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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000369/FIN-2018012-01 | 2018Q3 | Mcguire McGuire | Failure to Translate Seismic Mounting Requirements for 125 VDC Vital Batteries into Installation and Replacement Procedure | The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to translate the mounting requirements for seismic qualification contained in NLI technical calculation C-017-074-2, Vital Instrumentation & Control Batteries & Racks Equipment Qualification Calculation, Rev. 0, into their battery replacement and installation procedure IP/0/A/3061/003, 125 Volt Vital Battery Maintenance and Repair, Rev. 23 |
| 05000259/FIN-2018010-01 | 2018Q2 | Browns Ferry | Licensee-Identified Violation | The Browns Ferry Nuclear Plant, Unit 3, Renewed Facility Operating License, DPR-68, License condition 2.C(7) required, in part, that TVA Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c)... Specifically, 10 CFR 50.48(c)incorporated by reference National Fire Protection Association Standard 805 (NFPA 805), and NFPA 805 section 2.4.2.2.2, Other Required Circuits, required in part, Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. Contrary to the above, since June 22, 2016, when the NFPA 805 requirements went into effect, the licensee did not implement and maintain in effect all provisions of the approved fire protection program, because the licensee did not correctly evaluate circuits that share common power supply for their impact on their ability to achieve nuclear safety performance criteria in accordance with NFPA 805.Significance: The team evaluated the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, for Mitigating Systems, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued May 2, 2018, and determined the finding to be of very low |
| 05000298/FIN-2018011-04 | 2018Q2 | Cooper | Incorrect Classification of Potential Safety-Related Components | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the inspectors identified three examples of the licensees failure to properly classify potential safety-related components in the emergency diesel generator ventilation system and RHR service water booster pump room cooling systems. |
| 05000298/FIN-2018011-03 | 2018Q2 | Cooper | Inadequate Design Basis Calculation for the EDG Rooms Temperature Distribution | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for the licensees failure to ensure design control measures provide for verifying or checking the adequacy of design of the emergency diesel generator room ventilation system by use of alternate or simplified calculation methods, or by a suitable testing program. Specifically, the licensee incorrectly extrapolated the results of the test program, which led to an incorrect room temperature profile. Additionally, the design calculation did not assume potential failures of the CO2 dampers. |
| 05000298/FIN-2018011-02 | 2018Q2 | Cooper | Failure to Ensure Adequate Design Control Measures are in Place Associated with RHR Service Water Booster Pump Room Cooling | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to incorporate malfunctions of the residual heat removal (RHR) service water booster pump (SWBP) room cooling temperature switch, which could cause environmental changes leading to functional degradation of system performance, into the design basis to verify the necessary protection system action be retained. |
| 05000298/FIN-2018011-01 | 2018Q2 | Cooper | Failure to Correct Extent of Condition of Surge Suppression Varistor Failures | An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, occurred when the licensee failed to correct conditions adverse to quality associated with the corrective actions identified in Condition Report RCR 2002-1665 to verify that installed surge suppressor varistors were appropriately sized and that design information was correctly reflected in controlled drawings for the reactor protection system, diesel generator control circuits, and high pressure coolant injection control circuits. |
| 05000250/FIN-2018001-02 | 2018Q1 | Turkey Point | Failure of radiation workers to notify Radiation Protection upon a spill of radioactively contaminated water | A self-revealing Green NCV of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for failure of radiation workers to notify Radiation Protection (RP), in accordance with procedure RP-AA-100-1002, Radiation Worker Instruction and Responsibilities, step 4.13.4, Spills and Observed Leaks, when a spill of radioactively contaminated water occurred. Specifically, on January 22, 2018, during a line-up of the 4D demineralizer resin fill isolation valve on the auxiliary building roof, two radiation workers (non-licensed operators) removed the weather-protective enclosure over the valve to verify its position. Upon removalof the enclosure, approximately half a gallon of highly contaminated water spilled onto the auxiliary building roof. The workers then attempted to clean up and decontaminate the area on their own with a water hose, rather than notify RP. This action spread the contamination into a larger area and into the site storm drain system |
| 05000250/FIN-2018001-01 | 2018Q1 | Turkey Point | Failure to conduct post maintenance testing in accordance with ASME OM code | A Green NRC-identified NCV of 10 CFR 50.55a, Codes and Standards, was identified for the failure to adequately perform post maintenance testing on valve CV-4-2906, 4B emergency containment cooler (ECC) air-operated outlet valve, in accordance with the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants. |
| 05000395/FIN-2018010-06 | 2018Q1 | Summer | Potential Unjustified Activation Energy for Barton Transmitters | The contractor, Impell Corporation, changed the activation energy for the Barton transmitters from 0.5 eV to 0.78 eV. The 0.78 eV was based upon an academic paper documenting experimental work, apparently, performed for the early space program and apparently first published in 1965. The paper cautioned the reader that the methods used were experimental and were not validated. A 0.5 eV activation energy for electronics was documented by the Electric Power Research Institute (EPRI) report NP-1558, which attributed it to electron migration of aluminum. The report was available to the licensee at the time of the change. Reports published by the Institute of Electrical and Electronics Engineers (IEEE) indicated that activation energies for various electronic failure modes could range from 0.5-0.66. Impell did not document an independent failure modes and effects analysis to justify the activation energy that they used. The licensee did not find the original qualification activation energies to be in error or non-conservative. The licensee chose to use less limiting activation energies that may not have been proven to be justified. In addition, the licensee was unable to demonstrate acceptable margins for extrapolation uncertainty. FSAR Section 3.11.2.1.3 stated that the environmental qualification of Class 1E equipment is in conformance with RG 1.89, Rev. 1. The RG in Section C.5.c stated that the aging acceleration rate and activation energies used during qualification testing and the basis upon which the rate and activation energy were established should be defined, justified, and documented. NUREG 0588 Section 5(2), Qualification Documentation, specified, in part that a certificate of conformance by itself is not acceptable unless it is accompanied by test data and information on the qualification program. The licensee captured this issue in their corrective action program as CR-18-00500, and determined that the NRC challenged the qualified life for Barton installed as IPT00456 based on an activation energy. VC Summer engineering does not agree with the NRC, nor do the OEMs Barton, Weed/Foxboro and Rosemount who have reviewed their prior research and state that it is suitable and adequate for our applications. The team must determine whether the activation energy used for the Barton transmitters was appropriate and, if not, whether the licensee had the responsibility to verify the information provided by their vendors and contractors. |
