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	Quarter	Site	Title	Description
05000250/FIN-2010009-02	2010Q1	Turkey Point	Failure to report Unit 3 spent fuel pool operation with degrading Boraflex.	The inspectors identified an apparent violation of 10 CFR Part 50.73(a)(2)(B), when a condition prohibited by Technical Specifications was not reported to the NRC after testing of Boraflex panels in 2004 in the Unit 3 spent fuel pool revealed degradation greater than assumed in criticality analyses. Because the FPL program for determining degradation of cells was a sampling program, the state of other cells could not be determined. When identified to the licensee by the NRC, condition report 2009-30043 was written to evaluate and report the non-compliance with Technical Specifications to the NRC. The finding was more that minor because it impacted the regulatory process which depends on plant activities being properly reported. The inspectors evaluated this finding against NRC IMC 0609 Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones. The inspectors determined that IMC 0609, Appendix M is required to determine the level of safety significance of this finding because the existing SDP guidance is not adequate to provide reasonable estimates of the finding significance within the established SDP timeliness goal of 90 days. NRC staff is currently reviewing this finding to determine the level of safety significance or enforcement aspect of the issue. (4OA2) (IR# 05000250, 251/2009005 dated January 28, 2010) FPL failed to provide notification to the NRC in accordance with the requirements of 10 CFR 50.73 when testing and evaluation of Boraflex panels in the Unit 3 SFP racks revealed Boraflex degradation beyond minimum design values specified in the UFSAR. The NRC considers the failure to provide the required notification to be a significant matter because it impacted the NRCs ability to review and assess FPLs corrective actions for managing SFP Boraflex degradation. In accordance with the Enforcement Policy, a base civil penalty in the amount of $70,000 is considered for a Severity Level III violation. (IR# 05000250/2010009 dated June 21, 2010
05000250/FIN-2010009-03	2010Q1	Turkey Point	Failure to maintain FSAR description of Unit 3 spent fuel pool activities	The inspectors identified an apparent violation of 10 CFR Part 50.71(e) requirements to periodically update the final safety analysis report so that the report contains effects of changes made to the facility such that the FSAR is complete and accurate. As of December 2009, changes made to manage the Unit 3 spent fuel pool since 2001, including neutron attenuation testing methods and results, use of computer programs such as RACKLIFE, and the use of alternate means of assuring that the spent fuel remains shutdown, such as rod control cluster assembly inserts and water holes, were not described in the FSAR. When identified to the licensee by the inspectors, the licensee documented the condition in condition report 2009-34470, and informed the NRC (in letter L-2009-295, dated December 31, 2009) of plans to make appropriate updates to the FSAR descriptions by March 15, 2010. The finding was more that minor because it impacted the regulatory process which depends on plant activities being properly documented. The inspectors evaluated this finding against NRC IMC 0609 Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones. The inspectors determined that IMC 0609, Appendix M is required to determine the level of safety significance of this finding because the existing SDP guidance is not adequate to provide reasonable estimates of the finding significance within the established SDP timeliness goal of 90 days. NRC staff is currently reviewing this finding to determine the level of safety significance or enforcement aspect of the issue. (4OA2) (IR# 05000250, 251/2009005 dated January 28, 2010) A Non-cited Violation 05000250/201009-03 was identified for failure to update the FSAR in accordance with 10 CFR 50.71(e) so that the report accurately reflects significant changes made to the facility. (IR# 05000250/2010009 dated June 21, 2010).
05000250/FIN-2010010-01	2010Q4	Turkey Point	Licensee-Identified Violation	10 CFR Part 50.73(a)(2)(B), states that the licensee shall report (to the NRC), any condition which was prohibited by the plants technical specifications. Technical Specification 5.5.1.1.c requires in part that a nominal 9.0 inch center-to-center distance for the Region II storage rack cells shall be maintained. Contrary to the above, in 2005 a condition prohibited by Technical Specifications was not reported to the NRC after the licensee identified the top of the cell walls in four Unit 3 and two Unit 4 SFP storage cells were damaged and no longer maintained the required separation that met the nominal licensing basis center-to-center distance of 9.0 inches. This issue was determined to be low safety significance because these cells have never been used to store fuel since they are not accessible due to interference from the discharge piping. The affected and adjacent cells have been administratively removed from service. The analysis results confirmed that the Unit 3 storage rack had significantly more damage than the Unit 4 racks, however they were structurally adequate to maintain their integrity during seismic events even with the damaged cells. These flow-damaged cells were discussed and documented in condition reports (CRs) 2005-12609, 2005-33602, 2006-13611, 2008-7031, and 2010-6225. The licensee has a monitoring plan in the CAP program to inspect these damaged cells once a year during the fuel inventory to ensure these storage racks still are structurally adequate while developing a permanent solution.
05000261/FIN-2014002-05	2014Q1	Robinson	Defective Motor Operated Potentiometer causes failure of the DSDG during surveillance testing	An URI was identified regarding the trip of the DSDG, on December 31, 2013, during monthly surveillance testing. The URI is being opened to provide for additional inspection of the equipment issues that led to the failure and to review the results of the vendors analysis of a defective motor operated potentiometer to determine if a performance deficiency exists. On December 31, 2013, during monthly testing of the DSDG in accordance with licensee procedure OST-910, Dedicated Shutdown Diesel Generator (Monthly), the output breaker tripped open on overcurrent while the operators were attempting to adjust DSDG output voltage. Operators in the field noted erratic voltage indication prior to the failure. Engineering identified that the likely cause was a failure of the motor operated potentiometer (MOP). The licensee replaced the MOP with a new part from stock and performed post maintenance testing. The MOP that was removed was sent offsite for forensic analysis. During examination, the licensee identified a manufacturing defect for the MOP. The licensees extent of condition investigation found the same manufacturing defect on the MOP installed in the DSDG and in a MOP in storage. The licensee replaced the MOP in the DSDG with a MOP that was verified to be acceptable. Engineering has sent the defective components back to the vendor for additional analysis. This issue will be identified as URI 05000261/2014002-05; Defective Motor Operated Potentiometer causes failure of the DSDG during surveillance testing.
05000261/FIN-2014003-01	2014Q2	Robinson	Failure to Identify and Correct Degraded Wire Labels in the Reactor Protection Relay Cabinets	A self-revealing Green non-cited violation (NCV) was identified for the licensees failure to promptly identify and correct degraded wire labels in the reactor protection cabinets, which were a condition adverse to quality, as required by 10 CFR Part 50, Criterion XVI, Corrective Action. This resulted in an automatic reactor trip. Immediate corrective actions included inspection of both trains of relay racks to identify and remove any potential foreign material. The licensee also tested both trains of reactor protection relays to verify no foreign material was present. Additionally, the licensee plans to replace the wire labels in the reactor protection and safeguards relay racks during refueling outages 29 and 30. The licensee documented the issue in the corrective action program as CR 654789. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the degraded wire labels became lodged between contact 2-6 on relay LC-496A1-X(B), which set up the half-trip condition to cause a reactor trip, during the surveillance testing. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because although the finding caused a reactor trip, it did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a crosscutting aspect of identification in the area of problem identification and resolution because the licensee failed to implement a corrective action program with a low enough threshold for identifying issues in that the licensee process did not recognize, during review of the work requests for the degraded wire labels, that this issue should have been entered into the corrective action program as a nuclear condition report.
05000261/FIN-2014003-02	2014Q2	Robinson	Licensee-Identified Violation	Section 50.49 of 10 CFR, Environmental Qualification of electric equipment important to safety for nuclear power plants, states that each licensee shall establish a program for qualifying specified electric equipment. Section (a)(3) of 10 CFR 50.49 specifies the environmental qualification requirements for post-accident monitoring equipment. Section (f) of 10 CFR 50.49 requires, in part, that each item of electric equipment important to safety must be qualified by testing an identical item of equipment under identical conditions. Contrary to the above, since May 1992, the licensee failed to maintain the qualification of the limit switches for CVC-204B, letdown line isolation, in accordance with the tested configuration of the equipment which rendered the Post Accident Monitoring Instrumentation function inoperable. The licensee documented this condition in AR 640902 and AR 633207. The cause was determined to be associated with a human performance event in which the licensee failed to use the proper heat shrink insulators per procedure CM-309, Sealing Low Voltage Electrical Splices for Environmentally Qualified or Safety Related Splices. Following discovery of this condition, the licensee replaced the non-environmental qualified splice and returned the equipment to the test configuration. Using IMC 0609, Appendix A, issued June 19, 2012, The SDP for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not represent an actual loss of function of one or more non-Technical Specification Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours.
