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05000395/FIN-2018003-01Licensee-Identified Violation2018Q3This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. V.C. Summer Operating License condition 2.c(18) states in part that the licensee 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 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to the above, on January 10, 2017, the licensee failed to implement an established procedure, FPP-025, Fire Containment, Revision 6D, to ensure fire door DRAB/319 closes and latches on its own power.
05000395/FIN-2018003-02Minor Violation2018Q3The inspectors concluded that rotation of the safety-related SW pipe support, SWH-4021, was a condition adverse to quality (CAQ) identified in CR-04-01705. The inspectors also concluded that the failure to correct this CAQ was a minor violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, which states in part that CAQs are promptly identified and corrected. Screening: The inspectors determined that the degradation of SWH-4021 was minor based on the absence of any deformed components on the pipe support, and that the respective train of the SW system remained overall operable. The licensee corrected the CAQ during the quarter using WO 1813577, Return SWH-4021 to design requirements. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000395/FIN-2018411-01Security2018Q2
05000395/FIN-2018001-01Failure to Perform an Adequate Risk Assessment With Consequent Reactor Trip2018Q1A self-revealed, Green NCV was identified for the licensees failure to adequately assess risk in accordance with 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, involving repairs to a non-safety related inverter, XIT-5905. This NCV closes LER 05000395/2017-005-00: Automatic Reactor Trip Due to Main Turbine Trip.
05000395/FIN-2018010-06Potential Unjustified Activation Energy for Barton Transmitters2018Q1The 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-2018010-05Potential High Radiation Dose Areas with Unqualified Components2018Q1The 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-2018010-04Unjustified Qualified Life for ASCO Valves2018Q1The 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-2018010-03Inadequate Radiation Harsh Environmental Qualification of Reactor Building Spray Pump A2018Q1During 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-2018010-02Failure to Verify the Seismic Qualification of Valcor Solenoid Operated Valve XVX06050A2018Q1Calculation 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-2018010-01Failure to Justify Activation Energy for Valcor SOV XVX06050A2018Q1The 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.
05000395/FIN-2017004-04Licensee-Identified Violation2017Q4TS 6.8.1, Procedures and Programs, requires, in part, that written procedures shall be implemented covering the activities recommended in Appendix A of Regulatory Guide 1.33, Rev. 2, Section 8, Procedures for Control of Measuring and Test Equipment and for Surveillance Tests, Procedures and Calibrations. Contrary to this, on October 23, 2017, the licensee identified they had failed to correctly implement STP-220.002, Turbine Driven Emergency Feedwater Pump and Valve Test, Rev. 9, to return the TDEFW pump governor speed control manual adjustment knob to the required position during the surveillance test conducted on the previous shift. The inspectors reviewed IMC 0609 Appendix A and Attachment 4 for Mitigating Systems to determine the finding was of very low safety significance, Green, because there was no design deficiency, loss of system, and the loss of function for the single train was less than the TS LCO action time and less than 24 hours. The licensee has documented this problem in their CAP as CR-17-05588.
05000395/FIN-2017007-01Failure to Verify the Adequacy of Design for the EFW system when Supplied by SW2017Q4The 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
05000395/FIN-2017007-02Failure to Establish a Testing Program for Inverter XIT5904 and Time Delay Relay in the EFW/SW Crosstie Valve Actuation Circuitry2017Q4The 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-2017007-03Failure to Identify a CAQ for Power Shield Catalog #609903-T501N2017Q4The 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-2017007-04Licensee-Identified Violation2017Q4Title 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
05000395/FIN-2017004-03Failure of an Emergency Feedwater Auto Start Actuation Signal2017Q4A self-revealing, Green, non-cited violation (NCV) of Technical Specification (TS) 3.3.2 was identified involving the failure of the C main feedwater pump to trip and resultant loss of an emergency feedwater auto start actuation signal. The licensee entered the issue in their corrective action program as condition report, CR-17-01611. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined that the PD was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone by adversely affecting the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the equipment reliability attribute was impacted because a failure of the C main feedwater pump to trip when required rendered an EFW auto start actuation signal inoperable. The inspectors used IMC 0609, Significant Determination Process, Attachment 4, and Appendix A Exhibit 2, and determined that the finding was of very low safety significance, Green, because there was no design deficiency or loss of function. Specifically, EFW auto start capability remained operable for other functions to maintain short term heat removal capability. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, and determined the cause of this finding involved the cross-cutting area of Human Performance and the aspect of problem identification and resolution, P.2, because the licensee had previous indications of water intrusion and feedwater pump control issues and failed to thoroughly evaluate to address the cause.
05000395/FIN-2017004-02Failure to Implement Corrective Actions to Restore Compliance for Previous NRC-identified Green NCV 05000395/2005007-012017Q4The inspectors identified a Green finding associated with a cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to ensure that conditions adverse to quality as noted in a previous NRC-identified Green NCV, 05000395/2005007-01, EFW Flow Control Valves Are Susceptible to Plugging by Tubercles or Other Debris from Service Water, were corrected. The licensee entered the issue in their corrective action program as condition report, CR-17-04630. The inspectors determined that the failure to promptly identify and correct the conditions adverse to quality (CAQ) for a design in which the emergency feedwater (EFW) flow control valves were susceptible to plugging by tubercles or other debris from the service water (SW) system was a performance deficiency (PD). The inspectors reviewed IMC 0612, Appendix B and determined the PD was more than minor and therefore a finding, because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and the respective attribute of design control because the EFW flow control valves were susceptible to plugging by SW debris. This finding had been evaluated and screened to a low safety significance (Green) and documented in the previous NRCidentified Green NCV, 05000395/2005007-01. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.3 of the NRC Enforcement Policy. The inspectors used IMC 0310 and determined this finding has a cross-cutting aspect of resolution in the area of Problem Identification and Resolution because the organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance and restore compliance (P.3).
