV) of 10 CFR 50.59, Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing local operator manual action (OMA) changes to Technical Specifications (TS) Bases 3.5.3 and abnormal operating procedure (AOP)-R.02 that were specified in Engineering Document Change (EDC) 22487. The 10 CFR 50.59 performed to support EDC 22487 was inadequate in that the 50.59 did not identify that prior NRC approval was required for implementation of the changes. Specifically, the licensee revised AOPR. 02, Shutdown LOCA, and TS Bases 3.5.3, ECCS - Shutdown, to include OMAs to cool the residual heat removal (RHR) system suction piping as part of RHR realignment to establish emergency core cooling system (ECCS) flow in the event of a loss-of-coolant-accident (LOCA) while RHR was aligned to the reactor coolant system for shutdown cooling in operational Mode 4. The new OMAs added for cooling the RHR suction piping had, in effect, changed the intent of the note in TS limiting condition for operation 3.5.3, and were beyond the scope of what the NRC had previously reviewed and approved in Technical Specification Change 07-05. The licensee entered this issue into the corrective action program as problem evaluation report 535471. The finding was determined to be more than minor because prior NRC review and approval was required before changing the AOP and the TS Bases to include the OMAs for cooling the RHR suction piping as part of ECCS realignment in the event of a Mode 4 LOCA. The inspectors reviewed this issue, in accordance with Inspection Manual Chapter 0612 and the NRC Enforcement Policy, and determined that traditional enforcement was applicable to this issue because it impacted the ability of the NRC to perform its regulatory oversight function. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The inspectors determined that this finding was of very low safety significance because, since implementation of EDC 22487, the OMAs to cool the RHR suction piping would not have been required if a LOCA had occurred during the times that RHR shutdown cooling was in service in Mode 4. The finding was reviewed by NRC management and because the violation was determined to be of very low safety significance, was not willful or repetitive, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation, consistent with the NRC Enforcement Policy. The violation was not screened for associated cross-cutting aspects because it involved traditional enforcement. | |
| 05000327/FIN-2012002-05 | 2012Q1 | Green | H.5 | Self-revealing | Turbine Driven Auxiliary Feedwater Pump Inoperable Due to Overspeed Trip | A self-revealing NCV of Unit 1 TS 6.8, Procedures & Programs, was identified for the licensees failure to provide adequate procedures for a maintenance activity involving the required inspection of an Essential Raw Cooling Water (ERCW) pipe leak in the Unit 1 Turbine Driven Auxiliary Feedwater Pump (TDAFW) room. This resulted in water intrusion into the governor control cabinet of the TDAFW Pump, which caused the pump to be inoperable due to an electrical overspeed trip caused by fluctuating speed indications. This issue was entered into the licensees corrective action program as Problem Evaluation Report (PER) 470310. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 1 TDAFW pump was made inoperable due to water intrusion into an AFW control system cabinet. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since it did not represent an actual loss of safety function of a single train for greater than the associated TS allowed outage time. The finding was determined to have a crosscutting aspect in the Work Control component of the Human Performance cross-cutting area since the licensee failed to appropriately plan and coordinate work activities, consistent with nuclear safety. |
| 05000327/FIN-2012002-06 | 2012Q1 | Green | NRC identified | Change to Fire Protection Program Which Adversely Affected Safe Shutdown Without Prior NRC Approval | The inspectors identified a non-cited violation of Sequoyah operating license conditions 2.C. (16) and 2.C. (13) for Units 1 and 2 respectively, for a change made to the Sequoyah fire protection program which was determined to adversely affect safe shutdown (SSD), without prior NRC approval. Specifically, in lieu of protecting the cables and equipment to ensure that one train of equipment required for SSD was free of fire damage, the licensee made a change to the Sequoyah fire protection program in 2002 that added new operator manual actions (OMAs) to achieve SSD, without prior NRC approval. The evaluation performed in 2002 for the new OMAs was not adequate to support the conclusion that adding the OMAs did not adversely affect post-fire SSD because the evaluation only addressed OMA feasibility and did not address defense-in-depth. The licensee entered this issue in the corrective action program as problem evaluation report 324757 to track resolution. This finding was determined to be more than minor because it affected the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire). The inspectors evaluated this finding in accordance with Inspection Manual Chapter 0609, Significance Determination Process (SDP), Appendix F, Fire Protection Significance Determination Process. The inspectors performed a Phase 1 and Phase 2 SDP screening assessment using IMC 0609, Appendix F, Attachment 1. Based upon the SDP Phase 2 