05000261/FIN-2014005-01	2014Q4	Robinson	Failure to Protect Diesel Driven Equipment from Effects of Extreme Cold Temperatures	The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1, for failure to establish procedural guidance to protect diesel driven equipment important to safety from the effects of extreme cold temperatures. Specifically, the licensees cold weather procedures failed to include actions to maintain fuel oil temperatures above the diesel fuel oil cloud point for the dedicated shutdown diesel generator (DSDG) and/or the engine driven fire pump (EDFP). The licensee entered this into the corrective action program (CAP) as AR 715032 and took immediate corrective actions to revise station procedures to protect the diesel driven equipment during periods of extreme low temperatures. The failure to establish procedural guidance to protect diesel-driven equipment important to safety from the effects of extreme cold temperatures was a performance deficiency. This issue was more than minor because if left uncorrected this finding would have the potential to lead to a more significant safety concern. Specifically, failure to maintain the fuel oil temperatures for the DSDG and/or the EDFP greater than the measured cloud point, may impact the operation of the equipment during extreme low temperature conditions, due to the associated fuel oil transfer system becoming non-functional. A detailed risk assessment was performed by a regional Senior Reactor Analyst in accordance with NRC IMC 0609 Appendices A and F. The latest NRC Robinson SPAR risk model was used to quantify the internal events risk and a calculation was performed to estimate the fire risk. The major analysis assumptions included: both the EDFP and the DSDG were simultaneously considered unavailable without recovery for a 1-day exposure interval, DSDG fire scenarios were considered for the emergency switchgear room (ESWGR), the cable spreading room, and the main control room, where fire could cause a loss of offsite power and the emergency diesel generators (EDGs), compartment total ignition frequency data from the Robinson NFPA 805 project was used and a bounding Conditional Core Damage Probability for the fire scenarios of 1.0. The dominant sequence was a fire in the ESWGR which remained unsuppressed long enough to cause a loss of offsite power and the EDGs requiring use of alternate shutdown which failed due to the performance deficiency impact on the DSDG resulting in station blackout, and core damage due to an unmitigated reactor coolant pump seal loss of cooling accident. The risk was mitigated by the low likelihood of the initiators occurring during the specific cold weather vulnerability periods. The risk due to the performance deficiency was determined to be an increase in core damage frequency of <1E-6/year, a GREEN finding of very low safety significance. The performance deficiency had a cross-cutting aspect of Evaluation in the area of Problem Identification and Resolution because the licensee failed to thoroughly evaluate the effects of cold weather on the fuel system for diesel driven equipment to ensure that resolutions address the extent of conditions commensurate with their safety significance (P.2).
05000261/FIN-2014008-01	2014Q2	Robinson	Failure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with the Steam Generator Tube Leak	The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take adequate corrective action to prevent repetition of a significant condition adverse to quality regarding steam generator tube leakage due to poor maintenance practices. Specifically, on February 27, 2014, the C steam generator showed indications of a primary to secondary tube leak due to foreign material that was introduced during the fall 2013 refueling outage. As immediate corrective actions, on March 7, 2014, the licensee shutdown the plant and repaired the leak. This violation was entered into the licensees CAP as nuclear condition reports (NCRs) 683695, 683593, and 683591. The licensees failure to implement appropriate corrective actions to address poor worker practices to prevent recurrence of a steam generator tube leak was a performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone equipment performance attribute 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. Specifically, foreign material entered the steam generator and damaged a steam generator tube, which increased the likelihood of a steam generator tube rupture. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At-Power, dated June 19, 2012. The finding screened as Green per Section D of Exhibit 1, Initiating Events Screening Questions, because testing showed that the affected steam generator tube could sustain three times the differential pressure across the tube during normal full power and that the steam generator did not violate the accident leakage performance criterion. The performance deficiency does not have a cross cutting aspect because the last revision of the root cause evaluation was completed in 2011 and it is not indicative of current licensee performance.
05000261/FIN-2016002-01	2016Q2	Robinson	Licensee-Identified Violation	Section 50.55a(h)(2) of 10 CFR states in part, for nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of Institute of Electrical and Electronic Engineers (IEEE) Std. 6031991 and the correction sheet dated January 30, 1995. The Robinson FSAR (current licensing basis) Section 3.1.2.20, states in part that, reactor protection is designed to meet all presently defined reactor protection criteria and is in accordance with the proposed Institute of Electrical and Electronic Engineers (IEEE) 279 Standard for Nuclear Plant Protection Systems, August 1968. IEEE-279, Section 4.2, requires that any single failure within the protection system shall not prevent proper protection system action when required. Contrary to this requirement, from initial startup, until April 13, 2016, when using a FRBV (i.e., FRBV in the open position in Modes 1, 2, and 3), and a MSLB occurred, the protection system would not provide the proper system protection action. Specifically, with a single failure of the FRBV to close, the protective system action to isolate feedwater could not be accomplished. This would cause an increase in secondary mass available for release in containment structure, resulting in a higher peak containment pressure that would challenge the containment design pressure. As corrective actions, the licensee implemented a standing instruction and placed caution tags on the FRBVs to ensure the valves remain closed/isolated while operating in Modes 1, 2, and 3. Additionally, the licensee completed an engineering change to update the containment analysis and licensing basis. The licensee entered this issue into the CAP as CRs 2012658, 2020495, and 2018710. The failure to meet the single failure criterion for feedwater isolation following a main steam line break inside containment was a performance deficiency (PD). Significance Determination Process (SDP) screening in accordance with NRC IMC 0609.04 determined that the PD affected the secondary short term heat removal safety function of the mitigating systems cornerstone. The finding was determined to represent a loss of function and a detailed risk assessment was performed per NRC IMC 0609 Appendix A. The bounding analysis assumed a conditional core damage probability of 1.0, a 14 day exposure period estimated from surveillance and outage schedules, and main steam line break inside containment (MSLBIC) initiating event probability and main feedwater regulating valve bypass (MFWRVBV) failure to close probabilities from the NRC SPAR model data. The dominant sequence was an MSLBIC with a failure to close of the MFWRVBV which was assumed to lead to core damage and large early release. The risk was mitigated by short exposure period and the low likelihood of the MSLBIC and the failure to close of the MFWRVBV. The bounding analysis determined that the PD represented a risk increase of < 1.0E-7/year, a GREEN finding of very low safety significance for both core damage frequency and large early release frequency.
05000280/FIN-2007005-04	2007Q4	Surry	Fibrous Material Left in Unit 1 Containment	On 11/28/07, during Unit 1 Containment Close-out walkdown, the inspectors identified that loose bat insulation had been placed in a 15\' X 5\' penetration in the \'C\' Loop Room. The insulation had not been found by the licensee during their containment readiness verification walkdown. When the inspectors notified the licensee, they removed two 55 gallon bags which were approximately 45 lbs of fibrous insulation. This issue was documented in the corrective action program as Condition Report CR025641. The walkdown was performed by the inspectors to verify that the containment walkdown conducted by the licensee was in accordance with procedural requirements. Later investigation found that the insulation had been there for a number of years. The inspectors reviewed the affected start-up procedure, 1-GOP-1.7 Rev 2, Unit Startup, RCS heatup from ambient to HSD and determined that even though the procedure had several steps in attachment 3 to ensure containment was clear of debris and fibrous material, the associated licensee walkdown failed to reveal the presence of the loose insulation. The licensee performed a more thorough walkdown of containment and verified that no other loose material was present. The issues associated with the fibrous material left in containment and the effects on the containment sump are identified as an unresolved item (URI) pending additional inspection and review from the NRC. This URI is designated 05000280/2007005-04, Fibrous Material left in Unit 1 Containment.
