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05000280/FIN-2018003-0130 September 2018 23:59:59SurryNRC identifiedFailure to Control a Modification on the Containment Spray SystemAn NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to control the modification for installation of a drain line on the Unit 1 A Containment Spray (CS) pump bearing housing. This resulted in the drain line being blocked by boric acid and inoperability of the Unit 1 A CS pump.
05000281/FIN-2018002-0130 June 2018 23:59:59SurryNRC identifiedFailure to Follow Preventative Maintenance Procedure Results in Additional Failures of Emergency Bus Undervoltage and Degraded Voltage RelaysAn NRC-identified Green NCV of Surry Technical Specification (TS) 6.4.D was identified for the failure to follow procedure ER-AA-102, Preventative Maintenance Program, which resulted in the Unit 2 H emergency bus degraded voltage (DV) relay failure on March 13, 2018, and the Unit 2 J emergency bus under voltage (UV) relay failure on May 21, 2018, while the unit was operating at rated thermal power (RTP).
05000280/FIN-2018010-0131 March 2018 23:59:59SurryNRC identifiedFailure to implement the 10 CFR, Part 50, Appendix R, III.G.3 requirements consistent with fire protection license condition 3I.The NRC identified a Green finding and associated non-cited violation (NCV) of the requirements consistent with license condition 3.I, Surry Units 1 and Unit 2. Specifically, the licensee failed to adequately protect fiberglass pipe that is susceptible to fire damage and required for safe shutdown. By not protecting the pipe, the licensee did not ensure the alternative shutdown methodology was implemented with the independence as defined by the 10 CFR 50 Appendix R section III.G.3 requirements.
05000280/FIN-2017004-0131 December 2017 23:59:59SurrySelf-revealingInadequate Instructions for Corrective Maintenance on Unit 1 C RC Hot Leg Sample ValveA self-revealing, non-cited violation (NCV) of Surry Technical Specification (TS) 6.4.A.7 was identified for the failure to have detailed written procedures with appropriate instructions in design change packages (DCPs) 78-001, 80-007, and 84-369 when replacing 1-SS-HCV-101C, the Unit 1 C reactor coolant (RC) hot leg sample valve. This resulted in 1-SS-HCV-101C developing a through-wall leak on the tube to valve socket weld. Additionally, due to the reactor coolant system (RCS) boundary leakage, Unit 1 required an unplanned shutdown per TS 3.1.C.3 on August 9, 2017. This issue was documented in the licensees corrective action program (CAP) as condition report (CR) 1075404. During the shutdown, the licensee made an American Society of Mechanical Engineering (ASME) code repair by cutting and capping the tubing to stop the leak. 1-SS-HCV-101C will be restored to normal system configuration during the next refueling outage in April 2018.The inspectors determined that the failure of the licensee to have the instructions necessary to properly install the C RCS loop hot leg sample valve and tubing as required by Surry procedure SUI-0001 was a performance deficiency (PD). Specifically, DCPs 78-001,80-007, and 84-369 did not have instructions necessary to ensure the 1-SS-HCV-101C and the associated tubing was properly mounted to absorb the stresses applied to the valve and tubing during normal operation of the valve. As a consequence of the insufficient supports, 1-SS-HCV-101C experienced a through-wall leak on a socket weld on August 9, 2017, which subsequently required an unplanned shutdown of Unit 1. Using Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the PD was more than minor because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated October 7, 2016; the finding was determined to adversely affect the Initiating Events Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power dated June 19, 2012, and determined that it screened as Green because the deficiency did not cause a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding did not have a cross-cutting aspect because it is not considered current licensee performance.
05000280/FIN-2017004-0231 December 2017 23:59:59SurryNRC identifiedFailure to Identify a Non-Functional Flood Control BarrierAn NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI was identified for the licensees fa ilure to identify a condition adverse to quality related to the material condition of the machinery equipment room (MER) 5 flood dike. Specifically, the inspectors identified on November 13, 2017, several bolts on the connecting plates of the dike that were visually not flush, and found to be loose. As a result, the licensee declared the MER 5 flood dike non-functional and the D and E main control room (MCR) chillers inoperable. This issue was documented in the licensees CAP as CR 1083839. As immediate corrective action, the licensee torqued all structural bolts to 12 ft-lbs and floor anchor nuts to 55 ft-lbs per WO 38103865619.The inspectors determined that failure to identify a condition adverse to quality associated with the material condition of the MER 5 flood dike was a PD. Specifically, the inspectors identified on November 13, 2017, several loose bolts on the connecting plates of the MER 5 flood dike. As a result, the licensee declared the MER-5 flood dike non-functional and the D and E main control room (MCR) chillers inoperable. The inspectors determined that the PD was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that WO 38103734871 and drawing 11548-FC-6L had fastener torque specifications and a re-torque requirement for the MER 5 dike after it was re-assembled; and failed to identify a non-functional MER 5 flood dike. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012; the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using IMC0609, Appendix A, SDP for Findings at-Power dated June 19, 2012, the inspectors determined that a detailed risk evaluation was required. A detailed risk evaluation of the PD was performed in accordance with IMC 0609 Appendix A by a regional Senior ReactorAnalyst (SRA) using input from the licensees full scope Probabilistic Risk Assessment model. The result of the bounding analysis was an increase in core damage frequency due to the performance deficiency of <1E-6/year, a Green finding of very low safety significance.This finding has a cross-cutting aspect in the evaluation component of the problem identification and resolution area, P.2, because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, ETE SU-2017-0044, written for the May, 2017, non-functional MER 5 flood dike, did not thoroughly evaluate gasket type and bolting torque, when evaluating if epoxy was required for the assembly of the MER 5 flood dike.
05000280/FIN-2017007-0130 September 2017 23:59:59SurryNRC identifiedFailure to Evaluate Design Maximum Ambient Temperature Effect on Main Steam Valve HouseThe NRC identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly evaluate the heat-up of the Main Steam Valve House (MSVH), which contains the auxilliary feedwater pumps as well as other safety-related mitigating systems. The violation was entered into the licensees corrective action program as Condition Reports 1077007 and 1077684 and the licensee conducted a preliminary calculation and evaluation to determine the actual temperature increase and determined that the equipment located in MSVH 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 decreased margin for reliability and capability of mitigating systems contained in the MSVH. The inspectors 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, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance.