05000395/FIN-2017004-01Failure to Accomplish Station Procedures for Severe Weather2017Q4The NRC identified a Green finding (FIN) for the failure of the licensee to accomplish operations administrative procedure, OAP-109.1, Guidelines for Severe Weather, Rev. 4H, for adequate control of sandbags used for ground level plant building access door protection during a permissible maximum precipitation (PMP) or other adverse rainfall events. The licensee entered the issue in their corrective action program as condition reports, CR-17-05632 and CR-17-05783.The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the performance deficiency (PD) was more than minor and therefore a finding because the PD affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and the respective attribute of protection against external factors (i.e., flooding). Specifically, without the sandbag container sealed, the licensee would not be able to expedite the sandbags into the protected area (PA), and degradation of the sandbags would have prevented use as specified. The inspectors reviewed IMC 0609, Attachment 4, Appendix A, and Exhibit 4, for the significance determination, and determined the finding was of very low safety significance, or Green, because the finding does not involve the total loss of any safety function, identified by the licensee through a probabilistic risk analysis (PRA), individual plant examination of external events (IPEEE), or similar analysis, that contributes to external event initiated core damage accident sequences (i.e., flooding). Specifically, the time afforded the licensee via weather forecasting would have allowed other measures to mitigate ingress of flood waters into plant areas.The inspectors reviewed IMC 0310, Aspects Within Cross-Cutting Areas, and determined the cause of the finding involved the area of human performance and the aspect of H.1:Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety, because the licensee did not ensure that resources were available to check the security seals on the containers and the licensee did not ensure the sandbags were capable of meeting its intended function.
05000395/FIN-2017003-03Failure to Adequately Assess the Risk for an Activity with Consequent Loss of Core Cooling2017Q3The inspectors identified a Green, NCV of 10 CFR 50.65(a)( 4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, involving the licensees failure to perform an adequate risk assessment for an activity involving restoration of the B train emergency bus to the normal supply and a subsequent loss of the B train residual heat removal (RHR) pump and a consequent loss of core cooling. The issue was entered into the licensees CAP as condition report, CR- 17- 03696 The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it was associated with the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform an adequate risk assessment resulted in performance of an activity causing a vulnerability to operability of the running RHR pump and a consequent loss of core cooling. The finding was screened for risk significance using NRC IMC 0609.04 and routed to NRC IMC 0609, Appendix G. A detailed shutdown risk assessment was performed by a regional senior risk analyst using NRC IMC 0609 Appendix G and Attachments 1 and 2. The major analysis assumptions included: treatment of the PD as a loss of RHR with an initiating event likelihood of 1.0 using NRC IMC 0609 Appendix G, Attachment 2, worksheet 9 (loss of RHR in plant operating state 2); and recovery credit was applied for closing the alternate feeder breaker. The dominant sequence was a loss of RHR, failure to recover decay heat removal prior to reactor coolant system (RCS) boiling, failure to initiate RCS injection before core damage and failure to restore power to the B train safety bus. The RCS conditions of time to boil and time to core uncovery and availability of mitigating equipment limited the risk. The detailed risk evaluation determined that the PD represented an increase in core damage frequency <1.0E -6 a GREEN finding of very low safety significance. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined that this finding had a cross -cutting aspect in the area of Work Management (H.5), because the licensee did not perform an adequate risk assessment in accordance with their procedure.
05000395/FIN-2017003-02Violation of NRC Examination Security as Required by 10 CFR 55.492017Q3The inspectors identified a Green, NCV of 10 CFR 55.49, Integrity of examinations and tests, for the licensees failure to ensure the paper strip chart recorders were advanced following the completion of Job Performance Measures (JPMs) and simulator scenarios to prevent examination compromise when the same examination was administered to multiple crews on the same day. While observing simulator JPMs, the inspectors identified that simulator staff was not advancing any of the strip chart recorders after each JPM. This left examination material out and visible to the next operator performing the JPM. The inspectors informed the licensee of this issue and they immediately started advancing each chart recorder until no exam 3 related material was visible. An initial review by the licensee indicated that not advancing the paper had been a standard practice for this exam period. In accordance with the NRC Enforcement Policy, this violation was classified as Severity Level IV Violation (Section 6.4.d). This Severity Level IV violation is being treated as a Non -Cited Violation, consistent with Section 2.3.2.a of the Enforcement Policy. This violation is in the licensees corrective action program under CR 17 -04424
05000395/FIN-2017003-01Inadequate Procedures for Inspection of Fire Barriers2017Q3The inspectors identified a Green, NCV of Operating License Condition 2.C.(18), Fire Protection Program, for failure to adequately establish a surveillance procedure for fire penetration inspections. The licensee entered the problem into their corrective action program (CAP) as condition report, CR -17-04029. The inspectors used IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the performance deficiency ( PD ) was more than minor and therefore a finding because it was associated with the Mitigating System Cornerstone and the respective attribute of protection against external factors (fire). This finding had a credible impact on safety and adversely affected the cornerstone objective because STP -728.043 failed to specify adequate inspection of penetrations added by modification of which one penetration (5045) was degraded. The inspectors used IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and determined that the finding impacted the fire confinement finding category. Based on review of respective Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, dated February 28, 2005, the inspectors determined that the degradation rating was low based on the size of the degradation identified. Specifically, the separation of foam seal material was approximately 3 from the mating conduit surfaces and there was greater than 12 of foam seal material when it was initially installed. As a result, this finding was determined to be of very low safety significance (Green) based on the guidance in IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013
05000395/FIN-2017002-02Failure to Provide NRC Staff Complete and Accurate Information2017Q2The inspectors identified a severity level (SL) IV NCV of 10 CFR 50.9(a), Completeness and accuracy of information, involving licensee document,RC-13-0142, dated October 14, 2013. This document was a response to a request for additional information involving a license amendment request (LAR) to adopt NFPA 805 and contained an approval request, L12, associated with oil misting from the reactor coolant pumps. The licensee entered this violation into their corrective action program as CR-17-03961. The inspectors determined that the licensees failure to provide complete and accurate information associated with approval request, L12, was a violation of 10 CFR 50.9(a). Because this violation of 10 CFR 50.9(a) impacted the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using traditional enforcement (TE). Since the TE violation is associated with a previous Green reactor oversight process violation, and the misinformation was identified after the NRC relied on it for issuing a previous operating license amendment, the TE violation was determined to be a SL IV, NCV, consistent with the language of the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information. This violation involved TE; therefore a cross-cutting aspect was not assigned.