05000280/FIN-2008002-01	2008Q1	Surry	Loss of thermal barrier cooling due to a failure to follow procedures	A self-revealing finding of very low safety significance that constituted a non-cited violation (NCV) of Technical Specification 6.4.D was identified. Licensee personnel failed to follow procedure 2-IPM-CC-F-207A and caused cooling water flow to the thermal barrier of the Unit 2 Reactor Coolant Pump (RCP) 1A to be isolated for approximately 15 minutes. The finding was entered into the corrective action program as Condition Report 093555. Licensee corrective actions included re-opening the valve, restoring cooling flow to the thermal barrier, and providing training station wide on procedure adherence. The failure to follow procedure 2-IPM-CC-F-207A was a performance deficiency. The finding is more than minor because it is associated with the human performance attribute of the Initiating Event Cornerstone, and adversely affected the cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The finding, evaluated per the SDP in IMC 0609, Appendix A, is of very low safety significance (Green) because the finding would not have resulted in exceeding the Technical Specification limit for RCS leakage, due to operation of the RCP seal injection system. This finding has a cross-cutting aspect in the area of human performance work practices (H.4.b) because personnel failed to follow a written and approved procedure. (Section 1R22
05000280/FIN-2008002-02	2008Q1	Surry	Failure to Follow Start-up Procedure which resulted in Leaving Loose Fibrous Insulation in Containment	An NRC-identified, non-cited violation (NCV) of very low safety significance was identified for the failure to follow start-up procedure 1-GOP-1.7, revision 2, Unit Startup, RCS Heat Up from Ambient to HSD, which resulted in leaving loose fibrous insulation in containment. This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the IMC 0609, \"Significance Determination Process,\" Phase 1 Worksheets, the finding is determined to have very low safety significance (Green) since it only affected the mitigating systems cornerstone and did not represent a loss of system safety function. The cause of this finding had cross-cutting aspects associated with work practices of the Human Performance area in that the licensee did not provide the appropriate oversight of contractors conducting the containment walk downs (H.4.c). The finding was entered into the corrective action program as Condition Report 02564. Corrective actions to remove the fibrous material from containment prior to startup and to establish the extent of condition and potential impact on Unit-2 were adequate. (Section 4OA5
05000280/FIN-2008002-03	2008Q1	Surry	Licensee-Identified Violation	10 CFR part 50.65(a)(4), requires, in part, that before performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activity. Contrary to the above, on February 4, 2008, the licensee tagged out and drained the emergency switchgear room ventilation coolers 1-VS-AC-6 and 2-VS-AC-6, rendering them inoperable, without properly assessing the risk. The components were erroneously thought to be included in the recently added chilled water piping replacement risk term. The licensee recognized the error on February 4, 2008, prior to releasing work. The omitted components were selected in the Safety Monitor program and risk for both units increased to a slightly elevated (Yellow) risk condition. In accordance with Manual Chapter (MC) 0612, Appendix E, example 7.e, the issue is more than minor. The finding was evaluated per MC 0609, Appendix K, and found to be of very low safety significance (Green) because the change in risk had existed for only a short period of time prior to being corrected and the necessary compensatory actions were in-place. This finding was entered into the licensees corrective action program as CR 090374
05000280/FIN-2008002-04	2008Q1	Surry	Licensee-Identified Violation	Surry Power Station (SPS) Operating License Condition 3.I states, in part, that the Licensee shall implement and maintain in effect the provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report. Branch Technical Position (BTP) Chemical Engineering Branch (CMEB) 9.5-1, which incorporated the guidance of Appendix A to BTP Auxiliary Systems Branch (ASB) 9.5-1 and the technical requirements of Appendix R to 10 CFR Part 50, established the regulatory and licensing requirements for the fire protection program at SPS. Section 9.10.1 of the UFSAR states, in part, Compliance with these criteria is contained in the following documents: Fire Protection Program document. Section 6.1.o of VPAP-2401, Fire Protection Program, Rev. 28, states that penetration seals must provide equal or greater fire rating than that of the fire barrier. Contrary to the above, the licensee failed to have any sealant providing a fire rating in two fire penetrations in the block walls that separate the Unit 1 and Unit 2 Main Control Room HVAC rooms (Fire Area 5) from the north stairwell (Fire Area 68). This violation is of very low safety significance because the violation did not affect ignition frequencies, detection, or suppression system performance. This issue was entered into the licensees corrective action program as CR 090704
05000280/FIN-2009006-01	2009Q4	Surry	Failure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves Nor Set Goals and Monitor Under 10CFR50.65(A)(1)	The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants, for failure to demonstrate effective preventive maintenance of Unit 1 low head safety injection (LHSI) cold leg check valves in accordance with 10CFR50.65(a)(2) and not establish goals and monitor against those goals in accordance with 10CFR50.65(a)(1). The finding is more than minor because it affected the Barrier Integrity cornerstone objective of providing reasonable assurance that physical design barriers (e.g., reactor coolant system (RCS)) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding affected the LHSI cold leg check valves, which provide an isolation barrier from the high pressure RCS when the SI System is in standby to ensure that the integrity of the reactor RCS boundary is maintained. The finding is also associated with the cornerstone attribute of reactor coolant system equipment and barrier performance. The inspectors determined that this performance deficiency was a separate consequence of the degraded performance associated with the LHSI cold leg check valves. Because of this characterization, the inspectors determined that this issue should not be processed through the Significance Determination Process. Therefore, in accordance with the guidance in NRC Inspection Procedure 71111.12, Appendix D, this issue was determined to be a maintenance rule Category II finding and is of very low safety significance (Green). Based on the assessment performed by the team on the current licensees implementation of 10CFR50.65, the results of the licensees extent of condition review for this finding, and because this finding occurred on November 18, 2007, the team determined that this finding was not indicative of current licensee performance and, therefore, no Cross Cutting Aspect was assigned to this issue. This issue was entered in the licensees CAP as CR02560. The licensee restored compliance by establishing goals and monitoring the system performance against those goals in accordance with 10CFR50.65(a)(1)
05000280/FIN-2009007-02	2009Q2	Surry	Availability of Portable Ventilation Fans for Use by the Fire Brigade	The team identified a URI, involving the handling of portable ventilation equipment needed for pre-fire-fighting smoke removal. Appendix A to Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976,\" dated August 23, 1976, Section B.5, states that the need for good organization, training, and equipping of fire brigades at nuclear power plant sites requires effective measures be implemented to assure proper discharge of these functions. Section B.1 states that guidance for the good organization, training, and equipping of fire brigades is contained in NFPA-27, Private Fire Brigades. The licensee committed to NFPA 27-1975 in their FPP. It is generally accepted that the minimum fire brigade equipment consists of personnel protective equipment such as turnout coats, boots, gloves, hard hats, emergency communications equipment, portable lights, portable ventilation equipment, portable extinguishers and SCBAs. It is preferable that the fire brigade equipment be brought to the scene of a fire at the time of the initial response. However, as a minimum, if any of this equipment is not initially brought with the fire brigade, it should be in a location where it can be promptly obtained. Consistent with this principle, NFPA 27-1975, Section 72, Equipment Storage, states that storage space for the brigade equipment should be provided so that it can be promptly obtained for use and be properly maintained. The team found that the licensee had chosen to store the portable ventilation fans at four locations. Two of those locations were the station administration building and the protected area administration building. The fans at those locations did not have flexible air ducts stored with them, so they could not be used for all fire situations. Two of the locations were at the emergency response building: one on a fire engine and one on the B.5.b emergency response truck. Those fans did have flexible ductwork and ductwork adapters stored with them. There were no dedicated plant personnel to transport this equipment to the fire scene. The licensee stated their practice was to have a non-fire brigade person drive the fire engine to a point close to the fire brigade staging area. The idea was to have the portable ventilation fans and other equipment on the fire engine promptly available for use by the fire brigade. Another option was to have a non-fire brigade person deliver one of the fans stored at the administration buildings to the fire brigade. The team questioned how this concept would work in an actual emergency that may take place at times when minimum staff was on site. All plant personnel may not know the locations of the fans. The emergency response building is outside the owner controlled area which may preclude the equipment from being promptly obtained in some circumstances. Apparently, forethought had not been given to designating reliable power outlets for the fans at the various FAs? Portable ventilation fans are needed at Surry because no special smoke exhausting systems are installed at the plant. Existing ventilation systems in the plant are not designed for smoke removal. This condition was recognized in the SER dated September 19, 1979, as evidenced by Section 4.4.1, Smoke Removal, which states that no special smoke exhausting systems are provided at the plant. It further states that when normal ventilation systems cannot be used (for smoke removal), the fire brigade will use the portable ventilation units with flexible ducting available at the plant for smoke removal. In order to allow time for answering these questions and evaluating the situation, a URI is established: 05000280, 281/2009007-02, Availability of Portable Ventilation Fans for Use by the Fire Brigad
05000280/FIN-2010002-02	2010Q1	Surry	Emergency Plan Minimum Staffing	An unresolved item (URI) was identified by the inspectors relating to maintenance of the required minimum onsite manning in accordance with the licensees Emergency Plan. On January 4, 2010, the licensee identified issues relating to the Emergency Plan minimum manning requirements for maintenance personnel. They subsequently initiated CR364061 in their CAP and the respective root cause evaluation, RCE000999, for appropriate corrective actions. The inspectors reviewed RCE000999 and require additional information from the licensee to appropriately characterize a performance deficiency which may be greater than minor. This issue is identified as URI05000280, 281/2010002-02, Emergency Plan Minimum Staffing
05000280/FIN-2010004-01	2010Q3	Surry	Licensee-Identified Violation	10 CFR 50.54(q) states in part that a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b) and the requirements in appendix E of this part. Contrary to this, between early December 2006 and January 2010, the licensee identified that the staffing was reduced for mechanical maintenance and electrical maintenance personnel on shift to below the minimum shift staffing requirements of the Emergency Plan without a 50.54(q) review. The violation was determined to be of very low safety significance because, the licensee demonstrated non-designated coincidental coverage for the shift staffing positions in question, no degradation of the planning standard existed and the criteria for a white finding was not met. The licensee corrected the deficiency when it was discovered and entered it into the corrective action program as condition report CR364194.