05000280/FIN-2017002-0130 June 2017 23:59:59SurryNRC identifiedFailure to Have Work Instructions Impacting MER 5 Flood BarrierAn NRC-identified, NCV of Surry Technical Specification (TS) 6.4.A.7 was identified because the mechanical equipment room (MER) 5 flood dike was not installed in accordance with the manufacturers installation procedures after it was removed for maintenance. Specifically, work order (WO) 38103734871, procedure GMP-013, Removal and Installation of Flood Protection Dikes and Secondary Flood Shields and Placing MER 3 in Extended Access, Revision 22, and drawing 11548-FC-6L, Flood Protection Dike Details MER 5 Turbine Building Unit 2, Revision 0, did not provide instructions, procedures, or drawing specifics that took into account the manufacturer instructions of using epoxy to ensure a water tight seal; and failed to use the materials as listed in drawing 11548-FC-6L during the reinstallation of MER 5 flood dike. The issue was documented in the licensees corrective action program (CAP) as condition reports (CR) 1068357, 1068357, and 1068528.The inspectors determined that not having and following work instructions and drawings appropriate to the reinstallation of MER 5 flood dike is a performance deficiency (PD). This PD is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, on May 2, 2017, the licensee failed to ensure WO 38103734871, procedure GMP-013, and drawing 11548-FC-6L had detailed manufacturer instructions to use epoxy to ensure a water tight seal and failed to use the materials as listed in drawing 11548-FC-6L. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power,dated June 19, 2012, and IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding is a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), in this case the main control room (MCR) chillers in MER 5, in which the SSC in question maintained its operability. This finding has a cross-cutting aspect in the area of human performance associated with teamwork, in that, individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, while preparing for and performing MER 5 flood dike reinstallation using WO 38103734871, procedure GMP-013, and drawing 11548-FC-6L, the licensee utilized a new foam material, but the differentdepartments in the organization (specifically Supply, Engineering, and Maintenance) failed to work together to evaluate the supplied manufacturer material and any specific requirements needed for installation (H.4)
05000280/FIN-2017404-0130 June 2017 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000280/FIN-2017008-0131 March 2017 23:59:59SurryNRC identifiedFailure to verify or check the adequacy of a design change in the Recirculation Spray Service Water Valve Pits.Green: The inspectors identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify or check the adequacy of the design of bulkheads in the recirculation spray service water motor operated valve pits. Specifically, the design allowed for unsealed penetrations in bulkheads and the licensee failed to demonstrate that the unsealed penetrations would not adversely affect the ability of the bulkheads to provide adequate train separation during a postulated pipe rupture. The licensee entered the issue into the CAP as Condition Report (CR) 1060189 and sealed the penetrations. This performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 3 capability to maintain train separation between the Recirculation Spray Service Water header motor operated valves was adversely affected due to the presence of degraded penetrations through the flood barriers. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.
05000280/FIN-2017001-0131 March 2017 23:59:59SurryNRC identifiedFailure to Maintain Requalification Examination IntegrityGreen . An NRC- identified NCV of 10 CFR 55.49, Integrity of examinations and tests, was identified for the licensees failure to adhere to the requirements of TR -AA -730, Licensed Operator Biennial and Annual Operating Requalification Exam Process, Revision 9. TR -AA- 730 was the procedure that the licensee used to implement industry standard ACAD 07- 001, Guidelines for the Continuing Training of Licensed Personnel. ACAD 07 -001 is a methodology which can be used to fulfill 10 CFR 55.59(c), Requalification program requirements and 10 CFR 55.4, Systems approach to training (SAT). This violation has been entered into the licensees corrective action program (CAP) as condition report (CR) 1058649. T he performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern with the administration of the operating exams. The inspectors assessed the significance in accordance with Manual Chapter 0609, Significance Determination Process, Appendix I, Licensed Operator Requalification Significance Determination Process (SDP). The finding was determined to be of very low safety significance (Green) because there was no evidence that a licensed operator had actually gained an unfair advantage on an examination required by 10 CFR 55.59. The finding was directly related to the cross -cutting aspect of Complacency in the cross -cutting area of Human Performance because the training staff was aware of the TR -AA -730 requirements for annual operating exam scenario overlap, but justified an alternative method of exam security that was used in the past. (H.12)
05000280/FIN-2016004-0131 December 2016 23:59:59SurrySelf-revealingChange of Surveillance Frequency Caused the Charging Service Water Header to Become Biologically FouledA self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI was identified because the surveillance procedure frequency used to flush the service water (SW) piping in Mechanical Equipment Room (MER)-3 and MER-4 was changed from two weeks to four weeks without sufficiently considering the effects of river conditions on biological growth and without getting management permission to change the periodicity. As a result of the periodicity change, the B charging (CH) and main control room (MCR) SW header became blocked with biological growth and was declared inoperable on September 22, 2016, during the performance of 0-OSP-VS-012, High Flow Flush of SW Strainers and Piping in MER 3 and MER 4. As immediate corrective action, the licensee cleaned the clogged SW strainer and completed the backflushing of the SW header. The SW flushing periodicity was restored to a two week frequency to be seasonally and risk assessed and reduced as heavy fouling season ends. This issue was documented in the licensees corrective action program (CAP) as CR 1048251. 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 because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not affect the design or qualification of the charging pump service water pump system and it did not represent a loss of system safety function. This finding has a cross-cutting aspect in conservative bias aspect of the human performance area, H.14, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowed.
05000281/FIN-2016004-0231 December 2016 23:59:59SurrySelf-revealingInadequate Design Change Post Maintenance Testing Causes Water Intrusion into Station Service Transformer and a Reactor TripA self-revealing finding was identified because the test requirements section of the station service transformer (SST) design change (DC) was not comprehensive in that it did not test that the isolated phase bus ducting terminal boxes were constructed to prevent water intrusion into the boxes. This was discovered during a significant rainfall event partially caused by Hurricane Matthew, which filled up the A SST terminal box with water and eventually shorted the A phase of the main generator causing a Unit 2 main generator, main turbine, and subsequent reactor trip on October 9, 2016. As corrective action, sealant was applied to the SST terminal boxes on all seams and bolt holes; and weep holes with drain assemblies were installed on each box. This issue was documented in the licensees CAP as CR 1049987. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined the PD was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone, and it 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. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated October 7, 2016, the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because although the deficiency did cause a reactor trip, it did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the Operating Experience aspect of the Problem Identification and Resolution area, P.5, because the licensee did not evaluate and implement relevant external operating experience.
05000280/FIN-2016003-0130 September 2016 23:59:59SurryLicensee-identifiedLicensee-Identified ViolationThe licensee identified a non-cited violation of very low safety significance of 10 CFR 72.150,Instructions, Procedures, and Drawings. Title 10 CFR 72.150, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established PBF5101, Fuel/Insert/Component Movement Authorization, Revision 17, as the implementing procedure for dry fuel storage fuel loading, an activity affecting quality. Procedure PBF5101 contains instructions for fuel handlers to move specific fuel assemblies from specific spent fuel pool locations into specific dry shielded canisters (DSCs) and DSC locations. Contrary to the above, on July 18, 2016, the licensee failed to follow PBF5101. Specifically, the licensee was utilizing a PBF5101 labeled for DSC25 during the loading of DSC24. This resulted in three fuel assemblies being incorrectly loaded into DSC24. The license entered the issue into its corrective action program under AR 02144237, dated July 18, 2016, and initiated actions to perform an apparent causal evaluation. The inspectors identified that DSC24 and DSC25 have identical design characteristics and therefore there was no actual safety significance to this event. Consistent with the guidance in Section 2.2 of the NRC Enforcement Policy, ISFSIs are not subject to the Significance Determination Process and, thus, traditional enforcement will be used for this issue. However, the inspectors determined that the violation significance could be informed by the significance determination process as no similar violations existed in the enforcement policy violations examples. The inspectors determined that the violation could be evaluated using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 3, Barrier Integrity Screening Questions. This resulted in the violation screening as Severity Level IV.