results, the inspectors were not able to screen out the issue. Further evaluation was conducted during a Phase 3 analysis performed by the senior reactor analyst. The analyst determined that the risk significance of the OMAs associated with the change made to the Sequoyah fire protection program was very low (3.13E-8, Green). The main contributors to the low risk results were: (1) the low likelihood of a human failure given the extensive time available to take the OMAs, (2) the redundancy of opposite train equipment for severely damaging fires in certain fire areas, and (3) the low number of fire areas and fire sequences that were ultimately analyzed. The inspectors determined that there was no cross-cutting aspect associated with this finding because the change to the fire protection program occurred in 2002 and was not reflective of current licensee performance. | |
| 05000327/FIN-2012002-07 | 2012Q1 | Green | Licensee-identified | Licensee-Identified Violation | Unit 1 facility operating license DPR-77 condition 2.(C).13 requires that TVA shall implement and maintain in effect all provisions of the approved fire protection program referenced in Sequoyah Nuclear Plants Final Safety Analysis Report and as approved in applicable NRC Safety Evaluation Reports. The Sequoyah Fire Protection Report Part II, Section 14.6, LCO 3.7.12.a requires that with one or more required fire barrier penetration non-functional, within one hour restore the inoperable equipment or: establish a continuous fire watch on at least one side of the affected penetration, or verify the operability of fire detectors on at least one side of the non-functional fire barrier and establish an hourly fire watch patrol. Contrary to the above, on December 27, 2011, the licensee failed to complete the required actions for establish a fire watch when two fire barrier penetrations were breached during a planned maintenance activity. This problem was entered into the licensees corrective action program as PER 484654. The finding was screened using Inspection Manual Chapter 0609, Appendix F Fire Protection Significance Determination Process, and was determined to be of very low safety significance | |
| 05000327/FIN-2012003-01 | 2012Q2 | Green | P.6 | NRC identified | Failure to Evaluate Fire Drill | The inspectors identified a non-cited violation of Units 1 & 2 Technical Specification 6.8.1.f for the licensees failure to implement procedures required for fire protection program implementation. Specifically, the licensee failed to evaluate six minimum critical objectives on December 5, 2011, during a fire drill as required by TVA-SPP-17.16, Conduct and Evaluation of Fire Drills, revision 0. This issue was entered into the licensees corrective action program as Problem Evaluation Reports (PERs) 538996, 568242, and 568248. The performance deficiency was determined to be greater than minor because it was associated with the protection against external events 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 lack of adequate drill performance could negatively affect the fire brigades capability to combat a fire. Findings associated with performance of the fire brigade are not evaluated using IMC 0609, Attachment F, Fire Protection Significance Determination Process, and Appendix M, Significance Determination Process Using Qualitative Criteria, as described in NRC Inspection Manual Chapter 0609.04, Table 3b, Phase 1 - Initial Screening and Characterization of Findings. The NRC concluded that the finding was of very low safety significance (Green) because the defense-in-depth attribute of the fire brigade was minimally affected, in that, the evaluated crew was only one of four crews of the site fire brigade team, the other crews had adequately been evaluated, and that the overall condition of the fire detection and suppression systems has been satisfactory. The finding was determined to have a crosscutting aspect in the area of Problem Identification and Resolution because of inadequate oversight and self-assessment of fire operations department activities, specifically fire brigade training. |
| 05000327/FIN-2012003-02 | 2012Q2 | Green | H.7 | NRC identified | Turbine Throttle Valve Reactor Trip Function Degraded | The inspectors identified a Green NCV of Unit 1 TS 6.8, Procedures & Programs, for the licensees failure to provide adequate procedures for maintenance and surveillance activities involving the main turbine throttle valves and the associated solid state protection system (SSPS) function which provides a reactor trip on turbine trip signal. The failure to include applicable torque requirements for set screws associated with the limit switch lever arm assembly resulted in one of the four turbine throttle valve position limit switches being in an inoperable condition such that the SSPS function of reactor trip on turbine trip, which involves a four-out-of- four logic, was inoperable and could not have functioned if required. This issue was entered into the licensees corrective action program as Problem Evaluation Reports (PERs) 419594 and 518647 The finding was determined to be greater than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding resulted in the inability of the SSPS to provide the required reactor trip signal upon closure of all four turbine throttle valves above 50 percent RTP. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since the trip is not credited in any Updated Final Safety Analysis (UFSAR) Chapter 15 accident analysis and the redundant reactor trip on turbine trip function that is based on low auto stop oil pressure was unaffected. The cause of this finding was determined to have a cross-cutting aspect in the area of Human Performance, Resources component, and the aspect of complete and accurate procedures and work packages. The procedures for performing maintenance and surveillance activities associated with the turbine throttle valves and associated SSPS function were not adequate to assure nuclear safety due to the failure to include applicable torque requirements for the components associated with the valve limit switch assembly. |