05000280/FIN-2010005-01	2010Q4	Surry	Failure to Correct Multiple Conditions Adverse to Fire Protection	A self-revealing apparent violation (AV) of Condition 1.B to the Surry Unit 1 and Unit 2 Updated Facility Operating Licenses, DPR-32 and DPR-37, was identified for the licensees failure to take corrective action for degraded conditions adverse to the fire protection program. Specifically, in 2003-2004, three breakers with loads including the Unit 2 1B Refueling Water Storage Tank (RWST) chiller motor, the Unit 1 2B charging component cooling water pump, and the Unit 2 B hydrogen recombiner were identified as being oversized with respect to the Surry design standard for breaker sizing and cable protection. The failure to take corrective action on the affected breakers led to a fault on the Unit 2 RWST Chiller Motor 1B on October 11, 2010, and a resulting fire which damaged the electrical cable and motor controller. The fire was promptly extinguished by the fire brigade. The licensee entered this issue into the CAP (CR 398628) and isolated the remaining breakers to prevent additional failures. The inspectors found that the failure to take action to correct multiple oversized breakers constituted a performance deficiency. The finding is more than minor because it adversely affected the external factors attribute (fire) of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 1B RWST chiller motor and the Unit 2 B hydrogen recombiner breakers were the most susceptible to fire due to their size; also a cable fault could potentially damage safety related cables routed nearby. In addition, the Unit 1 2B charging component cooling water pump is safety related and was also unprotected. The inspectors reviewed IMC 0609, Appendix F, Attachment 1, and determined the category of post fire safe shutdown was affected and the finding required a phase 3 analysis. The significance of this finding is to be determined pending completion of the phase 3 evaluation. This finding has a cross cutting aspect in the work control component in the Human Performance area because the licensee did not appropriately plan work activities by incorporating risk insights. Specifically, although work orders were planned in 2006 they were neither prioritized consistent with their safety significance nor scheduled and completed in a timely manner. (H.3(a)). (Section 1R15.b.1
05000280/FIN-2010005-02	2010Q4	Surry	Failure to Follow Procedure Results in Inadvertent Actuation of Safety Injection	A self-revealing Green NCV of TS 6.4, Unit Operating Procedures and Programs, was identified for the failure to follow procedure 1-OPT-ZZ-001, ESF Actuation with Undervoltage and Degraded Voltage 1H Bus. Specifically, on October 26, 2010, a test lead was incorrectly installed in the Unit 1 relay room for the logic circuit associated with the A train of Consequence Limiting Safeguards (CLS). This resulted in an inadvertent safety injection, isolated component cooling water supply to the standby residual heat removal (RHR) train, and automatically initiated several safety-related components including emergency diesel generator (EDG) #1. Operators entered AP- 10.20, Response To Spurious Safety Injection With RCS Temperature Less Than 350F, and terminated the safety injection in approximately three minutes. The licensee entered this issue into the CAP (CR 400908). Failure to install the test leads as required by procedure 1-OPT-ZZ-001, is a performance deficiency. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding, evaluated in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 3, identified the finding is of very low safety significance (Green) because the finding did not lead to a loss of decay heat removal. This finding has a cross cutting aspect in the work practices component in the Human Performance area, because human error prevention techniques were not properly used commensurate with the risk significance of the assigned task (H.4(a)). (Section 1R22
05000280/FIN-2010005-03	2010Q4	Surry	Heavy Load Lift of the 1B RCP Motor Over Exposed Reactor Fuel	An NRC-identified Green NCV of Technical Specification (TS) 6.4, Unit Operating Procedures and Programs, was identified. Personnel failed to follow the defined heavy load shipping path inside containment as specified in procedure, GMP- 001, Heavy Load Rigging and Movement, which resulted in the movement of the 1B reactor coolant pump motor over exposed reactor fuel. The licensee has entered the issue into the CAP (CR 404106). Transport of the 1B reactor coolant pump motor over the exposed reactor core is a performance deficiency. The finding is more than minor because it can reasonably be viewed as a precursor to a significant event, the heavy load traveled over exposed irradiated fuel with the reactor vessel head removed. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, the inspectors conducted a Phase 1 SDP screening and determined the finding required a Phase 2 analysis. The Phase 2 analysis determined the finding is of very low safety significance (Green) because: (1) there is a low probability of dropping the load based on a study in NUREG-1774 performed for crane operating experience; (2) the polar crane was in working condition and had no known deficiencies that would have affected the cranes ability to lift the load; and, (3) the duration of the heavy load lift over the exposed reactor core was very short. In addition, in accordance with NRC IMC 0609, Appendix H, Containment Integrity SDP, the finding would not contribute to LERF due to the time since the reactor was shutdown. The finding has a cross-cutting aspect in the work practices component of the Human Performance area because plant supervisors failed to properly supervise workers executing procedure steps (H.4(c))
05000280/FIN-2010005-04	2010Q4	Surry	Inadequate Risk Evaluation for Leaving Common ESGR HELB Door Open	A licensee identified AV of 10CFR50.65 (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, was revealed after the licensee discovered that 2-BS-DR-21, common emergency switchgear room (ESGR) door was blocked open for two hours without clear communication to licensed operators. The licensee did not adequately assess the increase in operational risk that resulted in the required risk management actions of fire and environmentally qualified watches not being established. The licensee immediately corrected the condition by shutting the HELB door and having security control personnel access. The issue was entered into the licensees CAP as CR397720. The failure to adequately assess the increased risk associated with blocking open the common ESGR door and to take the required risk management actions is a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, both Unit 1 and Unit 2 plant risk were not evaluated and risk management activities were not put in place when the common ESGR door was blocked open for maintenance and unable to perform its function as a fire barrier, a halon suppression pressure boundary, a main control room pressure boundary, and a HELB boundary. In accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, this finding will require a phase 3 analysis. The significance of this finding is to be determined pending completion of the phase 3 evaluation. The inspectors determined that this finding had a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating risk insights (H.3(a)). (Section 1R15.b.2
05000280/FIN-2010005-05	2010Q4	Surry	Licensee-Identified Violation	10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of designs such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to this, the licensee identified in CR398124 that they failed to verify the adequacy of design changes that implemented a new drainage path through the primary shield wall. The design changes were inadequate because the impact to the sump from Microtherm insulation known to be installed on the reactor vessel head had not been evaluated or tested. The licensee removed all Microtherm from the Unit 1 reactor head and plans on removing all Microtherm from the Unit 2 reactor head during the outage in Spring 2011. The violation was determined to be of very low safety significance because the licensee was able to show that the debris loading was bounded by testing
05000324/FIN-2008002-02	2008Q1	Brunswick	Review the Significance of the STORM Drain Stabilization Pond Evaporation Pathway Dose Compared to Doses from All Other Pathways	An unresolved item (URI) was identified regarding the significance of the Storm Drain Stabilization Pond (SDSP) evaporation pathway dose in regard to meeting the requirement of the Offsite Dose Calculation Manual that dose assessments are required to be consistent with the methodology provided in Regulatory Guide 1.109, Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I. Specifically, RG 1.109 specifies that exposure pathways that may arise due to unique conditions at a specific site should be considered if they are likely to provide a significant contribution to the total dose. A significant pathway is considered one whose additional dose increment is equal to or greater than ten percent of the total from all pathways. Based on the preliminary assessment of doses to the public in 2007 from the SDSP via the evaporation pathway as compared to the 2006 annual effluent release data, the potential exists that this previously unevaluated pathway exceeded ten percent of the total dose for 2007 and should be included in the ODCM. This item is unresolved pending NRC review and evaluation of the final dose assessment for the SDSP evaporation pathway and the total public dose for 2007 that will be reported in the 2007 Radioactive Effluent Release Report. URI 05000325,324/2008002-02, Review the Significance of the Storm Drain Stabilization Pond Evaporation Pathway Dose Compared to Doses from All Other Pathways