05000395/FIN-2017002-01Failure to Implement Corrective Actions to Restore Compliance for Previous NRC-identified Green NCV 05000395/2013003-032017Q2The inspectors identified a Green finding with a cited violation of Operating Licensee Condition 2.C.(18) for failure to ensure that conditions adverse to fire protection as noted in a previous NRC-identified Green NCV, 05000395/2013003-03, Failure to Adequately Design, Install and Maintain Oil Collection Devices for Reactor Coolant Pump Motors, were corrected. Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for (1) a failure to ensure an adequate design for the oil lift pump enclosure, and (2) a failure to have oil collection components for internally leaked oil dripping from the motor air discharge ductwork flange. The licensee entered the issue in their corrective action program as condition report CR-17-03962.The inspectors determined that the failure to implement corrective actions for the oil collection system to restore compliance was a performance deficiency (PD). The inspectors used IMC 0612 and determined that the PD was more than minor and therefore a finding because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. This finding has a credible impact on safety because the failure to adequately install, maintain and design the oil collection system presented a degradation of a fire confinement component which has a fire prevention function of not allowing an oil leak to reach hot surfaces. This finding had been evaluated and screened to a low safety significance (Green) and documented in the previous NRC-identified Green NCV, 05000395/2013003-03. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.3 of the NRC Enforcement Policy. The inspectors used IMC 0310 and determined this finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance and restore compliance (P.3).
05000395/FIN-2017201-01Security2017Q1
05000395/FIN-2017001-01Licensee-Identified Violation2017Q1TS 3/4.5.2, ECCS Subsystems T avg > 350 o F states in part that two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with one OPERABLE centrifugal charging pump in Modes 1, 2 and 3. TS 1.18, OPERABLE OPERABILITY, definition states in part that a subsys tem shall be OPERABLE when all necessary auxiliary equipment that are required for the subsystem are capable of performing their related support functions. Contrary to this, from June 20, 2015, through July 20, 2016, safety -related subsystem chiller, XHX0001A, an auxiliary component supporting the A train charging pump (XPP0043A) was incapable of performing its safety function resulting in the inoperability of XPP0043A for greater than the allowed action times of TS 3/4.5.2. A review of IMC 0609, Appendix A, determined the finding was of very low safety significance (Green) 14 because the finding was not a design deficiency and it did not result in a loss of function. The licensee has documented this problem in their CAP as CR -15- 04395
05000395/FIN-2017201-02Security2017Q1
05000395/FIN-2017001-02Licensee-Identified Violation2017Q1V.C. Summer Operating License condition 2.c(18) states in part that the licensee 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 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on January 10, 2017, the licensee failed to establish procedures, SAP -131A, Fire Protection Program Surveillances and Compensatory Measures, Rev 3., FPP - 015, Shift Inspection, Rev. 7, and FPP -025, Fire Containment, Rev. 6A, to ensure the fire doors listed in TR07800- 020, NFPA 805 Monitoring Program Phase 1: Scoping, Rev. 0, were appropriately identified for adequate licensee actions concerning surveillances and degraded conditions. NRC used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, Attachment 1, dated September 20, 2013, to perform a Phase 1 analysis and determined that the finding was of very low safety significance (Green) based on the response for Question 1.3.1A, in which the reactor was able to reach and maintain safe shutdown. The licensee has documented this problem i n their CAP as CR- 17-00150.
05000395/FIN-2016007-02Failure to Correct a Condition Adverse to Quality Associated with a Previously Issued NCV2016Q4The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct a condition adverse to quality associated with a previously issued NCV, 05000395/2012004-02, Inadequate Installation of Unit 1 Service Water Piping and Related Pipe Support. The licensee entered the issue in the correction action program as condition report (CR)-16-04621. The PD is more than minor because if left uncorrected, the reduction in design margin of the pipe support could affect the Unit 1 SW systems ability to mitigate a seismic event. Specifically, Unit 1 service water (SW) piping and support had been impacted by the reduction in design margin and without formally updating the associated drawings and calculations or restoring to the original design, future modifications to the system could impact the systems ability to mitigate a seismic event. Using Manual Chapter 0609 Attachment 04, Initial Characterization of Findings, Table 2, dated October 07, 2016, the finding was determined to adversely affect the External Event Mitigating Systems. The inspectors screened the finding using Inspection Manual Chapter (IMC) 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green (very low safety significance) because the service water system maintained its functionality to mitigate a seismic event. The inspectors determined that the finding had a cross-cutting aspect in the area of PI&R because the licensee did not take effective corrective actions to address this issue in a timely manner (P.3).
