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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000400/FIN-2018003-01 | 2018Q3 | Green | P.5 | Self-revealing | Failure to Implement Adequate Periodic Exercising of Turbine Trip Solenoid Operated Valves | A self-revealing Green finding was identified for the licensees failure to establish and implement adequate preventive maintenance (PM) for exercising the turbine electro-hydraulic auto-stop trip (AST) solenoid operated valves (SOVs) in accordance with procedure AD-EG-ALL-1202, Preventive Maintenance and Surveillance Testing Administration. As a result of the failure to exercise the SOVs at the weekly vendor recommended frequency, three of the four SOVs experienced mechanical binding (sticking) which rendered the turbine emergency trip system incapable of tripping the main turbine within the time response requirements of Technical Specifications. |
| 05000400/FIN-2018002-07 | 2018Q2 | Severity level Minor | Self-revealing | Minor Violation | A minor, self-revealing violation of TS 6.8.1.a, Procedures and Programs,was identified for failure to follow procedure AD-OP-ALL-0200, Clearance and Tagging. On April 7, 2018, while the plant was in Mode 3 at 0 percent power, the licensee isolated breaker DP-1A-1 circuit 28 in accordance with clearance OPS-1-18-5015-DEH MODS-0093. Isolating this breaker caused an unexpected auto start signal for both motor driven auxiliary feedwater (MDAFW) pumps for a loss of last running main feed pump despite the 1B main feedwater pump still being in operation. Both MDAFWs started and operators manually secured the 1B main feedwater pump to maintain proper feedwater flow to the steam generators. TS 6.8.1.a, requires, in part, that written procedures be implemented covering activities referenced in Regulatory Guide 1.33, Revision 2, dated February 1978, including safety-related activities carried out during operation of the reactor plant. Procedure AD-OP-ALL-0200, Section 5.5, step 4, states Clearance impacts must be evaluated to ensure that effects on systems and components outside of the boundary are identified and are acceptable, or properly dispositioned. Contrary to this requirement, the licensee did not identify that the isolation of breaker DP-1A-1 circuit 28 would cause the MDAFWs to auto start in Mode 3 when developing clearance OPS-1-18-5015-DEH MODS-0093. Screening: The violation is minor because the impact to the plant was minimal; the unit was in Mode 3 throughout the event, the reactor remained subcritical, and feedwater flow to the steam generators was not lost. Enforcement: Because the performance deficiency is minor, it will not be subject to enforcement action in accordance with the NRCs Enforcement Policy. The licensee entered this issue into their CAP as NCR 02196873. The associated LER is closed. | |
| 05000400/FIN-2018002-06 | 2018Q2 | Green | H.5 | NRC identified | Failure to Implement Viable Compensatory Actions with Seismic Monitoring System Out of Service for Planned Preventive Maintenance | An NRC-identified Green NCV of 10 CFR 50.54(q)(2) was identified for the licensees failure to follow and maintain the effectiveness of its emergency plan that meets the requirements of the risk-significant emergency planning standard 10 CFR 50.47(b)(4). Specifically, the licensee failed to implement viable compensatory actions while conducting planned preventive maintenance that rendered both seismic monitoring systems unavailable for 53.5 hours resulting in a loss of emergency assessment capability for declaring a Notification of Unusual Event under Emergency Action Level (EAL) HU2.1 for a seismic event. |
| 05000400/FIN-2018002-05 | 2018Q2 | Green | H.8 | NRC identified | Failure to Follow Secondary Water Chemistry Plan for Elevated Levels of Secondary Water Impurities | An NRC-identified Green NCV of TS 6.8.4.c, Secondary Water Chemistry, was identified for the licensees failure to follow secondary water chemistry control requirements in accordance with procedure CSD-CP-HNP-0002, Harris Secondary Water Chemistry Strategic Plan. . Specifically, the licensee remained at 100% power for approximately 10 hours after entering secondary water chemistry Action Level 3 due to elevated chlorine and sulfates chemical impurity concentrations, which was contrary to the procedure requirements to downpower the unit to below 5% power as quickly as safe plant operation permits. This unit downpower delay allowed additional time for the chemical impurities to adversely affect the steam generators. |
| 05000400/FIN-2018002-04 | 2018Q2 | Green | H.12 | Self-revealing | Failure to Implement Adequate Steam Generator Blowdown Demineralizer Control Procedures | A self-revealing Green NCV of Technical Specifications (TS) 6.8.1.a, Procedures and Programs, was identified for licensees failure to establish and implement adequate steam generator blowdown demineralizer control operating procedures resulting in exceeding secondary water chemistry Action Level 3 criteria for impurities in the steam generators. Specifically, the licensee did not implement adequate isolation valve controls between the demineralizer resin regeneration system and the feedwater system during resin regeneration activities. This open path allowed leakage of sulfates and chlorides into the feedwater system. The level of these impurities exceeded the secondary chemistry Action Level 3 threshold and resulted in an unplanned shutdown. |
