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05000280/FIN-2000008-01Willful failure of an individual to report an arrest in accordance with VPAP-01052000Q2The inspectors identified a non-cited violation for the failure to comply with the requirements of the Physical Security Plan (PSP). Specifically, an individual willfully failed to report an arrest in accordance with VPAP-0105. Based on the individuals position in the organization, lack of management involvement and no other similar events involving the individual, this finding was determined to be of very low significance.
05000280/FIN-2006005-01Proceduralized Departures from TS2006Q4No Color: The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments. Specifically, the licensee implemented proceduralized departures from the approved station technical specifications (TS) without the required NRC approval in procedures AP-13.0, Turbine Building Flooding, revision 13, and FCA 6.01, Uncontrollable Turbine Building Flooding, revision 2. This finding was evaluated using traditional enforcement since it impacted or impeded the regulatory process in that the licensee improperly used the 10 CFR 50.59, Changes, Tests, and Experiments, process to incorporate operator actions inconsistent with the TS. This finding was of more than minor safety significance because the procedure changes improperly bypassed the required NRC review and approval prior to implementation. The unapproved procedural actions would only be involved at the end of a very rare accident sequence. Given the time during the accident sequence in which these actions were to be accomplished, the actions were not a determent to core damage. Therefore, the violation was of very low safety significance. The finding is identified as Severity Level IV because the noncompliance is not considered to be of more than very low significance based on risk.
05000280/FIN-2006010-01Failure of Exercise Critique to identify a RSPS weakness as a DEP PI opportunity Failure2006Q1The NRC identified an AV for failure of the licensees exercise critique process to properly identify a weakness associated with a risk-significant planning standard (RSPS) that was determined to be a Drill/Exercise Performance (DEP) Performance Indicator (PI) opportunity failure during a full-scale exercise. The AV is associated with emergency preparedness planning standards 10 CFR 50.47(b)(14) and 10 CFR 50.47(b)(4), and the requirements of 10 CFR 50, Appendix E, IV.F.2.g. This finding was not entered into the licensees corrective action program. The failure of the licensees exercise critique process was a performance deficiency. This finding was greater than minor because it was associated with the Emergency Preparedness Cornerstone. The finding affects the associated cornerstone objective to ensure 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. The finding was an identified weakness that demonstrated a level of performance that could preclude effective implementation of the Emergency Plan in an actual emergency. This finding was also determined to potentially have greater significance because the licensees exercise critique process failed to properly identify a weakness associated with a RSPS that was determined to be a DEP PI opportunity failure during a full-scale exercise. (Section 1EP1)
05000280/FIN-2006501-01Protective Actions for Sever Reactor Accidents2006Q4No Color: The inspectors identified a Severity Level IV non-cited violation (NCV) of 10 CFR 50.54(q) for implementing a change which decreased the effectiveness of the emergency plan without prior NRC approval. The licensee implemented an Emergency Plan change that modified the default Protective Action Recommendation (PAR) for the General Emergency classification to evacuate to 5 miles in all directions. The finding was evaluated using the NRCs Enforcement Policy because licensee reductions in the effectiveness of its emergency plan impacted the regulatory process, in that, NRC approval was not requested prior to licensee implementation of the change. This finding is of more than minor concern because the change made may be overly conservative in such a way as to place members of the public at unnecessary risk during evacuation of an area unaffected by a radiological release which would be more appropriately recommended for sheltering. The finding was determined to be a noncited Severity Level IV violation in accordance with Supplement VIII of the Enforcement Policy because it involved licensee failure to meet an emergency planning requirement not directly related to assessment and notification.
05000280/FIN-2007005-04Fibrous Material Left in Unit 1 Containment2007Q4On 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-2007005-07Control of Heavy Loads2007Q4The inspectors identified that the licensee failed to incorporate a heavy load lift analysis into their UFSAR. Failure to update the UFSAR to reflect aspects of heavy load lifts involving the reactor vessel head and include information from a reactor vessel head drop analysis was a violation of 10 CFR 50.71(e). The NRC has found industry uncertainty regarding the licensing bases for handling of reactor vessel heads, and as a result issued EGM 07-006, Enforcement Discretion for Heavy Load Handling Activities, on September 28, 2007. NEI has informed NRC of industry approval of a formal initiative that specifies actions each plant will take to ensure that heavy load lifts continue to be conducted safely and that plant licensing bases accurately reflect plant practices. The NRC staff believes implementation of the initiative will resolve uncertainty in the licensing bases for heavy load handling, and enforcement discretion related to the uncertain aspects of the licensing basis is appropriate during the implementation of the initiative. During inspection of heavy load lifts, the inspectors determined that the licensee implemented interim actions prior to the specified lifts in accordance with the industry initiative, thereby meeting the following criteria to warrant enforcement discretion: 1) The licensee had neither a single-failure-proof crane nor a load drop analysis (generic or plant-specific) that bounded the planned lifts with respect to load weight, load height, and medium present, so the licensee conducted the head lift at the minimum practicable height and flooded the refueling cavity with water during the head movement to limit the maximum potential impact velocity of the head. The licensee maintained the bottom of the head less than 15 feet above the refueling cavity water surface when the head was lifted above the guide studs. Once the cavity was fully flooded (greater than 23 feet above the reactor vessel flange), the reactor vessel head was allowed to be lifted more than 15 feet above the water surface as necessary to lift the head above immovable structures around the refueling cavity. 2) Included the movement of heavy loads as a configuration management activity in administrative controls established to implement 10 CFR50.65(a)(4). Therefore, consistent with EGM 07-006, we are exercising enforcement discretion for the above violation in accordance with Section VII.B.6 of the NRC Enforcement Policy and are not issuing enforcement action for the violation.