05000395/FIN-2016010-03Failure to ensure credited equipment to support the 50 54hh requirements were adequate2016Q4The NRC identified a Green, non-cited violation (NCV) of the 10 CFR 50.54(hh)(2) requirements. Specifically, the team identified aspects of the implementation strategy that were inconsistent to support the stated commitments. The licensees failure to ensure that credited components needed to implement the strategy were adequate for circumstances consistent with the stated commitments was a performance deficiency (PD). This PD was determined to be more than minor because of the adverse impact to the Mitigating Systems cornerstone objective. Specifically, the PD had the ability of impacting the availability and reliability of the credited strategy in response to conditions postulated to meet the 10 CFR 50.54hh requirements. The team screened the issue as Green using IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that further screening was necessary consistent with IMC 0612, Appendix L, B.5.b Significance Determination Process. Dated December 24, 2009. In this instance, the finding was determined to be of Green significance since no additional strategies were impacted. The licensee initiated CR-16-05266 to address the NRC concerns.
05000395/FIN-2016004-01Failure to Establish Procedures for Corrective Actions to Address Conditions Adverse to Fire Protection2016Q4Green. The inspectors identified a Green, non-cited violation (NCV) of the V.C. Summer Nuclear Station Operating License, Condition 2.C (18), Fire Protection Program, for the failure to establish procedures requiring corrective action for conditions, including significant and repetitive, adverse to fire protection. The licensee immediately notified the corrective action program (CAP) supervisor and entered the problem into their CAP as condition report CR-16-05270. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the performance deficiency (PD) was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. Specifically, the failure to establish corrective action program requirements specific to fire protection with appropriate definitions for significant and repetitive would result in corrective actions not commensurate with the significance of the adverse condition. The inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, Attachment 1, dated September 20, 2013, to perform a Phase 1 analysis and determined that the reactor oversight process (ROP) finding was of very low safety significance (Green) based on the response for Question 1.3.1A, in which the reactor was able to reach and maintain safe shutdown. While the licensee does not have the required corrective actions defined, they have generally addressed conditions adverse to fire protection within the existing corrective action program. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of human performance and the aspect of resources, H.1, because the licensee leadership failed to ensure that adequate procedures were in place to address significant and repetitive conditions adverse to fire protection.
05000395/FIN-2016004-05Licensee-Identified Violation2016Q4V.C. Summer Operating License condition 2.c(18) states in part that the licensee 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 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on September 14, 2016, the licensee failed to implement the requirements of procedure, Fire Protection Procedure, FPP-025, Fire Containment, Rev. 4, to ensure that fire door and SPB, DRIB/107, did not remain open in excess of 1 hour. NRC IMC 0609.04 and NRC IMC 0609 Appendix A screening determined that the finding represented a loss of the short term heat removal safety function within the Mitigating Systems Cornerstone and required a detailed risk evaluation. A bounding analysis was performed by a regional SRA using the VC Summer SPAR model. The finding was modelled as a Steam Line Break Outside Containment (SLBOC) initiating event assessment. A 3 hours and 22 minutes bounding exposure was utilized. No recovery was assumed. Equipment impacts from potential HELBs were determined using the results from Gothic Analyses performed to assess the temperature, pressure and relative humidity increases in mild environment spaces in the Intermediate Building due to the various HELB boundary breaches associated with the finding. For non-Loss of Offsite Power (LOOP) conditions, HELB impacts in mild areas were minimal. For LOOP conditions, the HELB was assumed to impact the chilled water system such that one train of safety related equipment was assumed failed as a bounding impact. The dominant sequence was a SLBOC initiator with a failure to isolate the break and a failure of high pressure injection impacted by loss of chilled water leading to loss of core heat removal. The risk evaluation determined that the finding represented a risk increase of < 1.0E-6/year, a GREEN finding of very low safety significance. The licensee has documented this problem in their CAP as CR-16-04703
05000395/FIN-2016004-02Failure to Promptly Identify and Correct a Condition Adverse to Quality for B Emergency Diesel Generator Exhaust Manifold Weld Indications2016Q4Green. A self-revealing, Green NCV of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the failure to promptly identify and correct a condition adverse to quality (CAQ) involving welded joint indications in the B emergency diesel generator (EDG) exhaust manifold. The licensee immediately removed the EDG from service to perform repairs, and the issue was entered into the licensees CAP as Condition Report CR-16-05421. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor and therefore a finding, because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and the respective attribute of equipment performance. Specifically, the B EDG was declared operable but degraded or nonconforming due to a circumferential weld failure and resulting separation of an exhaust manifold joint causing a small reduction in EDG power. The inspectors used IMC 0609, Significant Determination Process, Attachment 4, dated October 7, 2016, and Appendix A Exhibit 2, dated July 1, 2012, and determined the finding was of very low safety significance or Green because the finding was not a design deficiency or loss of function. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of problem identification and resolution and the aspect of resolution, P.3, because the licensee failed to take effective corrective actions commensurate with an issues safety significance in that they failed to promptly identify and correct a CAQ involving welded joint indications in the B EDG exhaust manifold.