05000280/FIN-2016001-0131 March 2016 23:59:59SurryNRC identifiedFailure to Perform a 10 CFR 50.59 Evaluation for Blocking Ventilation to Main Steam Valve HousesAn NRC-identified finding of very low safety significance and an associated Severity Level IV NCV of 10 CFR 50.59, Changes, Tests, and Experiments, was identified when the licensee failed to perform and maintain a written evaluation to demonstrate that a procedure change did not require a license amendment. Specifically, the licensee implemented a change to procedure 0-OP-ZZ-021, Severe Weather Preparation, Revision 12, to allow installation of tarpaulins over the main steam valve house (MSVH) ventilation louvers thereby changing the Updated Final Safety Analysis Report (UFSAR) facility design without maintaining supporting calculations. The licensees failure to perform a 10 CFR 50.59 evaluation was a performance deficiency (PD). The inspectors determined that the PD 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. Specifically, the change allowed the ventilation of the MSVH to be blocked and the lack of engineering calculations resulted in a condition where there was a reasonable doubt about the operability of the auxiliary feedwater (AFW) pumps for their required mission time. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012; the finding was determined to adversely affect the Mitigating Systems Cornerstone. 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 because the PD did not affect the design or qualification of the AFW system and it did not represent an actual loss of system safety function. Using IMC 0310, Aspects within the Cross-Cutting Areas, dated December 4, 2014, the inspectors determined that the finding had a cross-cutting aspect in the procedure adherence component of the human performance area, H.8, because the licensee failed to follow processes, procedures and work instructions for the 50.59 applicability review when changing the severe weather preparation procedure. Additionally, the failure to perform a 10 CFR 50.59 evaluation was determined to be more-than-minor in accordance with the guidance in the NRC Enforcement Manual for traditional enforcement violations, because the MSVH louvers were actually covered and there was a reasonable likelihood that the lack of MSVH ventilation could affect the operability of the AFW pumps for their required mission time. The failure constitutes a violation of 10 CFR 50.59, which impacts the regulatory process and therefore, was evaluated through the traditional enforcement process. The SDP, which was used to evaluate this performance deficiency, does not specifically consider the impact on the regulatory process. Thus, although related to a common regulatory concern, it is necessary to address both the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated performance deficiency.
05000281/FIN-2015004-0131 December 2015 23:59:59SurrySelf-revealingInsufficient Gasket Crush on Pressurizer Spray Valve Body to Bonnet JointA self-revealing, Green non-cited violation (NCV) of Surry Technical Specification (TS) 6.4.A.7 was identified because 2-RC-PCV-2455A, the Unit 2 A pressurizer (PZR) spray valve, developed a body to bonnet mechanical joint leak as a result of the failure of the joint upper gasket to adequately seal the joint. The gasket inadequately sealed the body to bonnet joint due to a misalignment of the cage and the cage spacer assembly with the valve body. This misalignment caused the reactor coolant system (RCS) allowable unidentified leak rate to approach the TS limit on July 13, 2015, and subsequently required an unplanned Unit 2 shutdown. This issue was documented in the licensees corrective action program (CAP) as condition report (CR) 1002302. The inspectors concluded that the failure of the licensee to have the instructions necessary to successfully accomplish the purpose of 0-MCM-0414-13, Copes-Vulcan 4 inch, 1500 pound Control Valve, Model D-1000 with Bellows Overhaul, Revision 3, as required by Dominion procedure SPAP-0504, Technical Procedure Writers Guide, Revision 9, and to correctly measure and resolve the upper gasket crush on A PZR spray valve, was a performance deficiency (PD). Using IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the PD was more than minor because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012; the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not cause a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding has a cross-cutting aspect in the consistent process aspect of the human performance area, H.13, because the licensee did not use a systematic approach to evaluate all available data in deciding to return the A PZR spray valve to service during the spring 2014 refueling outage (RFO).
05000281/FIN-2015004-0231 December 2015 23:59:59SurrySelf-revealingInadequate Testing Procedure Causes an Emergency Bus to DeenergizeA self-revealing, Green NCV of Surry TS 6.4.A.7 was identified because the Unit 2 H emergency bus was lost during performance of 2-PT-2.33A, Emergency Bus Undervoltage and Degraded Protection Test H Train, on September 16, 2015. An inadequate procedure allowed steps in the procedure to continue without verification that a tripped relay had not reset. Specifically, 2-PT-2.33A did not have instructions necessary to validate the state of the normally energized undervoltage (UV) relays once power was restored to the relay. This allowed an UV relay to remain in a deenergized state when the next relay was tested. As a consequence, the two of three coincidence was met for the Unit 2 H emergency bus to deenergize and automatically start and load the #2 emergency diesel generator (EDG) onto the Unit 2 H bus. This issue was documented in the licensees CAP as CR 1009999. The inspectors concluded that the failure of the licensee to have the instructions necessary to successfully accomplish the purpose of 2-PT-2.33A, as required by Dominion procedure SPAP-0504, was a PD. Using IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was more than minor because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012; the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not involve the complete or partial loss of a support system that contributes to the likelihood, or cause, an initiating event and affected mitigation equipment. This finding has a cross-cutting aspect in the documentation aspect of the human performance area, H.7, because the licensee did not create and maintain complete and accurate documentation to validate that an emergency bus UV relay had been restored to its normal energized state during testing.
05000280/FIN-2015004-0331 December 2015 23:59:59SurrySelf-revealingCharging Pump Service Water Pump Failure due to Inadequate Preventative MaintenanceA self-revealing Green NCV of Surry TS 6.4.D was identified because the preventative maintenance cleaning of the six inch service water (SW) piping upstream of the SW rotating strainers was deferred with insufficient technical justification. Specifically, the licensee did not follow procedure ER-AA-PRS-1010, Preventative Maintenance Task Basis & Maintenance Strategy, and provide justification for a differing disposition when they deferred the cleaning of the six inch SW header three times. A lack of maintenance on this piping allowed excessive biofouling and subsequent blockage of the SW rotating strainer to occur. This was discovered when the Unit 1 and 2 A charging service water (CHSW) pumps experienced a zero flow rate during performance of 0-OPT-VS-001, Control Room Air Conditioning System Pump and Valve Inservice Testing, Revision 43, on July 24, 2015. This issue was documented in the licensees CAP as CR 1003878. The inspectors concluded that the failure of the licensee to provide technical justification to defer the preventative maintenance of the six inch SW header was a PD. Using IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the PD was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not affect the design or qualification of the charging pump service water pump system and it did not represent a loss of system safety function. This finding has a cross-cutting aspect in work management aspect of the human performance area, H.5, because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, ER-AA-102, Operability Determination, Revision 15 was not followed to ensure the management of risk commensurate to the work and the need for coordination with different groups was obtained.