| 05000327/FIN-2012003-03 | 2012Q2 | Severity level IV | Licensee-identified | Licensee-Identified Violation | The violation referenced in Section 4OA5.2 was identified by the licensee and meets the criteria of Section 2.3.2.b of the NRC Enforcement Policy for characterization as a Non- Cited Violation. This issue is in the licensees CAP as PER 452027, MPC Forced Helium Dehydration Calculation Methodology. This is a Severity Level IV violation and is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.b of the NRC Enforcement Policy; specifically, the violation was identified by the licensee, the issue was placed into the licensees CAP, the violation was not repetitive as a result of inadequate corrective action, and the violation was not willful. Documents reviewed are listed in the Attachment. | |
| 05000327/FIN-2012004-01 | 2012Q3 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50 Appendix B, Criterion V, required, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to this, on August 25, 2011, and again on September 6, 2012, the licensee failed to accomplish NPG-SPP-09.17, Temporary Equipment Control, an activity affecting quality, in accordance with the documented instructions. Specifically, Appendix A, Step C states, Specific Case Mechanical Engineering Evaluation is required if Temporary Equipment is placed inside primary containment during unit Operating Modes 1 through 4 at SQN. However, this step was not performed and material was brought into primary containment without a proper evaluation. This problem was entered into the licensees corrective action program as PERs 599247 and 604614. The finding was screened using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding to be of very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. | |
| 05000327/FIN-2012005-01 | 2012Q4 | Green | H.9 | NRC identified | Failure to Implement Freeze Protection Program Requirements | A NRC-identified Green non-cited violation (NCV) of Unit 1 and 2 Technical Specification 6.8.1.a for the licensees failure to follow station procedures to adequately implement freeze protection requirements. Specifically, inspectors found a number of requirements improperly executed with no specific follow-up of those requirements contained within periodic instructions used to verify program implementation. The licensee placed the issue into the CAP and corrected the identified deficiencies. The inspectors determined that the failure to adequately implement all requirements of the licensees freeze protection program procedures was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating System Cornerstone attribute of Protection against External Factors and adversely affected the cornerstone objective in that specific measures required for freeze protection were not properly implemented and station procedures did not maintain those expected conditions. The inspectors determined the finding was of very low safety significance (Green) as the site had not experienced significant freeze conditions yet this season. The cause of this finding was related to the cross-cutting aspect of ensuring personnel training is adequate to assure nuclear safety. |
| 05000327/FIN-2012005-02 | 2012Q4 | Green | H.2 | NRC identified | Failure to Implement Fire Protection Impairment Requirements | The inspectors identified a Green non-cited violation of Units 1 & 2 Technical Specification 6.8.1.f for the licensees failure to implement procedures required for fire protection program implementation. The inspectors found multiple examples of where fire watches were not conducted in accordance with procedure NPG-SPP- 18.4.6, Control of Fire Protection Impairments, Revision 1, when required. The licensee entered this issue into the CAP program as PERs 635934 and 635934. Failure of the licensee to implement the requirements of procedure NPG-SPP-18.4.6, Control of Fire Protection Impairments, Revision 1, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) 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 failure to perform compensatory measures (fire watches), could have potentially compromised the ability to safely shutdown the plant in the event of a fire in any of the fire zones where the fire watches were required. The significance of this finding was evaluated in accordance with the IMC 0609 Attachment 4, Phase 1- Initial Screening and Characterization of Findings, which required further evaluation in accordance with Appendix F, Attachment 01, Part 1, Fire Protection SDP Phase 1 Worksheet. The finding was assigned to the fire prevention and administrative controls category and represented a low degradation level. The inspectors concluded that the finding was of very low safety significance (Green) based on a qualitative screening and the low degradation rating. The finding was determined to have a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area (H.4(c)) since the licensee failed to ensure that there was adequate supervisory and management oversight of fire watches. |