05000395/FIN-2016010-06Licensee-Identified Violation2016Q4The licensee identified an example of a failure to meet the 2.C.18 commitments. The licensee was required to implement applicable aspects of the NFPA 805 requirements in order to achieve the risk reductions specified in RG 1.174, An Approach for Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. To accomplish this, the licensee committed to various changes to the facility. In this instance, the licensee committed to ensure the PRA developed to meet the 2.C.18 requirements was completed consistent with Table S-2, Implementation Item 22 and SER Section 2.7.2. Contrary to the above, the licensee failed to implement the stated requirement specified by license condition 2.C.18 and docketed correspondence from the licensee to the NRC. Specifically, the licensee deviated from the stated commitment without NRC approval by which formed the basis for the team to evaluate the finding using traditional enforcement (TE) based upon the guidance in NRC Enforcement Policy. The team reviewed the NRC Enforcement Guidance, Part II, Section 2.2, Actions Involving Fire Protection, to aid assessing the significance of the issue and determined the issue to be a SL IV. A cross cutting aspect was not assigned based upon the TE determination. Based upon the identification by the licensee, the issues have been entered into the licensees corrective action program as CR-16-00321, CR-16-01132, CR-16-01602, CR- 16-04828, and CR-16-04829.
05000395/FIN-2016010-01Failure to Meet the Quality Requirements Specified By NFPA 8052016Q4The NRC identified a SL IV, non-cited violation (NCV) of the 10 CFR 50.48(c), National Fire Protection Association Standard (NFPA) 805, requirements. Specifically, the team identified the licensees inability to ensure licensing basis information was maintained consistent with administrative procedures to support the NFPA 805 Section 2.2(j) and NFPA 805, Section 2.7 requirements. The licensees failure to meet the quality requirements specified by NFPA 805 Section 2.2. (j) and NFPA 805, Section 2.7, Program Documentation, Configuration Control and Quality was a performance deficiency (PD). This PD was determined to be more than minor because it affected the regulatory process. In this instance, the licensee failed to ensure information to support the NFPA 805 licensing commitments was controlled in the manner specified by the requirement. This information served as the basis for the NRC to perform its regulatory function and had the ability to impact the credited analysis relied upon to reach and maintain safe and stable conditions in case of a fire. As a result, the team evaluated the finding using the traditional enforcement (TE) process based upon the guidance in NRC Enforcement Policy and NRC Enforcement Guidance. The team reviewed the NRC Enforcement Guidance, Part II, Section 2.2, Actions Involving Fire Protection, to aid assessing the significance of the issue and determined the issue to be a SL IV. A cross cutting aspect was not assigned based upon the TE determination. The licensee initiated CR-16-05060, CR-16-05074, CR-16-05160, CR-16-05276, and CR-16-05278 to address the NRC concerns.
05000395/FIN-2016007-01Failure to Implement Corrective Actions and Restore Compliance for Previous NRCIdentified SLIV NCV2016Q4The inspectors identified a cited Severity Level (SL) IV violation of Operating Licensee Condition 2.C.(18) for failure to ensure that conditions adverse to fire protection as noted in a previous NRC-identified SLIV NCV, 05000395/2016001-01, Failure to Implement Adequate Administrative Controls Following a Departure from National Fire Protection Association (NFPA) 80-1973 and Provide NRC Staff Complete and Accurate Information, were promptly corrected. Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for (1) the noncompliance with 10 CFR 50.9 to provide staff complete and accurate information and (2) fire doors DRIB/105A&B currently do not meet self-closing requirements in accordance with the current NFPA 805 licensing basis and no actions were specified in licensees corrective action program to restore compliance. The licensee entered the issue in their corrective action program as condition report (CR)-16-04701. The inspectors determined that the performance deficiency was more than minor because it impacted the ability of the NRC to perform its regulatory oversight function and was dispositioned using traditional enforcement. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.2. a of the NRC Enforcement Policy. This violation involved traditional enforcement and a cross-cutting aspect was not assigned to this violation.
05000395/FIN-2016004-04Failure to Update FSAR with a New Design Function for the Equipment and Floor Drain System2016Q4SL IV. The NRC identified a severity level IV (SL IV) non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.71(e) for the licensees failure to update the final safety analysis report (FSAR) with the latest information developed, regarding the design functions of the equipment and floor drain system. Specifically, the licensee failed to update the FSAR to reflect the high-energy line break (HELB) steam propagation barrier (SPB) function of the floor drain system following installation of new floor drain orifices used as SPBs. The licensee entered this issue into their corrective action program as CR-16-06003. The inspectors treated the noncompliance with 10 CFR 50.71(e) as traditional enforcement because not having an updated FSAR hinders the licensees ability to perform adequate 10 CFR 50.59 evaluations and impacts the NRCs ability to perform its regulatory function such as license amendment reviews and inspections. This was determined to be a SL-IV violation of 10 CFR 50.71(e) because it was similar to the NRC Enforcement Policy, Section 6.1.d.3, SL IV example of, a licensee fails to update the FSAR as required by 10 CFR 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.
05000395/FIN-2016010-04Failure to seek or gain approval for risk-informed changes constituted a self-approved change which is inconsistent with the NFPA 805 requirements2016Q4The NRC identified a SL IV, non-cited violation (NCV) of the 10 CFR 50.48(c), NFPA 805, requirements. Specifically, a Risk Informed Change was made that was inconsistent with Transition License Condition 2.C.18.(c).1 which stated in part: Before achieving full compliance with 10 CFR 50.48(c), ...risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval. In this instance, the team identified the licensee failed to seek or gain NRC approval for riskinformed changes that had a more than minimal risk impact to the fire protection program during the post-safety evaluation issuance period date of February 11, 2015. The licensees failure to obtain NRC approval prior to making any changes to the 2.C.18 license requirements was a performance deficiency (PD). This PD was determined to be more than minor because it impacted the regulatory process. Specifically, the team determined that risk-informed changes made to a commitment specified by license condition 2.C.18.(c).1, which was based upon docketed correspondence from the licensee, required NRC approval. The licensee deviated from the stated commitments without NRC approval which formed the basis for the team decision to evaluate the finding using traditional enforcement (TE) based upon the guidance in NRC Enforcement Policy. The team reviewed NRC Enforcement Guidance, Part II, Section 2.2, Actions Involving Fire Protection, to assess the significance of the issue and determined the issue to be a SL IV. The licensee initiated CR-16-01490 and CR-16-05291.