05000280/FIN-2008002-01Loss of thermal barrier cooling due to a failure to follow procedures2008Q1A 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-2008002-02Failure to Follow Start-up Procedure which resulted in Leaving Loose Fibrous Insulation in Containment2008Q1An 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-2008002-03Licensee-Identified Violation2008Q110 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-2008002-04Licensee-Identified Violation2008Q1Surry 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-2008003-01Licensee-Identified Violation2008Q2TS 6.4.A requires, in part, that detailed written procedures with appropriate check-off lists and instructions be provided for normal startup, operation, and shutdown of a unit, and of all systems and components involving nuclear safety of the station. TS 6.4.D requires, in part, that all procedures specified in TS 6.4.A be followed. Contrary to the above, in the fall 2006 Refueling Outage containment close-out, the licensee had not identified the debris located in the Containment Air Recirculation Fan Duct Ring in the basement of Unit 2 Containment during the performance of procedure 2-GOP-1.7. This debris was discovered on an engineering walkdown performed on 5/9/2008 as an extent of condition for NCV 05000280/2008002-02, Failure to Follow Start-up Procedure which Resulted in Leaving Loose Fibrous Insulation in Containment. This violation was entered into the licensees corrective action program as CA024916, CR028438: SURR Containment walkdown to verify the design input for GSI- 191 project. The debris has been removed from containment. This finding is of very low safety significance because the amount of debris removed did not have the capability to overwhelm the partially installed replacement ECCS and CSS strainers
05000280/FIN-2008004-01Inadequate Design control for the EDG Ambient Air Temperature Limit2008Q3The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control, for a change in the EDG ambient air temperature operating limits, from 100oF to 120oF, that was made without an adequate design analysis. The licensee entered the issue into their corrective action program (CAP) for resolution using condition report (CR) 102488. The inspectors concluded that the licensees failure to perform the necessary analysis to support the increase of the EDG ambient air temperature operating limit from 100oF to 120oF was a performance deficiency. The finding, more than minor in accordance with MC 0612, Appendix E, examples 3j and k, is associated with the design control attribute of the Mitigating System Cornerstone. The cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences was adversely affected. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4 the inspectors concluded that the finding is of very low safety significance (Green) because the condition did not represent an actual loss of safety function due to the ambient temperature exceeding 100oF but not exceeding 105oF. The finding also was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. A cross-cutting aspect was not assigned to the issue because it is not indicative of recent performance. (Section 1RO1)
05000280/FIN-2008004-02SW Silting in the Charging Pump Lube Oil coolers and TCVs2008Q3On February 19, 2008, charging pump 2-CH-P-1C was declared inoperable due to a high thrust bearing temperature (CR 091548) due to plugging of the TCV with silt. Fourteen (14) instances of SW silting, (five in Unit 1 and nine in Unit 2) were identified in ACE 13823 for the Unit 1 and 2 charging pump lube oil coolers and associated TCVs. Based on available information, the silting had not caused the charging pumps to become inoperable until the February, 2008 event. ACE 13823, performed due to fouling of 2-CH-E-5A (CR 102051), identified that the lowest flow velocities (1.7 ft/sec) occur in the SW lines to the charging pump lube oil coolers. During winter months, the flow rates are even lower. As a result of the low flow, silt/mud accumulates in the lube oil coolers and TCVs. The ACE stated, past corrective action dealt only with cleaning the heat exchanger rather than a permanent fix to the silting problem. ACE 13684, performed in response to fouling of 2-CH-E-5C (CR 091566), also identified low flow rates to be the cause of fouling of the TCV. Procedure PI-AA-200 (Rev. 2), Corrective Action; defines a Significant Condition Adverse to Quality (SCAQ) as: a condition Adverse to Quality that has, or if left uncorrected could have, an undesirable effect on plant safety, regulatory position, or environmental impact. It further states, because of the high regulatory or safety consequences associated with this type of condition the cause of the condition must be determined and corrective action taken to preclude repetition. Service water silting of the Unit 1 and 2 charging pump lube oil coolers and associated TCVs is a SCAQ which had an undesirable effect on plant safety and caused HHSI pump 2-CH-P-1C to be inoperable. The issues associated with the failure to take corrective action to prevent recurrence of SW silting in the charging pump lube oil coolers and TCVs is identified as an unresolved item (URI) pending additional inspection and review from the NRC. This URI is designated 05000280, 281/2008004-02, SW Silting in the Charging Pump Lube Oil Coolers and TCVs.