| 05000395/FIN-2018010-05 | 2018Q1 | Summer | Potential High Radiation Dose Areas with Unqualified Components | The NRC opened a URI to determine if a performance deficiency exists. The licensee did not perform analysis to determine the radiation exposure to shielded components adjacent to electrical and blank penetrations on the outboard side through containment. As a result, many mild environment components may be adversely affected. The inboard side of the penetrations is exposed to rad levels approaching 9X107 rads and the out board side is shielded by thin steel plates with electrical pass-thru holes. The inspectors noted that there were many areas of the plant identified as mild environments with unanalyzed penetrations. For example, the inspectors observed that the two trains for the plant service water were adjacent to unanalyzed penetrations. The components adjacent to the outboard side of the penetrations may be unqualified for service conditions expected during the most severe DBA as required by 10 CFR 50.49(e)(4). NUREG-0588 Section 1.4 "Radiation Conditions Inside and Outside Containment," required, in part, that "(8) Shielded components need be qualified only to the gamma radiation levels required..." and that "(12) Equipment that may be exposed to radiation doses below 104 rads should not be considered to be exempt from radiation qualification, unless analysis supported by test data is provided to verify that these levels will not degrade the operability of the equipment below acceptable values. The licensee provided a white paper for this issue that asserts that consideration of radiation streaming was not part of their licensing basis, thus enforcement would be addressed through a backfit analysis in accordance with 10 CFR 50.109. The team must determine whether the site licensing basis required consideration of radiation streaming and whether a backfit analysis would be appropriate in lieu of enforcement. The licensee captured this issue in their corrective action program as CR-18-00684 and determined that the process for qualification of equipment used was found acceptable per the VCS SER. Further evaluation will be performed under this CR but currently all components are qualified to their expected operating conditions and will perform their design functions. At worst, the EQ life of components may be reduced. All equipment in penetration areas are operable. |
| 05000395/FIN-2018010-04 | 2018Q1 | Summer | Unjustified Qualified Life for ASCO Valves | The NRC opened a Unresolved Item (URI) to determine if a performance deficiency was more than minor. In 1993, the licensees contractor, Impell Corporation, re-analyzed the qualified life established by ASCO qualification report AQR-67368 and a field notification from ASCO dated 10/27/1989. Impell erroneously used the heat rise temperatures from the field notification for both the AQR-67368 test samples accelerated aging temperature and the actual service temperatures in various plant locations. Replacing the actual test specimens documented accelerated aging temperature with an assumed temperature was not justified. As a result, when using the actual temperature identified in the qualification report, many of these solenoids are currently beyond their qualified lives. The licensee provided an alternate heat rise test report less limiting than the ASCO testing to justify that the ASCO valves were within their service lives, report 8058-001-2000-RA-0001-R00, Environmental Qualification Temperature Test of ASCO 206 and NP Series Solenoid Valves, dated June 2000. The teams evaluation must determine whether the alternate report is applicable to the licensee, and, if so, whether the test report indicated that the ASCO testing was invalid to conclude that the valves are currently within their qualified lives. 10 EnclosureNUREG-0588 Section 4(6) and Regulatory Guide 1.89, Rev. 1, Regulatory Position 5.c, required, in part, that the aging acceleration rate and the basis upon which it was established be described, documented, and justified. The team determined that the failure to justify the aging acceleration rate was a performance deficiency. However, a review of the additional information is warranted to determine if the performance deficiency is more than minor. The licensee entered the performance deficiency into their corrective action program as CR-18-00175 and determined that preliminary calculations indicated that the ASCO valves are currently operable based on the additional information provided for review. |
| 05000395/FIN-2018010-03 | 2018Q1 | Summer | Inadequate Radiation Harsh Environmental Qualification of Reactor Building Spray Pump A | During the review of EQDP-H-MO1-G03 for RB spray Pump A, the team noted that the pump was qualified for a maximum harsh environment of 1x106 radiation absorbed dose (rad); however, the total integrated dose (TID) was expected to be greater than 6.1x106rad TID over its 40 year life. Tab F1 of the EQDP, containing the equipment qualification report of the motors dated June 1977, stated that the maximum integrated radiation dose justified by the report over the 40 year operating life of the motor was 1x106 rads. The EQDP stated that component data shows that all components are suitable for the rated 1x106 rads integrated dose with the exception of (a) unfilled polyester resin and (b) the Dacron felt. In all cases, the polyester resins are filled to various degrees with glass or similar products. Such filling of the resin results in a significant increase in the radiation resistance of the combination -- as high as 9x108 rads. The Dacron felt by itself, at a threshold resistance of 8.6x105 rads, approaches the required radiation resistance but the felt is designed to be saturated with the impregnating epoxy resin and occurs only in this state. No specific data is available on the radiation resistance of the combination (Dacron filled epoxy), but the evidence indicates that the combination will exceed the required 1x106 rads. The team noted that the expected TID dose over the 40 year life of the RB spray pump A motor exceeded the original qualification provided in this test report. In order to ensure the pump was qualified for its radiation environment, the licensee had Impell Corporation perform Calculation 0980-036-030, Qualified Radiation Levels for GE Motors, Rev. 0, in August 31, 1988, which concluded that the motor was qualified for 1.5x107rads. The re-analysis was not based on partial type testing of the motor or a similar motor in accordance with NUREG-0588, but only reinterpreted the same material information previously provided by GE. The team noted that the reanalysis made different assumptions than GE did on the material characteristics of an unknown polyester resin fill material and Dacron felt. For the polyester resin, Impell could not determine what the fill material was or how much fill was used, but determined that it had a higher radiation resistance. For the Dacron felt, Impell assumed that the Dacron would not be a weak link in radiation resistance because of the epoxy. These assumptions were used to justify increasing the radiation qualification of the RB spray pump motor. The team determined that the original qualification of 1x106 rads was appropriate and was not proven to be inadequate by Impell because of the uncertainties documented by GE, and the lack of actual type testing information for the motor to support the Impell assumptions. FSAR Section 3.11.2 states that the licensee is committed to NUREG 588 Category II requirements. Section 2.1.2 of NUREG 588 states The choice of the methods selected is largely a matter of technical judgment and availability of information that supports the conclusions reached. Experience has shown that qualification of equipment subjected to an accident environment without test data is not adequate to demonstrate functional operability. In general, the staff will not accept analysis in lieu of test data unless (a) testing of the component is impractical due to size limitations, and (b) partial type test data is provided to support the analytical assumptions and conclusions reached. Section 2.1(3)(a) of NUREG 588 states Equipment that must function in order to mitigate any accident should be qualified by test to demonstrate its operability for the time required in the environmental conditions resulting from that accident. The team determined that the basis for raising the radiation qualification was not justified and that the qualification test report did not demonstrate that RB spray pump A was qualified over its 40 year operating life. Corrective Actions: On February 16, 2018, the licensee entered this issue into their corrective action program as CR 18-00707 and performed an immediate determination of operability to verify that the pump could still perform its intended safety function. 9 EnclosureCorrective Action Reference: CR 18-00707 Performance Assessment: The licensees failure to justify that RB spray pump A could perform its function under the radiation conditions expected during an accident in accordance with Section 2.1(3)(a) of NUREG 588 was a PD. The PD was determined to be more than minor because it adversely affected 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. Specifically, the failure to qualify the pump to expected radiation conditions adversely affects the pumps capability to perform its intended safety function during a design basis accident. The team used inspection manual chapter (IMC) 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for mitigating systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating structure, system, and component (SSC), and the SSC maintained its operability. Since the underlying cause of the issue occurred on August 31, 1988, the team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance. Enforcement: Title 10 CFR 50.49 (e)(4) requires, in part, that the electric equipment qualification program must include and be based on radiation, and the radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects. Contrary to the above, since August 31, 1988, the licensee failed to qualify RB spray pump A to the total dose expected during normal operation over the installed life of the pump and during the most severe DBA. This violation is being treated as an NCV, consistent Section 2.3.2 of the Enforcement Policy |