05000395/FIN-2016004-07Licensee-Identified Violation2016Q4TS 6.8.1, Procedures and Programs, requires in part that written procedures be maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33, Rev. 2, Section 9, Procedures for Performing Maintenance. Contrary to this, on October 4, 2016, the licensee determined they had failed to establish station procedures to adequately remove ventilation access covers for fire damper inspections during plant modes that would not present a HELB challenge to those components qualified for mild EQ conditions. NRC IMC 0609.04 and NRC IMC 0609 Appendix A screening determined that the finding represented a loss of the short term heat removal safety function within the Mitigating Systems Cornerstone and required a detailed risk evaluation. A bounding analysis was performed by a regional SRA with the following major analysis assumptions: a fifty one minute exposure period, a HELB frequency of 7.7E-3/year from the NRC VC Summer SPAR model, and a conditional core damage probability given a HELB of 1.0. The risk increase due to the finding was <1.0E-6/year, a GREEN finding of very low safety significance. The risk was mitigated by the short exposure. The licensee has documented this problem in their CAP as CR-16-05696.
05000395/FIN-2016010-05Failure to meet corrective action requirements consistent with NFPA 805 Section 2.6.32016Q4The NRC identified a Green non-cited violation (NCV) of the V.C Summer Nuclear Station, Unit 1, Renewed Facility Operating License (FOL) Condition 2.C.18 requiring the licensee to implement and maintain in effect all provisions of the approved FPP that complied with 10 CFR 50.48 (c), National Fire Protection Association Standard NFPA 805. The NRC safety evaluation report (SER) dated February 11, 2015, relied upon an adequate corrective action program to implement the NFPA 805 requirements. NFPA 805 Section 2.6.3, Corrective actions. Specifically, the team identified a failure to adequately classify and correct conditions adverse to quality (CAQ) in a timely manner. The licensees failure to properly assign an action level commensurate to ensure corrective actions were addressed consistent with the NFPA 805, Section 2.6.3 was a PD. The PD was more than minor because, if left uncorrected, it could lead to more significant safety concern. Specifically, the inadequate application of the corrective action program can lead to deficiencies degrading SSCs which can adversely impact the FPP requirements and lead to a more significant safety concern. The finding was screened in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and IMC 0609, Attachment 4, Characterization of Findings dated October 7, 2016. A determination was made using IMC 0609, Significance Determination Process, dated April 29, 2015. Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012 was applicable since the administrative controls in this instance were not associated with transient or hot work activities. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance (Green) because it did not represent an actual loss of safety function. The team assessed the finding against the IMC 0310, Cross-cutting Aspects, dated December 4, 2014, requirements and determined that cross-cutting was applicable. In this instance, the cause of this finding was determined by the team to have a cross-cutting aspect of the Resolution component (P.3) of the Problem Identification and Resolution (PI&R) area. This was selected based upon the inability organization to adequately identify and take effective corrective actions to address issues in a timely manner commensurate administrative procedures to meet the NFPA 805, Section 2.6.3 requirements. The licensee initiated CR-16-05306 and CR-16-05160, Action 1 related to this issue.
05000395/FIN-2016004-03Failure to Accomplish Procedure for Foreign Material Exclusion Control Involving Failure of a Safety-Related Breaker2016Q4Green. The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to accomplish safety-related (SR) station administrative procedure, SAP-0363, Foreign Material and Debris Control, Revision 8H, for foreign material exclusion (FME) control during a SR breaker refurbishment. A subsequent breaker failure occurred due to foreign material. The licensee immediately initiated corrective actions to repair the breaker, and the licensee entered condition report, CR-16-03099, in their CAP. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone by adversely affecting the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the equipment reliability attribute was impacted because foreign material rendered the SR breaker nonfunctional causing inoperability of the pressurizer backup group 2 heaters for greater than the Technical Specification limiting condition for operation. The inspectors used IMC 0609, Significant Determination Process, Attachment 4, dated October 7, 2016, and Appendix A Exhibit 2, dated July 1, 2012, and determined that the finding required a detailed risk evaluation. A regional senior risk analyst performed a bounding risk evaluation in accordance with NRC IMC 0609 Appendix A using the VC Summer SPAR model. The finding was modelled as a transient initiator with a loss of the B EDG as a surrogate for the group 2 pressurizer heaters for a 94 hour exposure interval. The dominant sequence was a transient initiator with a consequential loss of offsite power without recovery, failure of the A EDG without recovery leading to a station blackout and loss of core heat removal after battery depletion. The risk was mitigated by the available normal and group 1 pressurizer heaters. The bounding assessment determined that the performance deficiency represented an increase in core damage frequency of < 1.0 E-6/year, a GREEN finding of very low safety significance. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of Problem Identification and Resolution and the aspect of work management, H.5, because the licensee failed to ensure the planning and execution of the respective work order for breaker refurbishment followed SAP-0363 for FME control to support nuclear safety-related work.