05000280/FIN-2008004-03Licensee-Identified Violation2008Q3Criterion XV, Nonconforming Materials, Parts, or Components, Appendix B, to 10 CFR Part 50, requires that measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures. Contrary to the above, during the 2008 Unit 2 Refueling Outage 21, the licensee installed two Rosemount pressurizer level transmitters (2460 and 2461) that had not been identified by the licensee as being part of a 10 CFR Part 21 notification in response to NRC Bulletin Nos. 90-01 and 90-01, Supplement 1, regarding a loss of fill-oil. This violation is of very low safety significance because the pressure level transmitter was still operable and within the performance criteria for measuring pressurizer level. This violation was entered into the licensees corrective action program as CR 105348.
05000280/FIN-2008004-04Licensee-Identified Violation2008Q310 CFR 50 Appendix B, Criterion III, Design Control requires, in part, that design changes, including field changes, shall be subject to design control measures commensurate to those applied to the original design. Contrary to the above, on August 30, 2008 the licensee installed a design change to the 1B ESW pump exhaust line without adequately evaluating each element of that change commensurate to the evaluations applied to the original design. This violation is of very low safety significance because the screen was removed. This violation was entered into the licensees CAP as CR 107948.
05000280/FIN-2008005-01Inadequate Work Instructions Result in Actuation of Unit 1 Safety Injection Train B2008Q4A Green self-revealing non-cited violation (NCV) of Technical Specification 6.4.A.7 was identified for failure to provide adequate work instructions for corrective maintenance on the safety injection (SI) system. The inadequate work instructions led to an inadvertent actuation of the Unit 1 B train of safety injection on October 29, 2008. The proposed corrective actions are to provide guidance/restrictions in the work planning process to assure appropriate reviews are obtained, commensurate with the safety significance of the work. The finding is greater than minor because it had an actual safety impact by causing a SI and if left uncorrected could lead to a more significant safety issue. The finding is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone 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 is determined to be of very low safety significance (Green) based on a Phase 3 SDP analyses performed by a regional Senior Reactor Analyst. This finding has a cross-cutting aspect in the area of human performance, decision making, because the decision to continue with the planned work was made without a complete understanding of either the effects of the job steps or the worst case possible unintended consequences (H.1(b)). (Section 4OA3.1
05000280/FIN-2008005-02Inadequate Review of Vendor Information Led to Unit 1 Manual Reactor Trip2008Q4A Green self-revealing Finding was identified for failure to provide adequate vendor oversight for non-safety related work, which led to the incorrect installation of balance weights on the Unit 1 main turbine. As a result, the turbine experienced high vibrations during startup on April 20, 2008, which required the insertion of a manual turbine and reactor trip. The licensee entered the deficiency into the corrective action program for resolution (CR 096233). Corrective actions to correct the balance move, implement peer review requirements, and procedural changes that require independent verification of the balance move location and weight have been implemented. A violation of regulatory requirements was not identified. The finding is greater than minor because it had an actual impact on safety, it led to a plant trip, and is associated with the human performance attribute of the Reactor Safety Initiating Events cornerstone and adversely affected the cornerstone 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 per Attachment 4 of the SDP, screened to very low safety significance (Green) because it did not contribute to both an initiating event and the likelihood of a loss of mitigating equipment or functions. The cause of the finding is related to the cross-cutting element of human performance work practices. Human error prevention techniques such as peer checks were not invoked by the licensee (H.4(a)). (Section 4OA3.2
05000280/FIN-2008005-03Licensee-Identified Violation2008Q4Technical Specification 6.4 A.1 requires, in part, that detailed written procedures with appropriate check-off lists and instructions shall be provided for the operation of components involving nuclear safety of the station. Licensee procedure GMP-012, Roving Flood Watch Responsibilities, requires that the water tight door to mechanical equipment room (MER) #3 be closed or monitored. Contrary to the above, on October 20, 2008, the watertight door to MER #3 was found open and unattended. This issue was identified in condition report 115052. This finding is of very low safety significance based on the results of a Phase 3 significance determination process
05000280/FIN-2008005-04Licensee-Identified Violation2008Q4Technical Specification 3.1.A.3.b requires the pressurizer safety valves (PSV) as found lift pressure setpoint be within +/- 3% of nominal 2485 psig. Contrary to the above, on May 2, 2008 the Unit 2 PSV 2-RC-SV-2551C as found lift pressure was 4.7% below the nominal value and on May 9, PSV 2-RC-SV-2551B as found lift pressure was 3.5% below nominal. This issue was identified in CR 097633 and ACE 013757. This finding is of very low safety significance because the PSVs were capable of performing their safety function, and operation with the low as-found lift setpoints were within the limits assumed in the accident analysi
05000280/FIN-2008006-01Failure to Evaluate and Use Limiting Case 4160 VAC Bus Frequency and Voltage in Design Calculations (Section 1R21.2.12)2008Q1The inspectors identified two examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate variations of emergency diesel generator (EDG) output frequency in electrical design loading calculations, and failure to consider worst case 4160 VAC bus voltage in safety related motor starting calculations. This finding was entered into the licensees corrective action program as condition reports (CR) 091493 and 091494. Planned corrective actions included revision of the EDG loading calculations to incorporate the most limiting voltages and frequencies. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in any EDG being inoperable based upon additional analysis that showed that the EDGs had sufficient margin to accommodate the increased loading due to worst case acceptably high EDG output frequency; and all safety related motor loads remained operable since they were still capable of starting with the revised worst case low voltage values. (Section 1R21.2.12
