Semantic search
Jump to navigation
Jump to search
| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000327/FIN-2018003-01 | 30 September 2018 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Sequoyah Unit 1 Operating License Condition 2.C(16) and Sequoyah Unit 2 Operating License Condition 2.C(13) require in part that TVA shall implement and maintain in effect all provisions of the approved fire protection program. The Sequoyah fire protection report describes how the licensee complies with applicable sections of 10 CFR 50, Appendix R, including Section III.L.1 which states in part that alternative or dedicated shutdown capability provided for a specific fire area shall be able to achieve cold shutdown conditions within 72 hours and maintain cold shutdown conditions thereafter. Contrary to the above, since implementation of the Sequoyah Fire Protection Program, the licensee failed to maintain all aspects of the approved program. Specifically, in August 2018, the licensee discovered that the sites ability to achieve cold shutdown conditions within 72 hours would be challenged due to an inadequate evaluation of the RHR pumps functionality during certain Appendix R fire scenarios. |
| 05000327/FIN-2018411-01 | 30 September 2018 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2018001-03 | 31 March 2018 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Sequoyah Unit 1 and Unit 2 Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (ABGTS), requires two ABGTS trains be operable in modes 1, 2, 3, and 4. Contrary to the above, from March 3-7, 2017, the licensee blocked open door A212 resulting in the inoperability of the auxiliary building secondary containment enclosure boundary and thus inoperability of both trains of the ABGTS. Significance/Severity Level: Green. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building.Corrective Action Reference: CR1269767 |
| 05000327/FIN-2018001-01 | 31 March 2018 23:59:59 | Sequoyah | Self-revealing | Essential Raw Cooling Water Pumps Inoperable due to Frozen Motor Bearing Cooling Lines | A self-revealing Green NCV of Technical Specification 5.4.1, Procedures, was identified when Sequoyah/TVA did not establish, implement and maintain applicable procedures recommended in Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, the essential raw cooling water (ERCW) pump motor maintenance procedure, 0-MI-MRR-067-002.0, Removal/Disassembly/Reassembly Instruction for ERCW Pumps does not contain specific direction for the slope of the motor bearing cooling supply and return lines for the motor reassembly. |
| 05000327/FIN-2018001-02 | 31 March 2018 23:59:59 | Sequoyah | Self-revealing | Improper Calibration of Reactor Trip Instrumentation Results in a Condition Prohibited by Technical Specifications | A self-revealing Green finding and associated NCV of Sequoyah Unit 1 Technical Specification 5.4, Procedures, was identified on June 25, 2016, when the licensee did not implement procedures to calibrate Delta-T/Tavg Channel IV with the correct test equipment input impedance settings, which resulted in Delta-T/Tavg Channel IV being out of technical specifications allowed tolerances. |
| 05000327/FIN-2017008-03 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Potential Inadequate Use of Thermal Aging and the Arrhenius Methodology | Introduction: The inspectors identified a URI for the licensees use of the Arrhenius methodology without consideration for the limits of extrapolation and confidence bounds for statistical uncertainties. Description: The licensee did not consider the limits of extrapolation specified for Category 1 qualification in NUREG 0588 Section 4 and IEEE 323-1974 Section 6.5, Determination of Qualification. NUREG-0588, Section 4(5) required, in part, that known material phase changes and reactions should be defined to insure that no known changes occur within the extrapolation limits, (staff position: claims that conservative extrapolation limits have been implemented must be supported). Standard IEEE 323-1974, Section 6.5, specified, in part, that the qualified life shall be based upon the known limits of extrapolation of the time dependent environmental effects if an accelerated aging test was used to determine the mathematical model. Ancillary quality standards to IEEE 323-1974 and nuclear industry EPRI reports specified that 9 extrapolating beyond the extrapolation limits could invalidate the results of the Arrhenius methodology. The ancillary standards used for qualification of the various examples specified the limits of extrapolation to be no greater than 30 oC from the test data used to determine activation energies. In addition, the licensee did not consider adequate confidence bounds to account for the statistical uncertainties present when using the Arrhenius methodology. The inspectors noted that the uncertainties grow exponentially when exceeding the extrapolation limits. The Limitorque MOV motor life line appeared to have been extrapolated from test data at 240 C to 50 C, which is 190 C from the test data. The silicone rubber cable, life line appeared to have been extrapolated from test data at 210 C to 51.67 C, which is 188.3 C from the test data. The ASCO Valves, life line appeared to have been extrapolated from test data at 266 C to 40 C, which is 226 C from the test data. The Target Rock Valves, life line appeared to have been extrapolated upward from the test data. The Westinghouse RHR Motor rewind, life line appeared to have been extrapolated from test data at 180 C to 58.6 C, which is 121 C from the test data. 275 C is 216 C from test data The ancillary quality standards used in the qualification of these examples included IEEE 98-1972, IEEE Standard for the Preparation of Test Procedures for the Thermal Evaluation of Solid Electrical Insulating Materials; IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data; IEEE 117-1974, IEEE Standard Test Procedure for Evaluation of Systems of Insulating Materials for Random-Wound AC Electric Machinery, and other quality standards. IEEE 98-1972, Section 10, Temperature Exposures, specified, in part, that the lowest test temperature shall be chosen so that the extrapolation necessary to establish the temperature index will not be more than 25 C. IEEE 101-1972, Section 1.3, Extrapolation, specified, in part, that extrapolation of the (qualified life) line below the range of test temperatures may cause erroneous predictions if the chemical reactions controlling the insulation aging are different at lower temperatures or if other conditions affecting the aging or the mode of failure are different. Therefore, the methods outlined in this guide are applicable only if all of the assumptions behind the use of the Arrhenius equation are met (identified in references). IEEE 117-1974, Section 3.3.1, Thermal Aging, specified, in part, for any system being evaluated, tests are made for at least three different temperatures. The lowest test temperature should be no more than 25 C above the system temperature rating. The highest temperature test should be at least 40 C above lowest temperature test, and temperature points should be selected to give approximately equal temperature intervals. The average life at the highest temperature shall be no less than 100 hours. The inspectors are concerned that the licensee did not meet the aforementioned Category 1 requirements in their licensing basis. The licensee has captured these concerns in their corrective action program as CR 1366022. The inspectors need further information from the licensee and NRC technical staff to evaluate the concerns. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-03, 05000328/2017008-03, Potential Inadequate use of thermal aging and the Arrhenius methodology) |