05000281/FIN-2015004-0431 December 2015 23:59:59SurrySelf-revealingInadequate Procedure Causes Main Turbine and Reactor TripA self-revealing Green NCV of Surry TS 6.4.A.7 was identified because Unit 2 tripped during performance of 2-OP-TM-001, Turbine Generator Startup to 20% - 25% Turbine Power, on July 21, 2015. An inadequate procedure allowed the main turbine (MT) governor valves to open rapidly during MT overspeed protection controller (OPC) testing, increasing MT first stage pressure above the P-2 and P-7 reactor protection system (RPS) permissive step points, and subsequently causing a reactor trip. Specifically, 2-OP-TM-001 did not have the minimum level of information needed to ensure that there was no speed error between MT speed and the setter position before initiating the OPC test. This allowed the test to be conducted with a speed error that caused the governor valves to open rapidly at the end of the test and subsequently cause a reactor trip. This issue was documented in the licensees CAP as CR 1003328. The inspectors concluded that the failure of the licensee to have the minimum level of information needed to ensure task critical actions in 2-OP-TM-001 and for operators to avoid error traps in conducting the MT OPC test, as required by Dominion procedure SPAP-0504, was a PD. Using IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was more than minor because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012; the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, SDP for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not involve the complete or partial loss of a support system that contributes to the likelihood, or cause, an initiating event and affected mitigation equipment. This finding has a cross-cutting aspect in the documentation aspect of the human performance area, H.7, because the licensee did not create a complete procedure for testing the MT overspeed protection.
05000280/FIN-2015003-0130 September 2015 23:59:59SurrySelf-revealingCharging Pump Cubicle Floor Drain Backflow Preventer Failures during Unit 1 Safeguards Building FloodingA self-revealing, Green finding was identified because the instructions section of the procedure used to test floor drain back flow preventers (BFPs) did not include the instructions necessary to successfully fulfill the purpose of the procedure. A lack of testing methodology instructions allowed BFPs to be installed in the Unit 1 and Unit 2 charging (CH) pump cubicle floor drains that would not prevent backflow into the cubicles during low flow conditions. This was discovered when the Unit 1 and 2 CH pump cubicles filled with approximately two inches of water during the Unit 1 Safeguards building basement flooding event on May 20, 2015. This issue was documented in the licensees corrective action program (CAP) as condition reports (CRs) 580231 through 242. The inspectors concluded that the failure of the licensee to have the instructions necessary to successfully fulfill the purpose of 0-MPM-1900-02, Flood Protection Floor Drain Back Water Stop Valve Replacement as required by Dominion procedure SPAP-0504, Technical Procedure Writers Guide, and to correctly test the CH pump cubicle floor drain BFPs to prove functionality, was a performance deficiency (PD). Specifically, 0-MPM-1900-2 did not have instructions on the flow rate to fill the test stand and to observe that the BFP seats at a specified flow rate. Using IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the lack of complete testing instructions for BFPs allowed BFPs to be installed in the CH pump cubicle floor drains that would not seal during all flooding scenarios; and once cocked to the side during low flow, then had the potential to pass much higher flow rates into the CH pump cubicles. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency involved the degradation of equipment specifically designed to mitigate a flooding initiating event, but did not involve the total loss of any safety function. This finding has a cross-cutting aspect in the documentation aspect of the human performance area, H.7, because the licensee did not have an adequate test procedure to ensure that the floor drain BFPs would seal during low flow backflow conditions.
05000280/FIN-2015003-0230 September 2015 23:59:59SurrySelf-revealingFailure to Follow Procedure during Maintenance Results in Service Water Header InoperabilityA self-revealing, Green NCV of Technical Specifications (TS) 6.4.D was identified for failure to follow procedure WM-AA-101, Work Order Planning, Revision 1. Specifically, the licensee inappropriately revised a work order which resulted in the actuator and hand wheel assembly on 1-SW-495, the 1D Service Water (SW) header inlet isolation valve, being rotated incorrectly. The incorrect rotation resulted in the 1D SW header being inoperable from November 19, 2013, the time the 1D SW header was placed in service following 1-SW-495 replacement, until the issue was corrected on April 11, 2014. This issue was documented in the licensees CAP as CR 544361 The inspectors determined that the failure to follow procedure WM-AA-101, Work Order Planning, Revision 1, was a performance deficiency that was within the licensees ability to foresee and correct and should have been prevented. The inspectors determined that the finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the rotation of the actuator and hand wheel assembly of 1-SW-495 resulted in the inoperability of the 1D SW header from November 19, 2013 until April 11, 2014. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, and IMC 0609 Appendix A, SDP for Findings at-Power, dated June 19, 2012, the inspectors determined that a detailed risk evaluation was required because the finding represented an actual loss of system function for greater than the TS allowed outage time for both the main control room (MCR) air conditioning system and the charging SW system during the two periods where only one SW header was operable. The finding had a cross-cutting aspect in human performance, work management, H.5, because the organization did not appropriately control or implement the maintenance activity associated with 1-SW-495 and also did not identify the need for coordination with other groups when the scope of the planned work was changed.
05000280/FIN-2015003-0330 September 2015 23:59:59SurryNRC identifiedFailure to Verify Adequacy of Class 1E 125VDC Branch Circuit Breaker DesignThe team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify or check the adequacy of design of the Class 1E 125 volt direct current (VDC) molded case circuit breakers (MCCBs). The licensee entered the issue into their CAP as CRs 559872 and 59875 and performed an immediate determination of operability, which determined the Class 1E 125VDC switchgear to be operable. The licensees failure to assure the quality levels of MCCBs through the specification of requirements known to promote high quality, such as requirements for design, for the de-rating of components, for manufacturing, quality control, inspection, calibration, and test, as specified by IEEE 279, Section 4.3, was a performance deficiency. 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 adequately assess the electrical rating of electrical components could prevent the Class 1E 125VDC circuits from performing their safety 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, or component, which 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.