05000280/FIN-2008006-02Failure to Use Appropriate Acceptance Criteria for Testing Battery Voltage at the One Minute Mark (Section 1R21.2.13)2008Q1The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for incorrect acceptance criteria in test procedure 1-EPT-0106-01, Main Station Battery 1A Service Test. This finding was entered into the licensees corrective action program as condition report 091906. Planned corrective actions included revision of the main station battery test procedures to include the correct voltage at the one minute mark. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the station batteries to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in station batteries being inoperable based upon a recent review of station battery discharge test results. The inspectors determined that the lack of a thorough evaluation of condition report 022112, which addressed deficiencies in station battery test procedures such that resolutions addressed causes, was a significant cause of this performance deficiency. Failure to thoroughly evaluate problems such that resolutions address causes is directly related to the Corrective Action Program component of the cross-cutting area of Problem Identification and Resolution and the aspect of thorough evaluation of problems (P.1(c)). (Section 1R21.2.13
05000280/FIN-2008006-03Failure to Use Limiting Case High dP In 2-FW-MOV-260A Design Calculations (Section 1R21.2.19)2008Q1The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to evaluate the most limiting differential pressure (dP) for opening valve 2-FW-MOV-260A, auxiliary feedwater (AFW) cross-tie motoroperated valve (MOV). This finding was entered into the licensees corrective action program as condition report 091698. Planned corrective actions included internal inspection of the valve and revision of the evaluation that identified the most limiting dP for opening. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective ensuring the availability, reliability, and operability of the AFW system to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design and plant modifications. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in 2-FW-MOV-260A being inoperable based upon additional analysis which showed that the MOV had sufficient margin to accommodate opening against the worst case high dP. The inspectors determined that the lack of control or understanding of the actual margin to maximum allowable dP to open 2-FW-MOV-260A was a significant cause of this performance deficiency. Failure to maintain design margins is directly related to the Resources component of the cross-cutting area of Human Performance and the aspect of maintenance of plant safety by the maintenance of design margins (H.2(a)). (Section 1R21.2.19)
05000280/FIN-2008006-04Licensee-Identified Violation2008Q1The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs. • 10 CFR 50, Appendix B, Criterion III, Design Control, requires that that measures shall be established to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. UFSAR Table 15.2-1 lists the AFW pumps as being components that will not fail during a tornado since they are protected by tornado resistant structures. Contrary to this, turbine drive AFW pumps 1/2-FW-P-2 were not completely protected in that the steam exhausts from the turbines could have been blocked by tornado missile damage. This was identified in the licensees corrective action program as CR 001132. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because it involved a severe weather initiating event and did not degrade more than one train of a multi-train safety system
05000280/FIN-2009002-01Licensee-Identified Violation2009Q110 CFR 50, Appendix B, Criterion III, Design Control; requires, in part, that design changes be subject to design control measures commensurate with those applied to the original design. Contrary to the above, design control measures commensurate with the original design were not implemented for a design change implemented in the 1980s to the safety-related motor operated valve (MOV) control circuits for the condenser inlet isolation valves. The condenser inlet isolation valve closing circuits were modified to actuate on a turbine building flooding signal The circuit modification did not include electrical isolation via qualified isolation devices, per IEEE-279, Criteria for Protection Systems for Nuclear Power Generating Stations, between the safety-related closing circuit for the MOVs and the non-safety related circuit in the flood control panel. This was identified in the licensees CAP as CR 320789. A regional Senior Reactor Analyst performed a Phase 3 evaluation under the Significance Determination Process and concluded that the performance deficiency was of very low safety significance. The most credible accident sequence involved a seismic event which caused the loss of automatic isolation of the condenser inlet motor operated valves via the panel in question and caused a condenser pipe break. Manual closure of the valves which would terminate the turbine building flood, though possible, failed. Consequently, all electric power to critical mitigation equipment was lost and core damage ensued. Critical assumptions included the low initiating event frequency of such an event and that operator actions could terminate the flooding
05000280/FIN-2009003-01Inadequate Work Instructionsfor Installations of a Design Change2009Q2A self-revealing Green non-cited violation (NCV) of Technical Specification6.4, Unit Operating Procedures and Programs; was identified for the failure to provide adequate work instructions for installation of design change (DC) SU-08-0001, for engine-driven emergency service water pump 1-SW-P-1A. Corrective action to remove the modification from the A pump was completed and reasonable compensatory measures established for all three pumps pending removal / alteration of the exhaust piping modifications. The licensee entered this issue into the CA program as CR 337337.The finding, associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone, is more than minor because it adversely affected the cornerstone objective to ensure the availability, reliability, and operability of 1-SW-P-1A to perform its safety function during a design basis event. Evaluated using a Phase IISDP risk analysis per Appendix A of MC-0609, the finding was determined to be of very low safety significance (Green) due to availability of the two remaining ESWPs which provided full mitigation capability for the safety functions required. A crosscutting aspect in the area of human performance work control was assigned to the findin