| 05000327/FIN-2017008-01 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Unjustified Qualified Life for Target Rock Power-Operated Relief Valves | The team identified a Green NCV of Title 10 Code of Federal Regulations 50.49(e)(5) Aging when the licensee failed to replace, refurbish, or demonstrate additional life for components that exceeded their qualified life. The licensee failed to justify changes to the accelerated aging calculations used for power operated relief valve harsh environmental qualification. The licensee entered this issue into their corrective action program as CRs 1365730 and 1366082, and performed operability determinations, which determined the systems were operable but non-conforming with 10 CFR 50.49. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that Target Rock power-operated relief valves were qualified for the duration that they were required to operate reduced the reliability of reactor coolant system in the harsh environments of design basis accidents. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the team determined that the finding was of very low significance (Green) because it was a design deficiency that affected the design or qualification of a mitigating system, however, the mitigating system maintained its operability. The team determined there was no cross-cutting aspect associated with this finding since it was not indicative of current licensee performance. |
| 05000327/FIN-2017008-02 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Inadequate Qualification for Unit 1 Reactor Lower Compartment Cooler Motors | The team identified a Green NCV of Title 10 Code of Federal Regulations 50.49(f) Electrical Equipment Qualification when the licensee failed to perform an adequate similarity analysis for the environmental qualification of their Reliance 75 horsepower reactor lower compartment cooling fan motors. The licensee entered this issue into their corrective action program as CR1366056 and performed an operability determination, which determined the reactor lower compartment cooling fan motors were operable but non-conforming in accordance with 10 CFR 50.49. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to ensure the qualification of the reactor lower compartment cooling fan motors adversely affected their reliability and capability in the harsh environment of a design basis accident, which in turn adversely affected the reliability and capability of other environmentally qualified components that rely on the containment cooling system. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the team determined that the finding was of very low significance (Green) because it was a design deficiency that potentially affected the design or qualification of a mitigating system; however, the mitigating system maintained its operability. The team determined there was no cross-cutting aspect associated with this finding since it was not indicative of current licensee performance. |
| 05000327/FIN-2017008-04 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Potential Inadequate Determination of Failure Modes for Qualified Life for Foxboro/Weed Instrument | Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for failure modes and the degradation leading to them in the determination of the qualified life for the Foxboro/Weed Instrument transmitters documented in the licensees EQ Binder IPT-002. Description: The inspectors reviewed EQ Binder IPT-002 for Foxboro/Weed Instrument flow transmitter qualification. In reviewing qualification test report QOAACIO, Rev. A, the inspectors identified two concerns with the qualification. a. The inspectors noted that the qualification test report identified that polysulfone had the most limiting activation energy, 0.72 eV, for the Weed instrument assembly. However, the 0.72 eV was not being used by the licensee to determine the qualified life in the Arrhenius calculations in accordance with NUREG-0588 Section 4(6) and RG 1.89, Rev. 1, C.5.c. When inspectors questioned the licensees use of a less limiting activation energy (0.78 eV for resistors), the licensee determined that the 0.72 eV was based on the degradation and failure modes associated with the tensile strength material property for polysulfone. The licensee consulted the component manufacturer, and determined that creep was the correct material property to be evaluated for end of life. The activation energy for polysulfone creep was identified as 3.81 eV. The inspectors have challenged the licensees determination that creep is the only material property that can produce a failure of the sealing function for polysulfone. The inspectors noted that the requirements in IEEE Std. 323-1974, Section 5 require, in part, that assurance be provided that any extrapolation or inference be justified by allowances for known potential failure modes (i.e. loss of sealing function) and the mechanisms leading to them (i.e. the degradation in various material properties). If the degradation associated with creep is not the only degradation mechanism that could lead to a loss of sealing function over time, inspectors question if the degradation of other material properties would have more limiting activation energies than the licensees current activation energy for the Foxboro/Weed Instrument transmitter (0.78 eV). b. The 0.78 eV activation energy used by the licensee for qualified life was derived from an academic white paper that documented experiments performed in the early space program. The white paper specified that its experimental methods were not validated. The vendor that qualified the transmitter subsequently used this experimental information to determine the qualified life of the transmitters. This activation energy appeared not to be valid in the range of service temperatures that the transmitters are expected to age in prior to a DBA. The inspectors identified that these experimental tests did not follow any identifiable quality standard. The tests were conducted as early as 1963, the white paper was published in 1968, and the inspectors could not identify any subsequent verification of these experimental methods. Although no failure modes and effects analysis was evident, the table of 11 components in the qualification appeared to identify other components that could have much more limiting activation energies that were identified by qualification, as low as 0.5 eV. The licensee has captured these concerns in their corrective action program as CR1366039, CR 1363427. The inspectors need further information from the licensee and NRC technical staff to evaluate the concerns. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-04, 05000328/2017008-04, Potential Inadequate Determination of Failure Modes for Qualified Life for Foxboro/Weed Instrument) |
| 05000327/FIN-2017008-05 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Potential Inadequate Justification for Eliminating Preventative Maintenance for ASCO Valves | Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for eliminating the replacement of components that have a shorter life than the qualified life of the ASCO NP-1 valves assemblies. Description: The inspectors reviewed records in environmental qualification data package (EQDP), SQNEQ-SOL-005, Revision 47. After the valve manufacturer (ASCO) stopped providing rebuild kits, the licensee eliminated the replacement schedule for subcomponents that had a shorter life than the valve assembly. The licensee changed the inputs to the accelerated aging calculation and recalculated the life of these subcomponents from the approximate eight-year replacement schedule to approximately 32.5 years. The licensee changed the activation energies from 0.94eV and 0.96eV for ethylene-propylene-diene-monomer (EPDM) and Viton-A, respectively, to 1.1eV for both. Both, EPDM and Viton-A are rubber elastomers used within the ASCO valve assemblies. The licensees written justification in the EQDP referenced a review of several studies for each elastomers, which identified less limiting activation energies than the activation energies ASCO selected in their qualification test reports. Each of these studies used different material degradation mechanisms and end of life failure mechanisms to derive different activation energies. The conclusion in the EQDP, for the change justification analysis, stated in part, that these studies show that Sequoyah's original values were, in many cases, very overly conservative. The inspectors identified that the qualification of record, report AQR-67368, selected the qualification testing criteria based on the maintenance requirements (replacement schedule) specified in Appendix C of the report. The activation energies determined applicable in the ASCO test reports, (EPDM 0.94 eV and for Viton 0.96 eV) were determined by material testing and do not