| 05000395/FIN-2018010-02 | 2018Q1 | Summer | Failure to Verify the Seismic Qualification of Valcor Solenoid Operated Valve XVX06050A | Calculation VCS-0423-DC-1, Valcor Voltage and Current Reducing Resistors, Rev. 0, dated September 10, 1981, located in Tab E1 of EQDP-H-VO4-V01 for solenoid operated valve XVX06050A, indicated a 300 ohm resistor was in series with the valve and that it reduced the voltage in the coil to approximately 32VDC at minimum conditions. The team questioned if the valve was seismically qualified at the lower voltage since the seismic qualification in test report QR 52600-515, Section 4.2.5, Seismic Vibrations, stated that it was performed at 108VAC. The team noted that the Valcor SOV was not installed in the same configuration that it was seismically qualified. The failure to ensure the valve was seismically qualified, as configured, did not ensure that damage would not occur during a seismic event. FSAR Section 3.10 stated that seismic qualification must be done in 7 Enclosureaccordance with IEEE 344-1971. Section 3.2.2.2 of IEEE 344-1971 states the device being tested should demonstrate its ability to perform its intended safety function and sufficient monitoring equipment should be used to evaluate its performance. The team determined that the licensee did not demonstrate the seismic qualification of valve XVX06050A in its current plant configuration at reduced voltage. Corrective Actions: On February 15, 2018, the licensee entered this issue into their corrective action program as CR 18-00686 and performed an immediate determination of operability to verify that the valve could still perform its intended safety function. Corrective Action Reference: CR 18-00686 Performance Assessment: The licensees failure to verify the adequacy of the seismic design and qualification of valve XVX06050A in accordance with IEEE 344-1971 was a performance deficiency (PD). The PD was determined to be more than minor because it adversely affected the Design Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to verify the adequacy of design for seismic qualification of the valve resulted in the valve being installed in an unqualified configuration. The team used inspection manual chapter (IMC) 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for barriers, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. Since the underlying cause of the issue occurred on August 30, 1988, the team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance. Enforcement: Title10 CFR Part 50, Appendix B, Criterion III Design Control, requires, in part, that The design control measures shall 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. Contrary to the above, since August 30, 1988, the licensee failed to verify valve XVX06050A was seismically qualified in its current configuration in accordance with IEEE 344-1971. This violation is being treated as an NCV, consistent Section 2.3.2 of the Enforcement Policy. |
| 05000395/FIN-2018010-01 | 2018Q1 | Summer | Failure to Justify Activation Energy for Valcor SOV XVX06050A | The qualification of the Valcor SOVs, completed in 1979, used the 10oC rule to determine the accelerated aging rate, which was equivalent to a 0.831 eV activation energy derived for Valcors ethylene propylene rubber (EPR). The inspectors determined that 0.831 eV for EPR, although realistic, it was not the most limiting identified for EPR. Valcor originally qualified the SOVs for 40 years at 120oF, however many of the valves are normally energized and will see temperatures exceeding 120oF. The SOV, XVX06050A, is a normally energized open valve that de-energizes to close on a containment isolation phase A signal and opened post-accident for hydrogen analyzing in the reactor building. In 1988, Impell Corporation, the licensees contractor, reanalyzed the qualification and determined that DuPont Tefzel insulation was the most limiting component instead of EPR and that a 50% loss of tensile strength was the limiting failure mechanism at 0.95 eV activation energy. To extrapolate a new activation energy, Impell estimated data points from a rudimentary log life plot that did not have any actual test data points. Impell obtained the plots from a DuPont Tefzel design handbook which also contained the log life plot for the elongation to break failure parameter of Tefzel, which appeared more limiting than tensile strength. Because the new activation energy extrapolation did not use actual test data, the extrapolation of that data was less limiting than the original qualification activation energy, and the elongation to break failure parameter was not evaluated, the team determined the new activation energy was not justified. FSAR Section 3.11.2.1.3 stated that the environmental qualification of Class 1E equipment is in conformance with RG 1.89, Rev. 1. Section C.5.c of the RG stated that the aging acceleration rate and activation energies used during qualification testing and the basis upon which the rate and activation energy were established should be defined, justified, and documented. The licensee did not find the original qualification activation energy to be in error or non-conservative. The licensee chose to develop an activation energy from less limiting log life plots, which was non-conservative. In addition, without actual data for the log life plots, the licensee was unable to demonstrate acceptable margins for uncertainty. The team determined that the valve would have exceeded its qualification based on the original qualification and unjustified use of the new activation energy. Corrective Actions: On February 19, 2018, the licensee entered this issue into their corrective action program as CR 18-00754 and performed an immediate determination of operability to verify that the valve could still perform its intended safety function. Corrective Action Reference: CR 18-00754 6 EnclosurePerformance Assessment: The failure to justify the basis upon which the activation energy of Valcor SOV XVX06050A was established in accordance with RG 1.89 Section C.5.c was a performance deficiency (PD). The PD was determined to be more than minor because it adversely affected the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to justify the activation energy used for Tefzel adversely affected the reliability of the solenoid to maintain its qualification over the entire 40 year qualified life of the plant. The team used inspection manual chapter (IMC) 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for barriers, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. Since the underlying cause of the issue occurred in 1988, the team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance. Enforcement: Title 10 CFR 50.49 (e)(5) states Equipment qualified by test must be preconditioned by natural or artificial (accelerated) aging to its end-of-installed life condition. Consideration must be given to all significant types of degradation which can have an effect on the functional capability of the equipment. If preconditioning to an end-of-installed life condition is not practicable, the equipment may be preconditioned to a shorter designated life. The equipment must be replaced or refurbished at the end of this designated life unless ongoing qualification demonstrates that the item has additional life. Contrary to the above, since August 30, 1988, the licensee failed to age Valcor SOV XVX06050A to its end of life condition and to replace the equipment at the end of its designated life. This violation is being treated as an NCV, consistent Section 2.3.2 of the Enforcement Policy. |
| 05000260/FIN-2017004-01 | 2017Q4 | Browns Ferry | Inadequate Determination of Operability for the HPCI System | Two examples of an NRC-identified NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" were identified for the licensee's failure to properly implement operability evaluation requirements for degraded High Pressure Coolant Injection (HPCI) components. Specifically, from September 23 to September 28, 2017, the operability evaluations for degraded Unit 2 and 3 HPCI injection valves 2/3- FCV-73-44 did not provide reasonable assurance of operability as per the sites operability review procedures.The performance deficiency was determined to be more-than-minor because it impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. As an immediate corrective action, the licensee later performed maintenance to open and inspect these valves. Subsequently the licensee initiated condition reports and a Performance Assessment Worksheet to assess the training for such evaluations. The violation was entered into the licensee's corrective action program (CAP) as CR 1341458. The inspectors determined that the finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution area (P.2), because the organization concluded Technical Specification operability prior to thoroughly investigating these issues commensurate with their potential safety significance. |