05000280/FIN-2009003-02Licensee-Identified Violation2009Q2Technical Specification 6.4.B.1 requires, in part, that the entrance to each high radiation area be barricaded and conspicuously posted as a high radiation area. Contrary to this, on May 3, 2009, the entrance to a high radiation area was found not barricaded and conspicuously posted. Approximately two hours after the swing gate and posting was replaced it was discovered that two of the wire ties used to secure the sign to the gate had failed resulting in the posting facing the wrong direction and again failing to meet the requirements of TS 6.4.B.1. This was identified in the licensees CAP as CR 333359 and apparent cause evaluation ACE017570. This finding is of very low safety significance because it did not lead to an overexposure, there was not a substantial potential to result in an overexposure, and the ability to assess dose was not compromised
05000280/FIN-2009003-03Licensee-Identified Violation2009Q2Technical Specification 6.4.A requires in part that detailed written procedures with appropriate instructions are provided for corrective maintenance operations which could have an effect on the safety of the reactor. Contrary to this, inappropriate work instructions were provided to repair ESWP 1-SW-P-1A which caused the ESWP to be found inoperable on June 6, 2009. This was identified in the licensees CAP as CR 337320. This finding is of very low safety significance because the ESWP was not out-of-service for more than the allowed TS outage time and the two remaining ESWPs were available
05000280/FIN-2009003-04Licensee-Identified Violation2009Q2Technical Specification 3.14 requires, in part, that all three ESW pumps be operable when both units are operating with the exception that one pump may be inoperable up to seven days for maintenance and testing. Contrary to this, on May 31, 2009, the A ESWP was determined to be inoperable for approximately 22 days and on May 12, 2009, two ESW pumps were inoperable for approximately 40 minutes, however, one of the pumps was available. This was identified in the licensees CAP as CR 337337. This finding is of very low safety significance because the dominant core damage sequence was loss of service water, and mitigated by the availability of two ESWPs
05000280/FIN-2009004-01Inadequate compensatory measures for the impairment of fire detection systems2009Q3The inspectors identified a Green NCV of the Surry operating license, section 3.I Fire Protection, for an inadequate procedure that resulted in compensatory continuous fire watches in MERs #3 and #4 being inadequate (CR 342078). This finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external events and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, to analyze this finding because the condition had an adverse affect on the Fixed Fire Protection Systems element of fire watches posted as a compensatory measure for fixed fire protection system outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding (i.e. fire watches) is expected to display nearly the same level of effectiveness and reliability. Using Manual Chapter 0609, Appendix F, this finding was determined to be of very low safety significance (Green). A cross-cutting element was not assigned to this finding because the most significant contributor to the performance deficiency is not reflective of current performance
05000280/FIN-2009004-02Failure to provide an adequate basis for operability of ESW pump 1-SW-P-1B2009Q3The inspectors identified a Green finding for the incorrect operability determination for emergency service water pump 1-SW-P-1B on August 1, 2009, after vibrations had increased 391% in the vertical plane (CR 343396). A violation of regulatory requirements was not identified. The pump, declared inoperable on August 2, was replaced within the Technical Specification allowed outage time. The finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or the loss of a single train of ESW for greater than the allowed outage time. This finding has a cross-cutting aspect in the area of human performance, decision making, because the licensee failed to use conservative assumptions in their operability decision for 1-SW-P-1B (H.1.b)
05000280/FIN-2009004-03Inadequate Tornado Protection for Engine-Driven Emergency Service Water Pumps 1-SW-P-1A/B/C2009Q3The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control. The design change for the emergency service water pumps (DC-SU-08-0001) was not adequate to protect the diesel-driven emergency service water pumps from damage resulting from a tornado missile as required by the UFSAR (CRs 337720, 337337, 341557). Pending resolution, interim compensatory measures have been established to provide assurance the pumps will be capable of performing their safety function. The finding, associated with the design control attribute of the mitigating systems cornerstone, is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, was determined to be of very low safety significance (Green) because of the extremely low initiating event frequency for a tornado. A phase III risk analysis was performed because the finding screened potentially risk significant for a severe weather initiating event. This finding has a cross-cutting aspect in the area of human performance, resources, because the licensees design documentation for DC SU-08-0001 and ET-S-08-0032 was not complete and accurate which led to the installation of inadequate modifications on ESWPs 1-SW-P-1A/1B/1C (H.2.c)
05000280/FIN-2009004-04Inadequate Work Instructions lead to packing failure of ESW Pump 1-SW-P-1B2009Q3A self-revealing Green NCV of Technical Specification 6.4, Unit Operating Procedures and Programs; was identified for the failure to provide adequate work instructions for maintenance on 1-SW-P-1B, a safety-related component, which led to the failure of the pumps packing gland on August 26, 2009 and required the pump be removed from service and repacked (CR 346268). The finding is associated with the equipment performance attribute of the mitigating systems cornerstone and is more than minor because it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or loss of a singe train of ESW for more than its allowed outage time. This finding has a cross-cutting aspect in the area of human performance, resources, in that a complete and accurate procedure was not available to assure nuclear safety during replacement of 1-SW-P- 1B (H.2.c)
05000280/FIN-2009004-05Failure to promptly identify and correct a ground on safety bus 1H2009Q3A Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for failure to promptly identify and correct a condition adverse to quality related to a ground on emergency safety bus 1H. This resulted in the degraded condition being allowed to exist for 72 days prior to de-energizing the containment recirculation fan and correcting the adverse condition (CR 336041). This finding is more than minor because it adversely impacted the equipment performance attribute of the reactor safety mitigating system cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, was determined to be of very low safety significance (GREEN). The finding screened to a Phase II assessment on the assumption that a second ground would result in a complete loss of the safety bus and its safety function. The Phase II analysis was performed for the core damage sequence Loss of a 4.16Kv Bus (1J or 1H) utilizing an increased initiating event likelihood (IEL) value of 1 due to the degraded condition of the 1H bus. The duration of the degraded condition was 72 days. The finding was not greater than Green because full mitigation capability of the opposite train remained available. This finding has a cross cutting aspect in human performance, decision making, in that the licensee did not use conservative assumptions in their decision making process (H.1.b)