| 05000400/FIN-2018002-03 | 2018Q2 | Severity level IV | NRC identified | Failure to Adequately Document Changes to the Emergency Plan | The inspectors identified multiple examples of a Severity Level IV (SL-IV) NCV of 10 CFR 50.54(q)(3), for changes to the licensees radiological emergency plan (E-Plan) associated with protective action recommendation (PAR) procedures and emergency response equipment that failed to demonstrate that the changes would not reduce the effectiveness of the E-Plan. Specifically, the licensee did not provide an adequate analysis to demonstrate that the removal of the sheltering in-place PARs was not a reduction in effectiveness of the E-Plan. Additionally, the licensee did not perform an analysis demonstrating that the removal of a temporary diesel generator providing a backup source of power to the Technical Support Center (TSC) did not reduce the effectiveness of the E-Plan. | |
| 05000400/FIN-2018002-02 | 2018Q2 | Green | P.6 | NRC identified | Inadequate Fire Brigade Performance Assessment of Announced Fire Drill | An NRC-identified Green NCV of 10 CFR 50.48(c) and National Fire Protection Association (NFPA) Standard 805, Section 3.4.3, Training and Drills, was identified for the licensees failure to adequately assess the fire brigade performance during an announced fire drill conducted March 21, 2018. Specifically, the inspectors identified several fire brigade performance deficiencies, improvement items, and lessons learned that were not identified and documented in the licensees corrective action program during the fire drill critique as required by the licensees fire drill administrative control procedure. |
| 05000400/FIN-2018002-01 | 2018Q2 | Green | P.1 | NRC identified | Failure to Promptly Identify and Correct a Condition Adverse to Quality For a Through-Wall Leak in the ESW Screen Wash Piping | An NRC-identified Green NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Actions, was identified for the licensees failure to promptly identify and correct a condition adverse to quality involving through-wall leakage in the B train ESW screen wash piping. Specifically, on April 30, 2018, operators failed to initiate a work request or condition report after security personnel reported through-wall leakage in the B train ESW screen wash piping. No further follow-up or corrective actions were taken until May 3, 2018, when NRC inspectors identified the same through-wall piping leakage during a plant walkdown inspection and reported the degraded condition. |
| 05000400/FIN-2018410-01 | 2018Q1 | Green | NRC identified | Security | ||
| 05000400/FIN-2018001-01 | 2018Q1 | NRC identified | Adequacy of Fire Brigade Response During Fire Dril | The inspectors identified an URI during the March 21, 2018, announced fire drill that was observed. The drill involved an electrical failure inside the A transfer panel located in the RAB 286 elevation A cable spread room. The fire scenario assumed the electrical failure caused an explosion and fire in the room. The inspectors noted several performance weaknesses during the drill:The fire brigade leader directed three fire brigade members into the fire hot zone to fight the fire as the attack team. Since there is a 5-member fire brigade, only two fire brigade members remain, one of which is the fire brigade leader (who also serves as the site incident commander (SIC)), to be part of the designated 2-out rescue team, required when fighting internal building fires. This 2-out rescue team is responsible, if necessary, for providing assistance or rescue for any or all of the attack team members. The inspectors were concerned that this fire brigade strategy could result in challenges with fire brigade leader command and control, and with the effectiveness of conducting rescues. The fire brigade leader could be hampered in his primary role of directing a site fire response while serving as a rescue team member. Adding to this complication, in locations where radios are not allowed inside some buildings with electrical sensitive equipment during firefighting, as was the case for this fire drill, it would be difficult for the fire brigade leader to communicate and coordinate with the control room or others during a rescue situation. Regarding the actual rescue activity, its effectiveness could be challenged since a two-person rescue team would be faced with potentially assisting/removing three attack members out of the hot zone. Based on discussions with licensee fire brigade training personnel following the drill, theinspectors learned that this 3-in, 2-out deployment was the current manner in which all internal building firefighting strategies and fire training was based upon.The fire brigade leader allowed the 3-man attack team to enter the fire hot zone with permission to commence firefighting prior to the 2-man rescue team arriving at the fire scenes pre-established incident command post and available for implementing rescue. The inspectors later learned that the rescue team, including the fire brigade leader, had arrived at the incident command post approximately five minutes after the attack team had entered the fire area. This delay involved the fire brigade leader completing his thermal protective clothing dressout in the locker room. The inspectors were concerned that under actual circumstances, if the 2-man rescue team were not ready and prepared to fulfill their rescue responsibilities upon entry of the attack team into the fire hot zone, the effectiveness of the rescue team could be challenged.The inspectors observed that no fire hose or other form of fire suppression was pulled or readily available for the 2-man rescue team to take with them should they have needed to