appear to inspectors to be very overly conservative (unrealistically low) or lacking in technical merit. In addition, the licensee did not consider the effect the various formulations for EPDM and Viton-A elastomers. The different formulations could non-conservatively affect the activation energies reviewed in their justification. The licensee subsequently replaced the accelerated aging rate used in ASCO qualification test report AQR-67368 with the more thermally severe acceleration rate used in AQS-21678/TR, ASCO Qualification Test Report, dated 7/1/1979. AQS-21678/TR does not appear to meet Category 1 requirements, yet its accelerated aging rate was used to replace the Category 1 qualification-aging rate in AQR-67368. The AQS-21678/TR report, specified, in part, that coils and elastomeric components shall be replaced every 4 years as noted in the Valve Design Specification Sheets. The thermal 12 aging in AQR-67368 simulated a minimum of 2,000 cycles (~4.5 cycles/hr.), which was more limiting than the once every 6 hours (96 cycles) specified in AQS-21678/TR. The inspector noted that the test program specified in AQS-21678/TR used IEEE 382-1972, which did not meet Category 1 qualification requirements. The test program in AQR-67368 used IEEE 382-1980, which did meet Category 1 requirements. The forward to IEEE 382-1980, stated, in part, that the testing in the report satisfies the latest issued requirements and standards, which were the NUREG-0588 Category 1 requirements issued for comment December 1979 and published in July 1981. The licensee also used a lower self-heating temperature than which was specified in ASCO letter ASCO Solenoid Valve Coil Heat Rise Data, dated 5/8/1986. The licensee determined that the ASCO heat-rise data in the above document was too conservative and used heat-rise data from the first testing done by the Franklin Research Center (FRC). Later, FRC completed NUREG/CR-5141 RV, Aging and Qualification Research on Solenoid Operated Valves, dated 4/1/1988, which was referenced in the EQDP. The NUREG specified that: Aging in forced air ovens significantly limited heat rise from self-heating. The NUREG further specified that the qualified lives of the subcomponents were significantly reduced after accounting for the differences between forced air vs less turbulent air flow and the actual temperature measurements made by ASCO vs approximate temperature measurements made by other testing. Additional difference between ASCO heat rise testing and other testing including FRCs was that ASCO drilled holes in the valves to measure the actual subcomponent temperature while others only measured externally near the subcomponents to avoid damaging the valves. This produced lower temperature readings than ASCOs. The inspectors noted that even applying the 1.1 eV currently used, the life of the elastomers appeared to be approximately 4 to 8 years, not the 32.5 years identified in the EQDP The ASCO qualification testing used nitrogen during the qualification testing. The NUREG/CR 5141 RV specified, that oxygen exposure from plant compressed air systems produced more degradation than did the nitrogen used in the ASCO qualification testing. The licensee determined that radiation margin from the qualification tests could be used to mitigate the differences between these two gases. The inspectors determined that both ASCO qualification reports, AQS-21678/TR and AQR-67368, used nitrogen instead of air, which limited the aging degradation. The inspectors question whether radiation margin can be applied to account for the difference between oxidizing gases and inert gases. In addition, the ASCO report AQR-67368, specified, that Viton elastomers significantly degraded above 18E6 test dose not 200E6 test dose used for the margin. The inspectors are concerned that the licensee failed to meet the Category 1 requirements as specified in NUREG-0588 and IEEE 323-1974. Category 1 specified proof of conservative extrapolations, and use of the most limiting activation energy. Additionally, that users of IEEE 382-1972 must meet Category 1 requirements. Furthermore, the inspectors noted that the licensee was made aware of similar deficiencies and NRC staff positions in Technical Evaluation Report (TER)-C5257-532, Implementation Guidance for New and Corrective Equipment Environmental Qualification, dated 4/22/1983. The licensee has captured these concerns in their corrective action program as CR 1366024. The inspectors need to evaluate: (1) the licensees justification for changing the activation energy; (2) the licensees assessment of how AQS-21678/TR met Category 1 requirements; (3) the adequacy of the licensees heat rise data; and (4) the licensees evaluation of elastomer degradation from oxygen vs nitrogen gas, and their use of apparent radiation dose margin to account for these differences. This URI is opened to determine if a performance deficiency exists. (URI 05000327/2017008-05, 05000328/2017008-05, Potential Inadequate Justification for Eliminating Preventative Maintenance for ASCO Valves) |
| 05000327/FIN-2017008-06 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Potential Unjustified Qualified Life for ASCO Solenoid Operated Valves | Introduction: The inspectors identified a URI to review the adequacy of the licensees justification for changing the activation energy and calculating a new qualified life for ASCO NP-1 valves assemblies. Description: The manufacturer, ASCO, conservatively established a 1.0 eV activation energy for the valve coil assemblies. The activation energy appeared to be determined by test and realistic coil failure modes. The conservative methodology used by ASCO, that used the most limiting activation energy, met the requirements in 10 CFR Part 50. By memorandum dated 8/19/2004, the nuclear utility user group for environmental qualification (NUGEQ), to which the licensee was a member, provided information supporting the use of revised activation energy values from 1.0 eV to a less limiting 1.37 eV. The memorandum (memo) specified that NUGEQ was tasked to revise the activation energy values for ASCO NP series SOVs to a less limiting one. The inspectors determined that the data and conclusions reported by NUGEQ did not appear to be justified by design control measures in accordance with 10 CFR Part 50 Appendix B Criterion III and 50.49. Adequate design control measures were specified in the Category 1 specifications established in NUREG-0588 Section 4, Aging and IEEE 323-1974 Section 6.3.3, Aging, as supplemented by RG 1.89 revision 1, Regulatory Position 5 Aging. The NUGEQ memo specified that they obtained their data through the research of information acquired from various sources. The use of the 1.37 eV value was for significantly increasing the qualified life of the ASCO coils. The inspectors are concerned that this did not meet the requirement to prove conservative extrapolations and use of the most limiting activation energies. Based on the inspectors review, NUGEQ did not demonstrate that the more limiting activation energies were unrealistic and could be discounted. The memo specified that the information NUGEQ used to derive 1.37 eV was based on emailed recollections of past DuPont testing. The DuPont email appeared to be supported by some identifiable test data, but was not quality related, was not commercial grade dedicated, and performed without any identifiable design control measures. In addition, the memo disregarded other coil components with more limiting activation energies by discounting the failure modes associated with them and the coil. The manufacturer ASCO found these discounted failure modes relevant to the coil safety functions. The inspectors are concerned that the licensee disregarded realistic, more limiting, failure modes without proper justification. The design control requirements in NUREG-0588 Section 4(5) specified, in part, that known material phase changes and reactions should be defined to insure that no known changes occur within the extrapolation limits, (staff position: claims that conservative extrapolation limits have been implemented must be supported), and Section 4(6) required, the aging acceleration rate used during qualification testing and the basis upon which the rate was established should be described and justified, (staff position: testing of the equipment should be conducted using the most limiting (lowest) activation energy of the components). Additionally, RG 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Revision 1, Regulatory Position 5.c, Section 6.3.3, Aging, of IEEE Std. 323-1974, specified, in part, that the aging acceleration rate and the basis upon which it was established be described, documented, and justified. The licensee has captured these concerns in their corrective action program as CR 1366020. The inspectors need to review the licensees analysis and justification for discounting realistic failure modes, changing the activation energy, and calculating a new qualified life for ASCO NP-1 valves assemblies. This URI is opened to determine if the performance deficiency for not providing adequate justification for changing the activation energy, is more than minor. (URI 05000327/2017008-06, 05000328/2017008-06, Potential Unjustified Qualified Life for ASCO Solenoid Operated Valves) |