05000280/FIN-2009004-06Ineffective action for ELU performance deficiencies2009Q3The inspectors identified a Green NCV of Surry operating licenses, section 3.I Fire Protection, for failure to promptly identify and correct a condition adverse to fire protection in regard to Appendix R emergency lighting unit performance failures, due to inadequate configuration control of the emergency lights defeat switch. Failure to reposition the switch following maintenance and or inadvertent switch manipulation has over time led to numerous Appendix R emergency lights being discovered non-functional. Corrective action to address the failure to restore the switch following maintenance has been taken and actions to prevent inadvertent manipulation are being evaluated (CR 352214). The finding is more than minor because it adversely affected the external factors attribute (fire) of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability and availability of the emergency lighting units (ELUs) was affected. The finding, evaluated per MC-0609, Appendix F, Fire Protection Significance Determination Process, was determined to be of very low safety significance (Green). The finding affected post fire safe shutdown and was assigned a low degradation rating because the issue did not have a significant impact on safe shutdown operations because there was not a simultaneous wide spread failure of the ELUs. This finding has a cross-cutting aspect in the area of problem identification and resolution, because the licensee did not take adequate corrective action in a timely manner to address an adverse trend in ELU functionality (P.1.d)
05000280/FIN-2009004-07Failure to remove blocking devices from pipint supports2009Q3The inspectors identified a Green NCV of Technical Specification 6.4, Unit Operating Procedures, associated with blocking devices not being removed from piping supports following maintenance due to procedure issues related to procedure adequacy and adherence. The blocking devices were removed upon discovery and appropriate corrective actions established to address the issue (ACE 017736). The finding is more than minor because if left uncorrected the performance deficiency could potentially lead to more significant safety concerns. The finding is associated with the procedure quality attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding, evaluated per MC-0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because operability of a safety system, though challenged, was never lost. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensees corrective actions were not effective in identifying additional blocked spring hangers on safety-related systems or preventing further configuration control issues associated with spring hanger blocking devices (P.1.d)
05000280/FIN-2009004-08Licensee-Identified Violation2009Q3Surry 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. CM-AAFPA- 100, Fire Protection/Appendix R (Safe Shutdown) Program, Rev. 10, Attachment 2, Section 3.1.2.o.2 states that penetration seals must provide equal or greater fire rating than that of the fire barrier. Contrary to the above, on June 5, 2009, the licensee breached a fire barrier penetration seal that separates MER #4 (Fire Area 54) from the Turbine Building (Fire Area 31). The breach existed until it was discovered and sealed on July 29, 2009. This issue was identified in the licensees corrective action program as CR 340416. This violation is of very low safety significance because the violation did not affect ignition frequencies, detection, or suppression system performance
05000280/FIN-2009005-01Inoperability of MCR isolation damper 1-VS-MOD-103D due to failure to promptly identify and correct internal hydraulic leakage2009Q4A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI, was identified for the failure to correct a condition adverse to quality which led to main control room isolation damper 1-VS-MOD-103D being inoperable for approximately19 hours on September 21-22, 2009 (CR 349075). The actuator was repaired and is scheduled for replacement in 2010.The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective, as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated perMC-0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a crosscutting aspect in the area of human performance, resources, in that equipment and other resources were not made available to assure nuclear safety by minimizing preventative maintenance deferrals (H.2.a)
05000280/FIN-2009005-02Failure to perform an adequate operability determination for main control room isolation damper 1-VS-MOD-103D2009Q4A self-revealing Green Finding was identified for the incorrect operability determination of main control room isolation damper 1-VS-MOD-103D. The damper, declared operable and left in-service following loss of power to its hydraulic pump on September 21, 2009 (CR 349003), failed to close on demand, on September 22, 2009. The damper was inoperable for approximately 19 hours (CR 349075) before power was restored to the pump, the damper closed, and the actuator repaired. The finding, associated with the performance attribute of the barrier integrity cornerstone, is more than minor because it adversely affected the cornerstone objective as it relates to control room integrity, to provide reasonable assurance physical design barriers protect public health and safety. The finding, evaluated perMC-0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not result in a loss of safety function or the loss of a singe train of the control room isolation boundary for more than its allowed outage time. This finding has a cross-cutting aspect in the area of problem identification, corrective action program, in that an adequate operability assessment that thoroughly evaluated the degraded condition of 1-VS-MOD-103D was not performed (P.1.c)
05000280/FIN-2009005-03Licensee-Identified Violation2009Q410 CFR 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 November 12-13, 2009, the licensee altered the evaluated work sequence for restoring the Unit 2J safety bus and 2A battery and DC electrical buses without re-evaluating the risk. This event was documented in the licensees corrective action program as CR 357883. The finding is of very low safety significance (Green) because the change in risk had existed for only a short period of time prior to being corrected.
05000280/FIN-2009005-04Licensee-Identified Violation2009Q4Technical Specification 6.4 requires, in part, that detailed written procedures with appropriate check-off lists and instructions shall be provided and followed for the operation of components involving nuclear safety of the station. Licensee procedure GMP-012, Roving Flood Watch Responsibilities, requires that the water tight door to mechanical equipment room (MER) #3 be closed or monitored. Contrary to the above, on October 5, 2009, the watertight door to MER #3 was found open and unattended. This was identified in the licensees CAP as condition report 350894. A regional Senior Reactor Analyst performed a Phase 3 evaluation under the Significance Determination Process and concluded that the finding was of very low safety significance (Green). The dominant accident sequence involved an internal flood in the Mechanical Equipment Room #3 that was not isolated. Eventually, the flooding would have caused a sustained loss of all electrical power to the facility resulting in core damage. The exposure period used in the evaluation was less than two hours.
05000280/FIN-2009005-05Licensee-Identified Violation2009Q4Technical Specification 6.4 requires, in part, that detailed written procedures with appropriate check-off lists and instructions be provided and followed for maintenance which could have an effect on the safety of the reactor. Contrary to the above, procedures for maintenance of the ESW pumps did not contain appropriate instructions to assure water tight integrity of the ESW pump house was maintained, and three unsealed penetrations in the emergency service water building floor were identified by the licensee on September 24, 2009. This was identified in the licensees CAP as condition report 349378. A regional Senior Reactor Analyst performed a Phase 3 evaluation under the Significance Determination Process and concluded that the finding was of very low safety significance (Green). The dominant accident sequence involved a postulated tornado-induced non-recoverable Loss of Offsite Power that failed all the Emergency Service Water Pumps, due to the performance deficiency. The accident sequence included numerous combinations of motor operated valves in the Service/Circulating Water System failing which drained the intake canal without the ability to makeup the lost inventory. The Auxiliary Feedwater System would have provided secondary side heat removal. However, without the Service Water System, cooling to the Reactor Coolant Pump seals would not have been maintained and a Seal Loss of Coolant Accident would have occurred, leading to core damage. The exposure period evaluated was one year, and no recovery of the Emergency Service Water Pumps was considered.
05000280/FIN-2009006-01Failure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves Nor Set Goals and Monitor Under 10CFR50.65(A)(1)2009Q4The 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-2009007-01Qualification of Fire Barrier Floor/Wall Penetration of Aluminum Conduit Through Sleeve2009Q2The team identified an unresolved item (URI) involving the qualification documentation for wall and floor fire barrier penetration seals. While inspecting the wall and floor fire barrier penetration seals, the team requested the licensees documentation for the qualification of a particular penetration seal configuration. That configuration was for one aluminum schedule 40 conduit (of various sizes as applicable) penetrating a 6 in. diameter floor or wall sleeve where the floor or wall was of poured concrete construction and the sleeve void around the conduit was filled with foamed silicon to the thickness of the floor or wall. The documentation package requested should establish a 3-hour fire rating, since the rated fire barrier walls and floors were required to have a 3-hour rating. In response, the licensee presented Impell Corporation Calculation No. 1250-111-C01, Penetration Seal Configuration Documentation Package, 10 in. Dow Corning Q3-6548 Silicone RTV Sealing Foam/North Anna and Surry, Rev.1. The qualification package or calculation was based on a tested configuration similar to that described above, except that the conduit was 3 in. or 4 in. galvanized steel. The team informed the licensee that this calculation was not valid to qualify aluminum conduit due to the lower melting temperature and greater heat conductance of aluminum as compared to steel. The licensee later transmitted supplemental information which included a fire barrier penetration seal fire test report for large diameter aluminum conduits through a sleeve. This new information was not a formal calculation comparing it to any installed penetration seal configuration at Surry. Moreover, certain aspects of the test data such as the temperature rise on the unexposed surfaces may not meet the licensing basis. At the time of issuance of this report, the team did not have sufficient information to determine the design criteria of that penetration seal. The team was aware that the fire barrier penetration seal configurations in question could probably be qualified by existing nuclear industry penetration seal testing data; therefore, there was no immediate safety concern. The licensee Initiated CR 339567 with an action item to establish a valid qualification package for the penetration configuration described above. URI 05000280, 281/2009007-01, Qualification of Fire Barrier Floor/Wall Penetration of Aluminum Conduit Through Sleeve, was established to track this issue until the final qualification package is reviewed
05000280/FIN-2009007-02Availability of Portable Ventilation Fans for Use by the Fire Brigade2009Q2The 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-2009007-03Failure to Establish Maintenance for Backup Battery for the Halon 1301 System in ESGRs2009Q2The team identified a performance deficiency and Green NCV for failing to implement a maintenance program for the backup batteries for the Halon 1301 system for the emergency switchgear rooms to ensure on a continuing basis that 24-hour backup power was available as required by the fire protection program (FPP) and Units 1 & 2 Operating License Condition 3.I, Fire Protection. The licensee entered this finding into their corrective action program, and demonstrated that the backup battery had sufficient capacity in the short term until the long term corrective actions can be implemented. The licensees failure to implement a maintenance program to help ensure that the backup battery for the Halon 1301 system continued to meet its licensing basis requirement of providing backup power for 24 hours is a performance deficiency. The finding is more than minor because the backup battery actually degraded on several occasions in the past, and the finding is associated with the reactor safety, mitigating systems, cornerstone attribute of protection against external factors, and affected the objective of ensuring reliability and capability of systems that respond to initiating events. The finding was determined to be of very low safety significance because it represented a low degradation of the fixed fire suppression systems. A cross-cutting aspect was not identified in relation to this finding since the cause was not representative of current license performance