05000395/FIN-2016010-02Failure to identify unsealed cabinet Specified by NFPA 805 Section 3.3.1.2(1)2016Q4The NRC identified a Green, non-cited violation (NCV) for the failure to include potentially high-risk fire scenarios in the current fire protection program. In this instance, the team identified unsealed electrical cabinets credited as being sealed. The licensees failure to identified and assess the applicable electrical cabinet as a firescenario in its FSA database was a performance deficiency (PD). This PD was determined to be more than minor because of the adverse impact to the Mitigating Systems cornerstone objective. Specifically, the PD resulted in an incomplete fire risk model. The licensee performed an analysis of the performance deficiency using their fire probabilistic model and the results were that the PD represented a risk increase of <1.0E-6/year in core damage frequency and <1.0E-7/year in large early release fraction. The licensees results were reviewed by a regional senior reactor analyst (SRA). Additionally, a bounding analysis was performed by the regional SRA in accordance with NRC IMC 0609 Appendix F which concluded that the core damage frequency risk increase due to the PD was <1.0 E-6/year, a GREEN finding of very low safety significance. The team assessed the issue consistent with IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the finding to have a cross-cutting aspect of Field Presence (H.2) in the Human Performance area because the licensee did not ensure that senior managers and supervisory staff maintained the proper amount of oversight of contractors and supplemental personnel in the performance work activities relevant to fire protection program implementation. The licensee has fire watches in place as a compensatory measure and has entered this issue into their corrective action program as CR-16-05287.
05000395/FIN-2016004-06Licensee-Identified Violation2016Q410 CFR 50, Appendix B, Criterion V, requires in part that activities affecting quality shall be accomplished by documented instructions of a type appropriate to the circumstances. Contrary to this, on August 4, 2009, the licensee failed to accomplish documented work instructions contained in ECR50585 to install floor drain orifices to act as SPBs to protect SR components from the effect of a HELB. NRC IMC 0609.04 and NRC IMC 0609 Appendix A screening determined that the finding represented a loss of the short term heat removal safety function within the Mitigating Systems Cornerstone and required a detailed risk evaluation. A bounding analysis was performed by a regional SRA using the VC Summer SPAR model. The finding was modelled as a Steam Line Break Outside Containment (SLBOC) initiating event assessment. A forty five day bounding exposure was utilized. No recovery was assumed. Equipment impacts from potential HELBs were determined using the results from Gothic Analyses performed to assess the temperature, pressure and relative humidity increases in mild environment spaces in the Intermediate Building due to the various HELB boundary breaches associated with the finding. For non-Loss of Offsite Power (LOOP) conditions, HELB impacts in mild areas were minimal. For LOOP conditions, the HELB was assumed to impact the chilled water system such that one train of safety related equipment was assumed failed as a bounding impact. The dominant sequence was a SLBOC initiator with a failure to isolate the break and a failure of high pressure injection impacted by loss of chilled water leading to loss of core heat removal. The risk evaluation determined that the finding represented a risk increase of < 1.0E-6/year, a GREEN finding of very low safety significance. The licensee has documented this problem in their CAP as CR-16-04716.
05000395/FIN-2016003-02Failure to Prescribe Work Instructions for a Temporary Repair on a Safety-Related Component2016Q3The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to prescribe instructions for a temporary repair of the safety-related C component cooling water (CCW) pump outboard bearing. The licensee entered condition report, CR-16-04576, in their corrective action program for appropriate response. The inspectors determined that the failure to prescribe documented work instructions of a type appropriate to the circumstances for the temporary repair of the C CCW pump outboard bearing was a performance deficiency (PD). The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it impacted the Mitigating Systems Cornerstone by adversely affecting the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design control attribute was impacted because not prescribing instructions that follow vendor instructions for temporary repairs on the safety-related pump resulted in improper repairs causing reasonable doubt in operability. The inspectors evaluated the finding in accordance with IMC 0609, Significant Determination Process, Attachment 4 and Appendix A, and determined that the finding was of very low safety significance, Green, because it did not represent an actual loss of a safety-related train since the C CCW pump was operable but degraded. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined the cause of this finding involved the cross-cutting area of Human Performance and the aspect of resources, H.1, because the licensee failed to ensure instructions were adequate and available to support nuclear safety-related work.
05000395/FIN-2016003-01Failure to Meet HRA Entry Requirements (Two Examples)2016Q3The inspectors identified two examples of a Green, self-revealing, non-cited violation (NCV) of Technical Specification (TS) 6.12.1, High Radiation Area. TS 6.12.1 requires that entries into high radiation areas (HRAs) be controlled with issuance of a radiation work permit (RWP) and that individuals entering these areas be made knowledgeable of the dose rates. Contrary to that, on two separate occasions, workers made entries into HRAs without being issued an appropriate RWP and without being knowledgeable of area dose rates. Specifically, on March 28, 2016, a worker tagging a pump on the auxiliary building (AB) 400-01 slab entered a HRA without the required radiological briefing and appropriate RWP. Also, on April 18, 2016, a worker performing dry cask welding operations in the fuel handling building entered a HRA without the required radiological briefing and appropriate RWP. The licensee entered these events into their corrective action program as condition reports CR-16-01528 and CR-16-01863. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure and the ability to assess dose was not compromised. Therefore, the finding was determined to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Avoid Complacency (H.12) because in both examples there were repostings, radiation areas were upgraded to HRAs due to changing radiological conditions, and prior to entry the workers failed to stop and get updated conditions and to adhere to the postings.
05000395/FIN-2016003-03Licensee-Identified Violation2016Q3V.C. Summer Operating License condition 2.c(18) states in part that the licensee 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 of which Chapter 3, Section 3.2.3, Procedures, states, Procedures shall be established for implementation of the fire protection program. Contrary to this, on September 14, 2016, the licensee failed to implement the requirements of procedure, Fire Protection Procedure, FPP-025, Fire Containment, Rev. 4, to ensure that fire door and SPB, DRAB/514, remained operable/functional. The inspectors used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and performed a Phase 1 analysis to determine the finding was of very low significance or Green. The fire confinement program element was not of low degradation, the non-suppression probability was 0.1, the fire frequencies related to the affected fire zones AB01.21.02 and FH01.04 were 2.79E-3 and 3.98E-4 respectfully, and the duration of the component inoperability was approximately 1 hour or 0.000114, which resulted in screening check frequency of 3.63E-8 that was less than the screening criteria of 1E-6. The licensee has documented this problem in their CAP as CR-16-05073.
05000395/FIN-2016002-01Failure to Adequately Manage Risk of Maintenance Activities Following Risk Model Updates2016Q2The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, involving the licensees failure to develop and implement specific risk management actions (RMAs) for a yellow risk condition associated with solid state protection system (SSPS) surveillance testing. The issue was entered into the licensees corrective action program (CAP) as condition report (CR)-16-02504. The inspectors identified a performance deficiency (PD) for the failure to manage the increase in risk associated with A train SSPS surveillance testing which was indicative of the lack of programmatic requirements for assessing and managing risk subsequent to equipment out of service (EOOS) model updates. The inspectors reviewed inspector manual chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because (1) it was associated with the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences, and (2) if left uncorrected the PD would have the potential to lead to a more significant safety concern. Specifically, the failure to manage the increase in risk jeopardizes the availability of remaining safety systems to combat the consequences of an initiating event. The inspectors reviewed IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined that the finding was of very low safety significance, Green, because the incremental core damage probability (ICDP) for the SSPS surveillance test was less than 1E-6. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined that this finding had a cross-cutting aspect in the area of Work Management (H.5), because the licensee did not develop specific RMAs for a yellow risk condition which was indicative of the lack of programmatic requirements for assessing and managing risk subsequent to EOOS model updates. (Section 1R13)
05000395/FIN-2016001-01Failure to Implement Adequate Administrative Controls Following a Departure from NFPA 80-1973 and Provide NRC Staff Complete and Accurate Information2016Q1The inspectors identified a Severity Level IV, non-cited violation (NCV) of 10 CFR 50.9(a), Completeness and accuracy of information, and an associated Green non-cited violation of V.C. Summer, Operating License Condition 2.C.(18) for a NFPA 80-1973 code deviation that was not discussed in the licensees NFPA 805 license amendment request (LAR), and would adversely affect the ability to achieve and maintain safe shutdown in the event of fire. The associated engineering evaluation relied on inadequate administrative controls to ensure the associated replacement doors in the intermediate building, DRIB/105A&B, were kept closed as a basis for not following NFPA 80-1973 which required the fire doors be self-closing. The licensee entered the violations into their corrective action program as condition reports CR-15-04027 and CR-16-00242 respectively. The inspectors identified a reactor oversight process (ROP) performance deficiency (PD) for the failure to provide adequate administrative controls to allow departure from NFPA 80-1973 requirements, which resulted in replacement of a self-closing fire door with two non-self-closing fire doors, DRIB/105A&B, that adversely affected the ability to achieve and maintain safe shutdown in the event of fire since they were found open on multiple occasions. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the ROP PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors, such as fire, 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 used IMC 0609, Significant Determination Process, Appendix F, Fire Protection Significance Determination Process, Attachment 1, dated September 20, 2013, to perform a Phase 1 analysis and determined that the ROP finding was of very low safety significance (Green) based on the response for Question 1.4.3.A in which the combustible loading on both sides of DRIB/105A&B was less than 120,000 BTU/ft2. Furthermore, the inspectors determined that the associated fire zone area (IB 7) with multiple equipment trains used a pre-action sprinkler system and automatic fire detection. The inspectors also determined that the licensees failure to include the departure from NFPA 80-1973 in their NFPA 805 license amendment request was a violation of 10 CFR 50.9(a). Because this violation of 10 CFR 50.9(a) had the potential to impact the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using traditional enforcement (TE). Since the TE violation is associated with a Green ROP violation, and the misinformation was identified after the NRC relied on it for issuing a previous operating license amendment, the TE violation was determined to be a Severity Level IV violation, consistent with the language of the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 14, 2014, and determined the cause of this finding involved the cross-cutting area of problem identification and resolution, P.3, because the licensee failed to ensure that adequate administrative controls were in place after the fire doors were found open multiple times.
05000395/FIN-2016001-02Licensee-Identified Violation2016Q1TS 3/4.5.2, ECCS Subsystems Tavg > 350 oF states in part that two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with one OPERABLE centrifugal charging pump in Modes 1, 2 and 3. TS 1.18, OPERABLE OPERABILITY, definition states in part that a subsystem shall be OPERABLE when all necessary auxiliary equipment that are required for the subsystem are capable of performing their related support functions. Contrary to this, from July 24, 2015, through September 17, 2015, safety-related subsystem chiller, XHX0001A, an auxiliary component supporting the A train charging pump (XPP0043A) was incapable of performing its safety function resulting in the inoperability of XPP0043A for greater than the allowed action times of TS 3/4.5.2. A review of IMC0609, Appendix A, determined the finding was of very low safety significance (Green) because the finding was not a design deficiency and it did not result in a loss of function. The licensee has documented this problem in their CAP as CR-15-04395.
05000395/FIN-2015406-01Security2015Q4
05000395/FIN-2015004-03Licensee-Identified Violation2015Q410 CFR 50, Appendix B, Criterion XVI states in part that conditions adverse to quality shall be promptly identified and corrected. Contrary to this, on August 4, 2015, the licensee discovered degradation of the B SW intake screen allowing the introduction of fish into the B SW pump bay. The licensee had initiated WO0601588 in 2006 to repair/rebuild the screen but failed to correct. A review of IMC0609, Appendix A, determined the finding was of very low safety significance (Green) because the finding was not a design deficiency and it did not result in a loss of function. The licensee has documented this problem in their CAP as CR-15-03574.
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