enter the hot zone to rescue the attack team. When questioned about this, the inspectors were told that on the same fire hose that the attack team was using, a 1-1/2 inch gated wye valve had been connected, and the rescue team could have connected another 50-foot, 1-1/2 inch fire hose to it and used that hose as a rescue hose. However, the inspectors determined this was inadequate since to get to this hose connection, the rescue team would have to enter into the hot zone prior to reaching it. In addition, the inspectors learned that the use of this 1-1/2 inch gatedwye valve to create two hose streams for either attack or rescue that essentially splits the available flow capacity through a single 1-1/2 inch hose station nozzlewas allowed in multiple fire pre-plan strategies. At the conclusion of the inspection, the inspectors were continuing to assess whether the use of these gated wye valves had been formally reviewed by the licensee in the past to ensure that the flow capacity of fire hose streams would not be adversely impacted by their use during a fire.Planned Closure Actions: Pending completion of additional evaluations needed to determine whether the above fire brigade issues of concern represented performance deficiencies and if so, whether the performance deficiencies were of more than minor significance, this issue was identified as an unresolved item.Licensee Actions: The licensee initiated an NCR to address the inspectors concerns. In addition, until a more thorough review of their fire brigade program could be performed against their NFPA 805 fire program requirements, an operator standing instruction (#18-009, Fire Brigade 2-Out Response) was developed and implemented. This standing instruction directed the following specific fire brigade required actions:The brigade attack team will consist of two fire members to ensure the fire brigade SIC is not normally utilized as one of the 2-out members. If a runner is needed based on the fire area, the SIC may serve as a 2-out member, but this should be the exception.The 2-out members will establish a ready method of suppression that is accessible outside the fire zone. This should be the identified backup hose in the fire pre-plan. This hose does not need to be charged but should be flaked out and ready for use.The attack team will not enter the fire area, except when search and rescue is necessary, until the 2-out team is in the area with the suppression method ready for use.The inspectors determined that the licensees interim actions were adequate to ensure the fire brigade response would be effective if called upon pending resolution of the issues. Corrective Action Reference: NCR 02194468NRC Tracking Number: URI 05000400/2018001-01 | ||
| 05000400/FIN-2017003-03 | 2017Q3 | Green | Licensee-identified | Licensee-Identified Violation | Section 50.54(q)(2) of 10 CFR requires, in part, that a licensee shall follow and maintain the effectiveness of an emergency plan which meets the planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR Part 50, Appendix E . Section 50.47(b)(4) of 10 CFR requires that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, from April 2010 to May 2017, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, the licensee's emergency classification scheme action levels for Category F Fission Product Barrier EAL , contained declaration threshold values for the containment high range radiation monitor , which were lower than the correct values due to an improper methodology used in calculating the loss of fuel clad barrier and potential loss of containment barrier threshold values and rendered the EALs ineffective. The licensee implemented compensatory actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented until a permanent change can be made to the license. The issue was entered into the licensees CAP as NCR 02123373. The inspectors evaluated this issue as an ineffective EAL per IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process , Figure 5.4 -1. The inspectors concluded that the violation was of very low safety significance (Green). Although the incorrect EAL would alone render an early EAL classification of a General Emergency (GE) based upon the specific radiation monitor, other EALs would provide a GE classification in an accurate and timely manner aligned with the incorrect threshold values of the containment high range radiation monitor . | |
| 05000400/FIN-2017003-01 | 2017Q3 | Severity level IV | NRC identified | Incomplete and Inaccurate Emergency Action Level Submittals | The NRC identified a Severity Level (SL ) IV non- cited violation (NCV) of 10 CFR 50.9, Completeness and accuracy of information, for failure to provide complete and accurate information for prior approval of a new emergency action level (EAL) scheme. The documents submitted to the NRC were, Shearon Harris Nuclear Power Plant, Unit 1 Changes to the Emergency Action Level Scheme, dated April 25, 2010, and License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99- 01, Revision 6, dated April 30, 2015. The submit ted documents specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, which contained declaration EAL threshold values for the containment high range radiation monitor that were lower than the correct values due to use of a n improper calculation methodology. The calculation methodology that was used was not in accordance with the license. It was used to calculate the loss of fuel clad barrier and potential loss of containment threshold values. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017 -017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented. Additionally, the licensee plans to submit a license amendment request to update the EAL scheme. The licensee entered this violation into their corrective action program (CAP) as nuclear condition report (NCR) 02155272. The inspectors evaluated the underlying technical issue and determined that the licensees failure to maintain the effectiveness of its emergency plan was a performance deficiency. The issue was documented as a Green licensee- identified violation (LIV) in Section 4OA7 of this report. The reactor oversight process (ROP) , significance determination process , does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which impeded the NRCs ability to regulate, using traditional enforcement to adequately deter non- compliance. Using the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, this issue was determined to be a SL IV violation. Though the NRC would have questioned the issue with a request for additional information, it would not have resulted in substantial further inquiry. 3 Additionally, the associated technical violation was determined to be of very low safety significance. Traditional enforcement violations are not assessed for cross -cutting aspects | |
| 05000400/FIN-2017003-02 | 2017Q3 | NRC identified | Review of Removal of the Technical Support Center (TSC) Temporary Diesel Generator | The inspectors conducted a detailed review of NCR 02123373, Emergency Action Level Document Calculation Assumptions. The inspectors chose the sample because the EAL issue initially appeared to be potentially more significant than finally determined. The inspectors evaluated the following attributes of the licensees actions: complete and accurate identification of the problem in a timely manner evaluation and disposition of operability and reportability issues consideration of extent of condition, generic implications, common cause, and previous occurrences classification and prioritization of the problem identification of root and contributing causes of the problem 19 identification of any additional condition reports completion of corrective actions in a timely manner 2. The inspectors conducted a detailed review of NCR 00520918, Loss of Offsite Power Impact on Technical Support Center (TSC). The inspectors chose the sample because it was discovered that on July 17, 2017, the licensee had removed a temporary diesel generator that was intended to provide a back -up reliable power source to the TSC until a permanent solution was implemented. The inspectors evaluated the following attributes of the licensees actions: complete and accurate identification of the problem in a timely manner evaluation and disposition of operability and reportability issues consideration of extent of condition, generic implications, common cause, and previous occurrences classification and prioritization of the problem identification of root and contributing causes of the problem identification of any additional condition reports completion of corrective actions in a timely manner b. Findings 1. Incomplete and Inaccurate Emergency Action Level Submittals Introduction: The NRC identified a Severity Level IV NCV of 10 CFR 50.9 , Completeness and accuracy of information, for failure to provide complete and accurate information for prior approval of a new EAL scheme. The documents submitted to the NRC were, Shearon Harris Nuclear Power Plant, Unit 1 Changes to the Emergency Action Level Scheme, dated April 25, 2010, and License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99- 01, Revision 6, dated April 30, 2015. The first submittal to the NRC in 2010 was not complete and accurate in all material respects , and the submittal in 2015 was a missed opportunity to identify the errors made in the first submittal in 2010. Description : On May 10, 2017, Shearon Harris identified the hot operating mode EAL thresholds were calculated incorrectly using a NUREG -0654 methodology vice the required NEI 99- 01 Rev. 6 method, as specified in the current facility licensing basis. When employing the NUREG -0654 methodology to calculate the EAL threshold values, the reactor coolant system (RCS) inventory was assumed to be released at a 50 gallons per minute (gpm) RCS leak rate and activity of 300 micro -Curies per gram (ci/gm) dose equivalent iodine (DEI), over a six -hour period of time. In comparison, when employing the NEI 99- 01 Rev. 6 methodology, the assumption as part of calculating the EAL threshold values was that the entire RCS inventory was released instantaneously at an activity of 300ci/gm DEI. Both of the licensees submittals to the NRC, specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, contained declaration EAL threshold values for the containment high range radiation monitor for loss of fuel clad barrier and potential loss of containment , that were significantly lower than the correct values , due to use of the improper calculation methodology. The submittal dated April 30, 2015, was submitted to provide a complete change to the EAL scheme. This submittal was a missed opportunity by the licensee to identify that the wrong methodology to calculate the EAL threshold values had been used. 20 These submittals were not correct in material content and impacted the NRC s regulatory processes. The NRC evaluated the licensees failure to provide complete and accurate information to determine if there were any unresolved issues. The inspectors concluded that the incomplete and inaccurate information in the license submittal was material to the NRC because, had the NRC staff known the actual methodology used was inaccurate, the staff would have required the licensee to modify the EAL threshold values . The licensee appropriately revised the EAL threshold values utilizing the correct calculation methodology. The licensee issued NC R 02123373, dated May 10, 2017, for EAL thresholds that were calculated without using the correct methodology described in the facility licensing basis. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision - makers of the proper application of the EAL scheme and revised threshold values to be implemented until a permanent change is made to the license. Additionally, the licensee issued NCR 02155272, dated October 3, 2017, for the incomplete and inaccurate EAL submittal, specifically addressing and resolving the completeness and accuracy issues identified by the inspectors. The final significance determination of the underlying technical issue for the licensees failure to maintain the effectiveness of its emergency plan was documented in NRC Inspection Report 05000400/2017003, Section 4OA7, as a Green LIV. Analysis : The inspectors evaluated the underlying technical issue and determined that the licensees failure to maintain the effectiveness of its emergency plan was a performance deficiency. The issue was documented as a Green LIV in Section 4OA7 of this report. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which impeded the NRCs ability to regulate, using traditional enforcement to adequately deter non- compliance. Using the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report , this issue was determined to be a SL IV violation. Though the NRC would have questioned the issue with a request for additional information, it would not have resulted in substantial further inquiry. Additionally, the associated technical violation was determined to be of very low safety significance. Traditional enforcement violations are not assessed for cross -cutting aspects . Enforcement : Section 50.9 of 10 CFR states, in part, that, information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on April 25, 2010, and on April 30, 2015 , information was submitted by the licensee to the NRC that was not complete and accurate in all material respects. Specifically, the submitted documents specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, contained EAL declaration threshold values for the containment high range radiation monitor , that were lower than the actual correct values , due to use of an improper calculation methodology. This was not in accordance with the license. It was used to calculate the loss of fuel clad barrier and potential loss of containment thresholds values. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017 -017 to inform operators and emergency response organization decision -makers of the proper application of the EAL scheme and appropriate threshold values to be implemented. Additionally, the licensee plans to submit a license amendment request to update the 21 EAL scheme. Because this violation was not repetitive or willful, and was entered into the licensees CAP as NC R 02155272, it is being treated as a SL IV NCV, consistent with Section 2.3.2 .a of the NRC Enforcement Policy. ( NCV 05000400/2017003- 01, Incomplete and Inaccurate Emergency Action Level Submittal s) 2. Adequacy of Process for Removal of the TSC Temporary Diesel Generator Introduction: The inspectors opened an Unresolved Item (URI) to complete a review of the licensees removal of a temporary diesel generator on July 17, 2017, that was previously installed to provide reliable backup power to the TSC in the event of a Loss of Offsite Power (LOOP) coincident with a Loss of Coolant Accident (LOCA) event. This temporary diesel generator was originally intended to be installed until a reliable backup power source could be implemented under a permanent modification. Description : The licensee initiated NCR 00520918 on March 1, 2012, to address the consequences of a LOOP/LOCA event on the T SC functionality. Since the TSC is designed with two sources of electrical power, both from offsite power sources, it was recognized that a complete loss of offsite power to the TSC could result in long term TSC operational concerns. Specifically, with t he loss of both offsite power sources, the TSC emergency ventilation system, which provides required radiation protection for event response personnel, would be non- functional, as well as other critical TSC equipment following the loss of short -term (~1 -2 hour s) back -up battery power supplies. The inspectors noted that the operability/functionality section of NCR 00520918 stated that the TSC was functional based on the (current) availability of both of the offsite power sources; however, should a LOOP event occur, then the TSC would be considered non -functional since offsite power would be rendered non -functional. This statement demonstrated the licensees understanding of the vulnerability of continued TSC functionality during a LOOP event. In recognition of this vulnerability, the NCR implemented a short -term solution for procuring and installing a temporary diesel generator in late 2012 under modification EC 85350. The inspectors noted that an emergency preparedness change review evaluation was conducted in accordance with 10 CFR 50.54(q) under action request 00568695. This change request stated that it was necessary to provide the infrastructure for an additional reliable power source for the TSC habitability systems. NCR 00520918 stated that the long- term solution was to provide a permanent backup power