05000280/FIN-2015003-0430 September 2015 23:59:59SurryLicensee-identifiedLicensee-Identified Violation10 CFR 50, Appendix B, Criterion III requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those SSCs are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, on January 27, 2015, the licensee discovered that abnormal procedure, 0-AP-37.01, Abnormal Environmental Conditions, used when there is a tornado watch or warning declared for Surry County or when hurricane force winds are expected in Surry County within 36 hours, did not have specific steps to shut the four total sliding missile shield doors on the Unit 1 and Unit 2 MSVHs. The shields are necessary to meet the design function of the MSVH for protection of the equipment inside the MSVH which includes the AFW pumps and other safety-related components in the main steam and AFW systems. This issue was discovered during a procedure revision walk-through. Using IMC 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, and IMC 0609 Appendix A, Significance Determination Process for Findings at-Power, dated June 19, 2012, the inspectors determined that a detailed risk evaluation was required because the finding could involve the total loss of any safety function, identified by the licensee through probability risk analysis (PRA) that contributes to external event initiated core damage accident sequences (i.e., severe weather event). A detailed risk assessment was performed by a regional SRA in accordance with NRC IMC 0609 Appendix A using the NRC Surry SPAR model. The major analysis assumptions included: a one year exposure period, the performance deficiency was modelled as a non-recoverable weather-related loss of offsite power (LOOP) with the Station Blackout DG and all AFW pumps on one unit failed, damage assumed if F2-F5 tornado winds occurred within the 100 square mile radius including the site, and no recovery credit for AFW or for closing the missile shield doors prior to damage. The dominant sequence was a success of the reactor protection system and the electric power system, late failure of AFW and failure of feed and bleed. The risk was mitigated by the low frequency of events requiring use of the sliding missile shields and the remaining mitigation equipment including the AFW unit cross-tie. The result of the risk evaluation was an increase in core damage frequency of <1.0E-6/year, a GREEN finding of very low safety significance. This issue was entered into the licensees CAP as CR 570365 and the abnormal procedure 0-AP-37.01 was revised with the correct operator actions.
05000280/FIN-2015008-0330 June 2015 23:59:59SurryNRC identifiedFailure to Perform Required 50.59 Evaluations and Failure to Update the UFSAR for Plant Changes Associated with RCP Seal Cooling During Fire EventsThe inspectors identified a Green NCV of 10 CFR 50.59 and 10 CFR 50.71(e) for the licensees failure to perform 50.59 evaluations; and failure to update the UFSAR for plant changes associated with reactor coolant pump (RCP) seal cooling during fire events. The licensee entered this issue into their corrective action program as condition report CRs 5813388. The licensees revision of fire safe shut down procedures; and the installation of a different reactor coolant pump seal package without completing the required 50.59 evaluations was a performance deficiency. Additionally, the licensees failure to update the UFSAR as required by 10 CFR 50.71(e) was a performance deficiency. The UFSAR did not adequately describe the charging cross-tie function; and did not adequately describe the fire protection programs procedural isolation of the RCP seals for the entire duration of an Appendix R event. In accordance with the Reactor Oversight Process, the performance deficiencies were more than minor because they were associated with the design control attribute of the Mitigating Systems Cornerstone. The performance deficiencies were also assessed using traditional enforcement because the NRCs ability to perform its regulatory function such as, license amendment reviews and inspections was affected. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green) because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. No cross cutting aspect was assigned because these performance deficiencies did not occur within the last three years.
05000280/FIN-2015008-0230 June 2015 23:59:59SurryNRC identifiedFailure to Implement In-service Testing and Inservice Inspections for Charging Cross-tie ComponentsThe inspectors identified a Green NCV of 10 CFR 50.55(a) for the licensees failure to implement in-service testing (IST) and in-service inspections (ISI) for charging cross-tie components. The licensee entered this issue into their corrective action program as CRs 581385 and 581386. The licensee failed to scope the charging cross-tie manual isolation valves and piping into the ISI and IST programs. This was a performance deficiency that resulted in the subsequent failure to perform ISI and IST activities required by the ASME OM Code-2004 and 10 CFR 50.55a(f) and (g). The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone. Specifically, the sites failure to perform required inspections and testing for charging cross-tie components, since 1989, resulted in a lack of reasonable assurance that the charging cross-tie function could perform its required function. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green) because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. No cross cutting aspect was assigned because the performance deficiency did not occur within the last three years.
05000280/FIN-2015201-0130 June 2015 23:59:59SurryNRC identifiedSecurity
05000280/FIN-2015008-0430 June 2015 23:59:59SurryNRC identifiedMultiple Design Deficiencies in the Fire Protection ProgramThe inspectors identified a Green NCV of Surrys Operating License, Condition 3.I, Fire Protection, for design control deficiencies in the fire protection program. The licensee entered this issue into their corrective action program as condition report CRs 581390. The licensees failure to adequately implement the design control requirements in the fire protection program as required by Topical Report, DOM-QA-1, Dominion Nuclear Facility Quality Assurance Program Description, Section 3.2, Design Control Program was a performance deficiency. The finding was more than minor because it was associated with the design control attribute and affected the Mitigating Systems cornerstone. Specifically, design control deficiencies resulted in a lack of assurance that the design control requirements were being adequately implemented within the fire protection program. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green) because it finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. No cross cutting aspect was assigned because the performance deficiency did not occur within the last three years.
05000280/FIN-2015201-0230 June 2015 23:59:59SurryNRC identifiedSecurity
05000280/FIN-2015002-0130 June 2015 23:59:59SurryNRC identifiedFailure to conduct a detailed visual examination of the concrete-liner interface for the Unit 1 containmentAn NRC-identified NCV of 10 CFR 50.55a, Codes and Standards, was identified for the licensees failure to conduct a detailed visual examination of the concrete-liner interface for the Unit 1 containment, per the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section XI, Subsection IWE 1241, Table IWE-2500-1, Category E-C, Item E 4.11. This issue was documented in the licensees CAP as CR 578448. The licensees failure to conduct a detailed visual examination of the concrete-liner interface of the Units 1 and 2 containment in accordance with the ASME BPVC Section XI, Subsection IWE 1241, Table IWE-2500-1, Category E-C, Item E 4.11, was a PD that was within the licensees ability to foresee and correct. Using Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the PD was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, detailed visual inspections of the containment metallic liner provides assurance that the liner remains capable of performing its intended safety function, and in the absence of such inspections, corrosive conditions could progress to challenge that capability. Using Manual Chapter 0609.04, Initial Characterization of Findings, dated June 19, 2012, the finding was determined to affect the Barrier Integrity Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that the finding was of very low safety-significance (Green) because the finding did not represent an actual open pathway in the physical integrity of the reactor containment. The team determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance.