| 05000259/FIN-2017004-03 | 2017Q4 | Browns Ferry | Licensee-Identified Violation | The following licensee-identified violation of NRC requirements was determined to be of very low safety significance and met the NRC Enforcement Policy criteria for being dispositioned as an NCV. 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, required, in part, thatmeasures 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, from May 2011 to September 26, 2017, Browns Ferry staff had not identified that a Reactor Building Crane degraded, adverse condition existed. The crane, which supports the safety-related movement function for irradiated fuel assemblies, had not been meeting the required Single Failure-Proof qualification as described in the Final Safety Analysis Report section 12.2.2.5 and licensee commitments to NRC NUREG 0554 since the last wire rope replacement in May 2011. Review for cause by the licensee determined that the Main Hoist Equalizer Arm had been resting on one of its stops, causing a loss of the shock absorbing Single Failure function. The finding screens to green per IMC 0609 Appendix A, Exhibit 3 as it was a only a qualification issue that did not cause mechanical damage to fuel, did not result in a loss of spent fuel pool water inventory, and did not affect SFP component placements. The licensee entered this issue into the CAP as CR 1341964. Immediate corrective action was to reestablish the Single Failure qualification per WO 119082082. |
| 05000395/FIN-2017007-04 | 2017Q4 | Summer | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures 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. Contrary to the above, since 2010, the licensee failed to evaluate the loading of the emergency diesel generators at the maximum voltage and frequency allowed by TS 3/4.8.1 in Calculation DC08360-006, Diesel Generator 1A and 1B Loading, Rev. 12, and to evaluate battery terminal voltage at the maximum battery cell-to-cell resistance allowed by TS 3/4.8.2 in Calculation DC08320-010, Class 1E 125 Volt DC System Voltages and Voltage Drop, Rev. 18. 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 SSC, and the SSC maintained its operability. The licensee entered these issues into their CAP as CRs 10-02395 and 10-02033. ATTACHMENT: SUPPLEMENTAL INFORMATION |
| 05000296/FIN-2017004-02 | 2017Q4 | Browns Ferry | Failure to Perform an IDO without delay for 3A EDG after Observing Indications of a Degraded Condition | The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to perform an immediate operability determination (IDO) for 3A Emergency Diesel Generator (EDG) upon discovering a degraded condition. Specifically, on December 19, 2017, the licensee failed to perform an IDO after identifying and confirming less than minimum cooling flow, thus leaving the EDG in an indeterminate state of operability.The performance deficiency is more than minor because it was associated with the equipment performance attribute and affected the associated cornerstone objective to ensure availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. As a corrective action, the licensee performed operations to restore flow within the acceptable range and performed an IDO. The violation was entered into the licensee's CAP as CR 1370601. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of H.13, Consistent Process, because the performance deficiency was caused by not following a consistent, systematic approach to making a decision concerning operability of the affected DG. |
| 05000395/FIN-2017007-03 | 2017Q4 | Summer | Failure to Identify a CAQ for Power Shield Catalog #609903-T501N | The NRC identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to identify that a deviation in equipment qualification of power shield relays in 480V switchgear XSW-1DB1 was a condition adverse to quality in their CAP. Specifically, the licensee failed to identify that Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was not qualified to meet its original total integrated dose limit of 100,000 rads as stated in the Asea Brown Boveri 10 CFR Part 21 notification letter. The licensee entered this issue into their CAP as CR-17-05391 and performed an evaluation to determine there was reasonable assurance that the power shield relay in purchase order NU-02SR750589 could perform its intended safety function. 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 identify that Power Shield catalog #609903-T501N in purchase order NU-02SR750589 was not qualified to the 1,350 rad TID specified in the equipment qualification database for zone AB-72 resulted in a reasonable doubt that the qualification requirements over the relays service life would be met. The team determined the finding to be of very low safety significance (Green) because the finding affected the design or qualification of a mitigating SSC and the SSC maintained its operability. The team determined that no crosscutting aspect was applicable because the finding did not reflect current licensee performance. |
| 05000395/FIN-2017007-02 | 2017Q4 | Summer | Failure to Establish a Testing Program for Inverter XIT5904 and Time Delay Relay in the EFW/SW Crosstie Valve Actuation Circuitry | The NRC identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, involving two examples. Specifically, the licensee (1) failed to establish a testing program to assure the adequacy of the shutdown setpoint of the safety-related inverters, and (2) failed to establish a testing program to assure the adequacy of the time delay relay in the emergency feedwater/service water (EFW/SW) crosstie valve actuation circuitry. The licensee entered this issue into their CAP as CRs17-05534 and 17-05536, and performed an operability determination to verify that the safety-related components remained operable. 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 establish a testing program for the low voltage DC setpoint of inverter XIT 5904 and for the time delay relay in the EFW/SW crosstie actuation circuitry could result in undetected degradation of the equipment to perform their intended safety functions. The team determined the finding to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding did not reflect current licensee performance. |
| 05000395/FIN-2017007-01 | 2017Q4 | Summer | Failure to Verify the Adequacy of Design for the EFW system when Supplied by SW | The NRC identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the emergency feedwater (EFW) pumps would be capable of taking suction from service water for an indefinite period of time as required by Updated Final Safety Analysis Report Section 10.4.9.2. The licensee entered this issue into their corrective action program (CAP) as condition report (CR) 17-05528 and performed an operability determination to verify the EFW pumps remained operable. 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 evaluate worst-case design conditions resulted in a reasonable doubt that the EFW pumps could provide cooling water to the steam generators and perform their design basis function. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, and component (SSC), and the SSC maintained its operability. The team determined that no crosscutting aspect was applicable because the finding did not reflect current licensee performance |
| 05000250/FIN-2017007-09 | 2017Q3 | Turkey Point | Failure to Update the UFSAR with the Latest Information Developed | The NRC identified a Severity Level-IV non-cited violation of Title 10 Code of Federal Regulations 71(e), Maintenance of Records, Making of Reports, for the failure to assure that the Updated Final Safety Analysis Report (UFSAR) contained the latest information developed, including all changes made in the facility or procedures as described in the UFSAR. The team determined that the licensee failed to update the UFSAR to include the latest information regarding several design features associated with turbine runback. For immediate corrective actions, the licensee entered this issue into their corrective action program as AR 2218695 to update the UFSAR. The NRC determined this violation was associated with a minor performance deficiency in accordance with the screening criteria in IMC 0612, Appendix E. Because the failure to update the UFSAR impacted the NRCs ability to perform its regulatory process, the team evaluated the violation using the traditional enforcement process. The team determined that this met the criteria for a SLIV violation because not accurately describing turbine runback design features in the UFSAR could have a material impact on licensed activities, and met the SLIV violation criteria in 6.1.d.3 of the NRC Enforcement Policy. The violation represented a failure to update the UFSAR as required by Title 10 Code of Federal Regulations Part 50.71(e), but the lack of up-to- date information has not resulted in any unacceptable change to the facility or procedures. Cross-cutting aspects are not assigned to traditional enforcement violations |