05000324/FIN-2008003-01	2008Q2	Brunswick	Inadequate Procedure for Performing Maintenance on the Control Room AC Subsystem	A self-revealing Green non-cited violation of Technical Specification 5.4.1 was identified for an inadequate procedure used to specify configuration controls during a maintenance activity. The configuration management program implementation procedure, ADM-NGGC-0106, was not clear in determining whether additional actions should be taken to ensure Control Room Air Conditioning (AC) operation while preventative maintenance was being performed on the CREV system. The three Control Room AC subsystems tripped inadvertently during the performance of this planned preventive maintenance activity due to the supply fan dampers drifting shut, resulting in Unit 1 and Unit 2 entering LCO 3.0.3. This issue was entered into the licensee\'s Corrective Action Program (CAP) as AR 281950. The finding was more than minor because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of equipment performance. The finding was determined to be of very low safety significance because it did not represent an actual loss of safety function for greater than the TS allowed outage time. The finding has a cross-cutting aspect in the area of Human Performance of complete documentation because the licensee did not provide an adequate procedure that provided clear guidance in identifying intrusive maintenance on the CREV system such that appropriate actions were taken to ensure proper operation during preventative maintenance. (H.2.(c)
05000324/FIN-2008003-02	2008Q2	Brunswick	Inadequate Calibration Procedure for the Conventional Service Water Relays	A self-revealing Green non-cited violation of Technical Specification 5.4.1 was identified for an inadequate procedure used for the calibration of the conventional service water pump logic relays in September 2007. Specifically, procedure 0PM-RLY- 001, PM GE HFA Relays, used to calibrate the conventional service water (CSW) pump relays was inadequate because the procedure was determined not to be applicable to the relay type. The incorrectly calibrated conventional service water pump relay resulted in improper operation of the conventional service water pump and could have affected proper emergency diesel generator operation during a Loss of Offsite Power (LOOP) Event. The finding is in the licensees Corrective Action Program (CAP) as AR 245864. The finding was more than minor because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of equipment performance. The finding was determined to be of very low safety significance because it did not contribute to improper emergency diesel generator operation. The finding has a cross-cutting aspect in the area of Human Performance of complete documentation because the licensee did not provide an adequate procedure to calibrate the CSW pump relays. (H.2.(c)
05000324/FIN-2008003-03	2008Q2	Brunswick	Failure to Conduct Adequate and Timely Evaluations of Onsite Groundwater Monitoring Well Tritium Concentration Trend Data	The inspectors identified a Green finding (FIN) for failure to properly evaluate the potential causes of increased tritium (H-3) concentrations in groundwater samples collected and reviewed in accordance with Brunswick procedure E&RC-3250, Environmental and Radiation Control. Specifically, the licensee failed to properly evaluate, and initiate actions to address increasing H-3 concentrations reported from 2003 through 2007 for quarterly samples collected from Environmental Sampling Station (ESS)-2C and ESS-16 monitoring wells. The failure to properly investigate the increasing H-3 concentrations resulted in the licensee continuing to attribute the subject results to a 1994 U2 radioactive liquid effluent waste line break without considering potential leakage of contaminated liquids from U2 storm drain piping. This issue has been entered in the licensees CAP as NCR 268357. The finding is more than minor because it is associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and adversely affects the cornerstone objective because it relates to effluent measurement and abnormal releases. The licensees failure to recognize the increasing groundwater tritium concentrations delayed actions to address and correct abnormal liquid releases within the switchyard area. Using the Public Radiation Safety Significance Determination Process, this finding was determined to be of very low safety significance (Green) because the performance deficiency did not result in offsite releases and resultant offsite doses to members of the public and was not a failure to implement the effluent program. Furthermore, the finding did not prevent the licensee from initiating appropriate corrective actions to determine extent of the contamination and to mitigate its effect on the surrounding environs. The cause of the finding was related to the cross cutting area of human performance, the component of work practices, and the aspect involving supervisory oversight of work activities, because the licensee failed to properly evaluate monitoring well sample data to determine the possible radiological effects of plant operation on the local groundwater. (H.4(c)
05000324/FIN-2008003-04	2008Q2	Brunswick	Evaluate Representativeness of Particulate Sampling for the Reactor Building Roof Vent Monitors, Turbine Building Wide Range Gas Monitors and Plant Stack Wide Range Gas Monitors	An unresolved item (URI) was identified regarding the representativeness of radioactive particulate sampling by the sampling skids used to monitor gaseous effluent releases from the turbine building ventilation system, reactor building roof vent, and plant stack. During plant walk-downs, the inspectors identified one or more T connections and/or elbows on the inlet side of the particulate filter on each of the specified monitoring skids. The licensee had no evaluation of the impact of these bends on the transmission of particles through the sampling lines. The licensee has contacted the vendor requesting documentation of particle transmission studies that may have been performed for the monitors. This item is unresolved pending NRC review and evaluation of any vendor supplied documentation on the particle transmission data obtained by the licensee. URI 05000325,324/2008003-01, Evaluate Representativeness of Particulate Sampling for the Reactor Building Roof Vent Monitors, Turbine Building Wide Range Gas Monitors, and Plant Stack Wide Range Gas Monitors
05000324/FIN-2008003-05	2008Q2	Brunswick	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 meets the criteria of Section VI.A.1 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. TS 5.4.1, Administrative Control (Procedures), requires that written procedures shall be establish, implemented, and maintained covering applicable procedures recommended in Regulatory Guide 1.33, Appendix A, November 1972. Regulatory Guide 1.33, Section I (1) states that maintenance that can affect the performance of safety-related equipment should be properly preplanned, and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, the licensees procedure MCP-NGGC-0401, Material Acquisition, Revision 23, a procurement procedure, was inadequate because it contained unclear maintenance actions which resulted in a premature declaration of operability of the Unit 1, B nuclear service water system. On March 28, 2008, work was complete on 1-SW-V20 and the nuclear service water header was declared operable. The licensee then discovered that valve 1-SW-V20 had not been cleared of the conditional release, which would have verified that the valve was fully qualified, because the maintenance personnel involved was unaware of their responsibilities in clearing the conditional release. Once this was discovered, the Unit 1 NSW header was declared inoperable until valve 1-SW-V20 was fully qualified. As a result, Emergency Diesel Generator #3 incurred unnecessary unavailability time. Because the finding is of very low safety significance and has been entered into the CAP (AR 272531), this finding is being treated as an NCV, consistent with Section VI.A of the Enforcement Policy
05000324/FIN-2012004-01	2012Q3	Brunswick	Failure to Maintain Secondary Containment Operable During an OPDRV Activity	The inspectors identified a Green non-cited violation (NCV) of TS 3.6.4.1, Secondary Containment because the licensee did not maintain secondary containment operable as required during a maintenance activity considered an operation with a potential for draining the reactor vessel (OPDRV). Once questioned by the inspectors, the licensee restored secondary containment, developed an Operation standing instruction (12-052) to treat the activity as an OPDRV and placed this issue into its corrective action program (CAP) as AR 562188. The failure to maintain secondary containment operable while Unit 1 was in Mode 4 with an OPDRV in progress was a performance deficiency. The finding was more than minor because it was associated with the configuration 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 because the Unit 1 secondary containment boundary was not preserved or maintained. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G since the reactor was in Mode 4 (cold shutdown). The finding was determined to be of very low safety significance (Green) according to IMC 0609 Appendix G, Attachment 1, Checklist 6, since a quantitative assessment (Phase 2 or Phase 3 evaluation) was not required. Specifically, the inspectors determined that the licensee maintained adequate mitigation capability for reactor vessel water level inventory and an event did not occur that could be characterized as a loss of control. The cause of this finding was directly related to the cross-cutting aspect of Accurate Procedures in the Resources component of the Human Performance area, because the licensee did not consider the recirculation pump seal replacement activity to be OPDRV based on procedural guidance that contains exclusions to what are considered OPDRV activities.