| 05000327/FIN-2017407-01 | 31 December 2017 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2017002-01 | 30 June 2017 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | Unit 1 and Unit 2 technical specifications LCO 3.7.10 required that if both trains of CREVS become inoperable than LCO 3.0.3 shall be immediately entered. Additionally, LCO 3.0.3 requires both units to be placed in Mode 3 within seven hours if the condition was not rectified. Contrary to the above, on August 10, with both trains of CREVS rendered inoperable, both units remained in Mode 1 for a period of approximately 24 hours. The finding was entered into the licensees CAP as CR 1201905. This finding was assessed using NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, and was determined to be of very low safety significance (Green) due to the finding only representing a degradation of the radiological barrier function provided for the control room. |
| 05000327/FIN-2017002-02 | 30 June 2017 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | Unit 1 and Unit 2 facility technical specifications LCO 3.6.10 required two operable EGTS systems in Modes 1 through 4. Contrary to the above, on August 2, 2016,during a system review, plant engineers noted a design flaw that could have resulted in one train of EGTS being rendered inoperable since initial plant operation. This problem was entered into the licensees CAP as CR 1198440 and CR 1200028. The TVA probabilistic risk assessment model does not consider the EGTS in core damage and large early release frequencies. The EGTS system is designed to maintain the shield building at a negative pressure and filter any leakage past the steel liner during a design basis event. With the EGTS inoperable, dose would still remain below 10 CFR 100 limits. The finding was screened using IMC 0609, Appendix A At Power Operation, and was determined to be of very low safety significance (Green). According to Exhibit 3, an issue related to degradation of the radiological barrier function of the reactor building is considered to be of very low safety significance. |
| 05000327/FIN-2017007-01 | 30 June 2017 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | The licensee identified examples of a failure to meet the requirements of the Facility Operating License (FOL) condition C.16 and C.13 for units 1 and 2 respectively. The license condition states, in part, that TVA shall implement and maintain in effect all provisions of the approved fire protection program referenced in the UFSAR and as approved by the NRCs Safety Evaluation Report (SER). Updated Final Safety Analysis Report (UFSAR), Section 9.5.1, Fire Protection System, states in part that the Fire Protection System and fire protection features are described in the Fire Protection Report (FPR). Contrary to the above, the licensee failed to maintain the requirements prescribed in the fire protection program. Specifically, on February 7, 2017, the licensee identified numerous discrepancies and errors in the Fire Hazards Analysis, including multiple instances of inaccurate and non-conservative manual action time requirements. The deficiencies were associated with inaccurate assumptions based on loop-stagnation, head vent flow rates, and containment cooling which was inconsistent with FPR, Part III, Section 1.1, Design Basis Evaluation. The finding was screened in accordance with IMC 0609 Appendix F, Significance Determination Process, and determined to be Green, by answering Yes on Task 1.3.1, Question 1.3.1. The issue into the CAP as CR 11 1259493, CR 1224829, and CR 1261866. Additionally, adequate compensatory actions have been implemented. |
| 05000327/FIN-2017007-02 | 30 June 2017 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | The licensee identified examples of a failure to meet the requirements of the Facility Operating License (FOL) condition C.16 and C.13 for units 1 and 2 respectively. The license condition states, in part, that TVA shall implement and maintain in effect all provisions of the approved fire protection program referenced in the UFSAR and as approved by the NRCs Safety Evaluation Report (SER). Updated Final Safety Analysis Report (UFSAR), Section 9.5.1, Fire Protection System, states in part that the Fire Protection System and fire protection features are described in the Fire Protection Report (FPR). FPR, Part III Safe Shutdown Capabilities, Section 3.0, Analysis of Safe Shutdown Systems, stated that a minimum set of plant systems and components are selected to ensure the plant is capable of reaching and maintaining the applicable safe shutdown state. Contrary to the above, the licensee failed to maintain the requirements prescribed in the fire protection program. Specifically, on February 13, 2017, the licensee identified non- conservative times for repair actions to achieve cold shutdown. The finding was screened in accordance with IMC 0609 Appendix F, Fire Protection Significance Determination Process, and determined to be Green, by answering Yes on Task 1.3.1, Question 1.3. The issue has been entered into the licensees corrective action program as CR 1261868, and CR 11261881. Additionally, adequate compensatory actions have been implemented. |
| 05000327/FIN-2017406-01 | 31 March 2017 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | |
| 05000327/FIN-2017001-01 | 31 March 2017 23:59:59 | Sequoyah | NRC identified | Degraded Fire Barrier Penetration | Green . The NRC identified a non -cited violation ( NCV ) of the facilitys operating license for the failure to identify a non functional fire barrier penetration and enter it into the corrective action program (CAP) when the initial damage to the fire barrier occurred. The licensee also failed to implement required compensatory measures for a nonfunctional fire barrier penetration contrary to the approved fire protection report . The licensee entered the issues into their CAP as Condition Report (CR) 1263322 . 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 in that there was no assurance the fire barrier would prevent the spread of fire through the cable penetration during a design basis fire. The finding was of very low safety significance (Green) due to fully functional automatic suppression systems on either side of the fire barrier . T he inspectors identified a cross- cutting aspect in the identification component of the Problem Identification and Resolution area because the licensee failed to enter the damaged fire barrier into their CAP after it was initially damaged . (P.1) |