| 05000250/FIN-2017007-08 | 2017Q3 | Turkey Point | Failure to Correct a Non-Conforming Condition Impacting Containment | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to take timely corrective action to maintain the unit 3 and 4 containment cathodic protection systems. These systems have been non-functional on both units since 2009. The cathodic protection systems purpose is to protect the containments interconnected liner, reinforcing bars, and tendon trumplates. For immediate corrective actions, the licensee entered the issue into their corrective action program as AR 2216534 and performed a prompt operability determination. The licensee concluded that the containment structure was operable but non-conforming and established plans to monitor the potentially impacted inaccessible areas through continued performance of the American Society of Mechanical Engineers (ASME) IWL and IWE programs until actions are taken to restore the Cathodic Protection System. The performance deficiency was determined to be more than minor, because it is associated with the Design Control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of maintaining the containment structural integrity and operational capability to provide reasonable assurance that the containment protects the public from radionuclide releases caused by accident or events. Specifically, the failure 6 to implement timely corrective actions to maintain the protection of the containments interconnected liner, reinforcing bars, and tendon trumplates affected the structural integrity and operational capability of the containment structure. The team determined the finding to be of very low safety significance because the finding was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance |
| 05000250/FIN-2017007-07 | 2017Q3 | Turkey Point | Failure to Identify ICW Pipe Corrosion | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to inspect Intake Cooling Water (ICW) piping in accordance with license renewal commitments. For immediate corrective actions, the licensee entered the issue into their corrective action program as AR 02218430 and AR 02218437, planned to perform localized corrosion wall thickness measurements to ensure the ICW system remained operable. The 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 reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, unmonitored corrosion affects the reliability of the ICW systems. The team determined the finding to be of very low safety significance because it did not represent an actual loss of function of one or more non-Tech Spec trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hrs. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee failed to implement a corrective action program with a low enough threshold for identifying issues (P.1). Specifically, individuals routinely failed to identify corrosion issues on CCW system area walk downs that exceeded proceduralized acceptance criteria of light surface rust specified in 0-ADM- 564, during the July 5, 2017, August 11, 2016, and April 11, 2016 CCW area walk downs. |
| 05000250/FIN-2017007-06 | 2017Q3 | Turkey Point | Failure to Verify the Adequacy of Design for the ECC and CCW Systems | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the Emergency Containment Cooler (ECC) unit 4A auto start circuitry would not result in exceeding the thermal limits of the CCW system during a design basis accident. Specifically, the licensee failed to verify that a single active failure of the motor starter auxiliary contacts would not result in exceeding the design basis limits for CCW as described in updated final safety analysis report (UFSAR) Section 9.3. For immediate corrective actions, the licensee entered the issue into their corrective action program as AR 2219505, performed a prompt determination of operability, and determined that the CCW system remained operable. 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, three ECC fans running during a during a design basis accident would result in exceeding the design basis temperature of 158.6 F for the 5 CCW supply and a significant reduction in margin for the SI pump lube oil cooler. The team determined the finding to be of very low safety significance because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC) and the SSC maintained its operability. This finding was not assigned a cross- cutting aspect because the issue did not reflect current licensee performance. |
| 05000250/FIN-2017007-05 | 2017Q3 | Turkey Point | Failure to Adequately Perform Discharge Testing on Battery 3B | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to perform surveillance testing on station battery 3B in accordance with the requirements of Institute of Electrical and Electronic Engineers (IEEE) 450-1987. For immediate corrective actions, the licensee entered this issue into their corrective action program as AR 2219948 and performed an extent of condition review, which determined that none of the station batteries were currently in a degraded condition, and placed surveillance procedure 0-SME-003.15 on administrative hold until the corrective actions are completed. The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the performance deficiency could result in masking degradation of the battery on future performance discharge tests and adversely affect the ability to trend when the testing periodicity should be increased to once a year as required by Technical Specifications (TS). The team determined the finding to be of very low safety significance because the finding did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hrs. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000250/FIN-2017007-04 | 2017Q3 | Turkey Point | Failure to Verify the Adequacy of Design for Component Protective Covers | The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design for the non-safety related component protective covers attached to safety related equipment. For immediate corrective actions, the licensee entered this into their corrective action program as AR 02220993 and removed visibly degraded protective covers. 3 The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute and of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as pow er operations. Specifically, the failure to ensure the quality and qualification of commercial components and assemblies to maintain adequate mounting to Class 1E equipment increased the likelihood of inadvertent component failures, and thus increased the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The team determined the finding to be of very low safety significance because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feedwater). This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000250/FIN-2017007-03 | 2017Q3 | Turkey Point | Failure to Verify the Adequacy of CCW isolation from Supplemental Cooling System (SCS) | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design of temperature set points used for isolation of the Component Cooling Water (CCW) from the CCW supplemental cooling system (SCS) during an accident. For immediate corrective actions, the licensee entered this into their corrective action program as AR 2218834, performed an operability determination, which determined the system is operable but non-conforming, and issued engineering change (EC) 289598 to account for uncertainties in the CCW SCS temperature isolation setpoint. 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 4 capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by not ensuring prompt isolation or adjusting the isolation setpoint to account for instrument uncertainties and temperature lag, the licensee failed to ensure that the SCS loop would be isolated at onset of an accident, which affected the reliability and capability of the CCW system when called upon. The determined the finding to be of very low safety significance because the findings were a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained their operability or functionality. The finding had a cross-cutting aspect in the area of Human Performance because the licensee failed to ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values (H.9). |