05000324/FIN-2012004-02	2012Q3	Brunswick	Failure to Maintain Secondary Containment Operable During an OPDRV Activity	A self-revealing Green NCV of 10 CFR 50.54(q)(2) was identified for the licensees failure to properly evaluate or consider the impact to emergency response facilities of design change ESR98-00436 which was implemented in 1999. This resulted in the loss of Emergency Response Facility Information System (ERFIS), Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), and all displays including radiation monitors for the emergency response facilities. Specifically, the licensee failed to ensure that adequate emergency response facilities and equipment were available as required by the Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3 revision 80 and 10 CFR 50.47(b)(8). This issue was captured in the licensees CAP as AR 542704. The licensees failure to properly evaluate or consider the impact to emergency response facilities of design change ESR98-00436 which was implemented in 1999 was a performance deficiency. Specifically, the licensee introduced a single point failure mode which did not meet the design requirements specified in their Design Basis Document (DBD 60) sections 3.6.7.2 and 3.6.7.3. This resulted in the licensees failure to ensure that adequate emergency response facilities and equipment were available as delineated in the Updated Final Safety Analysis Report (UFSAR) Section 7.7.1.9, and required by the Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3, revision 80, and 10 CFR 50.47(b)(8). The finding was more than minor because it adversely affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the Facilities and Equipment attribute was affected during the time when the ERFIS, ERDS, SPDS, and all displays including radiation monitors for the emergency response facilities were degraded, and as a result did not meet 10 CFR 50.47(b)(8) Planning Standard program element, adequate emergency facilities and equipment to support the emergency response are provided and maintained. The finding was assessed for significance in accordance with NRC IMC 0609, Appendix B Emergency Preparedness Significance Determination Process. Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure to comply; Loss of Risk Significant Planning Standard Function (RSPS), No; RSPS Degraded Function, No; Loss of Planning Standard Function, No; the result is a Green finding. The inspectors determined that this resulted in a very low safety significance finding (Green). No cross-cutting aspect was assigned to this finding because the performance deficiency occurred more than three years ago and is not reflective of current plant performance.
05000324/FIN-2012004-03	2012Q3	Brunswick	EDG2 Wiring on ASSD Switch	A wiring discrepancy was identified during inspection of the EDG 2 ASSD switch 2-DG-SS-A1. A contact in the circuit was determined to be bypassed that would have the potential to prevent proper isolation of the EDG2 control circuits from the Main Control Room (MCR) during an Appendix R fire event. The inspectors plan to review the licensees cause evaluation for this event and determine if a performance deficiency existed. This issue is being tracked as URI 05000325; 324/2012004-03, EDG2 wiring on ASSD switch.
05000324/FIN-2012004-04	2012Q3	Brunswick	Licensee-Identified Violation	A wiring discrepancy was identified during inspection of the EDG 2 ASSD switch 2-DG-SS-A1. A contact in the circuit was determined to be bypassed that would have the potential to prevent proper isolation of the EDG2 control circuits from the Main Control Room (MCR) during an Appendix R fire event. The inspectors plan to review the licensees cause evaluation for this event and determine if a performance deficiency existed. This issue is being tracked as URI 05000325; 324/2012004-03, EDG2 wiring on ASSD switch.
05000324/FIN-2012004-05	2012Q3	Brunswick	Licensee-Identified Violation	10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Licensee procedure ADM-NGGC-0107, Equipment Reliability Process Guideline, steps 9.4.9 and 9.4.10 required component experts and preventive maintenance (PM) optimization to determine if there was a cost effective PM to prevent failure and then to develop the PM model. Contrary to the above, the Unit 1 high pressure coolant injection (HPCI) ramp generator signal converter (RGSC) did not have the appropriate preventive maintenance to prevent failure. As a result, the Unit 1 high pressure coolant injection (HPCI) system failed the HPCI System Operability Test performed on April 30, 2012 and was declared inoperable. The licensee entered this issue into the CAP as NCR 534364. Corrective actions included replacing the RGSC and creating a PM task to replace the RGSCs on a specified frequency. Using IMC 0609, Appendix A, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined this finding required a Phase 2 analysis. The Phase 1 screened this Mitigating Systems Cornerstone finding to Phase 2 because the finding represented a loss of HPCI system and/or function. The inspectors, with the assistance of the regional Senior Risk Analyst, performed a Phase 2 analysis using the Saphire 8 Model. 109 hours of unavailability time was used for the analysis since HPCI was not required during the refueling outage from February 23, 2012 through April 29, 2012. Based on the results of the Phase 2 analysis, the inspectors determined the finding was of very low safety significance
05000324/FIN-2012005-01	2012Q4	Brunswick	Floor Drains Not Functioning Due to Plugging	The inspectors are opening an URI to review the licensees evaluation of the potential for adverse impact due to floor drain sock filter plugging in safety-related pump rooms and determine if there is a performance deficiency. On November 24, 2012, during a steam leak in the 2A Feedwater Heater Room, water did not adequately drain from the room through the floor drains due to plugging in the floor drain sock filters. The licensees immediate corrective actions included removing the sock filters so that the water could drain. The sock filters are also installed in safety-related pump rooms in the reactor building. The inspectors are opening an URI to review the licensees evaluation of the potential for adverse impact due to drain plugging in safety-related pump rooms and determine if there is a performance deficiency. The licensee entered this issue in the CAP as NCR 574261. This issue is being tracked as a URI: URI 05000325/2012005-01 and 05000324/2012005-01, Floor Drains Not Functioning Due to Plugging.
05000324/FIN-2012005-02	2012Q4	Brunswick	Emergency Diesel Generator 3 Slow Start	On October 14, 2012, the licensee was running EDG 3 for a zero oil pressure start test in accordance with Procedure 0PT-12.2.c, No. 3 Diesel Generator Monthly Load Test. The EDG reached rated speed at approximately 38 seconds after the EDG was started and then tripped. Surveillance Requirement 3.8.1.7 requires the EDG reach rated conditions within 10 seconds. Several seconds after reaching rated speed, the EDG began to coast down due to receiving a lockout signal since full rated conditions were not achieved within the nominal time delay of 45 seconds. The licensee replaced the overspeed start emergency boost cylinder and declared the EDG operable on October 17, 2012. The inspectors are opening an URI to review the licensee\'s evaluation of the cause of the EDG failure and determine if there is a performance deficiency. The licensee entered this issue in the CAP as NCR 567016. This issue is being tracked as a URI: URI 05000325/2012005-02 and 05000324/2012005-02, Emergency Diesel Generator 3 Slow Start.
05000324/FIN-2013003-01	2013Q2	Brunswick	Failure to Have Adequate Installation and Testing Instructions for the EDG Overspeed Boost Cylinder	An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to have adequate installation and testing instructions for the EDG control oil system overspeed boost cylinder and accomplish the installation and testing in accordance with these instructions. The licensee replaced the boost cylinder and returned the EDG to operable. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 567016. The inspectors determined that the failure to properly install the EDG 3 overspeed boost cylinder and properly test the boost cylinder, to ensure the boost cylinder can perform its design basis function, was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affects 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 correctly install and test the EDG 3 overspeed boost cylinder resulted in the failure of EDG 3 to start and EDG 3 being declared inoperable on October 14, 2012. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of human performance associated with the resources attribute because the licensee did not have complete, accurate and up-to-date design documentation, procedures, and work packages to install and test the EDG 3 overspeed boost cylinder.
05000324/FIN-2013003-02	2013Q2	Brunswick	Failure to Adequately Lubricate the 1B Residual Heat Removal Cooler Damper	An NRC-identified Green NCV of TS 5.4.1a, Procedures, was identified for the failure of the licensee to follow the procedure to properly lubricate the 1B RHR room cooler damper The licensee lubricated the damper and returned the room cooler to operable, and entered this issue into the CAP as NCR 607514. The inspectors determined that the failure of the licensee to properly lubricate the 1B RHR room cooler damper in accordance with Procedure 0PM-DMP500 was a performance deficiency. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and affects 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 lubricate the 1B RHR room cooler damper resulted in a failure of the cooler fan and damper, and the inoperability of the 1B RHR train. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of human performance associated with the work practices attribute because the licensee did not define and effectively communicate expectations regarding procedural compliance to Procedure 0PM-DMP500 and personnel did not follow this procedure.
05000324/FIN-2013003-03	2013Q2	Brunswick	Inadequate Work Order to Perform a Modification to the Control Room Emergency Ventilation System	An NRC-identified Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified, for the licensees failure to have an adequate instruction or procedure to perform a modification to the control room emergency ventilation system (CREV). The licensee took immediate action to return CREV to service and entered this issue into the CAP as NCR 578363. The inspectors determined that the failure of the licensee to have an adequate procedure for installing a jumper on the 2A CREV system was a performance deficiency. The finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and affects 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 failure to have an adequate procedure to install a jumper on the 2A CREV system resulted in the safety system functional failure of CREV. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At- Power, the inspectors determined the finding screened to a detailed risk evaluation because the finding represented a degradation of the radiological barrier function and smoke or toxic atmosphere function of the control room barrier. The regional SRA performed a Phase 3 analysis on the finding. A screening calculation was performed to estimate the impact the finding would have on the facility for conditions that would lead to plant shutdown, or failure of the filtering function of the ventilation system. The low likelihood of failure to recover the system, combined with the short time the deficiency existed, resulted in a finding of very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance associated with the work control attribute because the licensee did not appropriately coordinate work activities by incorporating the impact of changes to the work scope or activity on the plant when installing a ring lug jumper on the 2A CREV subsystem.