supply to the TSC , at which time the temporary diesel generator would be removed. While an action item was initiated to install this TSC permanent backup power source under modification EC 85145, the modification was later revised, removing the intended implementation of a permanent backup power source to the TSC. The inspectors were concerned that the TSC could have equipment and habitability issues during design basis LOOP/LOCA events when the normal TSC offsite power would be non- functional. In addition, the inspectors determined that the TSC temporary diesel generator was removed from the site on July 17, 2017, without implementing the originally intended reliable permanent backup power to the TSC and without conducting a 10 CFR 50.54(q) evaluation specific to its removal to demonstrate that this action did not reduce the effectiveness of implementing the emergency plan. The inspectors requested additional information from the licensee related to the documentation, basis, and process used for the removal of the TSC temporary diesel generator, and evidence that the TSC facility would still be capable of performing all of its intended functions during a LOOP/LOCA event. This issue of concern requires more information to 22 determine if a performance deficiency exists, and if the performance deficiency potentially constitutes a violation of regulatory requirements . Pending review of additional information from the licensee, this issue is identified a s URI 05000400/2017003 -02, Review of Removal of the Technical Support Center ( TSC ) Temporary Diesel Generator. | ||
| 05000400/FIN-2017002-01 | 2017Q2 | NRC identified | Evaluate Fire Protection Discrepancies in RHR/CS Pump Rooms | An unresolved item (URI) was identified by the inspectors during the walkdown of the A and B RHR and CS pump rooms, involving the use of unapproved non-fire retardant plastic sheeting to contain contamination on the A RHR piping. Additionally, the inspectors identified that the fire pre-plan for fire brigade response delineates a hose station that did not contain adequate fire hose length.Description: The inspectors identified two issues of concern during the fire protection walkdown of the A and B RHR and CS pump rooms as follows: 1) Use of Unapproved Plastic for Contamination Control: The inspectors noted that an approximately 30 foot section of the A RHR pump suction piping had been wrapped with multiple layers of plastic sheeting materials that included radiation protection yellow Caution Radioactive Materials stamped plastic sheeting overlaid with clear stretch wrap plastic. The section of RHR piping where this plastic was installedincluded the motor-operated RHR suction valves from the containment recirculation sump (valve 1SI-310) and the refueling water storage tank supply (valve 1SI-322). The inspectors were concerned that these valves could be adversely impacted from a potential fire involving this plastic material. The inspectors questioned whether this plastic was fire retardant material or had been evaluated and allowed under the licensees transient combustible control procedure. The licensee subsequently determined that none of the plastic material was fire retardant or met the requirements of National Fire Protection Association (NFPA) 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, and no previous transient combustible evaluation could be found that allowed the use of the non-fire retardant plastic in the RHR pump room. In addition, radiation protection personnel indicated that there could be other areas where this plastic was used since it was a typical practice to use the material to prevent the spread of contamination from leaking piping connections, valves, and valve packing. The licensee subsequently removed the plastic from the A RHR piping and initiated NCR 02132781 to evaluate this issue of concern. 2) Inadequate Fire Hose Length in Hose Station Described in Fire Pre-Plan: During review of the fire pre-plan procedure (FPP-012-02-RAB190-216) for the A and B RHR/CS pump rooms on the RAB 190 elevation, the inspectors noted that the procedure described two fire hose stations intended to be used during fire brigade response for a fire in either of the pump rooms. These two hose stations were the respective hose stations located just inside the access door to each of the two RHR/CS pump rooms. The procedure states that an extra 100 feet of hose would be needed to account for the additional distance for the hose from the opposite train pump room. However, the inspectors identified that even with the extra 100 feet added to the existing 100 feet that is already in each hose station, there would still not be adequate length for this second hose to reach the opposite train pump room with the fire. The inspectors measured the actual distance between the two locations and estimated the hose would have to be over 300 feet in length in order to be effective in fighting a fire in either of the rooms. A separate hose station on the 216 RAB elevation may provide adequate backup coverage. However, the inspectors were concerned that the issue with the fire pre-plan hose station use could cause confusion or pose an unnecessary delay in fire brigade response for a fire in either of the rooms. The licensee subsequently initiated NCR 02134163 to evaluate this issue of concern.Pending completion of additional evaluations needed to determine whether the above issues of concern represented performance deficiencies and if so, whether the performance deficiencies were of more than minor significance, this issue is identified as URI 05000400/2017002-01, Evaluate Fire Protection Discrepancies in RHR/CS Pump Rooms. | ||