05000280/FIN-2015002-0230 June 2015 23:59:59SurrySelf-revealingA MDAFW Pump Motor Outboard Bearing DamagedA self-revealing NCV of Surry Technical Specification (TS) 6.4.D was identified because the Unit 1 A motor driven auxiliary feedwater (MDAFW) pump motor outboard bearing thermocouple was improperly installed while installing a new motor on the MDAFW pump in November, 2013. The improper thermocouple installation in the bearing caused the bearing to fail while the pump was running on January 5, 2015. This issue was documented in the licensees corrective action program (CAP) as condition report (CR) 568663. The inspectors concluded that the failure of the licensee to use a procedure to remove and reinstall the A MDAFW pump motor thermocouples was a performance deficiency (PD). Using Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the PD was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and it 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 incorrect installation of the motor outboard bearing thermocouple eventually damaged the bearing and caused the A MDAFW pump to become inoperable. Using Manual Chapter 0609.04, Initial Characterization of Findings, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green because the deficiency did not affect the design or qualification of the AFW system and it did not represent a loss of system safety function. This finding has a cross-cutting aspect in the Challenge the Unknown aspect of the human performance area, H.11, because the individuals involved in removing and installing the thermocouples did not stop when faced with a work order that did not have the appropriate procedure reference for the action they were taking.
05000280/FIN-2015008-0130 June 2015 23:59:59SurryNRC identifiedFailure to Ensure a Functional Alternate Shutdown System Alignment during Appendix R Fire Events EventsThe inspectors identified a Green non-cited violation (NCV) of Surrys Operating License, Condition 3.I, Fire Protection, for the licensees failure to ensure a functional alternate safe shutdown flow path during an Appendix R fire. The licensee entered this issue into their corrective action program as condition report (CR) 580928. The licensees failure to ensure a functional alternate shutdown system alignment during an Appendix R fire event was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone. Specifically, Surry failed to implement appropriate corrective actions to mitigate the spurious closure and subsequent damage of more than one motor operated valve as identified in an engineering evaluation. The failure to re-open credited Appendix R MOV(s) would result in the loss of secondary heat removal and/or RCS make-up capability during Appendix R fire events. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green). A Region II senior risk analyst performed a bounding phase 3 analysis that determined the finding represented an increase in core damage frequency of < 1 E-6 /year. No cross cutting aspect was assigned because the performance deficiency did not occur within the last three years.
05000281/FIN-2015001-0131 March 2015 23:59:59SurryNRC identifiedFailure to Identify Charging Service Water Pipe LeakAn NRC-identified, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified because the licensee failed to promptly identify a condition adverse to quality associated with the material condition of the Unit 2 charging service water (CH/SW) piping. Specifically, the NRC resident inspectors identified a leak in the discharge piping of the Unit 2 A CH/SW pump on November 24, 2014. The licensee had previously identified a leak on the Unit 1 B CH/SW pump discharge piping on June 16, 2014. The issue was documented in the licensees corrective action program (CAP) as condition report (CR) 563166. The licensees failure to identify a condition adverse to quality associated with the material condition of the Unit 2 A CH/SW piping was a performance deficiency (PD). The inspectors determined that the PD was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not having compensatory actions in place for CH/SW Green Thread piping that has been prone to through-wall leaks, left the licensee susceptible to undetected leaks from the CH/SW piping systems. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to adversely affect the Mitigating Systems Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at- Power, dated June 19, 2012, and determined that it screened as Green because the PD did not affect the design or qualification of the CH/SW system and the leak rate did not represent an actual loss of system safety function. This finding has a cross-cutting aspect in the evaluation component of the problem identification and resolution, P.2, because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the license did not institute compensatory actions for a long-term corrective action on CH/SW piping that has had a recent history of developing through-wall leaks.
05000281/FIN-2014005-0131 December 2014 23:59:59SurryNRC identifiedUnit 2 Trip Due to Loose RPS Wire ConnectionAn NRC-identified, non-cited violation (NCV) of Surry Technical Specification (TS) 6.4, Unit Operating Procedures and Programs, Section A.7 was identified because Surry procedure 0-ECM-1801-01, Westinghouse Type BF BFD or NBFD65NR Relay Replacement did not include a torque value for the reactor protection system (RPS) relay terminal screws to a field wiring connection. Subsequently, Unit 2 tripped on October 13, 2014, when a field wire connection became loose from the terminal end of a RPS trip relay and caused a reactor trip breaker to open. The issue was documented in Surrys corrective action program (CAP) as condition report (CR) 561820. The licensees failure to specify a torque value in procedure 0-ECM-1801-01 was a performance deficiency (PD) that was within the licensees ability to foresee and correct. Specifically, the licensee removed the correct torque value from the procedure based on a licensee procedure action request (PAR) that was incorrectly implemented. The inspectors determined that the PD was more than minor because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the procedure that controlled the connection of electrical termination to RPS relays did not specify a torque value and therefore, left it up to the technician to determine the tightness of the connection. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power dated June 19, 2012, and determined that it screened as Green because the deficiency did not cause a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding has a cross-cutting aspect in the documentation component of the human performance area, H.7, because the organization failed to maintain complete, accurate and up-to-date documentation for the replacement of RPS relays.
05000280/FIN-2014005-0231 December 2014 23:59:59SurryLicensee-identifiedLicensee-Identified ViolationSurry TS 6.4.A.3 requires, in part, that detailed written procedures with appropriate instructions shall be provided for conditions that include: action to be taken for specific and foreseen malfunctions of systems or components including alarms, primary system leaks and abnormal reactivity changes. Contrary to the above, on September 4, 2014, the licensee discovered that the annunciator response procedure, 0-VSP-M4, Flood Control Panel Trouble, used to mitigate ESGR flooding, did not have specific steps to direct an operator on how to isolate a leak. This was revealed when the licensee discovered that their current PRA model assumed an incorrect low flow rate for a break in SW piping in mechanical equipment room MER-3. Consequentially, the licensee had to take compensatory actions to isolate the SW flow path by shutting 1-SW-846, MER-3 chiller SW to Unit 1 discharge tunnel, until the PRA model was analyzed with the new flow rate and 0- VSP-M4 was changed. This finding is of very low safety significance (Green) because the completed PRA analysis did not affect the design or qualification of the SW system and it did not represent a loss of system or train safety function. This issue was entered into the licensees CAP as CR 557706 and the annunciator response procedure was revised with the correct operator actions.
05000280/FIN-2014005-0331 December 2014 23:59:59SurryLicensee-identifiedLicensee-Identified ViolationSurry TS 6.4.D requires, in part, that procedures described in section 6.4.A shall be followed. Surry TS 6.4.A.1 requires, in part, that detailed written procedures with appropriate check-off lists and instructions shall be provided for conditions which include: normal startup, operation, and shutdown of a unit, and of all systems and components involving nuclear safety of the station. These requirements are implemented, in part, by Dominion procedure 1-OP-RS-001A, Outside Recirculation Spray System Alignment, Revision 9, and independently verified, in part, by using Dominion procedure PI-AA-500, Verification Practices, Revision 3. Contrary to the above, on November 13, 2013, Dominion personnel failed to independently verify, in accordance with PI-AA-500, 1-RS-92 and 1-RS-97, the Unit 1 A OSRS pump seal tank inlet and outlet isolation valves were tie-wrapped open when performing procedure 1-OP-RS-001A. Consequentially, on November 9, 2014, the licensee found both of these valves tie-wrapped shut while performing a check of all Unit 1 valve locking devices. The licensee declared the Unit 1 A OSRS pump inoperable until the valves were repositioned opened. This finding is of very low safety significance (Green) because it did not affect the design or qualification of the recirculation spray system and it did not represent a loss of system or train safety function. This issue was entered into the licensees CAP as CR 564824.