05000324/FIN-2013003-04	2013Q2	Brunswick	Failure to Implement Risk Management Actions During Elevated Risk	An NRC-identified Green NCV of 10 CFR 50.65(a)(4) was identified for the failure of the licensee to manage the increase in risk that resulted from the E6 bus outage. Specifically, between May 19, 2013 and May 21, 2013, the licensee did not manage the increase in risk on Unit 2 during the E6 bus outage by use of appropriate risk management actions (RMAs). Operations personnel took immediate actions to protect the equipment in the control room and in the field. The licensee entered this issue into the CAP as NCR 607741. The inspectors determined that the failure of the licensee to manage risk during the E6 outage by performing RMAs for the protected 2A RHR and RHRSW loops, the 2A and 2B core spray and the hardened vent was a performance deficiency. The finding was more than minor because if left uncorrected, the failure to perform RMAs when required could result in safety-related mitigating equipment being unavailable during already elevated plant risk, specifically the 2A RHR and RHRSW loops, the 2A and 2B core spray and the hardened vent. This finding was associated with the human performance attribute of the Mitigating Systems Cornerstone. Using IMC 0609, Appendix K, issued May19, 2005, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of RMAs, the inspectors determined the finding screened as very low safety significance (Green) since the incremental core damage probability was less than 1E-6. The finding has a cross-cutting aspect in the area of human performance associated with the work control attribute because the licensee did not appropriately plan work activities by incorporating risk insights during the E6 bus outage.
05000324/FIN-2013003-05	2013Q2	Brunswick	Inadequate Design Control for Allowable Jacket Water Leak Rate	An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the failure of the licensee to verify the adequacy of the design acceptance criteria for jacket water leakage to ensure EDG 3 could meet the design basis mission time of seven days. The licensees corrective actions include developing a plan to fill the EDG jacket water system to ensure operation of the EDG for seven days. The licensee entered this issue into the CAP as NCR 615491. The inspectors determined that the failure to ensure sufficient jacket water to the EDGs, with a jacket water leak, for the seven-day mission time, was a performance deficiency. The violation is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the calculational error assuming a fourday mission time versus a seven-day mission time results in a condition where there was reasonable doubt on the capability of an EDG when a jacket water leak exists. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding screened to a detailed risk evaluation because the finding represented an actual loss of function of at least a single Train of EDG for greater than the TS Allowed Outage time. The regional SRA performed a Phase 3 analysis on the finding. The time to failure of the EDG due to the leak precluded any internal risk impact, since it exceeded 24 hours to failure. A screening calculation was performed to estimate the impact the finding would have on an extended loss of offsite power from seismic or external flooding. The low likelihood of the seismic or external flood event occurring, combined with the short time the deficiency existed, resulted in a finding of very low safety significance (Green). The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. Engineering evaluation was performed on July 7, 2004.
05000324/FIN-2013003-06	2013Q2	Brunswick	Non-Conservative Calculation for Service Water Flow Rate to the Emergency Diesel Generators	The inspectors are opening an URI to review the revision to Calculation M-89-0008, Heat Balance on DG 2 Jacket Water Service Heat Exchanger, for service water flow rate required for EDG operability during a design basis event, to determine if the performance deficiency associated with this issue is more than minor. On January 14, 2013, the licensee was performing Procedure 0ENP-2705, Service Water Heat Exchanger Thermal Performance Testing, to measure the service water flow rate to the EDG 3 jacket water heat exchanger, and found flow to be in the range of 351 to 358 gpm. The expected flow rate is 900 gpm to 1100 gpm. After visual inspection, it was determined the EDG 3 service water outlet valve 2-SW-V208 was throttled to 1-1.25 turns instead of the required 2.25 turns specified in 0OP-39, Diesel Generator Operating Procedure. Diesel 3 was determined to be operable based on Calculation M-89-0008, which required the measured service water flow rate to be above 350 gpm at an ultimate heat sink temperature of 90F. The licensee performed a past operability evaluation and determined the service water outlet valve 2-SW-V208 was out of position since April 2010. The licensee determined the maximum service water inlet temperature experienced between April 2010 and January 2013 was 89.2F on August 2, 2011 and August 6, 2012. The licensee concluded that since the inlet temperature was below 90F in 2011 and 2012, and the service water flow rate was above 350 gpm, that EDG 3 had always been operable. The inspectors reviewed Calculation M-89-0008 and determined that the calculation assumed an EDG loading of 3500 kW instead of the EDG loading of 3850 kW allowed by TS 3.8.1, AC Sources Operating, Surveillance Requirement 3.8.1.11. The inspectors determined that the failure to have an adequate calculation for service water flow rate required for EDG operability was a performance deficiency. The inspectors are opening an URI to review the revision to Calculation M-89-0008 and determine if the performance deficiency is more than minor. The licensee entered this issue in the CAP as NCR 592035. This issue is being tracked as a URI: URI 05000325/2013003-06 and 05000324/2013003-06, Non-Conservative Calculation for Service Water Flow Rate to the Emergency Diesel Generators.
05000324/FIN-2013003-07	2013Q2	Brunswick	Failure to Update the UFSAR for the Removal of the Chlorine Detection System	An NRC-identified SLIV NCV of 10 CFR 50.71(e) was identified for the licensees failure to revise the UFSAR with information consistent with plant conditions. Specifically, from August 6, 2006 to the present, the licensee did not remove reference to or correct information to reflect current plant conditions related for the chlorine detection system used by the CREV in UFSAR Sections 6.4, Habitability System and 9.4.1, Control Building Ventilation System. The licensees corrective actions include revising the UFSAR. The licensee entered this issue into the CAP as NCR 614474. The inspectors determined the failure of the licensee to update the UFSAR after removing the chlorine detection function from the safety-related CREV as required by 10 CFR 50.71.e and in accordance with Procedure REG-NGGC-0101, Final Safety Analysis Report Revisions, was a performance deficiency. This issue is considered within the traditional enforcement process because it has the potential to impede or impact the NRCs ability to perform its regulatory functions. The inspectors used the Enforcement Policy, Supplement I Reactor Operations, to evaluate the significance of this violation. Similar to Enforcement Policy, Section 6.1, example d.3, the inspectors determined the violation was a SLIV violation since the erroneous information not updated in the UFSAR has not resulted in any unacceptable change to the facility or procedures.
05000324/FIN-2013003-08	2013Q2	Brunswick	Failure to Have an Adequate Procedure for Preventative Maintenance on a SCRAM Contactor Coil	A self-revealing Green NCV of TS 5.4.1a, Procedures, was identified for the failure of the licensee to have an adequate procedure incorporating a preventative maintenance schedule which specifies inspection or replacement of the RPS coil contactor 2-C72B-K1A that had a specific recommended lifetime. The licensee took action to manually open valve 2-E11-F009 and entered this issue into the CAP as NCR 599641. The inspectors determined that the failure of the licensee to have an adequate procedure incorporating a preventative maintenance schedule which specifies inspection or replacement of contactor 2-C72B-K1A was a performance deficiency. The finding was more than minor because if left uncorrected, the failure of the GE CR105 contactors could result in the failure of the Unit 1 and Unit 2 A and B RPS buses. The finding was also associated with the configuration control attribute of the Barrier Integrity Cornerstone. Specifically, the failure to perform a PM on contactor coil 2-C72B-K1A resulted in a loss of decay heat removal to the SFP on April 5, 2013. Using IMC 0609, Attachment 4, issued June 19, 2012, Initial Characterization of Findings, the inspectors determined that since this issue occurred during a refueling outage, that the finding should be processed in accordance with IMC 0609, Appendix G, issued February 28, 2005, Shutdown Operations Significance Determination Process. Using IMC 0609, Appendix G, Table 1, Losses of Control, the inspectors determined that the finding was of very low safety significance (Green) because the inadvertent change in RCS temperature due to loss of RHR divided by the change in temperature that would cause boiling was less than 0.2 (temperature margin to boil). The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. The PM was not implemented per vendor recommendations in 1990.