| 05000327/FIN-2016004-01 | 31 December 2016 23:59:59 | Sequoyah | NRC identified | Degraded Fire Barrier Penetrations | Green. The NRC identified a non-cited violation (NCV) of the facilitys operating license for the licensees failure to ensure that all fire barrier penetrations in fire zones boundaries protecting safety related areas are functional at all times. Specifically, on eight separate fire barrier penetrations, the licensee failed to recognize that the barrier had become damaged to the point of being nonfunctional. The licensee also failed to implement required compensatory measures for a nonfunctional fire barrier penetration contrary to the approved fire protection report (FPR). The licensee entered the issues into their corrective action program (CAP) as Condition Reports (CRs) 1229468, 1229470, 1243550, 1243970, 1243552, 1243554, 1243555, and 1243557. 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, with the fire barriers being damaged to the point of declaring the fire barrier penetrations nonfunctional, there was no assurance that the fire barrier would prevent the spread of fire through the cable penetration during a design basis fire. The inspectors performed the SDP using NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and assigned a High degradation rating, giving no credit for Barrier Protection in accordance with the Fire Barrier Degradation section. The inspectors concluded, that the finding was of very low safety significance (Green) due to fully functional automatic suppression systems on either side of the fire barrier (Question 1.4.3-C). Using Manual Chapter 0310, Aspects Within the Cross-Cutting Areas, the inspectors identified a cross-cutting aspect in the Identification component of the Problem Identification and Resolution area, because the licensee failed to enter the damaged fire barrier into their CAP after it was initially damaged (P.1) |
| 05000327/FIN-2016003-02 | 30 September 2016 23:59:59 | Sequoyah | Self-revealing | Isolation of Fire Suppression System to a Significant Portion of the Plant Site | A self-revealing non-cited violation (NCV) of the facility operating licenses DPR-77 and DPR-79 conditions 2.C.(16) and 2.C.(13), respectively, was identified for the licensees failure to properly implement the clearance process such that the fire suppression system was rendered non-functional for approximately 41 hours. The licensee inappropriately expanded an existing clearance on March 29, 2016 in order to attempt to reduce boundary valve leakage affecting existing maintenance on the fire suppression system within a valve pit. Subsequently on March 30, 2016 during fire system testing, technicians noted a lack of system pressure and it was ultimately concluded the clearance expansion had inadvertently isolated fire suppression water to a significant portion of the site. Upon discovery of the clearance error, the system was restored to a functional status. The licensee entered the issue into their corrective action program (CAP) as CR 1155763. The licensees failure to properly assess the system impact of a clearance revision for the High Pressure Fire Protection (HPFP) suppression header and enter the required FPR Operating Requirement (FOR) Action was a performance deficiency. The performance deficiency was 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. The inability to pressurize the HPFP system from either the electric or diesel-driven fire pumps rendered the fire suppression system inoperable. Based on the complexities of this particular event, the inspectors concluded that Appendix M, Significance Determination Process Using Qualitative Criteria, of IMC 0609 should be performed in lieu of a Phase 3 analysis. Under appendix M, the Senior Reactor Analyst (SRA) performed an initial bounding evaluation using qualitative methods. The licensee submitted a detailed analysis that estimated an upper bound for the risk of the finding which was less than 1E-6. The SRA performed a review of this screening analysis as part of this SDP evaluation. In addition to the SRA review, the resident inspectors performed an independent review of the licensees estimation of the success of actions used to recover the isolated fire header. To the extent reviewed, the methodology and results were determined to be acceptable for use in this SDP review of this Performance Deficiency. The SRA concurred with the submitted results of the licensees screening analysis, and has determined the finding to be GREEN. The inspectors determined that the finding had a cross cutting aspect of Procedural Adherence within the Human Performance area, because the licensee failed to consider the affect that changing a clearance order could have on the operability of the fire suppression system. |
| 05000327/FIN-2016003-01 | 30 September 2016 23:59:59 | Sequoyah | Self-revealing | Hydrogen Mitigation System Inoperable Longer than Allowed by Technical Specifications | A self-revealing NCV of Technical Specification 3.6.8, Hydrogen Mitigation System (HMS), was identified for the licensees failure to restore an inoperable train of HMS within the 7 day completion time or place the unit in Mode 3 within the action time of 6 hours. Each train of HMS has 34 hydrogen igniters and SR 3.6.8.1 defines an operable train as one that has at least 33 igniters operable. A review of the operating history revealed the A train HMS had only 31 operable igniters for a period of 91 days due to a mispositioned circuit breaker. Upon discovery of the unexpected condition, the circuit breaker was closed to restore operability to the HMS train. The licensee entered the issue into their CAP as CR 1179126. The licensees failure to preclude an inoperable HMS train for more than 7 days without a subsequent plant shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the Configuration Control attribute of Barrier Integrity cornerstone and adversely affected the cornerstones objective to ensure the structural integrity of the containment boundary. Specifically, the finding challenged containment integrity as hydrogen igniters have a high risk significance in ice condenser style containments. The finding was screened to Green based on the fact that the loss of igniters did not affect multiple igniters in adjacent compartments. The inspectors determined that the finding had a cross cutting aspect of Avoid Complacency within the Human Performance area because the licensee failed to implement appropriate error reduction tools while working near the HMS circuit breakers (H.12). |
| 05000327/FIN-2016002-01 | 30 June 2016 23:59:59 | Sequoyah | Self-revealing | Isolation of Fire Suppression System to a Significant Portion of the Plant Site | A self-revealing apparent violation (AV) of the facility operating licenses DPR-77 and DPR-79 conditions 2.C.(16) and 2.C.(13) was identified for the licensees failure to properly implement the clearance process such that the fire suppression system was rendered non-functional for approximately 48 hours. The licensee inappropriately expanded an existing clearance on March 29 in order to attempt to reduce boundary valve leakage affecting existing maintenance on the fire suppression system within a valve pit. Subsequently, on March 30, during fire system testing, technicians noted a lack of system pressure and it was ultimately concluded the clearance expansion had inadvertently isolated fire suppression water to a significant portion of the site. Upon discovery of the clearance error, the system was restored to a functional status after being isolated for approximately 48 hours. The licensee entered the issue into their corrective action program as condition report (CR) 1155763. 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. The inability to pressurize the high pressure fire protection (HPFP) system from either the electric or diesel-driven fire pumps rendered the fire suppression system inoperable. The finding could not be screened to Green and is pending a significance determination. The inspectors determined that the finding had a cross-cutting aspect of Procedural Adherence within the Human Performance area, because the licensee failed to consider the effect that changing a clearance order could have on the operability of the fire suppression system. (H.8). |