| 05000327/FIN-2012005-03 | 2012Q4 | Green | H.7 | NRC identified | Failure to Establish Adequate Procedures for Fire Protection Impairment Requirements | The inspectors identified a Green non-cited violation of Units 1 & 2 Technical Specification 6.8.1.f for the licensees failure to establish adequate procedures required for fire protection program implementation. Specifically, NPG-SPP-18.4.6, Control of Fire Protection, Revision 1 Impairments was determined to be inadequate because it did not provide any guidance on what a fire watch was supposed to do when they came to a protected door. The licensee entered this issue into the CAP program as PER 652672. Failure of the licensee to establish adequate procedures required for fire protection program implementation was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) 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 failure to establish adequate procedures required for fire protection program implementation caused compensatory measures (fire watches) to not be adequately completed and could have potentially compromised the ability to safely shutdown the plant in the event of a fire in any of the fire zones where the fire watches were required. The significance of this finding was evaluated in accordance with the IMC 0609 Attachment 4, Phase 1- Initial Screening and Characterization of Findings, which required further evaluation in accordance with Appendix F, Attachment 01, Part 1, Fire Protection SDP Phase 1 Worksheet. The finding was assigned to the fire prevention and administrative controls category and represented a low degradation level. The inspectors concluded that the finding was of very low safety significance (Green) based on a qualitative screening and the low degradation rating. The finding was determined to have a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area (H.2(c)) for failure to provide adequate procedures for individuals conducting fire watches. |
| 05000327/FIN-2012005-04 | 2012Q4 | Green | P.5 P.2(b) | NRC identified | Failure to Perform ISI General Visual Examinations of Containment Moisture Barrier Associated with Containment Liner Leak Chase Test Connection Threaded Pipe Plugs | 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 metal-to-metal pipe plugs installed in the containment liner channel weld leak chase test connections that provide a moisture barrier to the containment liner seam welds. Following the inspectors identification of this issue, the licensee conducted the visual examinations on all eight of the leak chase test connection upper cavities. These visual examinations revealed significant corrosion of the upper cavities, including one through-wall hole in the tubing leading down to the leak chase channels. Upon further inspection of the channels using a boroscope, the licensee noted water in the channels and corresponding corrosion. No through-wall condition was noted in any leak chase channel, and corrosion was limited to a very small percentage of the liner plate thickness. The licensee adequately evaluated the deficiencies prior to entering Mode 4 (Hot Shutdown) to ensure the integrity of containment was maintained. The issue was entered into the licensees CAP as problem evaluation report (PER) 636215. The failure to conduct a general visual examination of 100 percent of the moisture barriers intended to prevent intrusion of moisture against inaccessible areas of the containment liner at metal-to-metal interfaces which are not seal welded, was a performance deficiency that was within the licensees ability to foresee and correct. This finding was of more than minor significance because the failure to conduct required visual examinations and identify the degraded moisture barriers which allowed the intrusion of water into the liner leak chase channel, if left uncorrected, would have resulted in more significant corrosion degradation of the containment liner or associated liner welds. The finding was associated with the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, visual examinations of the containment metal liner provide assurance that the liner remains capable of performing its intended safety function. The inspectors used IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment. The inspectors identified a cross-cutting aspect in the Operating Experience component of the CAP cross-cutting area (P.2(b)). Specifically, the licensee failed to apply available Operating Experience from four other relevant industry issues to assure plant performance. |