| 05000250/FIN-2017007-02 | 2017Q3 | Turkey Point | Failure to Perform Design Verification for Under Frequency Trip of the Main Generator Breakers | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for failure to verify or check the adequacy of design of the under frequency trip feature of the main generator circuit breakers with regard to the effect of its operation on plant stability and the maintenance of critical safety functions. The licensee entered this issue into their corrective action program as AR 2220874 and AR 2224998. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions. Specifically, opening of the main generator breakers due to an under frequency condition on the offsite power system would cause the generator load to suddenly drop from full power to the level of the plant loads, and there was no verification that plant stability and critical safety functions would be maintained. The team evaluated the finding with Inspection Manual Chapter 0609, Appendix A, and determined the finding met the Support System Initiators screening criteria for requiring a detailed risk evaluation. The team determined that this issue increased the likelihood of the support system initiator loss of offsite power (LOOP). The regional senior risk analyst conducted a detailed risk evaluation with a one year exposure and determined the change in core damage frequency was less than 1E-6, which was of very low safety significance (Green). The team did not assign a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000250/FIN-2017007-01 | 2017Q3 | Turkey Point | Inadequate Verification of Electrical Protective Device Selective Coordination | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for failure to verify that coordination exists between the protective devices on safety related switchgear in order to minimize the probability of losing a safety related power bus. For immediate corrective actions, the licensee entered this issue into their corrective action program as Action Request (AR) 2220956 and performed an operability determination, which determined the system was operable, and was performing a reevaluation of the calculation to determine adequate coordination. 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 verify short circuits in non-safety related SSCs downstream of the safety related switchgear would not cause a lockout of the safety related bus affected its availability and reliability. The team determined the finding to be of very low safety significance because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained their operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000250/FIN-2017007-10 | 2017Q3 | Turkey Point | Potential failure of 125 Vdc Bus 3B Class 1E components | UFSAR Section 8.2.2.3.1 stated that the emergency power for vital instrumentation and controls is supplied by a station DC power system which contains five safety related 125Vdc batteries and four DC distribution panels. 125 Vdc distribution panel 3B supplies safety related power to several safety-related equipment including sequencers, reactor trip switchgear, inverter 3Y06, and control power to 480Vac load centers 3B and 3D and 4160 Vac switchgears 3AB01 and 4AB20. UFSAR Section 7.2 stated that the reactor protection system was designed in accordance with IEEE 279- 1968. Section 4.5 of IEEE 279-1968, Channel Integrity, requires all protection system channels be designed to maintain necessary functional capability under extremes of conditions relating to malfunctions. During the review of calculation 5177-265-EG-22, Circuit Breaker/Fuse Coordination Study, Rev. 8, the team questioned if there were instances where class 1E cables associated with DC Bus 3B (3D23) would not be adequately protected given a short circuit on the load side of the breakers. The failure to ensure the Class 1E protective devices would not allow the maximum available short circuit to permanently damage cabling to safety-related equipment associated with DC Bus 3B could result in additional loss of Class 1E equipment. Unresolved Item (URI) 05000250/2017007-01 and 05000251/2017007-01, Potential failure of 125 Vdc Bus 3B Class 1E components,) is opened for additional review to determine if the Class 1E cables on DC Bus 3B can withstand the maximum possible short circuit and to determine if a performance deficiency exists. |
| 05000348/FIN-2017007-02 | 2017Q2 | Farley | Untimely Corrective Actions for Check Valve Q2E21V0026 | The NRC identified a non-cited violation (NCV) of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to implement timely corre ctive actions to change the classification of check valve Q2E21V0026 (QV026) from cat egory C to category A/C in accordance with ASME OM Code-2001, Subsection ISTC -1300, Valve Categories. The licensee entered this issue into their corrective action program as condition report 10377744, reclassified the valve as category A/C in January 2017 to perform the leakage test during the next outage, and determined there was reasonable assurance the valve could perform its intended safety function until the outage. The performance deficiency was determined to be more than minor because it was associated with the structure, system, component, and barrier performance attribute of the Barriers Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to implement timely corrective actions resulted in the licensee not ensuring reverse flow to the refueling water storage tank (RWST) from the containment sump during the recirculation phase of safety injection (SI) would not exceed the plants dose rate limits. The team determined the finding to be of very low safety significance (Green) because the finding did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool, and the finding did not represent a degradation of the barrier func tion of the control room against smoke or a toxic atmosphere. The team determined the finding was indicative of present licensee performance and was associated with the cross cutting aspect of Conservative Bias in the area of Human Performance because the licensee failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14). |
| 05000348/FIN-2017007-01 | 2017Q2 | Farley | Failure to Translate Design Basis Time Requirement into the Time Critical Operator Action Program Procedure | The NRC identified a non-cited violation (NCV) of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate the design basis time limit for the alignment of the emergency core cooling system (ECCS) to cold leg recirculation into their time critical operator action procedure. Specifically, the licensee failed to translate the ECCS to cold leg recirculation alignment activity time requirement of 9 minutes and 25 seconds from calculation SM-94-0452-001, RWST Depletion During Injection Mode with LOCA Until Switchover to Recirculation, Version 5.0, and UFSAR Table 6.3-4, into procedure NMP- OS-014-001, FNP Time Critical Operator Action Program, Version 4.0. The licensee entered this issue into their corrective action program as condition report 10365952 and determined that operability was not impacted due to conservatisms in the calculation and recent operating crew simulator performance. 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 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 licensees failure to translate the correct design basis time requirement into their acceptance criteria in procedure NMP-OS-014-001 resulted in several unidentified periodic time validation failures without remediation, therefore adversely affecting the licensees capability and reliability of aligning safety-related equipment needed during a loss of coolant accident within the established design basis time limits. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design and qualification of a mitigating system, structure, or component (SSC), and the SSC maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding did not reflect current licensee performance. |
| 05000413/FIN-2017007-01 | 2017Q1 | Catawba | Failure to Translate Design Requirements into Operating Procedures for NW System | Green: The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate the limiting high pressure design requirement of the containment valve injection water (NW) system surge chamber into station proc edures. Specifically, the licensee failed to translate the NW surge chamber high pressure design limit of 85 psig from calculation CNC-1223.19-00-0004, NW system setpoint ca lculation, Rev. 7, into procedure OP/1/A/6200/019, Containment Valve Injection Water System, Rev. 36, to ensure the NW system could perform its intended safety function during a design basis accident. The licensee entered this issue into their corrective action program as action request 02096392, reviewed the issue for current and past operability, and issued an operations guide to limit the NW surge chamber pressures to 80 psig. The performance deficiency was determined to be more than minor because it adversely affected the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to translate the 85 psig NW surge chamber pressure limit into procedures resulted in exceeding the NW surge chamber high pressure limit, which could result in an inability of the safety re lated nuclear service water system to provide makeup water to the NW surge chamber and result in entrainment of nitrogen gas in the surge chamber outlet. The team determined the finding to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containmen t, containment isolation system, and heat removal components, and the finding did not involve an actual reduction in function of hydrogen igniters in the reactor containment. This finding was assigned a cross-cutting aspect of Evaluation in the Problem Identification and Resolution Area because the finding was indicative of present licensee performance, and the licensee did not thoroughly evaluate the issue identified in ARs 01912139 and 01912453 after the revision to the calculation was completed to ensure that the correct high pressure NW surge chamber design requirement would have been translated into procedures (P.2) |