| 05000327/FIN-2016007-01 | 30 June 2016 23:59:59 | Sequoyah | NRC identified | Failure to Implement the Design Change Process when Modifying the Safety-Related Fire Dampers | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to use the design change process to make modifications to the Emergency Diesel Generator EDG room inlet dampers as required by NPG-SPP-9.3, Plant Modifications and Engineering Change Control. The licensee entered the issue into the corrective action program and implemented compensatory measures, while implementing plans to modify each of the affected inlet and exhaust fire dampers. This performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee modified the dampers to include the wrong brackets, which could adversely affect the dampers ability to remain open to provide cooling during EDG operation and support EDG reliability and availability. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000327/FIN-2016007-02 | 30 June 2016 23:59:59 | Sequoyah | NRC identified | Failure to Install Safety-Related Components that are Designed to Withstand the Effects of a Design Basis Tornado | The NRC identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to install emergency diesel generator components that could withstand the effects of a design basis tornado as required by Section 3.1.2 of the Update Final Safety Analysis Report (UFSAR). The licensee entered the issue into the corrective action program and implemented compensatory measures to protect the affected components. This performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the capability of the equipment to withstand the effects of a tornado was not ensured. The team determined the finding to be of very low safety significance (Green) because of the low frequency of tornados/high winds and the potential for recovery by the operators on site. This finding was not assigned a crosscutting aspect because the issue did not reflect present licensee performance. |
| 05000327/FIN-2016007-03 | 30 June 2016 23:59:59 | Sequoyah | NRC identified | Inadequate Monitoring of the 480V Shutdown Transformers | The NRC identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have documented procedures in place to ensure effective monitoring of the 480V Shutdown Transformers as required by Section 5.3.2.(4) of IEEE 308-1971. The licensee entered the issue into the corrective action program and planned to put additional transformer testing/monitoring in place to detect degradation prior to equipment failure. This performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to perform adequate maintenance on the shutdown transformer, which could result in the inability to detect the deterioration of the shutdown transformer toward an unacceptable condition. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect present licensee performance. |
| 05000327/FIN-2016007-04 | 30 June 2016 23:59:59 | Sequoyah | NRC identified | Failure to Energize Hydrogen Igniters during Extended Station Blackout | The NRC identified a finding (FIN) for the licensees failure to meet their docketed commitment to revise the back-up generators to include supplying one train of containment hydrogen igniters per unit in response to Generic Safety Issue 189, Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen Combustion During a Severe Accident. The licensee entered this issue into their corrective action program and completed immediate corrective actions to revise procedure FSI-5.01, Initial Assessment and Flex Equipment Deployment, Rev. 0, to ensure the hydrogen igniters would be energized during an extended station blackout (SBO) event. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to energize the hydrogen igniters during an extended SBO event could result in containment failure. The team determined the finding to be of very low safety significance (Green) because the risk was mitigated by the low frequency of SBO conditions and the high likelihood of operator recovery given the obvious diagnosis of the performance deficiency. The team determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of operating experience (OE), in the area of Problem Identification and Resolution, because the licensee failed to effectively collect, evaluate, and implement relevant internal OE before implementing their new station procedures to use the FLEX diesels as the power supply to the hydrogen igniters. |
| 05000327/FIN-2016001-02 | 31 March 2016 23:59:59 | Sequoyah | Self-revealing | Inadvertent Safety Injection Due to Inadequate Main Steam Procedure | A self-revealing NCV of Units 1 & 2 Technical Specification, 5.4.1 was documented for the licensees failure to implement an adequate procedure associate with the startup of the main steam system. Specifically, the licensee caused an inadvertent safety injection which unnecessarily challenged the operators due to an inadequate draining of the main steam header during system start up. The licensee placed the issue into the CAP. The failure of the licensee to adequately drain condensate from the main steam header resulted in an inadvertent safety injection (SI) and was a performance eficiency. The finding was determined to be greater than minor because it adversely effected the Procedure Quality attribute of the Initiating Events Cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of this finding was evaluated in accordance with the Manual Chapter 0609 Appendix A, The Significance Determination Process for Findings At-Power. Although the unit was in Mode 3 at the time, this appendix was chosen because the plant did not meet the entry conditions for residual heat removal system operation. The inspectors concluded that the finding was of very low safety significance (Green) because no significant initiating event prompted this transient. The finding was determined to have a cross-cutting aspect in the operating experience component of the problem identification and resolution area, because the licensee failed to evaluate and implement relevant internal and external operating experience. |
| 05000327/FIN-2016008-02 | 31 March 2016 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, requires, in part, measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, between July 2000 and Sept 14, 2015, the licensee failed to implement corrective actions which would have identified a condition adverse to quality on the output switch on the Spare Vital Inverter 0-II. This issue was entered into the licensees corrective action program as CR 1081482. The finding was screened using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, and was determined to be of very low safety significance (Green). |
| 05000327/FIN-2016403-01 | 31 March 2016 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2016008-01 | 31 March 2016 23:59:59 | Sequoyah | Self-revealing | Failure to Maintain Design Control of MSIV Controls | A self-revealing non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to maintain design control of MSIV 1-FCV-1-22 controls. Specifically, inadequate design controls led to an under tightened electrical connection following replacement of the Unit 1 MSIV Control Room handswitch (1-HS-1-22A). The licensee entered the condition into the CAP as CR 1107656 and has corrected the loose wiring connection on 1-HS-1-22A. The finding was determined to be more than minor because it adversely impacted the Design Control 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. 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 will not be available. No cross-cutting aspect was assigned to this finding because the performance deficiency is not reflective of current licensee performance. |
| 05000327/FIN-2016403-02 | 31 March 2016 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2016001-01 | 31 March 2016 23:59:59 | Sequoyah | NRC identified | Inadequate Application of Flame Retardant on Cable Room Penetrations | The NRC identified a non-cited violation (NCV) of Unit 1 and 2 Technical Specification 5.4.1 for the licensees failure to adequately implement fire protection procedures. Specifically, the inspectors identified several cables located within a cable tray that penetrated the floor of the cable spreading room that were not adequately coating with fire retardant material as required by plant procedures. The licensee placed the issue into the corrective action program (CAP) and implemented a fire watch for the degraded condition. The inspectors determined that the failure to adequately implement all requirements of the licensees fire protection program procedures 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. The inspectors determined the finding was of very low safety significance (Green) because of the fire protection defense in depth concept provided other barriers to prevent the spread of fires. The cause of this finding was related to the procedural adherence component of the human performance area, because the licensee failed to properly install cable bundles through wall penetrations. |