| 05000400/FIN-2017002-02 | 2017Q2 | NRC identified | B ESCW Chiller Failure to Start | The inspectors opened a URI to facilitate the completion of inspection and determination of whether a performance deficiency was associated with the start failure of the B ESCW chiller on May 13, 2017. Description: On May 13, 2017, while attempting to start the B ESCW chiller, the motor compressor immediately tripped on C phase instantaneous overcurrent relay actuation. The chiller was declared inoperable and immediate troubleshooting was conducted to determine the cause of the trip. The licensees initial investigation did not identify any electrical or mechanical issues with the compressor motor, supply breaker and electrical bus, or other chiller control components. While the calibration of the C phase instantaneous overcurrent relay was checked and found to be within specification, the licensee determined the most probable cause of the trip was an intermittent failure of this relay. The relay was replaced and subsequent post-maintenance testing of the chiller was successfully performed without any other chiller operational problems being identified. The chiller was returned to operability early May 14, 2017, following the completion of this post-maintenance testing. At the end of the inspection period, the licensees investigation into the cause of the start failure had just completed. A URI is being opened for the NRC to review the licensees failure analysis and causal evaluation to determine whether the chiller start failure was reasonably within the licensees ability to predict or prevent and therefore a performance deficiency. This issue is being tracked as URI 05000400/2017002-02, B ESCW Chiller Failure to Start | ||
| 05000400/FIN-2017001-01 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as a NCV. Appendix B to 10 CFR Part 50, Criterion V, Instructions, Procedures, and Drawings, required in part, 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. Contrary to the above, on March 22, 2017, the licensee identified that an environmental protection feature, 1EE-E668 (FHB 261 floor hatch), was removed from service withou stationing a dedicated attendant as required by licensee procedure AP-046, Control of Environmental Protective Features. The floor plug was removed for two days (March 20-22, 2017) to support maintenance activities before the condition was identified by licensee operations personnel while performing rounds. Using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, inspectors determined that this violation was of very low safety significance (Green) because the finding did not impact the frequency of a fire or internal flooding initiating event and all structures, systems, and components remained capable of performing there intended safety functions. This issue was documented in the licensees CAP as AR 20110596. | |
| 05000400/FIN-2016004-01 | 2016Q4 | Green | Licensee-identified | Licensee-Identified Violation | TS limiting condition for operation (LCO) 3.3.3.6, Action C, Accident Monitoring Instrumentation, states in part that with the number of operable accident monitoring instrumentation channels for the radiation monitor(s), less than the minimum channels operable, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours, and either restore the inoperable channel(s) to operable status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 14 days. TS Table 3.3-10 indicates that a minimum of one channel of the Containment High Range Radiation Monitors (CHRRMs) is required to be operable. Contrary to the above, the licensee identified that they failed to identify the inoperability of the CHRRMs and take the required actions of LCO 3.3.3.6, Action C, from 1998 until an operability determination was completed in September 2016. Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, inspectors determined that this violation was of very low safety significance (Green) because the finding was a failure to comply with a non-risk significant planning standard and no planning standard function failure occurred since other parameters could be used to validate the indications from the CHRRMs. This issue was documented in the licensees CAP as AR 2063783. | |
| 05000400/FIN-2016003-01 | 2016Q3 | NRC identified | Subsequent Loss of Safety-Related Chilled Water System Results in a Loss of Safety Function | The inspectors opened a URI to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, associated with the A ESCW chiller failures. On July 15, 2016, the A ESCW chiller tripped on low oil pressure. Licensee investigation identified that oil was leaking from the threaded portion of a brass fitting located between a pressure switch and needle valve associated with PDS-01CY-9428ASA-HI. Upon removal, it was observed that significant radial cracking occurred in the threaded portion of the brass fitting. A like-for-like replacement was installed and the A ESCW chiller was returned to service. One week later, on July 22, 2016, the A ESCW chiller tripped again on low oil pressure. The investigation revealed that the same brass fitting had failed and the A ESCW chiller could not meet its mission time of 30 days of continuous operation in the event of a loss of cooling accident. During this 7-day period, the B ESCW chiller was inoperable for a period of time, which means the ESCW system would not have been able to meet its safety function. The licensees investigation into the cause of the subsequent failure is ongoing. A URI is being opened to determine whether the subsequent failure of the brass fitting was reasonably within the licensees ability to predict and therefore a performance deficiency. This issue is being tracked as URI 05000400/2016003-01, Subsequent Loss of Safety-Related Chilled Water System Results in a Loss of Safety Function. | ||