| 05000269/FIN-2016004-01 | 2016Q4 | Oconee | Failure to Perform Appropriate Evaluation of Motor Operated Valve Actuator Output Capability | Green. The NRC identified a non-cited violation (NCV) of Title 10 Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly determine the bounding degraded voltage to be assumed in the determination of motor operated valve (MOV) actuator output capability. Specifically, the licensee did not use appropriate transient voltages as input into the evaluation of the capability of the MOVs that are required to reposition in response to an accident signal. In response, the licensee entered the issue into their corrective action program as nuclear condition report (NCR) 2056895 and planned to formally revise their calculations to reflect the current plant configuration. This performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees programmatic failure to use bounding terminal voltage values in the evaluation of their automatically actuated, safety-related MOVs did not ensure they would be capable of mitigating accidents when powered from sources other than the 230kV switchyard, thus resulting in doubt on their capability to perform their intended safety function. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the most recent transient analysis that was performed for the sources other than the 230kV switchyard was performed in 2012. |
| 05000250/FIN-2016008-01 | 2016Q4 | Turkey Point | Licensee-Identified Violation | Turkey Point Nuclear Generating Station, Unit 4, Renewed Facility Operating License 3.D, Fire Protection, stated that Florida Power and Light (FPL) shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48 (c), National Fire Protection Association (NFPA) 805. NFPA 805, Section 2.4.2.2.2 (b), Common Enclosure Circuits, required circuits that share a common enclosure with circuits required to achieve nuclear safety performance criteria shall be identified for their impact on the ability to achieve nuclear safety performance criteria. Contrary to the above, since 2014, the licensee failed to identify circuits that impact the ability to achieve nuclear safety performance criteria as a result of the effects of fire on circuits that share a common enclosure with the Unit 4 4kV switchgear. The violation was determined to be of very low safety significance based on risk evaluation provided by the licensee and reviewed by NRC senior reactor analyst. The licensee entered this issue into their corrective action program as action request 2134673. |
| 05000269/FIN-2016004-02 | 2016Q4 | Oconee | Inappropriate Voltage Band in Lee Combustion Turbine Unit Operating Procedure | Green. The NRC identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to identify appropriate procedural updates that were needed to ensure the Lee combustion turbine (LCT) procedures were appropriate for the circumstances and maintained current. Specifically, the licensee did not include appropriate operational limitations in procedures associated with the LCTs. In response, the licensee generated NCR 2058763, verified the LCT automatic voltage regulator setpoint was, and had been, 13.8kV, and generated a corrective action to revise the affected procedures limits to 13.78kV, a value bounded by station analyses. This performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Oconees failure to limit the operating voltage band of the LCTs to an amount that was demonstrated as acceptable by analysis resulted in doubt on their capability to provide power to safety-related equipment during an accident. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the update to the procedures occurred in January and October 2007, after replacement of the LCTs. |
| 05000269/FIN-2016003-01 | 2016Q3 | Oconee | Failure to Translate Design Requirements to Prevent the Effects of Waterhammer | The NRC identified a finding for the licensees failure to translate the limiting flow rate design requirement into station procedures used to start and operate the alternate reactor building cooling (RBC) system, in accordance with the Duke Energy Carolinas Topical Report, Quality Assurance Plan (QAP). Specifically, the licensee failed to translate the limiting flow rate of 170 gallons per minute (gpm) into Procedure AP/0/A/1700/051, Alternate Reactor Building Cooling, Revision (Rev.) 2, to ensure prevention of waterhammer on the A reactor building cooling unit (RBCU) or connecting low pressure service water (LPSW) lines when starting the RBCU Hale pump. The licensee entered this issue into their corrective action program as Action Request (AR) 02049903 and revised Procedure AP/0/A/1700/051 to limit the RBCU Hale pump discharge flow to each affected unit to an initial fill rate of 120 gpm or less. The performance deficiency was determined to be more than minor because it adversely affected the protection against external factors 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, opening the RBCU Hale pump discharge valve four turns, as specified in the procedure, would have resulted in filling the alternate RBC system at approximately 600-700 gpm and exceeding the design flow rate of 170 gpm established to prevent equipment and piping damage as a result of waterhammer. This provided a reasonable doubt that the alternate RBC system had the capability to reliably perform its intended safety function and, in turn, that the protected service water (PSW) system had the capability to meet its 30-day mission time during a turbine building fire that resulted in a loss of offsite power. The finding was determined to be of very low safety significance (Green) because the finding would not have resulted in a fire that caused secondary fires outside of the originating fire area due to circuit issues and did not affect the ability to reach and maintain a stable plant condition within the first 24-hours of a fire event. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of design margin, in the area of human performance. Specifically, the licensee failed to operate and maintain the alternate RBC system equipment within design margins when they did not translate design requirements from Engineering Change (EC) 110008 and Calculation OSC-8107 into station procedures. |
| 05000287/FIN-2016003-02 | 2016Q3 | Oconee | Licensee-Identified Violation | Technical Specification (TS) 5.4.1., Procedures, states, in part, written procedures shall be established, implemented, and maintained covering activities described in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Procedure MP/0/A/3009/017, Visual PM Inspection and Electrical Motor Tests is used by the licensee during maintenance of electric motors. Contrary to the above, on April 25, 2016, the licensee did not adequately implement maintenance procedure MP/0/A/3009/017. Specifically, the licensee incorrectly wired the 3C RBCU motor control center contactor leads during maintenance causing 3C RBCU fan to operate in the reverse direction. On June 16, 2016, during an engineer walkdown, the engineer noted anomalies in the RBCU inlet temperature readings. On June 28, 2016, while investigating the temperature readings the licensee discovered that the 3C RBCU fan was operating in the reverse direction and declared the 3C RBCU inoperable. The 3C RBCU was inoperable when the plant entered Mode 4 on May 14, 2016 until June 28, 2016 when the 3C RBCU was repaired (approximately 45 days). Technical Specification 3.6.5, Reactor Building Spray and Cooling Systems, requires all three trains of RBCU operable while in Modes 1, 2, 3, and 4. On May 14, 2016, Unit 3 was starting-up from the refueling outage and entered Modes 4 through 1 with one train of RBCU inoperable. This action of changing modes with the 3C RBCU inoperable is prohibited by TS 3.0.4. The licensee entered this condition into their corrective action program as NCR 02041501. The licensee also restored 3C RBCU operability, trained/counseled technicians, and incorporated a procedure change which will enhance configuration control for the lifted leads aspect in the maintenance procedure for this activity. This finding was assessed using IMC 0609, Phase 1 screening worksheet of Attachment 4, Appendix A, and Appendix H, and was determined to be of very low safety significance (Green). |