05000280/FIN-2014007-0130 September 2014 23:59:59SurryNRC identifiedFailure to Perform Required Preventative Maintenance on Class 1E Molded Case Circuit BreakersThe team identified a Green non-cited violation of Technical Specification 6.4.A.7, Unit Operating Procedures and Programs, for the licensees failure to implement written procedures to perform periodic tests for the Class 1E 125 volt direct current thermal-magnetic molded case circuit breakers (MCCBs). The licensee entered the issue into their corrective action program as condition reports CR558445 and CR560488 and performed an immediate determination of operability, in which they determined that the MCCBs were operable but not fully qualified. The licensees failure to conduct periodic tests to detect the deterioration of the system and to demonstrate that components not exercised during normal operation of the station are operable, as required by IEEE 308-1970, Section 6.3, was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, absent testing to detect deterioration and to demonstrate continued operability, the likelihood that these MCCBs will unpredictably fail when called upon increases with time in service. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and Inspection Manual Chapter 0612, 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 or qualification of a mitigating structure, system, or component, which maintained its operability or functionality. The team determined that no crosscutting aspect was applicable because the finding was not indicative of current licensee performance.
05000280/FIN-2014007-0330 September 2014 23:59:59SurryNRC identifiedAdequacy of Class 1E 125VDC Branch Circuit Breaker DesignThe team identified an Unresolved Item (URI) regarding the adequacy of design of the Class 1E 125VDC power branch circuit breaker for the 1H 4160V Bus controls. The team reviewed the Class 1E 125VDC power distribution design to verify compliance with the licensing basis requirements in IEEE 308-1970, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations. The Surry licensing basis commitment to IEEE 308-1970 required the quality of the Class 1E power system design to be sufficient to ensure that multiple engineered safety features (ESF) would not lose power because of design vulnerabilities. Specifically IEEE 308-1970 stated, in part, The Class IE electric systems shall be designed to assure that any design basis event as listed in Table 1 will not cause: 1) A loss of electric power to a number of engineered safety features, surveillance devices, or protection system devices sufficient to jeopardize the safety of the plant. Table 1 stated, in part, that design basis events include Single act, event, component failure, or circuit fault that can cause multiple equipment malfunctions. The team identified design vulnerabilities in design basis documents and in the sampled branch circuitry. In Calculation EE-0499, DC Vital Bus Short Circuit Current, dated 11/30/1998, the licensee used AC power time current curve (TCC) data for HFB MCCBs (used in the 125VDC distribution system) instead of DC TCC data. In addition, in this calculation, the licensee did not de-rate components for the ambient temperature in the switchgear room. Furthermore, in 2009, the licensee replaced certain HFB MCCBs with model HFDDC MCCBs; however, did not evaluate the DC characteristics of these HFDDC MCCBs, and instead evaluated an AC model HFD MCCB. Because of these vulnerabilities the team questioned the coordination of the installed HFDDC breaker and whether it was adequate to protect the 1H branch circuit in the ambient temperature of the switchgear room. These calculational vulnerabilities were consistent across both trains A & B and for both Units 1 & 2. The licensee captured the inspectors questions in their corrective action program as CR559872 and CR559875. This issue is a URI pending further review of information provided by the licensee on November 4, 2014, and consultation with the Office of Nuclear Reactor Regulation to determine if this issue of concern constitutes a violation.
05000280/FIN-2014007-0230 September 2014 23:59:59SurryNRC identifiedFailure to Evaluate the Range of Conditions that Effect Canal Level ProbesThe team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate and quantify the system response times and accuracies over the range of conditions under which the service water canal level probes must operate. The licensee entered the issue into their corrective action program as condition report CR558429 and performed an immediate determination of operability, in which they determined the canal level probes to be operable but not fully qualified. The licensees failure to evaluate conditions that affected system response times and accuracy of the canal level probes, as required by IEEE 279-1968, Section 4.1, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, response time delays could allow the canal water level to fall below Technical Specification limits reducing the available heat removal required to mitigate Updated Final Safety Analysis Report chapter 14 design basis accidents. The team used Inspection Manual Chapter 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and Inspection Manual Chapter 0612, 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 or qualification of a mitigating structure, system, or component, which maintained its operability or functionality. The team determined that the finding was associated with the Design Margin cross-cutting aspect of the Human Performance area because recent modification designs for the canal probes were completed and approved without evaluating effects on the canal level probe response times and accuracies.
05000280/FIN-2014007-0430 September 2014 23:59:59SurryNRC identifiedQualification Basis for Safety-Related Molded Case Circuit BreakersThe team identified a URI regarding the licensees actions to maintain or extend the qualification basis for safety-related MCCBs installed in mild environments greater than vendor design life specifications. In 2004, the licensee received Westinghouse Electric Technical Bulletin TB-04-13, Replacement Solutions for Obsolete Classic MCCBs, UL (Underwriters Laboratory) Testing Issues, Breaker Design Life and Trip Band Adjustment, which was superseded in 2006 by TB-06-02, Aging Issues and Subsequent Operating Issues for Breakers That are at Their 20-Year Design/Qualified Lives; UL Certification/Testing Issues Update. These bulletins informed the licensee of MCCB aging and operating issues. Specifically, grease and red oil used in these breakers were found to be key limiting factors for continued operability within published specifications. As grease and red oil aged beyond 20 years, their lubrication properties were reduced, resulting in slower trip times beyond the published time-current curves. The bulletins further defined the design life of MCCBs in mild environments as 20 years. However, the inspectors noted that approximately 60 safety-related MCCBs installed in mild environments exceeded 20 years of service, and the licensee had not performed an engineering evaluation to justify continued operation beyond this design life. The affected MCCBs were associated with the Class 1E 125VDC distribution systems (switchgear) on both units. The licensee captured the inspectors questions in their corrective action program as CR558445 and CR560488. This issue is a URI pending further review, including consultation with the Office of Nuclear Reactor Regulation, to determine if this issue of concern constitutes a violation.