| 05000327/FIN-2015004-01 | 31 December 2015 23:59:59 | Sequoyah | NRC identified | Failure to Recognize and Submit for Approval a Reduction in Effectiveness of the Emergency Plan | The inspectors identified a Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations, Part 50.54(q), for changes to the licensees radiological emergency plan, effective December 18, 2014, that reduced the effectiveness of the plan and therefore, should have received NRC approval prior to making the change. Specifically, the effectiveness of TVAs Radiological Emergency Plan (Generic Part), Revision 104, was reduced by the inadvertent removal of the offsite telephone communications description for the Health Physics Network and Emergency Notification System communication tools, as well as the monthly testing of those devices. The licensees failure to recognize that Revision 104 reduced the effectiveness of the emergency plan was a performance deficiency. The licensee entered this issue into their corrective action program (CAP) as Condition Report (CR) 1093684 This finding is more than minor because it brings into question the thoroughness of the licensees review process when making changes to the emergency plan and adversely affects the procedure quality attribute of the emergency preparedness cornerstone objective. This finding is a violation of NRC requirements and because it has the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with IMC 0612, Appendix B. This finding is determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involves the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan is reduced. |
| 05000327/FIN-2015404-03 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2015404-02 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2015404-05 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2015007-02 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Failure to meet Design Basis Requirements to have Interlocks between Shared systems | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B Criterion III, Design Control, for the licensees failure to ensure that plant licensing and design basis for shared Class 1E electrical systems were controlled and maintained. The licensing and design basis of shared electrical systems required mechanical interlocks to prevent an operator error that could parallel these diverse power sources in accordance with IEEE 308-1971 and Regulatory Guides 1.81 and 1.6. A modification removed the kirk-key interlocks. The issue was entered into the licensees corrective action program as CR 1064736. The licensee has administrative controls in place to limit the risk of this configuration pending determination of corrective actions. The inspectors determined that the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and the removal of mechanical interlocks that separated diverse shared electrical systems adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to have a cross-cutting aspect in the conservative bias area of Human Performance (H.14) because the licensees decision making-practices did not emphasize prudent choices over those that are simply allowable. |
| 05000327/FIN-2015404-01 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2015404-04 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Security | |
| 05000327/FIN-2015007-01 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Failure to maintain control of and update safety related design out documents (electrical calculations | The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to control safety related calculations that reviewed equipment essential to the function of Class 1E electrical systems. The issue was entered into the licensees corrective action program as CRs 1059281 and 1064042. Planned corrective actions were to revise the calculations. The inspectors determined that the performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone. The failure to plan and control updates to safety related calculations to review the suitability of new molded case circuit breakers in Class 1E electrical systems adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the systems that respond to initiating events to prevent undesirable consequences. The finding was determined to have a cross-cutting aspect in the resolution area of Problem Identification and Resolution (P.3), because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. |
| 05000327/FIN-2015003-01 | 30 September 2015 23:59:59 | Sequoyah | Self-revealing | Inadequate Clearance Causes damage to A train SSPS | A self-revealing Green NCV of Unit 1 Technical Specification (TS) 6.8.1.a was identified for the licensees failure to adequately establish a clearance boundary during plant maintenance. Specifically, the licensee caused damage to a safety-related component during maintenance as a result of a failure to de-energize all electrical sources during maintenance troubleshooting activities. The licensee placed the issue into their corrective action program (CAP) and corrected the identified deficiencies. The inspectors determined that the failure to adequately implement clearance procedures was a performance deficiency. The inspectors determined that the performance deficiency was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding was of very low safety significance (Green) as the affected safety significant component was repaired within 24 hours. The cause of this finding was related to the cross-cutting aspect of leaders ensuring that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. |
| 05000327/FIN-2015003-02 | 30 September 2015 23:59:59 | Sequoyah | Self-revealing | Failure to Implement Work Risk Activity and Oversight of Supplemental Personnel Procedures | A self-revealing Green NCV of TS 6.8.1.a, Administrative Controls of Procedures and Programs, was identified for the licensees failure to implement procedures related to quality during the surveillance capsule relocation activity. Specifically, procedures NPG-SPP-07.3, Work Activity Risk Management, and NPG-SPP.07.7, NPG TCM Role and Oversight of Supplemental Personnel, were not appropriately implemented. The deficiency was entered into the licensees CAP as Problem Evaluation Report (PER) 1016839. This finding was determined to be greater than minor because it was associated with the Human Performance attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the failure to properly secure reactor vessel surveillance capsules and the subsequent damage to the reactor vessel pressure boundary, reactor internals and fuel filter screens. The proper higher risk categorization would have led to enhanced contractor oversight, and the ability to detect when the contractors were performing actions outside the approved procedure. These additional oversights would reasonably be expected to prevent the events that led to the surveillance capsule ejections, and eliminate any potential to cause damage to the reactor vessel pressure boundary, reactor internals, and fuel filter screens. The inspectors identified a cross-cutting aspect in the Human Performance Consistent Process cross-cutting area. Specifically, the licensee failed to consistently incorporate risk insights, as required by procedure NPG-SPP-07.3, which resulted in less than conservative classification for an infrequently performed activity inside the reactor vessel performed by contract personnel. |
| 05000327/FIN-2015007-05 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Failure to Identify Qualification Criteria Associated with Class 1E Electrical Component Static Performance Characteristics | The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of defined shelf life and design life characteristics of Class 1E electrical equipment. The issue was entered into the licensees corrective action program as CR 1064785. The inspectors determined that the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and the failure to ensure the Class 1E static and dynamic performance characteristics were identified and evaluated adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the SSCs that responds to initiating events to prevent undesirable consequences. The finding was determined to have a cross-cutting aspect in the change management area of Human Performance (H.3) because Leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. |
| 05000327/FIN-2015007-04 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Failure to identify and correct inadequate procurement evaluation processes | The inspectors identified three examples of a Green non-cited (NCV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a conditions adverse to quality that were associated with processes for evaluating Class 1E critical characteristics for molded case circuit breakers. The issue was entered into the licensees corrective action program as CRs 1064483, 1064744, 1064479, 1059273 and 1064731. Planned corrective actions were to update procedures to document critical thinking in evaluating CRs and include additional critical characteristics. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and the failure to identify and correct nonconformances in Class 1E equipment and the failure to resolve adverse conditions with evaluating Class 1E critical characteristics adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to have a cross-cutting aspect in the change management area of Human Performance (H.3) because Leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. |
| 05000327/FIN-2015007-03 | 30 September 2015 23:59:59 | Sequoyah | NRC identified | Failure to request a licensee amendment prior to removing interlocks from shared onsite emergency and shutdown AC electric systems | The inspectors identified a SLIV violation of 10CFR 50.59.c.(2).ii, Changes, tests and experiments, for the licensees failure to obtain a license amendment prior to implementing a change to the onsite emergency and shutdown AC electric systems supplying the shared Essential Raw Cooling Water (ERCW) systems. The change removed the kirk key interlocking system from the tie breakers that originally prevented an operator error that would parallel the Unit 1A and Unit 2A 480V AC motor control centers (MCCs). The issue was entered into the licensees corrective action program as CR 1076179. The licensee has administrative controls in place to limit the risk of this configuration pending determination of corrective actions. The inspectors determined that the performance deficiency was more than minor because there was a reasonable likelihood that the change required Commission review and approval prior to implementation and the failure to request approval impacted the regulatory process. Specifically, the departure from acceptance criteria identified in IEEE 308, RG 1.81, and RG 1.6 more than minimally increased the likelihood of occurrence of an ERCW power train malfunction. |
| 05000327/FIN-2015002-03 | 30 June 2015 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation | Unit 1 Technical Specifications Section 3.3.1.1 requires two channels of intermediate range nuclear instrumentation in Mode 2 to provide the input to the P-6 interlock. This interlock allows the operator to block the source range channels during a reactor startup. This is done to prevent damage to the detectors as power is elevated to levels that could damage the detector. Contrary to the above, between March 11 and March 27, the P6 interlock was inoperable due to a non-conservative bias, concurrent with Unit 1 being in Mode 2 on March 14 from 0558 to 1010. This problem was entered into the licensees corrective action program as CR 1005422. The finding was screened using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, and was determined to be of very low safety significance (Green). |
| 05000327/FIN-2015002-02 | 30 June 2015 23:59:59 | Sequoyah | NRC identified | Spilled Specimen Capsule | The inspectors identified an unresolved item (URI) associated with the control of specimen capsules inside the reactor vessel. During this refueling outage, the licensee noted that two specimen capsules had become dislodged from their location on the core barrel. This particular outage required a 10 year ISI inspection and the core barrel was removed as part of the outage plan. The two capsules were noted to have been moved during the last refueling outage. A formal root cause evaluation was performed. The root cause team included industry experts independent of the licensee organization. The root cause team noted several procedural violations during the previous capsule move. The team concluded that these significant errors led to the improper seating of the capsules in the specimen baskets and ultimately allowed the capsules to dislodge from the core barrel. A significant foreign object retrieval evolution was completed during the outage and the core barrel, lower internals, and lower vessel head were inspected by the licensee and the NRC. The specimen parts were collected and placed in storage containers and transferred to the spent fuel pool. The unit was restarted on May 15 without a full accountability of the specimen parts. The inspectors determined that more inspection of this issue is required in order to understand all aspects of the incident. This issue will be tracked as URI 05000327/2015002-02, Spilled Specimen Capsule. |
| 05000328/FIN-2015002-01 | 30 June 2015 23:59:59 | Sequoyah | Self-revealing | Failure to Adequately Follow Foreign Material Control Procedures | A self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to follow a foreign material exclusion procedure and precluded foreign material from entering the safetyrelated Essential Raw Cooling Water (ERCW) system. This resulted in wood debris within the ERCW that ultimately migrated to the 2B2 emergency diesel heat exchanger. Immediate corrective actions included removal of the foreign material and the performance of an engineering analysis to ensure the wood debris did not affect the system operability. The licensee placed this issue into their corrective action program as CR 1033792. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency allowed a piece of wood to enter into the 2B2 emergency diesel heat exchanger blocking 11 tubes. Because the finding was a deficiency affecting the design of a mitigating structure, systems, or components (SSCs) that was confirmed not to have resulted in the loss of operability, it was determined to be of very low safety significance (Green). The inspectors determined that no cross-cutting aspect will be assigned to this performance deficiency since it occurred in 2008 and is therefore not indicative of current licensee performance. |
| 05000327/FIN-2015001-01 | 31 March 2015 23:59:59 | Sequoyah | Self-revealing | Failure to Follow Procedure Results in an Inadvertent Sprinkler Deluge in the Cable Spreading Room | A self-revealing Green non-cited violation (NCV) of Technical Specification (TS) 6.8.1.f, Fire Protection Program Implementation, was identified for the licensees failure to follow a fire protection procedure. Specifically, the licensee failed to isolate the fire main from the cable spreading room (CSR) header during testing as required by procedure. This resulted in pressurization of the fire header to the cable spreading room which then caused a rupture of one of the sprinkler heads in the room. The licensee entered this issue into their corrective action program (CAP) as problem evaluation report (PER) 1001695. As immediate corrective actions, the licensee replaced the failed sprinkler head and conducted a formal review of the incident. The finding was determined to be more than minor because it was associated with the human performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the excessive amount of water sprayed in the CSR increased the likelihood of a plant transient due to the potential impact on non-waterproof junction boxes located in the CSR as well as safety-related instrument racks located in the auxiliary instrument room (AIR) directly below the CSR. Using Appendix A, Exhibit 1, Initiating Events Screening Questions, the finding was determined to be of very low safety significance because the deficiency did not cause a reactor trip nor a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding was determined to have a cross-cutting aspect in the avoid complacency component of the human performance area (H.12), because the technicians failed to properly implement appropriate error reduction techniques while performing a fire protection procedure. |
| 05000327/FIN-2015403-01 | 31 March 2015 23:59:59 | Sequoyah | Licensee-identified | Licensee-Identified Violation |