05000280/FIN-2010002-01Failure to identify a non-conservative error in the quarterly TS surveillance for the Unit 1 A battery2010Q1The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for failure to identify that a non-conservative error had been introduced into the Unit 1 A main station battery quarterly technical specification surveillance procedure (CR366388). The licensee modified the procedure to eliminate the non-conservative error. The inspectors determined the failure to identify a non-conservative error which was introduced into the TS quarterly surveillance procedure following the replacement of individual battery cells, was a condition adverse to quality and a performance deficiency which was reasonably within the licensees ability to foresee and correct, and should have been prevented. The finding was more than minor because if left uncorrected the non-conservative error in 1-EPT-0103-01 would have the potential to lead to a more significant safety concern. Specifically, this is because the error was large enough to mask cell degradation and an inoperable cell. The finding was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the safety related 125 VDC station batteries that provide class 1Ebackup power to risk significant components needed to prevent undesirable consequences during a loss of offsite power event. The finding was evaluated using MC-0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green) because operability of the Unit 1 A battery was not lost and the error was removed prior to the next quarterly surveillance. This finding had a cross cutting aspect in the area of problem identification and resolution because the licensee did not evaluate and communicate relevant external OE, including vendor recommendations, to affected internal stakeholders in a timely manner P.2(a). Specifically, the caveat to have cells on a float charge for 72 hours was not fully evaluated when the battery cells were replaced.
05000280/FIN-2010002-02Emergency Plan Minimum Staffing2010Q1An 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-2010003-01Failure to demonstrate that the reliability of systems or components were effectively controlled per 10 CFR 50.652010Q2The NRC identified a Green Non-Cited Violation of 10CFR50.65 a(2) for the licensees failure to demonstrate that the reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by was being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function. Specifically, the licensees MR program would not demonstrate that a system should remain in category a(2) as defined by regulatory requirements The inspectors determined the licensees MR program could not demonstrate that reliability of High Safety Significant (HSS) systems and Low Safety Significant (LSS) systems in stand-by were being effectively controlled through the performance of appropriate preventative maintenance, such that the systems or components remain capable of performing their function is a performance deficiency. Specifically, the monitoring established by the license did not effectively demonstrate that systems in a(2) were being effectively controlled through performance of appropriate preventative maintenance. This masking of poor equipment performance does not allow the licensee to determine if a system should be in increased monitoring of a(1) The finding was more than minor because it adversely affected the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of HSS and LSS systems to perform their functions when required. Specifically, multiple HSS and LSS systems could have a high probability of failure, because poor equipment performance would not be recognized by the licensee. This could prevent a poor performing system from being placed into the a(1) category when required and appropriate corrective action would not be taken The finding was evaluated using MC-0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined to be of very low safety significance (Green), because the finding did not involve an actual failure of equipment. This finding had a crosscutting aspect in the area of human performance and resources because the licensee did not ensure that personnel, procedures, and other resources were available and adequate to assure proper implementation of MR program. The MR personnel did not have the training required to implement the program within the required industry regulations and guidelines.
05000280/FIN-2010003-02Inadequate rigging practices result in damage to risk significant equipment2010Q2

A self-revealing Green Finding was identified for failure to adequately rig a 300 pound motor in the auxiliary building in accordance with the manufacturers recommendations on May 11, 2010. As a result, the motor slipped from its rigging and dropped approximately 15 feet onto the A component cooling water (CCW) pump motor below, damaging the motors cabling and electrical junction box. The CCW pump was declared inoperable (CR 380834), the damage was repaired, and the CCW pump restored to an operable status on May 15, 2010

Inspectors determined that the failure to implement adequate rigging practices in accordance with vendor recommendations as required by procedure MA-AA-101, Revision 5, Fleet Lifting and Material Handling constituted a performance deficiency and a finding which was reasonably within the licensees ability to foresee and correct and which should have been prevented. The finding is similar to MC 0612, Appendix E example 4.f, and is more than minor because it resulted in damage to and inoperability of a risk significant component. The finding is associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events which upset plant stability and challenge critical safety functions during shutdown as well as power operations because a loss of the component cooling water system would have resulted in a unit transient. The finding, evaluated per Attachment 4 of MC-0609, Phase 1 Initial Screening and Characterization of Findings, was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a plant transient and the loss of accident mitigation equipment. This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not make safety/risk significant decisions using a systematic process, especially when faced with uncertain decisions, to ensure safety is maintained (H.1(a)). Specifically, the rigging team made safety/risk significant decisions within lifting/rigging procedures that did not include a systematic process for evaluating each lift, especially loads <5000 lbs in the vicinity of risk significant equipment.

05000280/FIN-2010004-01Licensee-Identified Violation2010Q310 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.
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