| 05000327/FIN-2012005-05 | 2012Q4 | Green | P.2 | NRC identified | Failure to Adequately Evaluate and Qualify Molded Case Circuit Breakers | The inspectors identified a violation with several examples of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to implement design control measures that review for suitability of application of materials, parts, and equipment that are essential to the safety-related functions of the structures, systems, and components and that provide for verifying or checking the adequacy of design such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program, including qualification testing of a prototype unit under the most adverse design conditions. The licensee entered this issue into the CAP as PER 668367. Failure of the licensee to ensure measures used to review the suitability of application of materials, parts, and equipment essential to the safety-related functions of molded case circuit breakers, and measures to provide for the verification of checking the adequacy of design were in place was a performance deficiency. This performance deficiency was more than minor because it affected the design control attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, adequate measures were not implemented to ensure the station 120-VAC vital instrumentation boards had properly maintained their seismic qualification for their application. The inspectors assessed this finding for significance in accordance with NRC Manual Chapter 0609, Appendix A, Exhibit 2, Significance Determination Process (SDP) for Findings At- Power Mitigating Systems Screening Questions, and determined that it was of very low safety significance (Green) as the devices in question had been intrinsically qualified for this application as part of a complete panel test by the original vendor and the licensee determined that the SSC maintained its operability or functionality despite the identified non-conformances. The inspectors evaluated this finding and violation of NRC requirements in accordance with the NRC Enforcement Policy, Section 2.3.2, and found two conditions to not be met requiring a Notice of Violation be issued. First, inspectors found the licensee failed to restore compliance within a reasonable time after the original violation (05000327.328/2011002-01) was identified. The NRC Enforcement Manual, Section 3.1.2.A.1.b).1), further defines restoring compliance to include those actions taken to stop an ongoing violation from continuing. Second, the inspectors determined that the identified non-conformances represented a repetitive violation as a result of inadequate corrective action and that identification was by the NRC inspector. The lack of rigor in addressing the root of the prior violation which resulted in the inadequate corrective action further led the inspectors to identify a crosscutting aspect in the CAP component of the Problem Identification and Resolution area |
| 05000327/FIN-2012005-06 | 2012Q4 | Severity level IV | NRC identified | Failure to Submit a Technical Specifications Required Report | The NRC identified a Severity Level IV non-cited violation of 10 CFR 50.36(c)(5) for failure to submit the Technical Specification (TS) required U1R18 Steam Generator report within 180 days after the initial entry into Mode 4 following completion of an inspection performed in accordance with the Specification 6.8.4.k, Steam Generator (SG) Program. The licensee entered this into their CAP as PER 648658 and as a corrective action submitted the report on December 17th 2012 to the NRC. The inspectors concluded that the failure of the licensee to submit a TS required report was a performance deficiency. The inspectors evaluated this performance deficiency using the traditional enforcement process because the failure to submit a required report affected the NRCs ability to perform its regulatory function. Consistent with the guidance in Section 2.2.2 and Section 6.9.d of the NRC Enforcement Policy, the inspectors concluded the finding was a Severity Level IV violation because the licensee failed to make a TS required report that resulted in no or relatively inappreciable potential safety consequences. In accordance with section 07.03.c of NRC Inspection Manual Chapter 0612 cross-cutting aspects are not assigned to traditional enforcement violations. | |
| 05000327/FIN-2012005-07 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. The Quality Execution Procedure governing the licensees contractor for the steam generator replacement project, 11.01 paragraph 2.5.1a requires that work be performed in strict accordance with the Work Instructions provided in the applicable Work Package. Contrary to these requirements, in work packages 3575A/D for the 2-1 and 2-4 steam generator enclosure plugs, multiple steps for the welding of tapered shims were not performed as indicated in the work packages. The licensee found the omission in document review and closure, and performed rework to establish the requisite number of shims for the seismic support requirements of the subject enclosure plugs. The inspectors determined that the violation was not greater than of very low safety significance as it was identified and corrected with the reactor defueled in Mode 6. The issue is documented in the licensees CAP as the contractors Nonconformance Report (NCR) 1163, and TVA PER 655762 | |
| 05000327/FIN-2012005-08 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, states, in part, that the quality assurance program shall provide control over activities affecting the quality of the identified structures, systems, and components, to an extent consistent with their importance to safety and that activities affecting quality shall be accomplished under suitably controlled conditions to include the use of appropriate equipment. Contrary to these requirements, the licensees quality assurance controls failed to ensure that critical fasteners utilized in reassembly of the sites motor-operated valves, specifically those utilized in safety-related service, were of the appropriate grade, class, or type, to meet design output requirements. The inspectors determined that the violation was not greater than of very low safety significance as it was determined that fasteners utilized would permit the valves to perform their function, although they were not acceptable per design. The issue is documented in the licensees CAP as TVA PERs 506338, 518423, 622076, 623383, 624373, 626982, and 644661. | |
| 05000327/FIN-2012005-09 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | Facility operating license DPR-79 condition 2.C.(13) states that TVA shall implement and maintain in effect all provisions of the approved fire protection program referenced in Sequoyah Nuclear Plants Final Safety Analysis Report and as approved in NRC Safety Evaluation Reports contained in NUREG-0011, Supplements 1, 2, and 5, NUREG-1232, Volume 2, NRC letters dated May 29, and October 6, 1986, and the Safety Evaluation issued on August 12, 1997, for License Amendment No. 218. Contrary to the above, on June 28, 2011, the licensee did not implement and maintain in effect all provisions of the approved fire protection program. Specifically, Sequoyahs Fire Protection Report Part II, Limiting Condition for Operation (LCO) 3.3.3.8.a.1 & 3.7.11.2.a.1 require a fire watch to be established when the required number of operable fire detection instruments and the required number of spray and/or sprinkler systems are inoperable. On January 3, 2012, the licensee discovered that standing Fire Protection Impairment Permit (FPIP) FOR110249 and the associated Fire Protection Report Part II action statement had been incorrectly entered. The licensee entered this issue into the corrective action program as PER 485817. The finding was screened using Inspection Manual Chapter 0609, Appendix F Fire Protection Significance Determination Process, and was determined to be of very low safety significance (Green). | |
| 05000327/FIN-2012005-10 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | In place of the controls required 20.1601(a) and b of 10 CFR Part 20, TS 6.12.2 requires that entryways into HRAs with dose rates exceeding 1 rem/hour at 30 cm be provided with a locked or continually guarded door or gate that prevents unauthorized entry. Contrary to this, on December 16, 2011, during the dewatering of a spent resin liner, the entryway into the Auxiliary Building Railroad Bay was controlled using remote monitoring and surveillance. During the resin transfer an individual was assigned as the access controller and a camera was placed in the area to monitor the LHRA. The access controller monitored the LHRA using the camera in Laundry/Trash Sorting area adjacent to the Railroad Bay behind a closed door. The dose rates on the top of the liner were 1500 mrem/hr at 30 cm. Although no unauthorized entry occurred during this time period, workers could have potentially entered the area from the Auxiliary Building Door. Although this event involved the failure to maintain proper control of a LHRA, this finding is of very low safety significance because there was neither evidence of unauthorized worker entry into the affected areas nor any unexpected radiation exposures to licensee personnel. | |
| 05000327/FIN-2012007-01 | 2012Q1 | Green | H.4 H.5 | Self-revealing | Reactor Trip Due to Improper Preferred Inverter Maintenance | A self-revealing finding was identified for the licensees failure to properly implement work procedures during the performance of a preventive maintenance (PM) activity associated with the Unit 1 Preferred Inverter. The improper performance of selected steps with the system in an inappropriate configuration to support the activity caused an electrical transient and loss of the preferred power board which resulted in a turbine trip and automatic reactor trip. The licensee entered this issue into their corrective action program as PER 405141 and implemented corrective actions to include guidance for operations supervisory review of work documents prior to returning equipment to service. The inspectors reviewed IMC 0612, Appendix B and determined that the finding was more than minor because it adversely impacted the human performance attribute of the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, it resulted in sudden closure of all main turbine governor valves from 100% power, which ultimately led to an automatic reactor trip. The inspectors reviewed IMC 0609, Attachment 4 and determined that the finding was of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would not be available. This finding was determined to have a cross-cutting aspect in the area of human performance, the component of work control, and the aspect of work activity coordination, H.3(b), due to the failure to appropriately coordinate this work activity consistent with nuclear safety. Specifically, the necessary interdepartmental communication and coordination between operations and maintenance work groups was inadequate to assure proper performance and accomplishment of the work activity in accordance with the procedure, including establishing proper plant conditions to support the work activity as well as understanding the potential operational impact of the proposed maintenance. |