05000324/FIN-2013003-09	2013Q2	Brunswick	Residual Heat Removal A Heat Exchanger Bypass Valve 2-E11-F048A Stud Failure	The inspectors are opening an URI to review the licensees evaluation of the operability of A RHR heat exchanger bypass valve 2-E11-F048A and determine if the performance deficiency associated with this issue is more than minor. On March 29, 2012, during Unit 2 refueling outage B221R1, maintenance personnel were going to repack A RHR heat exchanger bypass valve 2-E11-F048A. A member of maintenance hit one of the four valve yoke to bonnet hold down 134 studs with his foot and the stud sheared off at the nut. A second yoke hold down stud sheared off at the nut when maintenance personnel tried to remove the nut. The licensees corrective actions included replacing the four studs. The licensee determined the failure mechanism of the two studs was low stress, high cycle fatigue caused by vibration of the valve during throttling operations. The inspectors determined that the performance deficiency associated with this issue was the failure of the licensee to evaluate the effects of vibration on valve 2-E11-F048A when the valve was used for throttling, which resulted in the two studs sheering. The inspectors are opening an URI to review the licensees evaluation of the operability of valve F048A and determine if the performance deficiency is more than minor. The licensee entered this issue in the CAP as NCR 598294. This issue is being tracked as a URI: URI 05000324/2013003-09, Residual Heat Removal A Heat Exchanger Bypass Valve 2-E11-F048A Stud Failure.
05000324/FIN-2013003-10	2013Q2	Brunswick	Notice of Enforcement Discretion for Replacement of the E8 Transformer	In accordance with the NRCs NOED process, the inspectors are opening a URI to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, including enforcement action determinations, associated with the NOED. The inspectors are opening the URI to determine if a performance deficiency exists. On April 15, 2013, due to the inoperability of Division II emergency buses E4/E8, the licensee requested the NRC not enforce compliance with TS 3.7.3, Required Action A.1; TS 3.8.1, Required Action B.3; TS 3.8.4, Required Action A.1; and TS 3.8.7, Required Action A1 until April 17, 2013 at 3:15 am. The licensee requested and was granted the NOED on April 15, 2013 at 2:05 pm. The LCO extension allowed the site time to complete the replacement of and test the E8 transformer to restore operability. The inspectors are opening an URI to determine if a performance deficiency exists. The licensee entered this issue in the CAP as NCR 601376. This issue is being tracked as a URI: (URI) 5000325/2013003-10; Notice of Enforcement Discretion for Replacement of the E8 Transformer.
05000324/FIN-2013005-02	2013Q4	Brunswick	Inadequate Design Control for Required Service Water Flow to the Emergency Diesel Generators	An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the failure of the licensee to verify the adequacy of design of the emergency diesel generator (EDG) service water flow. Specifically, from May 1, 1989, until October 28, 2013, Calculation M-89-0008, contained non-conservative values for EDG maximum loading, service water inlet temperatures, and heat exchanger fouling factor, resulting in a non-conservative calculation for required service water flow to the EDG jacket water heat exchanger, which called into question the operability of EDG 3. The licensee re-performed Calculation M-89-0008 and determined EDG 3 was operable. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 592035. The inspectors determined that the failure of the licensee to have an accurate calculation for required service water flow to the EDG jacket water heat exchanger was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the non-conservative calculation called into question the operability of EDG 3. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structures, systems, and components (SSC), the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the technical specification (TS) allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of human performance associated with the resources attribute because the licensee did not have complete, accurate and up-to-date design documentation for EDG service water flow. Specifically, due to the inspector?s questions, Calculation M-89-0008 required revision due to non-conservatisms in August 2013 and in November 2013.
05000324/FIN-2013005-04	2013Q4	Brunswick	Licensee-Identified Violation	The following finding of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a NCV. 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed by documented procedures and shall be accomplished in accordance with these procedures. Contrary to the above, from August 2010 until August 2013, when SRV pilot valve conical seating surface finish requirements were incorporated into licensee Procedure OCM-VSR509, Main Steam Relief Valves Target Rock Model 7567 Air Operators and Pilot Assembly, Disassembly, Inspection, and Reassembly, the licensee failed to prescribe procedural requirements for the SRV pilot valve conical seating surface finishes. This resulted in four of the eleven SRVs being out of tolerance, which was a violation of plant TS 3.4.3., Safety Relief Valves. The licensee took action to replace all of the pilot valves with valves that had the correct surface finish. This violation was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC that maintained functionality. The licensee entered this issue into their CAP as NCR 607846. The licensee revised procedure OCM-VSR509 as a corrective action to prevent recurrence.
05000324/FIN-2013010-01	2013Q4	Brunswick	Failure to Identify and Correct Flood Protection Degradation in Safety-Related Buildings	The NRC identified an AV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, with two examples. The first example involved the failure of the licensee to promptly identify and correct conditions adverse to quality associated with flood protection of multiple safety-related buildings. Specifically, the licensee failed to promptly identify or correct safety-related buildings that contained openings that would have adversely impacted their ability to mitigate external flooding of these buildings in the event of a design basis probable maximum hurricane (PMH). The second example involved the failure of the licensee to correct a significant condition adverse to quality. Specifically, the licensee failed to implement a corrective action to preclude repetition by not adequately developing an engineering program to mitigate the consequences of external events (flooding, high winds, and seismic) that ensured appropriate equipment classifications, with interfacing programs of maintenance rule (MR) and zero tolerance for equipment failures. This resulted in a violation of technical specification (TS) 3.7.2, Service Water (SW) System and Ultimate Heat Sink, and TS 3.5.2, Emergency Core Cooling System (ECCS) Shutdown, since the inoperability of the required number of service water pumps (SWPs) would violate TS 3.7.2, and TS 3.5.2 since SW cools the residual heat removal (RHR) system heat exchangers. The inspectors determined the failure to identify and correct the missing and degraded flood barriers in multiple safety-related buildings, and the failure to implement a corrective action to preclude repetition by not developing an engineering program to mitigate the consequences of external events that ensured appropriate equipment classifications, with interfacing programs of MR and zero tolerance for equipment failures, was a performance deficiency. The finding was more than minor because it was associated with the protection against external factors attribute (flood hazard) of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, nine of the ten Unit 1 and Unit 2 SWPs would be potentially inoperable and unavailable during specified PMH events. Because the finding involved reactor shutdown operations and conditions, IMC 0609, Appendix G, Shutdown Operations Significance Determination Process (SDP), Attachment 1, issued May 25, 2004, Phase 1 Operational Checklists for Both pressurized water reactors (PWRs) and boiling water reactors (BWRs), was used. The inspectors used Checklist 5, BWR Hot Shutdown: Time to Boil < 2 Hours, and determined the finding increased the likelihood that a loss of decay heat removal (DHR) will occur due to failure of the system itself or support systems, degraded the licensees ability to cope with a loss of offsite power (LOOP), degraded the licensees ability to add reactor coolant system (RCS) inventory when needed, and degraded the licensees ability to establish an alternate core cooling path if DHR could not be re-established for 24 hours. Further, the performance deficiency involved external events. Consequently a Phase 2 analysis could not be performed and the issue screened directly to a Phase 3 analysis. The significance of this issue is To Be Determined (TBD) and its final significance will be dispositioned in separate transmittal. The issue is not an immediate safety concern because the licensee has taken appropriate corrective actions. The finding has a cross-cutting aspect in the area of human performance associated with the field presence attribute because deviations from standards and expectations were not corrected promptly, and the licensee did not ensure supervisory and management oversight of work activities, including contractors. Specifically, licensee management failed to ensure degradation associated with flood protection of the safety-related buildings was identified and corrected.
05000324/FIN-2013010-04	2013Q4	Brunswick	Failure to Submit a Timely LER for Service Water System Inoperability	An NRC-identified AV of 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B), was identified for failure of the licensee to provide a written Licensee Event Report (LER) to the NRC within 60 days of identifying a condition which was prohibited by plant TS 3.7.2, SW System and Ultimate Heat Sink, and an event that could have prevented the fulfillment of a safety function of RHR. The licensees corrective actions included submitting LER 50-325 and 50-324/2013-003-00 on November 14, 2013. The licensee entered this issue into the CAP as NCR 629064. The inspectors determined the failure of the licensee to provide a written LER to the NRC within 60 days as required by 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B) was a performance deficiency. This violation involved a failure to make a required report to the NRC and is considered to impact the regulatory process. Such violations are dispositioned using the traditional enforcement process instead of the SDP. As discussed in the Enforcement Policy, the severity level of a violation involving the failure to make a required report to the NRC will be based upon the significance of and the circumstances surrounding the matter that should have been reported. This issue is being characterized as an AV in accordance with the NRC's Enforcement Policy, and its final significance will be dispositioned in separate future correspondence. Because this violation involves the traditional enforcement process, a cross-cutting aspect is not assigned to this violation.

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