05000281/FIN-2014003-0130 June 2014 23:59:59SurrySelf-revealingInadequate Amount of Packing in Pressurizer Spray ValveA self-revealing NCV of Surry Technical Specification (TS) 6.4.A.7 was identified because 2-RC-PCV-2455A, the Unit 2 A pressurizer (PZR) spray valves packing gland was repacked with the incorrect number of packing rings in May, 2008. When the Unit 2 A PZR spray valve bellows failed in March 2014, the amount of packing in the valve was insufficient to prevent packing leakage. This leakage approached the technical specification (TS) allowable unidentified reactor coolant system (RCS) leak rate on March 19, 2014, and subsequently required an unplanned unit shutdown. The issue was documented in Surrys corrective action program (CAP) as CR 542547. The failure of the licensees packing control program to list the correct number of packing rings in the packing control form for the repack of 2-RC-PCV-2455A, the Unit 2 A PZR spray valve, was a performance deficiency that was within the licensees ability to foresee and correct. Specifically, the licensee did not thoroughly evaluate decreasing the number of packing rings from five to four when packing control was shifted from the PZR safety valve overhaul procedure to the licensees Packing Control Program. As a consequence of the inadequate number of packing rings, the Unit 2 A PZR spray valve experienced a packing leak that approached the TS allowable unidentified RCS leak rate on March 19, 2014, which subsequently required an unplanned shutdown of Unit 2. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset stability and challenge critical safety functions during shutdown as well as power operations. Specifically, an incorrect number of packing rings listed on the packing control form eventually allowed packing leakage to approach the TS limit. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012; the finding was determined to affect the Initiating Events Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at- Power dated June 19, 2012, and determined that it screened as Green because the deficiency did not cause a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. Because the PD occurred outside of the nominal three-year period for present performance, no cross-cutting aspect has been assigned.
05000280/FIN-2014403-0130 June 2014 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000280/FIN-2014405-0431 March 2014 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000280/FIN-2014405-0531 March 2014 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000280/FIN-2014405-0131 March 2014 23:59:59SurryNRC identifiedSecurity
05000280/FIN-2014405-0631 March 2014 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000280/FIN-2014405-0231 March 2014 23:59:59SurryNRC identifiedSecurity
05000280/FIN-2014405-0731 March 2014 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000281/FIN-2014002-0131 March 2014 23:59:59SurrySelf-revealingRecirculation Spray Heat Exchanger Inlet Isolation Valve MOV Thermal Overload Not Properly ResetA self-revealing NCV of Surry Technical Specification (TS) 6.4.A.7 was identified because 1-SW-MOV-103D, the B and C recirculation spray heat exchanger (RSHX) inlet isolation valve, motor thermal overload was improperly reset after planned maintenance and became disengaged on November 29, 2013, rendering one service water (SW) flow path of the B and C recirculation spray (RS) subsystem inoperable. The issue was documented in Surrys corrective action program (CAP) as CR 533932. The licensees failure to include acceptance criteria for determining if a thermal overload was properly reset was a performance deficiency (PD) that was within the licensees ability to foresee and correct. Specifically, an inadequate procedure did not have electricians verify that the trip indication flag in the thermal overload had fully cleared the viewing window or provide some other criteria for acceptance. The inspectors determined that the PD was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and it 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 motor thermal overload was improperly reset after planned maintenance which resulted in rendering one SW flow path of the B and C RS subsystem inoperable thereby affecting the availability of the RS subsystem. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors screened the finding using Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power dated June 19, 2012, and determined that it screened as Green because the deficiency did not affect the design or qualification of the RS system and it did not represent a loss of system safety function. This finding has a cross-cutting aspect in the Documentation aspect of the human performance area, H.7, because the licensee did not create and maintain a complete and accurate procedure to ensure that MCC thermal overloads were properly reset.
05000280/FIN-2014405-0331 March 2014 23:59:59SurryLicensee-identifiedLicensee-Identified Violation
05000280/FIN-2013005-0231 December 2013 23:59:59SurryNRC identifiedFailure to Missile Protect Beyond Design Bases FLEX Modification to Low Head Safety Injection PipingAn NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile. Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a design change (DC) and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 feet elevation in the safeguards valve pit and connected directly to the LHSI piping. The issue was documented in the licensees corrective action program (CAP) as condition report (CR) 533401. The licensees failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the diverse and flexible coping strategies (FLEX) mechanical piping connection as part of DC SU-12-00022 was a performance deficiency (PD) that was within the licensees ability to foresee and correct. The inspectors determined that the performance deficiency (PD) was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, Significance Determination Process (SDP) for Findings at-Power, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process.
05000280/FIN-2013005-0131 December 2013 23:59:59SurryNRC identifiedApplication of ASME Section XI, Table IWB 2500-1, Item B10.10, Inspection Requirements and Note 1 ExemptionsThe inspectors identified an unresolved item related to the inspection of the reactor pressure vessel (RPV) component supports as required by ASME BPVC Section XI, for which additional information is needed to determine if the issue of concern represents a performance deficiency or a violation of the regulatory requirements. The code of record for the current ISI program at Surry Power Station Unit 1 is the 1998 Edition of the ASME BPVC Section XI with the 2000 addenda. This Code edition includes inspection requirements for both nuclear class 1 piping and vessel supports (Subsection IWF) and their attachment welds (Subsection IWB). Subsection IWB, Table IWB-2500-1, item number B10.10, describes the examination requirements for welded attachments for vessels, piping, pumps, and valves. Note 1 of Table IWB- 2500-1 states that attachment welds (weld buildup) on nozzles that are in compression under normal load conditions and provide only component support are excluded from the surface examination requirements. The note also provides additional conditions to identify what type welded attachment configurations require inspection. Table IWB- 2500-1 also references Figures IWB-2500-13, -14 and -15 to further describe the examination requirements. The inspectors noted that the scope of the Surry Unit 1 ISI program for the inspection of the nuclear class 1 RPV supports did include the requirements for the IWF portion of the ASME Section XI code required inspections. However, the inspectors identified that the licensee excluded the surface examination requirements for the RPV support attachment welds required by Table IWB-2500-1, item number B10.10 based on the exemptions provided by Note 1 of the table. The licensees position was that the surface examinations are not required based on the exclusion criteria provided in Note 1 for attachment welds under compressive loads during normal conditions and the configurations described in Figures IWB-2500-13, -14 and -15. The inspectors reviewed design basis documents for the Unit 1 RPV supports and identified that the normal loading conditions of the supports included both compressive and shear loads. The inspectors determined that additional information and discussion with the NRC Office of Nuclear Reactor Regulation (NRR) staff was required, in order to determine if the licensees interpretation and implementation of the exemptions in Table IWB-2500-1 were in compliance with the ASME BPVC Section XI. Therefore, the NRR and Region II staff agreed to submit a Task Interface Agreement (TIA), which could involve the submittal of a formal inquiry to the applicable ASME BPVC committee to request an interpretation of the examination requirements and exemptions in Table IWB- 2500-1 for welded attachments for vessels and piping. The NRC initiated TIA-2014-02 to determine the staffs position on whether the configuration of the RPV supports at Surry meets the exclusion criteria in ASME BPVC Section XI. This issue remains unresolved until the resolution of TIA-2014-02 to determine if the issue of concern represents a performance deficiency or a violation of regulatory requirements. This issue is identified as URI 05000280/2013005-01, Application of ASME Section XI, Table IWB 2500-1, Item B10.10, Inspection Requirements and Note 1 Exemptions.