| 05000400/FIN-2016002-02 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | Section 50.48(c) of 10 CFR and NFPA 805, 2001 Edition, Section 2.4.2.2.2(b), Common Enclosure Circuits, require that those circuits which share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required component, shall be identified to prevent propagating fires outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables. Contrary to the above, from October 1986 to September 2014, the licensee failed to meet the requirements of 10 CFR 50.48(c) and NFPA 805, Section 2.4.2.2.2(b), in that, the licensee failed to identify and provide adequate electrical fault protection for the turbine emergency oil pump control cables 11376C and 11376D. The cables could have created a common enclosure fire hazard under postulated situations which could have resulted in a secondary fire in other fire areas and could have adversely affected the capability to achieve safe and stable plant conditions. A fire-induced failure could have caused the loss of the required safe shutdown components. This violation was determined to be of very low safety significance (Green) based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. A detailed risk evaluation was performed in accordance with NRC IMC 0609 Appendix F, and NUREG/CR6850 Rev. 0 and 1, using inputs from the licensees NFPA 805 Fire PRA. The major analysis assumptions included a one-year exposure interval, and secondary fires occurring between the power supply and the fire induced hot short. The dominant sequence was a fire in the main control board causing a secondary fire in the B cable spreading room which if unsuppressed could result in the inability to achieve safe shutdown resulting in core damage. The quantitative screening approach resulted in a calculated delta core damage frequency of less than 1E-06, which screened this violation to Green (very low safety significance). This violation was documented in the licensees corrective action program as Condition Report 692766. | |
| 05000400/FIN-2016002-01 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | Section 50.48 of 10 CFR, Fire Protection, states that a fire protection program that is maintained to the requirements of National Fire Protection Association (NFPA) standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, is an acceptable method for complying with the requirements of Section 50.48. Section 3.8.1 of NFPA 805 states, in part, that alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble alarms to the control room or other constantly attended location from which required notifications and response can be initiated. Contrary to the above, from December 2015, to May 17, 2016, neither the licensees design reviews nor post-modification tests identified that the fire protection system installed on the 286-ft elevation of the turbine building did not transmit trouble alarms to the Harris main control room. Following installation and testing, the newly-installed Protecta WireTM system and fire detection panel, 1-SFD-E144, were placed in service in late December 2015. On May 17, 2016, while performing maintenance periodic test, MPT-I0052, Turbine Building Local Fire Detection Control Panel LFDCP-10 Test and 1-SFD-E144 Test of the fire detection system, the technicians performing the test recognized that the remote trouble alarm function would not cause an alarm in the control room. The licensee entered the issue concerning the inadequate remote alarm function into the corrective action program via AR 2030427 and implemented actions to incorporate and test the remote trouble alarm function into the EC package. The licensee also initiated corrective actions via AR 2033716 and AR 2038682 to address issues in the design review process. Using IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green) since the reactor would still be able to achieve and maintain safe shutdown. | |
| 05000400/FIN-2016001-01 | 2016Q1 | Green | NRC identified | NRC Biennial Written Examinations Did Not Meet Qualitative Standards | The inspectors identified a finding of very low safety significance associated with 10 CFR 55.59, Requalification, based on a determination that greater than 20 percent of the written examination questions administered to licensed operators during the 2014 biennial written examination were flawed. The licensee entered this issue into their Corrective Action Program (CAP) as Action Request (AR) 01940942, Inspection Procedure (IP) 71111.11B NRC Biennial LOCT Inspection Feedback, dated August 6, 2015. The inspectors determined that the finding was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding adversely affected the quality and level of difficulty of biennial written examinations, which potentially impacted the facilitys ability to appropriately evaluate licensed operators. The risk importance of this issue was evaluated using IMC 0609, Appendix l, Licensed Operator Requalification Significance Determination Process (SDP). The qualitative standards used by the inspectors were defined in TPP-306, Licensed Operator Continuing Training Program, and TRN-NGGC-0441, Licensed Operator Requalification Annual/Biennial Exam Development, and further described within NUREG-1021, Revision 9, ES-602, