Semantic search
Jump to navigation
Jump to search
| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000369/FIN-2018012-01 | 30 September 2018 23:59:59 | Mcguire McGuire | NRC identified | Failure to Translate Seismic Mounting Requirements for 125 VDC Vital Batteries into Installation and Replacement Procedure | The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to translate the mounting requirements for seismic qualification contained in NLI technical calculation C-017-074-2, Vital Instrumentation & Control Batteries & Racks Equipment Qualification Calculation, Rev. 0, into their battery replacement and installation procedure IP/0/A/3061/003, 125 Volt Vital Battery Maintenance and Repair, Rev. 23 |
| 05000369/FIN-2018003-01 | 30 September 2018 23:59:59 | Mcguire McGuire | NRC identified | Failure to Adequately Document the Basis for a Change to the Emergency Plan | The inspectors identified a SL IV NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(3), for changes made to the McGuire Nuclear Station (MNS) Radiological Emergency Plan (E-Plan) that failed to demonstrate the changes would not reduce the effectiveness of the E-Plan. Specifically, the licensee did not provide an adequate analysis to determine that the removal of specific procedure references was not a reduction in effectiveness of the MNS E-Plan |
| 05000369/FIN-2018411-01 | 30 September 2018 23:59:59 | Mcguire McGuire | NRC identified | Security | |
| 05000369/FIN-2018410-02 | 30 June 2018 23:59:59 | Mcguire McGuire | NRC identified | Security | |
| 05000369/FIN-2018410-01 | 30 June 2018 23:59:59 | Mcguire McGuire | NRC identified | Security | |
| 05000369/FIN-2018011-01 | 30 June 2018 23:59:59 | Mcguire McGuire | NRC identified | Question about treatment of well-sealed, robustly secured cabinets in the Fire PRA | Inspectors identified an unresolved item (URI) associated with how the site calculated fire frequencies of electrical cabinets in the fire PRA. The site retained floor-mounted electrical cabinets characterized as well-sealed and robustly-secured as ignition sources in the fire PRA. The guidance of NUREG-6850, which the site is committed to, instructs that electrical cabinets housing voltages less than 400V, and that are characterized as well-sealed and robustly secured not be counted as ignition sources. This is because the fire PRA should only consider fires that can propagate to other combustibles and targets, and by including ignition sources that cannot propagate to other combustibles and targets, the frequency of fires in other electrical cabinets that can actually propagate could be erroneously lowered. Inspectors noted that retaining floor-mounted cabinets characterized as well-sealed and robustly-secured appeared to not be in alignment with the sites NFPA 805 submittal, and associated SE, which each contain information specific to the question about how well-sealed robustly secured cabinets were treated in the fire PRA. The SE states, Regarding the counting of well-sealed, robustly-secured electrical cabinets having circuits less than 440V, the licensee stated in its response to PRA RAI 03.b.01 that it updated the FPRA to remove cabinets meeting this definition. The site has asserted that this statement in the SE is incorrect, and does not align with what the site submitted on the docket as a part of their NFPA 805 submittal. Planned Closure Action(s): This issue is being characterized as a URI, pending a decision from NRR as to the interpretation of the sites submittal regarding treatment of floor-mounted well-sealed and robustly-secured cabinets, and the accuracy of the associated SER |
| 05000369/FIN-2018002-01 | 30 June 2018 23:59:59 | McGuire | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: NAC-Magnastor Certificate of Compliance 1031, Amendment 2, Technical Specifications SR 3.1.1.2 requires, in part, that the transportable storage canister (TSC) be backfilled with helium in the range of 0.694-0.802 g/liter prior to transport operations. Contrary to the above, on June 4, 2018, the licensee transported Magnastor cask 45 to the independent spent fuel storage installation pad with the TSC backfilled to approximately 0.85-0.89 g/liter due to the use of out of tolerance flow meters during backfilling operations. Significance/Severity Level: The inspectors determined that traditional enforcement is applicable for this NCV as it involved requirements pertaining to ISFSI operations and therefore the reactor oversight process is not applicable. The NCV was determined to be a Severity Level IV violation as it did not involve willfulness, was identified by the licensee, and was determined to be of minimal safety significance as the over fill of helium did not exceed any design parameters of the TSC during the transport operations.Corrective Action Reference: This issue was entered into the licensees corrective action program as NCR 2211048, Potentially Exceeding Magnastor Helium Density Upper Range. |
| 05000369/FIN-2018001-02 | 31 March 2018 23:59:59 | McGuire | Self-revealing | Failure to Adequately Implement the Solid State Protection System(SSPS)Test Procedure Resulted in a Unit 1 Reactor Trip | On February 16, 2018, the Unit 1 reactor tripped from 100 percent power due to a human performance error. Specifically, during B train SSPS testing, technicians reached Step 12.4.6 in procedure PT/0/A/4601/008 B which instructed Technician 1 to Depress and hold UV TEST push button for Train B in back of breaker CABINET- 1(RTB/BYB), and Technician 2 to proceed with Steps 12.4.7 through 12.4.13. Technician 2 completed the actions, out of the area, while Technician 1 stayed in the area holding the UV TEST push button. Technician 1 proceeded, without the procedure, to Step 12.4.14, which states, Release UV TEST push button, and proceeded to Step 12.4.15, which instructs the technicians to verify that the reactor trip breaker (RTB) opened. Technician 2 erroneously opened the bypass breaker(BYB) cabinet to perform Step 12.4.15 and noticed that the breaker was closed. Technician 2 proceeded to trip the BYB causing a reactor trip because the BYB was supplying power to the control rods while the RTB was being tested. Technician 2s actions were done without the procedure in-hand and the procedure did not direct the technician to manipulate the breaker, regardless of position. Corrective Actions: Technicians were removed from dutyPost-trip fitness-for-duty testing was performedReactor Trip investigation performed Corrective Action Reference: NCR 02185409, Unit 1 Reactor Trip during 1B SSPS Testing |
| 05000369/FIN-2018001-01 | 31 March 2018 23:59:59 | McGuire | NRC identified | Non-conservative Change to Core Exit Thermocouple (CET) Acceptance Criteria in Procedure PT/1/A/4600/003 D, Monthly Surveillance Items | On December 27, 2017, the licensee made a procedure change to PT/1/A/4600/003 D, Enclosure 13.1, Main Control Board Instrumentation Checklist, to expand the acceptance criteria for operable CET readings. The existing acceptance criteria specified an upper limit of saturation temperature (TSAT) and a lower limit of TCOLD. The procedure change subtracted 20 degrees Fahrenheit from TCOLD by applying CET total loop uncertainty. The licensee surmised that a CET could be reading as low as 20 degrees Fahrenheit below TCOLD and still be operating as designed. The purpose of the change was to allow the acceptance of CETs reading below TCOLD, as a number of CETs were failed and the margin to the required number of operable CETs was challenged. The inspectors noted that this logic would only apply to CETs that may be exposed to significant bypass flow (water at or near TCOLD). However, the vast majority of CETs would be exposed to water at hot leg temperature (THOT) or higher (90 degrees Fahrenheit or more higher than the new lower limit). The new acceptance criteria would increase the probability that a CET reading significantly lower than the actual water temperature would be considered operable. Since a low reading CET is non-conservative with respect to calculating sub-cooling margin, over time enough low reading CETs could be accepted such that the average core exit temperature used to calculate sub-cooling margin would be non-conservatively biased. This could have an adverse impact on appropriate operator actions for loss of sub-cooling conditions during accident mitigation.Corrective Actions: The licensee made a procedure change to PT/1/A/4600/003 D, to require an NCR to be written for CETs reading below TCOLD to prompt an engineering evaluation of the acceptability of any low reading CET. Additionally, the licensee created an action register item to track completion of repairs/replacement of existing inoperable CETs for both units.Corrective Action Reference: The licensee entered this issue into their corrective action program (CAP)as NCR 2176763. |
| 05000369/FIN-2018010-01 | 31 March 2018 23:59:59 | Mcguire McGuire | NRC identified | Failure to Update Offsite Circuit Operability Limit for Single Busline Alignment | The inspectors identified a Green finding for the licensees failure to update calculations as required by procedure AD-EG-ALL-1117, Design Analyses and Calculations, Rev. 5. Specifically, the licensee revised calculation MCC-1381.05-00-0258, U1, 6.9kV, 4.16kV & 600V Auxiliary Power Systems Safety-Related Voltage Analysis, Rev. 6, to identify the effect of longer motor-driven auxiliary feedwater pump (CA pump) acceleration times on the switchyard voltage limits in place to ensure offsite power source operability. However, the licensee failed to update the previously analyzed condition of only one offsite circuit in service from the switchyard to the 4160V Class 1E buses via the unit step-up and unit auxiliary transformers (single busline alignment). As a result, there was no verification that the offsite circuit operability limit was adequate during single busline alignment |
| 05000369/FIN-2018010-02 | 31 March 2018 23:59:59 | Mcguire McGuire | NRC identified | Failure to Update the FSAR With Pertinent Design Information | The inspectors identified a Green finding and associated Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.71(e), for the licensees failure to update the final safety analysis report (FSAR) to include the design function of manually opening the residual heat removal (ND) system shutdown cooling suction valves. Consequently, the licensee failed to consider the design capability of the valves, time impacts on dose consequence analyses, and the implication of pressure locking. |
| 05000369/FIN-2017003-01 | 30 September 2017 23:59:59 | Mcguire McGuire | NRC identified | Technical Specification (TS) Required Shutdown Due to Reactor Coolant System (RCS) Leakage | Inspection Scope On February 23, 2017, McGuire Unit 2 was shut down, as required by Technical Specification (TS) 3.4.13, due to RCS pressure boundary leakage on a Unit 2 safety 12 injection (NI) pipe upstream of the connection to D reactor coolant system (NC) cold leg. The licensee determined that the preliminary cause of the NI pipe leak was thermal fatigue. The inspectors reviewed the LERs for accuracy and completeness and reviewed the associated corrective actions (NCR 2102868) to determine whether they were appropriate . The inspectors also reviewed the LERs and NCRs to identify any licensee performance deficiencies associated with the issue. Licensee Event Reports 05000370/2016- 001 -00 and -01 are closed. Documents reviewed are listed in the attachment. b. Findings Description : On February 20, 2017, the McGuire Unit 2 RCS unidentified leakage increased from 0.01 gallons per minute (gpm) to 0.09 gpm. Concurrently, containment particulate counts increased from 500 to 30,000 counts per minute (cpm) and humidity increased from 8 percent to 12 percent. The licensee also determined unidentified RCS leakage increased to 0.255 gpm on February 22 (dayshift) and 0.303 gpm on February 23 (nightshift). On February 23, 2017, at 1922 hours, operators commenced a Unit 2 shutdown upon positive identification of pressure boundary leakage on the NI pipe upstream of the connection to D NC cold leg. Following shutdown, the licensee also discovered a small pinhole pressure boundary leak on the body of valve 2NC -30, the pressurizer spray bypass valve. The leakage from 2NC -30 was a negligible contributor to the total unidentified leakage measured during unit operation. The licensee determined that the cause of the NI pipe leak was thermal fatigue damage caused by NC cross -loop flows. The cause of the 2NC -30 valve leak was a casting flaw attributed to a combination of defects during the manufacturing process that resulted in a through wall pinhole leak in the valve body. The inspectors reviewed the licensees analysis and determined that the licensee had appropriately evaluated the issues. Since the causes were not reasonably within the licensees ability to foresee and correct, the inspectors concluded that there was no performance deficiency associated with the issue. For the purpose of closing the associated licensee event reports and categorizing the associated technical specification violation, a detailed risk evaluation of the RCS pressure boundary leakage condition was performed by a regional senior reactor analyst using the NRC McGuire SPAR probabilistic risk assessment model. The evaluation was performed in accordance with NRC Inspection Manual Chapter 0609 Appendix A. The exposure period was 114 hours and conditional rupture probability values were taken from EPRI report TR111880. The detailed risk evaluation determined that the condition represented an increase in core damage frequency of <1.0E -6/year, a finding of very low safety significance. McGuire TS 3.4.13 limiting condition for operation (LCO) requires that RCS operational leakage shall be limited to No pressure boundary leakage when in Modes 1 through 4. Action Statement B of TS 3.4.13 requires the plant to be placed in Mode 3 within 6 hours and Mode 5 within 36 hours. The licensee entered Mode 3 on February 24, 2017 , at 0041 hours and entered Mode 5 on February 24, 2017 , at 1721 hours. The licensee made an event notification (EN) to the NRC on February 23, 2017, at 2201 hours (EN 52573) due to a shutdown of the plant required by TS. Enforcement: The McGuire TS 3.4.13 LCO requires that RCS operational leakage shall be limited to No pressure boundary leakage when in Modes 1 through 4. Contrary to the above, McGuire Unit 2 experienced RCS pressure boundary leakage while operating in Mode 1. Although a violation of the TS occurred, the violation was not attributable to 13 equipment failures that were avoidable by reasonable licensee quality assurance measures or management controls, therefore a performance deficiency was not identified. The inspectors utilized the Enforcement Policy examples of Section 6.1, and available risk -informed tools to assess the safety significance of the RCS pressure boundary leakage and related violation. Based on the fact that the leak rate was stable and within the capacity of the charging system and would not impact other systems used to mitigate a loss of coolant accident, the inspectors concluded the safety significance of the violation was very low and consistent with Severity Level IV. Additionally, the risk aspects were discussed and confirmed with a regional senior risk analyst. This issue was documented in the licensees corrective action program as NCR 2102868. The NRC exercised enforcement discretion in Enforcement Action (EA) -2017- 119, in accordance with Section 3.10 of the Enforcement Policy because the violation was not associated with a licensee performance deficiency. Specifically, the violation was not attributable to an equipment failure that was avoidable by reasonable licensee quality assurance measures or management controls and therefore inspectors concluded that there was no performance deficiency associated with the RCS boundary leakage. The violation will not be considered in the assessment process or the NRCs Action Matrix. |
| 05000369/FIN-2017002-01 | 30 June 2017 23:59:59 | Mcguire McGuire | Self-revealing | Inadequate Survey Results in Unposted HRA | Green . A self -revealing Green non- cited violation (NCV) of 10 CFR 20.1501(a)(2) was identified for the licensees failure to conduct an adequate area radiation survey in Room 619 of the auxiliary building (waste gas decay tank (WGDT) room). Specifically, on April 19, 2016 , a high radiation area (HRA) was identifi ed near WGDT A in the WGDT room when a worker entering the area received a dose rate alarm on his electronic dosimeter (ED) and follow -up surveys revealed dose rates as high as 110 mrem/hr at 30cm. Also, as a result of the licensees failure to perform a survey, the area was not barricaded and posted in accordance with plant Technical Specification (TS) 5.7.1, High Radiation Area. The licensee immediately barricaded and posted the area as an HRA, performed an apparent cause evaluation to determine additional long term actions and entered the issue into their corrective action program as Nuclear Condition Report (NCR) 02021742. The licensees failure to conduct an area radiation survey to evaluate the magnitude and extent of radiation levels near WGDT A was a performance deficiency. This finding was determined to be more than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, failure to identify, post and control HRAs could allow workers to enter HRAs without knowledge of the radiological conditions in the area and receive unintended occupational exposure. The finding was evaluated using Inspection Manual Chapter (IMC) 0609 Appendix C, Occupational Radiation Safety Significance Determination Process. The finding was not related to the a s low as reasonably achievable (ALARA) planning, did not involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross -cutting aspect of avoid complacency in the area of human performance because the possibility of significant dose rate changes in the WGDT room during startup was a latent issue for which the licensee failed to recognize and plan. (H.12) |
| 05000369/FIN-2017007-01 | 31 March 2017 23:59:59 | Mcguire McGuire | NRC identified | Failure to Translate Required Gasket Replacement Requirements into Limit Switch Maintenance Manual | Green. The team identified a green non-cited violation (NCV) of Title10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate requirements necessary for maintaining the environmental qualification of the pressurizer power-operat ed relief valve (PORV) NAMCO EA-180 limit switches into maintenance procedures. The licensee evaluated the impact of the incorrect guidance and determined that the PO RV limit switches remained operable. The licensee plans to correct the affected procedures. The licensee entered this issue into the corrective action program as NCR 02095333. This performance deficiency was more than minor because it was associated with the mitigating systems cornerstone attribut e of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining the PORV limit switches in their qualified condition impacted their reliability. The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued October 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 design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000369/FIN-2017001-01 | 31 March 2017 23:59:59 | McGuire | NRC identified | Failure to Adequately Control Transient Combustibles Using a Receptacle with a Self-Closing Lid | Green. An NRC-identified Green non-cited violation (NCV) of the McGuire Unit 1 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately implement fire protection procedures for a waste receptacle fitted with a self-closing fire lid used to store plastic hard hats in the Unit 1 B train electrical penetration room. The licensee took immediate corrective actions to empty the receptacle (Nuclear Condition Report (NCR) 2100090). The licensees failure to properly implement transient combustible control requirements for a waste receptacle equipped with a self-closing fire lid was a performance deficiency. The performance deficiency was more than minor because it affected the initiating events cornerstone attribute of protection against external factors, specifically fire, and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, a fire ignited in the overfilled receptacle without a functioning self-closing lid could damage safety related cabling running directly overhead. The inspectors determined the finding to be of very low safety significance (Green) because it did not affect the ability to reach and maintain cold shutdown conditions in that a postulated fire in the overfilled receptacle did not present the possibility of impacting more than one train of safe shutdown equipment. This finding had a cross-cutting aspect of procedure adherence in the area of human performance, because personnel did not follow procedural requirements of procedure AD-EG-ALL-1520. (H.8) |
| 05000369/FIN-2016003-01 | 30 September 2016 23:59:59 | McGuire | Licensee-identified | Licensee-Identified Violation | Technical Specifications 5.4.1.a, Procedures, requires, in part, that procedures for certain activities recommended in Regulatory Guide 1.33, Rev. 2, Appendix A, be established, implemented, and maintained. Administrative procedures for shift and relief turnover is one of the identified activities. Administrative procedure AD-OP-ALL-1000, Conduct of Operations, Rev. 4, implements the licensees shift and relief turnover standards. This procedure requires shift turnovers to contain detailed information on equipment and system status, alignments, and activities, to ensure watchstanders have a complete understanding of plant status. Contrary to the above, from August 10 to August 13, 2015, operators were not aware of the required nuclear service water system alignment which required a continuous vent (passing water flow) to be maintained in the condenser cooling water (RC) suction supply to the Unit 1 turbine driven auxiliary feedwater pump. The continuous vent mitigates the potential for air entrainment in the RC piping high point and is needed in order for the standby shutdown system to be functional during an Appendix R fire event when the suction of the turbine driven auxiliary feedwater pump is transferred from the auxiliary feedwater storage tank to the long term water supply provided by the RC system. This lack of operator awareness stemmed from a misunderstanding in the operator turnovers that the nuclear service water system was in a standby nuclear service water pond cooling alignment, which does not require the continuous vent to be maintained. The discrepancy was subsequently identified by oncoming shift operations personnel and the continuous vent was re-established on August 16, 2015, after removing material that obstructed the continuous vent line. As a result of not maintaining the continuous vent at the suction of the turbine driven auxiliary feedwater pump, the standby shutdown system was rendered non-functional for a period of eleven days, which was in excess of the 7-day limit allowed by Selected Licensee Commitments 16.9.7. This violation was determined to be of very low safety significance (Green) because it only affected the non-safety related Appendix R water supply to the turbine driven auxiliary feedwater pump. This violation was entered into the licensees corrective action program as NCR 01943414. |
| 05000369/FIN-2016002-01 | 30 June 2016 23:59:59 | Mcguire McGuire | NRC identified | Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak Chase Test Connections | An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50.55a, Codes and Standards, was identified for the licensees failure to perform general visual examinations of moisture barrier material in the reactor containment leak-chase channel test connections in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV Code), Section XI, Subsection IWE. The licensee performed the required examinations in Unit 1 during the March 2016 refueling outage and initiated corrective actions to revise the Containment Inservice Inspection (ISI) Plan. The licensee also planned to perform similar examinations in Unit 2 prior to the end of the first containment ISI period. Additionally, the licensee performed a containment operability determination to justify continuous operation of the Unit 1 and Unit 2 containment based on the results of all visual examinations, extent of condition activities, and the results of containment integrated leak rate tests. The licensee entered this issue into their corrective action program as action request (AR) 02038505. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME BPV Code was a performance deficiency (PD). The PD was of more than minor significance per IMC-0612, Appendix B, Issue Screening, because the current Containment ISI Plan did not adequately implement the ASME BPV Code requirements for the examination of moisture barriers, and if left uncorrected, it had the potential to lead to a more significant concern. The finding was of very low safety significance (Green) per IMC-0609 because it did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect of resolution in the problem identification and resolution cross-cutting area because the licensee did not take effective corrective actions to implement the ASME BPV code requirements in the Containment ISI Plan when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014-07. |
| 05000369/FIN-2016002-02 | 30 June 2016 23:59:59 | McGuire | NRC identified | Failure to Ensure Containment Equipment Hatch Was Properly Closed During Fuel Movements | An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for the licensees failure to adequately implement the commitments in Selective Licensee Commitment (SLC) 16.9.25, Refueling Operations Containment Equipment Hatch, which required the containment equipment hatch to be closed during the movement of non-recently irradiated fuel inside containment. Specifically, during reactor vessel fuel reload activities, the inspectors identified that the equipment hatch was left partially open due to the failure to properly tighten the bolts evenly around the hatch resulting in direct communication of the containment atmosphere with the environment. The licensee took immediate corrective action to suspend fuel movements and properly tighten the equipment hatch bolts prior to resuming fuel movements and entered the issue into their corrective action program as ARs 02018605 and 02018701. The PD was more than minor because it impacted the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that containment protects the public from radionuclide releases caused by accidents or events. Additionally, if left uncorrected, the PD would have the potential to lead to a more significant safety concern. Specifically, the radiological barrier functionality of the containment equipment hatch was degraded due to the gap opening which could have allowed direct access of radiological releases from the containment atmosphere to the outside environment during a potential fuel handling accident inside containment. The inspectors screened the finding in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Because the finding degraded the ability to close or isolate the containment, it required review using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. While the containment boundary function was considered degraded, the incident occurred eight days after the beginning of the refueling outage when short lived volatile radioisotopes had decayed sufficiently such that the potential radiological releases to the public would not likely contribute to the large early release frequency (LERF). Based on this, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of procedure adherence in the cross-cutting area of human performance because the licensee failed to follow containment equipment hatch closing procedures which explicitly required performing a visual inspection that the containment equipment hatch was sealed and secured with metal-to-metal contact with the containment hatch flange and had no visual gaps. |
| 05000369/FIN-2016001-01 | 31 March 2016 23:59:59 | Mcguire McGuire | NRC identified | Failure to Maintain Fire Extinguishers in Contaminated Radiation Control Zones in Accordance with the Fire Protection Program | An NRC-identified Green non-cited violation (NCV) of the McGuire Nuclear Station Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for failure to perform annual maintenance on fire extinguishers located in contaminated radiation control zones (RCZs). The licensee took immediate corrective action to replace the past due fire extinguishers and entered the issue into their corrective action program as action request (AR) 02009794. The performance deficiency (PD) was more than minor because if left uncorrected the PD could have the potential to lead to a more significant safety concern, in that, fire extinguishers located in any contaminated RCZs may not be functional for firefighting purposes due to lack of maintenance. Every fire extinguisher, five total, located in a contaminated RCZ, did not have its annual maintenance up-to-date. The longest duration without annual maintenance was six years for two of the five extinguishers. The finding was determined to be of very low safety significance (Green) within the mitigating system cornerstone because it would not affect the ability to reach and maintain a safe shutdown condition, in that, for each of the fire areas where the out-of-date extinguishers were present, there were also properly maintained fire extinguishers and hose stations outside of the RCZ. The out-of-date extinguishers were weighed and it was determined that they would have performed their function, if needed. The cause of the PD was directly related to the cross-cutting aspect of field presence in the cross-cutting area of human performance because the licensee failed to correct deviations from the FPP and ensure proper oversight of the vendor contracted to perform fire extinguisher maintenance. |
| 05000369/FIN-2015008-01 | 31 December 2015 23:59:59 | Mcguire McGuire | NRC identified | Failure to Completely and Accurately Translate the Safe Shutdown Analysis to Procedures | The NRC identified a Green non-cited violation (NCV) of McGuire Technical Specification 5.4.1.a, for Unit 1, for having an inadequate procedure to support safe shutdown for a fire in fire area (FA) 15/17. Specifically, the licensees deterministic safe shutdown analysis identified the need for a procedural action to de-energize PORV 1NC- 34A at power supply 1EVDA, breaker 8. This action was not translated to Enclosure 15 of McGuire fire safe shutdown procedure AP-45. This item was entered into the corrective action program (CAP) as action requests (ARs) 1979875 and 1983360, and the licensee initiated a procedure change to incorporate the missing action. The performance deficiency (PD) was more than minor because it was associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e. fire), and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event (Task 1.4.5-B). No cross cutting aspect was assigned because the finding did not represent current licensee performance. |
| 05000369/FIN-2015003-01 | 30 September 2015 23:59:59 | Mcguire McGuire | Self-revealing | Failure to Adequately Implement a Temporary Modification for a Leak Enclosure | A self-revealing Green finding (FIN) was identified for failure to adequately implement the modification procedural requirements of engineering directives manual (EDM)-601, Engineering Change Manual, for a temporary modification that installed a valve leak seal enclosure on main steam drain valve 2SM-27. Specifically, EDM-601 required the weight and vibration response of the enclosure to be evaluated as part of the installation. The failure to consider this resulted in vibration induced piping failure upstream of the valve and an unexpected rapid plant down power. The failure to adequately implement a temporary modification in accordance with EDM- 601 was a performance deficiency (PD). The PD was more than minor because it was associated with the design control attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability during power operations. Specifically, the performance deficiency resulted in a rapid down power to approximately 20 percent and subsequent actions to take the Unit 2 turbine generator offline to repair the leak. Using NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance because the it did not contribute to both the cause of a reactor trip and affect mitigation equipment. The finding had a cross cutting aspect of consistent process, as described in the human performance crosscutting area because the licensee failed to use a consistent, systematic approach to make de.cisions during implementation of a temporary modification. |
| 05000369/FIN-2015007-01 | 30 June 2015 23:59:59 | Mcguire McGuire | NRC identified | Failure to Verify Protection System DC Molded Case Circuit Breaker Ratings | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, consisting of two examples. In one example, the licensee failed to verify the adequacy of GE model TED molded case circuit breaker (MCCB) design. In the second example, the licensee failed to verify the adequacy of Eaton model HFB MCCB design. The licensee initiated Action Request (AR) 01929605 and AR 193674, which determined the systems were operable because upstream protective devices provided protection from a failed HFB and/or TED MCCBs, and that the HFB and TED MCCBs would be replaced with MCCBs that have adequate ratings. The licensees failure to design the Class 1E electric system MCCBs in accordance with IEEE 308-1971 Sections 4.1 and 5.3.5 was a performance deficiency. The team determined that the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the deficiency affected the design or qualification of a mitigating structure, system, or component (SSC), but the SSC maintained its operability or functionality. No cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000369/FIN-2015002-01 | 30 June 2015 23:59:59 | Mcguire McGuire | NRC identified | Failure to Establish Compensatory Actions for Obstructed Fire Sprinkler Spray Nozzle | An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for failure to evaluate and establish adequate compensatory measures for an impaired fire protection automatic water sprinkler system. Specifically, a solid deck scaffold platform was erected below a sprinkler system spray nozzle that would have obstructed the nozzle spray pattern protecting safe shutdown equipment involving the 2B2 component cooling water pump/motor. The licensee entered the issue into the corrective action program (CAP) as nuclear condition report (NCR) 01931412 and implemented immediate corrective actions to remove the scaffolding obstructing the sprinkler nozzle. The failure to evaluate scaffolding obstruction of a sprinkler system spray nozzle and implement required fire protection compensatory actions was a performance deficiency (PD). The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (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. Specifically, the failure to provide adequate compensatory actions for an obstructed sprinkler nozzle would have reduced the licensees ability to quickly extinguish fires in the area. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings. Using the guidance in IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, the finding was assigned a category of fixed fire protection systems. The inspectors determined the finding to be of very low safety significance (Green), because it was assigned a low degradation rating that was based upon meeting the criteria described in IMC 0609, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements. Specifically, less than ten percent of the sprinkler nozzles were nonfunctional, there were functional nozzles within five feet of the combustibles of concern, and the system was nominally code compliant. The finding had a cross-cutting aspect of procedure adherence in the human performance area, because the licensee failed to follow scaffolding erection procedures which explicitly required not erecting scaffolding that could obstruct sprinkler nozzles unless approved by a fire protection engineer and necessary compensatory actions were implemented. |
| 05000369/FIN-2015007-02 | 30 June 2015 23:59:59 | Mcguire McGuire | NRC identified | Failure to Perform Adequate Periodic Testing of Molded Case Circuit Breakers | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, consisting of two examples. In one example, the licensee failed to scope some Class 1E molded case circuit breakers (MCCBs) into the Class 1E MCCB testing program. In the second example, the licensees test procedure preconditioned the Class 1E MCCBs before testing their safety function. The licensee initiated Action Request (AR) 1936760 and AR 01934403, which determined the systems were operable because an engineering review of previous TED breaker testing and PM's has not shown a trend of degradation of the breakers ability to perform its function. In addition, the licensee planned develop a more extensive and adequate testing program. The licensees failure to perform adequate MCCB testing in accordance with IEEE 308- 1971, Section 6.3, Periodic Equipment Tests, was a performance deficiency. The team determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the deficiency affected the design or qualification of a mitigating structure, system, or component (SSC), but the SSC maintained its operability or functionality. No cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. |
| 05000369/FIN-2015404-01 | 31 March 2015 23:59:59 | Mcguire McGuire | NRC identified | Security | |
| 05000369/FIN-2014005-02 | 31 December 2014 23:59:59 | McGuire | NRC identified | Failure to Adequately Implement Containment Closeout Resulting in Loose Debris and Unanalyzed Materials Left in Containment | An NRC-identified Green NCV of Technical Specification 5.4.1.a, Procedures, was identified for the failure to properly implement containment cleanliness and material control closeout procedures in accordance with procedure PT/1A/4600/003F, Containment Cleanliness and ECCS Operability Inspection, prior to entering Mode 4, following the Unit 1 refueling outage. Specifically, a large amount of unanalyzed general loose debris, as well as scaffolding with aluminum walkboards and fibrous lead blankets, were left in containment that could either contribute to emergency core cooling system (ECCS) recirculation sump screen blockage or containment hydrogen generation during design basis accidents. The licensee placed this issue into their CAP and took corrective actions to remove the loose debris and unanalyzed materials and performed re-inspections of containment to identify any additional loose debris or unanalyzed materials left in containment. The failure to perform an adequate containment cleanliness and material control closeout following the Unit 1 refueling outage in accordance with procedure PT/1/A/4600/003F was a PD. The PD was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective in that, loose debris in containment could result in the debris being transported to the ECCS recirculation sump screens in the event of design basis accident and adversely affect the sump performance. In addition, the PD was associated with the configuration control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective in that, the failure to control scaffolding that contained unanalyzed amounts of aluminum in containment challenged the existing analysis for containment aluminum inventory limitations. The finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of safety function of the ECCS sumps, was not safety significant due to external events, and no actual open pathway in the physical integrity of containment occurred. The finding had a cross-cutting aspect of field presence in the human performance area because the licensee failed to ensure that adequate supervisory and management oversight of the containment closeout process was conducted to ensure proper performance of procedure PT/1/A/4600/003F prior to entering Mode 4. |
| 05000369/FIN-2014005-01 | 31 December 2014 23:59:59 | Mcguire McGuire | NRC identified | Failure to Adequately Control Transient Combustible Materials and Ignition Sources in Accordance with the Fire Protection Program | An NRC-identified Green NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately control fire ignition sources in the Unit 1 and Unit 2 exterior doghouses in accordance with the FPP requirements of Nuclear System Directive (NSD)- 313, Control of Transient Fire Loads. Specifically, temporary electric portable heaters were energized for several days without implementing required hourly fire watches, locating the energized heaters greater than prescribed separation distances from safety-related equipment, and preventing other transient combustible materials from being located near the heaters. The licensee placed this issue into their corrective action program (CAP) and took corrective actions to de-energize the heaters, distance the heaters away from safety related feedwater isolation valve electrical cables, and remove unnecessary transient combustibles from the area. The failure to control fire ignition sources in accordance with NSD-313 was a performance deficiency (PD) . The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective in that, a fire could have affected nearby safety-related feedwater isolation valve electrical cables which provide a shutdown mitigation function. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition. This finding had a cross cutting aspect of teamwork in the human performance area because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions. |
| 05000369/FIN-2014004-01 | 30 September 2014 23:59:59 | McGuire | NRC identified | 1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws | The licensee identified flaws with ultrasonic testing in the 1B and 1C cold leg safety injection pipe welds as part of their extent of condition from Unit 2 for MRP- 146, Thermal Fatigue. Further evaluation determined these flaws were a circumferential flaw with an axial component on the nozzle side for 1B and an axial flaw from the centerline of the weld into the base metal for 1C. The licensee completed examinations on all welds included in the MRP-146 program and found them to be within the acceptance criteria. The licensee also removed and repaired the 1B and 1C nozzles. Welding of the new components have been examined and have passed all quality assurance examinations. The licensee determined that the flaws were a result of thermal fatigue. The licensee has performed all required examinations and repairs and is completing a metallurgical analysis of the flaws. This is an unresolved item pending review of the licensees metallurgical analysis of the flaws to determine if there is a performance deficiency. This issue will be tracked as URI 05000369/2014004-01, 1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws. |
| 05000369/FIN-2014004-02 | 30 September 2014 23:59:59 | McGuire | NRC identified | Review NOED 14-2-002 Granting Exercise of Enforcement Discretion to Complete 1B EDG Repairs | The inspectors reviewed NOED 14-2-002 and related documents to determine the accuracy and consistency with the licensees assertions and implementation of the licensees compensatory measures and commitments. The inspectors independently verified the proper implementation of these compensatory measures which included deferring non-essential surveillances and other maintenance activities on the 1A EDG, TDCA pump, SSF, switchyard, and posting dedicated fire watches in selected risk significant areas. Additional inspection of this issue will be conducted as part of the NRCs review of the subsequent Licensee Event Report (LER) to be submitted by the licensee within 90 days. This LER will describe the circumstances of the 1B EDG failure, the root cause, and planned licensee corrective actions. This URI is identified as URI 05000369/2014004-02, Review NOED 14-2-02 Granting Exercise of Enforcement Discretion to Complete 1B EDG Repairs. |
| 05000369/FIN-2014003-01 | 30 June 2014 23:59:59 | Mcguire McGuire | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, the licensee to establish measures to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, in March 2008 and September 2012, the licensee failed to promptly identify a condition adverse to quality. A flaw in the Cold Leg 2D Nozzle 4-1 weld was missed during UT examinations of this component. During the most recent outage, the licensee reexamined this weld and identified an 85 percent through-wall flaw originating from the inner diameter in the weld. Destructive testing and analysis established that this flaw most likely existed since 2005. This violation was determined not to be greater than very low safety significance (Green) because it could not result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) and could not have likely affected other systems used to mitigate a LOCA. This violation was documented in PIP M-14-03544. |
| 05000369/FIN-2014403-02 | 30 June 2014 23:59:59 | Mcguire McGuire | Licensee-identified | Licensee-Identified Violation | |
| 05000369/FIN-2014003-02 | 30 June 2014 23:59:59 | Mcguire McGuire | Licensee-identified | Licensee-Identified Violation | TS 5.4.1.a requires that written procedures shall be established, implemented, and maintained as recommended in Regulatory Guide (RG) 1.33, Rev. 2, Appendix A, February 1978. Section 3.d of RG 1.33 recommends that appropriate procedures be prepared for energizing, filling, venting, draining, startup, shutdown, and changing modes of operation for the ECCS. Contrary to the above, the licensee failed to develop an adequate procedure to fill and vent the Unit 2 NV system following system draining and argon gas introduction associated with NV 2A mixed bed demineralizer valve modification work conducted December 1-12, 2013. On December 14, 2013, following this modification work, a large gas void was identified in the ECCS piping at high point vent valve 2NV-1056, located in the suction of the ECCS pumps during design basis accident conditions involving cold-leg recirculation. The licensee determined that the use of an inadequate fill and vent procedure during the system restoration from the modification work resulted in the accumulation of a significant amount of the gas at this location. This violation was determined to be of very low safety significance (Green) because the licensee provided reasonable evidence that the ECCS pumps would have been capable of performing their intended safety function had the gas void been ingested into the suction of the pumps. This violation was documented in the licensees CAP as PIP M-13-11181. |
| 05000369/FIN-2014403-01 | 30 June 2014 23:59:59 | Mcguire McGuire | NRC identified | Security | |
| 05000369/FIN-2014007-01 | 30 June 2014 23:59:59 | McGuire | NRC identified | Inadequately Sealed Safety Related Electrical Cabinet | An NRC-identified NCV of 10 CFR Part 50 Appendix B, Criterion XVI, Corrective Action, was identified when the licensee failed to promptly identify a condition adverse to quality associated with the inadequate sealing for safety related cabinet 1FWPNRWLP (Unit 1 Refueling Water Storage Tank (RWST) Channel 4 Level Instrumentation loop). Specifically, the licensee did not identify that the seal around a cable bundle entering the top of 1FWPNRWLP had degraded to the point where it would no longer protect against water intrusion into the cabinet. The licensee placed this issue into their CAP as PIP M-14-05643 and took corrective action by replacing the seal. The inspectors determined that the failure to promptly identify a condition adverse to quality associated with the inadequate sealing of 1FWPNRWLP was a performance deficiency. This performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the capability of the automatic RWST swap over function to respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined this finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification and did not represent an actual loss of system and/or function. The finding had a cross-cutting aspect of Procedure Adherence, as described in the Human Performance cross-cutting area because the licensee failed to adequately implement the walkdown process outlined in EDM-203 and promptly identify this degradation. |
| 05000369/FIN-2014002-03 | 31 March 2014 23:59:59 | McGuire | Self-revealing | Failure to Implement Adequate Design Control Measures for Rod Control Power Supply Replacement Resulting in Reactor Trip | A self-revealing finding (FIN) was identified for the licensees failure to implement adequate design control measures for the rod control power supply modification which resulted in the loss of 24VDC power in the 1AC rod control power cabinet. The inspectors determined that the licensees failure to implement adequate design control measures was more than minor because it affected the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective, in that, the insufficient margin in the rod control power supply OVP function caused a multiple drop rod event which resulted in a reactor trip. This finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance. |
| 05000369/FIN-2014002-02 | 31 March 2014 23:59:59 | Mcguire McGuire | NRC identified | Failure to Adequately Control the Use of Self- Extinguishing Fire Lids | An NRC-identified NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, FPP, was identified for the licensees failure to adequately control the storage of transient combustibles in waste receptacles equipped with self-extinguishing fire lids in accordance with the FPP requirements. The licensee took actions to correct all waste receptacles in the plant that were filled beyond the manufacturers specification or had loosely fitted lids. This condition was placed in the licensees corrective action program. The licensees failure to control the storage of transient combustibles in accordance with the requirements of NSD-313 was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective in that the self-extinguishing function was not retained which could allow the spread of the fire and adversely affect mitigating system equipment in the area. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown conditions. A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance. |
| 05000369/FIN-2013502-01 | 31 December 2013 23:59:59 | McGuire | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy for being disposition as a NCV.TS 3.3.2, Function 6.f, requires that all four instrumentation channels of the TDCA pump suction transfer function be operable in Modes 1, 2, and 3. TS 3.3.2, Condition N, specifies if one or more of the pressure switch instrumentation channels are inoperable, the channel must be restored to operability within 48 hours or the associated TDCA pump must be declared inoperable. Contrary to this requirement, from September 1993 to May 30, 2013, the channel associated with pressure switch 1CAPS5390 was inoperable and the licensee failed to declare the Unit 1 TDCA inoperable within the required TS completion time. This violation was determined to be of very low safety significance (Green) because the channel would still have been capable of actuating and aligning the TDCA to its assured water source within the timeframe necessary for the pump to perform its intended safety function. This violation was documented in the licensees CAP as PIP M-13-05935. |
| 05000369/FIN-2013008-01 | 31 December 2013 23:59:59 | Mcguire McGuire | NRC identified | Fuel Manipulator Crane Digital Modification | The fuel handling system consists of the equipment needed for the refueling operation of the reactor core. This equipment is comprised of manipulator cranes, fuel handling equipment and a fuel transfer system. The manipulator cranes have the potential to initiate a fuel handling accident as described in Updated Final Safety Analysis Report (FSAR) section 15.7.4. The postulated accident assumes a fuel assembly is damaged while being moved inside containment or the Spent Fuel Pool Building. Engineering changes EC 77048, Unit 1 Manipulator Crane Upgrade Project, Rev. 18 and EC 77051, Unit 2 Manipulator Crane Upgrade Project, Rev. 15, were implemented to address reliability and obsolescence problems confronting the fuel manipulator cranes in the reactor containment and spent fuel building. The power and control systems were upgraded, including the position sensors, motor drives, control consoles, and wiring. The existing analog controls were replaced using a digital programmable logic controller (PLC) with a graphic user interface. The PLC can now be programmed in advance with the refueling sequence, and the step-wise destinations of each fuel assembly. The PLC controls allow multi-axial travel (in the x-y dimensions) within established safe operation zones. The maximum crane bridge and trolley speeds were increased, but the existing acceleration limits were retained. The hoist slow-speed zones were reduced, but an adequate distance had been retained for the safe insertion of a fuel assembly into a storage or core location. The licensee performed a 10 CFR 50.59 evaluation in accordance with procedure NSD-209, 10 CFR 50.59 Process, Rev. 14, and determined the change could be implemented without prior NRC review and approval. After reviewing the 10 CFR 50.59 evaluation, the inspectors found that they did not have sufficient information to determine that NRC review and approval was not required prior to the implementation of the modification. The inspectors could not verify the licensees conclusions regarding the reliability and dependability of the software used to operate the manipulator crane. Specifically, the licensee did not address software failure modes and effects, and the software development processes described in section 5.3.3 of NEI 01-01, Digital System Quality, Rev. 1, in enough detail for the team to reach the same conclusion. In order to determine the adequacy of the licensees 10 CFR 50.59 evaluation and whether or not there is a violation of 10 CFR 50.59, Changes, Tests, and Experiments, this issue remains unresolved pending the inspectors review of additional information to be provided by the licensee to address the issues described above. The licensee entered this issue into their corrective action program as problem investigation program (PIP) report M-13-11029 to track the actions taken to address the teams concern. |
| 05000369/FIN-2013008-02 | 31 December 2013 23:59:59 | Mcguire McGuire | Licensee-identified | Licensee-Identified Violation | 10 CFR 50.71(e) requires, in part, that each person licensed to operate a nuclear power reactor, shall update periodically, the FSAR originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. This submittal shall include the effects of all changes made in the facility or procedures as described in the FSAR. Contrary to the above, since November 30, 2012, the site failed to include updates related to a modification that changed the actuation of the containment spray system from automatic to manual. Traditional enforcement is applicable because the violation could impact the regulatory process, and was evaluated using the NRCs Enforcement Policy. This violation was determined to be a Severity Level IV violation because the lack of up-to-date information did not result in an unacceptable change to the facility or procedures. This violation was documented in the licensees corrective action program as PIPs M-13-08057, M-13-08607, and M-13-08684. |
| 05000369/FIN-2013405-01 | 30 September 2013 23:59:59 | Mcguire McGuire | NRC identified | Security | |
| 05000369/FIN-2013003-01 | 30 June 2013 23:59:59 | Mcguire McGuire | Self-revealing | Failure to Implement Adequate Venting Instructions for Condensate Booster Pump Trip Instrumentation Resulting in Reactor Trip | A self-revealing finding was identified for the licensees failure to implement adequate instructions for venting condensate booster pump (CBP) emergency low suction pressure trip instrumentation which resulted in air entrainment causing a non-conservative shift in the trip setpoint. During a subsequent secondary side transient involving a heater drain tank pump trip, the non-conservative trip setpoint resulted in a premature trip of all three CBPs ultimately causing a reactor trip. The performance deficiency was more than minor because it affected the Procedure Quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective, in that, the inadequate venting allowed air entrainment in the instrumentation lines resulting in a reactor trip. This finding was determined to have very low safety significance (Green) because it did not contribute to the likelihood of both a reactor trip and that mitigation equipment or functions would not be available. No cross cutting aspect was identified. |
| 05000369/FIN-2013007-01 | 31 March 2013 23:59:59 | Mcguire McGuire | NRC identified | Modifications Result in Nonfunctional Fire Doors | An NRC identified Green non-cited violation of McGuires Selected Licensing Commitment 16.9.5, Fire Rated Assemblies was identified for the licensees inadequate implementation of modifications that resulted in nonfunctional fire doors. The licensee has entered the finding into the corrective action program as PIP M-13- 01454, declared the doors as nonfunctional and implemented fire watches for the fire areas of concern. The licensees inadequate implementation of fire door modifications that resulted in the failure to meet the requirements of Selected Licensee Commitment 16.9.5, Fire Rated Assemblies, was a performance deficiency. The performance deficiency was more than minor because it adversely affected the Mitigating Systems cornerstone attribute of Protection Against External Events. Specifically, the welding modifications performed on nine fire doors adversely affected their capability to provide the required 3-hours of fire resistance. In accordance with NRC IMC 0609 Appendix F, Part 1; Fire Protection Significance Determination Process Phase 1 Worksheet the inspectors determined the finding to be of very low safety significance (Green) because the fire doors would still provide a minimum of 20 minutes fire endurance protection A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance. |
| 05000369/FIN-2013002-02 | 31 March 2013 23:59:59 | Mcguire McGuire | Licensee-identified | Licensee-Identified Violation | TS 3.6.3 required that each containment isolation valve be operable in Modes 1, 2, 3, and 4. TS 3.6.3, Condition A, specified if one containment isolation valve is inoperable, the flow path must be isolated within 4 hours and verified isolated once per 31 days. Contrary to the above, from November 2, 2012, to November 4, 2012, with Unit 2 in Mode 4, manual containment isolation valve 2NV-1053 was inoperable and the licensee failed to isolate the flow path within 4 hours. This violation was determined to be of very low safety significance (Green) due to the small size of the piping and that a control room air-operated valve (i.e., 2NV-840) located downstream of 2NV-1053 could have been used to isolate the penetration. This violation was documented in the licensees CAP as PIP M-12-09347. |
| 05000369/FIN-2013002-01 | 31 March 2013 23:59:59 | Mcguire McGuire | Self-revealing | Failure to Revise Turbine Inlet Pressure Calibration Procedures During Implementation of High Pressure Turbine Replacement Design Modific | A self-revealing finding was identified for the licensees failure to follow the requirements of the station modification program manual EDM 601 during implementation of a high pressure turbine replacement modification revision. This resulted in Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC) calibration procedures not being revised with the proper setpoints. The performance deficiency (PD) was more than minor because it affected the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective in that AMSAC actuated causing a turbine trip. The finding was determined to have very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The cause of this finding was related to the cross-cutting aspect of the need for work groups to maintain appropriate interfaces and communicate, coordinate with each other during important work activities as described in the Work Control component of the Human Performance cross-cutting area because necessary revisions to the AMSAC input device calibration procedures were not adequately communicated. |
| 05000369/FIN-2013007-02 | 31 March 2013 23:59:59 | Mcguire McGuire | NRC identified | Appendix R Non- Compliance Could Have Potentially Affected Cold Shutdown (EA-13-050) | McGuire Operating License condition 2.C.(4) for Units 1 & 2, state, in part, that Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as approved in the NRC SER (NUREG - 0422) Supplement 5, dated April 1981. SER Supplement 5 stated, in part, that, in a letter dated January 9, 1981, the licensee committed to meet the requirements of 10 CFR 50, Appendix R, Section III.G. 10 CFR 50, Appendix R, Section III.G.1 stated, in part, that fire protection features shall be provided for structures, systems, and components important to safe shutdown. These features shall be capable of limiting fire damage so that systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station(s) can be repaired within 72 hours. Contrary to the above, from initial plant operation through March 22, 2012, the licensee failed to implement all provisions of the McGuire approved FPP in that features were not provided that were capable of limiting fire damage such that systems necessary to achieve and maintain cold shutdown could be repaired within 72 hours. Specifically, the licensee discovered that since initial plant operation, for fires in FA 16-18 and 15-17, circuits for the SG PORV block valves were not adequately protected, and as a result, postulated spurious actuation of the valves could cause valve damage. These valves would be unable to be repaired within 72 hours. The licensee entered this finding into the corrective action program (CR 12-02194) and the licensee has taken appropriate compensatory measures. Because the licensee committed to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c), the NRC is exercising enforcement and reactor oversight process (ROP) discretion for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, this issue was identified and will be addressed during the licensees transition to NFPA 805, was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, was not likely to have been previously identified by routine licensee efforts, was not willful, and it was not associated with a finding of high safety significance. |
| 05000369/FIN-2012005-02 | 31 December 2012 23:59:59 | Mcguire McGuire | NRC identified | Evaluation of the Occupational Radiation Dose Assigned to a Worker from a Piece of Contaminated Wire | While working in the reactor building an individual received a puncture wound in their hand from a piece of contaminated wire. Licensee attempts to decontaminate the wound were unsuccessful and the radioactive material from the contaminated wire remained inside the individuals hand. The licensee was reviewing that data and determining what dose to assign to the individual. The NRC will review the methodologies used once the licensee has completed its assessment to determine if a violation of regulatory requirements existed. This issue is identified as URI 05000369,370/2012005-02, Evaluation of the Occupational Radiation Dose Assigned to a Worker from a Piece of Contaminated Wire. |
| 05000369/FIN-2012005-03 | 31 December 2012 23:59:59 | Mcguire McGuire | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.7, High Radiation Area, required areas with radiation levels greater than 1,000 millirem (mrem) per hour at 30 centimeters (cm) from the radiation source or from any surface which the radiation penetrates to be provided with locked or continuously guarded doors to prevent unauthorized entry. Contrary to the above, on September 23, 2011, an area with radiation levels greater than 1,000 mrem per hour at 30 cm from the radiation source or from any surface which the radiation penetrates was not locked or continuously guarded to prevent unauthorized entry. The locking method for a LHRA door leading to the reactor head stand did not prevent unauthorized entry. The padlock used to secure retaining bolts on the doors was supposed to be installed through openings in the bolts preventing them from being removed. Instead, the padlock was installed around the bolts allowing them to be removed. Corrective actions included identifying other HRA, LHRA, and VHRA barriers with the unique locking mechanism, photographing the proper locking method, providing proper instructions to individuals during key issuance, and clarifying procedural guidance on the proper use of the locking mechanism. The corrective actions were documented under PIP M-11-07009. The violation was evaluated using the Occupational Radiation Safety Significance Determination Process and was determined to be not more than very low safety significance (Green) because this finding did not have a substantial potential for over-exposure because of additional controls and warnings present such as personal ED alarming devices and LHRA posting. |
| 05000369/FIN-2012005-01 | 31 December 2012 23:59:59 | Mcguire McGuire | NRC identified | Failure to Maintain Complete and Accurate Pre-Fire Plans | An NRC-identified Green non-cited violation (NCV) of the Unit 2 Facility Operating License, Condition 2.C.4, Fire Protection Program, was identified for failure to maintain prefire plans in areas that contain safety-related equipment. The inspectors identified that all copies of fire strategy plan view for the Unit 2 lower annulus and containment were missing from their pre-fire plans and unavailable to the Fire Brigade Leader and Operations personnel in the event of a fire in the Unit 2 reactor building. Corrective actions included replacement of the missing fire strategy plan views and additional review of the fire strategy books located in the Fire Brigade Leaders Kit, Control Room, and Emergency Preparedness office. This violation was entered into the licensees corrective action program (CAP) as Problem Investigation Program (PIP) M-12-08270. The performance deficiency (PD) was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events (Fire) and adversely affected the cornerstone objective, in that, it degraded the manual fire suppression capability. The finding was determined to be of very low safety significance (Green) because the fire brigade consisted of plant personnel familiar with the plant layout and associated fire hazards and appropriate fire-fighting equipment was available. The cause of the PD was directly related to the aspect of complete, accurate, and up-to-date procedures of the Resources Component in the cross-cutting area of Human Performance because the Fire Brigade Program Administrator failed to include all approved plan view updates into the fire brigade response strategies. |
| 05000369/FIN-2012008-01 | 30 September 2012 23:59:59 | Mcguire McGuire | NRC identified | Failure to Perform Required Extent of Condition Assessments for Quick Cause Evaluations in accordance with McGuires Quality Assurance Program | A finding of very low safety significance and associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the licensees failure to perform required extent of condition assessments for Quick Cause Evaluations (QCE) in accordance with McGuires Quality Assurance Program. Specifically, Nuclear System Directive (NSD) 212, Cause Analysis, requires in part that an Extent of Condition review shall be conducted as soon as possible when a QCE is performed. One example included the licensees failure to perform an extent of condition assessment for a QCE of the safety-related NSW system. To address this issue, the license entered PIP M-12-6309 into their CAP. The failure to perform the required extent of condition assessments for QCE in accordance with NSD 212 was considered a performance deficiency. The finding was determined to be more than minor because it adversely affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to evaluate events for extent of condition applicability for the Nuclear Service Water issue (PIP M-12-0106) was not only a failure to follow a procedure requirement, but allowed the station to be susceptible to the existence of similar discrepancies in other systems, units, organizations, programs, processes, components, or trains. The finding was determined to be of very low safety significance (Green) because the finding did not result in a loss of system safety function or a loss of safety function of a single train for greater than allowed technical specification allowed outage time. The team identified a cross-cutting aspect in the work practices component of the Human Performance area, because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures. |
| 05000369/FIN-2012004-01 | 30 September 2012 23:59:59 | Mcguire McGuire | NRC identified | Failure to Correctly Implement Technical Specifications Adversely Affects Requalification Operating Test Quality | An NRC-identified finding was identified associated with the quality of the simulator scenarios developed by the licensee for the licensed operator requalification annual operating test. The licensee failed to follow the Technical Specification (TS) rules of usage for concurrent inoperability as shown in TS Example 1.3-3. The licensee entered this issue into their corrective action program (CAP) as PIP M-12-4157. The performance deficiency (PD) was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective in that it impacted the licensees ability to evaluate and ensure operator performance. The significance determination was performed in accordance with Manual Chapter 0609, Appendix I, and determined to be of very low safety significance (Green). The cause of the finding was directly related to the crosscutting aspect of personnel training and qualifications in the Resources component of the cross-cutting area of Human Performance, in that the licensee failed to ensure the quality of the operating tests used to evaluate the knowledge, skills, abilities, and training provided to operators to assure nuclear safety. |