| 05000259/FIN-2016010-01 | 2016Q2 | Browns Ferry | Failure to Include Required Gasket Replacement in Limit Switch Surveillance Procedure | An NRC-identified non-cited violation (NCV) of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include vendor requirements for maintaining the environmental qualification of the main steam isolation valve (MSIV) limit switches in maintenance procedures. Specifically, not maintaining the MSIV limit switches in their qualified condition impacts their reliability. The licensee entered this issue into the corrective action program as CR 1160702. The licensee evaluated the impact of the incorrect guidance, and determined that all three units were affected, and that the MSIV limit switches remained operable, although they were in an unqualified condition. The licensee plans to correct the affected procedures. This performance deficiency was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance 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, not maintaining the MSIV limit switches in their qualified condition impacted their reliability. The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and 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. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000338/FIN-2016002-01 | 2016Q2 | North Anna | Inadequate Translation of Design Calculations into Compensatory Measures when Removing Missile Barriers Could Result in EDGs and SBO Diesel Inoperable | The NRC identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the inadequate translation of design calculations into compensatory measures when removing missile barriers. The inadequate translation of design calculations into compensatory measures when removing required passive missile shields is a performance deficiency (PD). The PD was more than minor because it was associated with the human performance attribute of the Mitigating System 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 (i.e. core damage). Specifically, the inadequate compensatory measure resulted in the licensee having to make required changes to the compensatory measures in order to resolve missile protection concerns. The inspectors performed the initial significance determination for the finding using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because the safety function of the onsite emergency diesel generators (EDGs) and the function of the station blackout (SBO) diesel were assumed to be completely failed due to inadequate compensatory missile protection measures for a high wind event. The finding has a cross-cutting aspect in the area of human performance associated with the conservative bias attribute because individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather that unsafe in order to stop (H.14). The licensee entered this issue into the corrective action program (CAP) as Condition Report (CR)1034958. |
| 05000339/FIN-2016002-02 | 2016Q2 | North Anna | Adequacy of Class 1E 120VAC Vital Bus Design | The NRC identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design for the protection devices at the 120VAC vital instrumentation buses. Specifically, the licensees failure to verify that the protective devices at the Unit 1 and Unit 2 120VAC vital instrumentation buses would isolate failed equipment when supplied by the voltage regulating transformer in accordance with IEEE 308-1971 was a PD. The licensee entered this issue into their CAP as CRs 1006865 and 1013278. At the time of the inspection, the licensee was evaluating the issue to determine appropriate corrective actions. This does not present an immediate safety concern because the performance deficiency is related to a non-conformance with a design standard upon which only one train would be affected by a postulated single failure and the other train would remain available and capable to respond to the design basis accident. The performance deficiency was determined to be more than minor because it adversely affected the Design Control 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, failing to isolate failed equipment could lead to tripping the associated 120VAC vital bus, causing unnecessary loss of other safety related equipment connected to the bus. The finding was determined to be of very low safety significance (Green), because it was a deficiency affecting the design or qualification of a structure, system, or component (SSC) and the SSC maintained its operability. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. |
| 05000338/FIN-2016002-03 | 2016Q2 | North Anna | Licensee-Identified Violation | 10 CFR Part 50.65 section (a)(2) requires, in part, that the performance or condition of a structure, system, or component is being effectively controlled through the performance of appropriate preventative maintenance, such that the structure, system, or component remains capable of performing its intended function. In accordance with this requirement, the licensee established procedure ER-AA-MRL-100, Implementing Maintenance Rule, to consistently apply requirements. One requirements of ER-AA-MRL-100 is monitoring equipment unavailability or out-of-service time on a rolling 12 month average. The allowed unavailability time for each charging pump is 438 hours. Contrary to the established goals, on May 5, 2016, the licensee identified Unit 1 C charging pump, 1-CH-P-1C, has accrued 517.3 hours of unavailability and Unit 2 C charging pump, 2-CH-P-1C, has accrued 588.0 hours of unavailability. By exceeding the established unavailability hour goals, licensee failed to control the condition of 1-CH-P-1C and 2-CH-P-1C to ensure the component remains capable of performing its intended function. Subsequently, both charging pumps are being evaluated under 10 CFR Part 50.65 (a)(1) for failure to meet established unavailability goals. The licensees failure to control the condition of 1-CH-P-1C and 2-CH-P-1C to ensure the components remained capable of performing their intended function was a PD. The PD was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed the significance determination for the finding using NRC Inspection Manual Chapter 0609, Appendix A, Attachment 2, dated July 1, 2012 and determined the risk significance was very low (GREEN), because the charging system maintained operability in accordance with TS. The licensee entered this condition into their CAP as CR1036685 and CR1036687. |
| 05000327/FIN-2016007-01 | 2016Q2 | Sequoyah | 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 | 2016Q2 | Sequoyah | 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 | 2016Q2 | Sequoyah | 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 | 2016Q2 | Sequoyah | 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. |
| 05000335/FIN-2016008-05 | 2016Q1 | Saint Lucie | Failure to Verify the Adequacy of Design of Unit 1 Electrical Equipment Room Ventilation System | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the Unit 1 electrical equipment room (EER) ventilation system design when performing a design calculation. The licensee entered the issue into the corrective action program and plans to re-balance flow rates in the EERs or revise the equipment qualification temperatures for equipment located in the EERs. 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 re-analysis of the ventilation system resulted in a reduction in temperature margin, which could impact the reliability and capability of emergency electrical equipment in the EERs. 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. |
| 05000335/FIN-2016008-01 | 2016Q1 | Saint Lucie | Failure to Consider Elevated Temperature Effects on MOV Actuator Output Capability | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to consider the impact of elevated ambient temperatures on motor operated valve (MOV) actuator output. The licensee entered the issue into the corrective action program and also evaluated the elevated ambient temperature effects on several affected station MOVs and determined the MOVs remained operable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control 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 did not ensure the capability of several MOVs scoped into their MOV program because they did not consider reduced actuator output torque due to elevated temperatures. 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 assigned a cross-cutting aspect of Evaluation in the Problem Identification and Resolution Area because the finding was indicative of current licensee performance, and the licensee did not thoroughly evaluate the issue identified in AR 2030822, such that the design issue of accounting for elevated temperature was resolved (P.2). |
| 05000335/FIN-2016008-02 | 2016Q1 | Saint Lucie | Failure to Update UFSAR to Reflect Station Blackout Coping Time Basis | The NRC identified a non-cited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, for the licensees failure to update the Updated Final Safety Analysis Report (UFSAR) to reflect the offsite power design characteristic group and emergency alternating current power configuration group for station blackout coping duration. The licensee entered the issue into the corrective action program in order to update the information. The failure to update the UFSAR was dispositioned using the traditional enforcement process because it had the potential to impact the regulatory process. The team determined the violation was more than minor because not accurately classifying the offsite power design characteristic group and emergency alternating current power design characteristic group could have a material impact on licensed activities. The team determined the violation to be a Severity Level IV violation because the lack of upto- date information has not resulted in any unacceptable change to the facility or procedures. This violation was not assigned a cross-cutting aspect because crosscutting aspects are not assigned to traditional enforcement violations. |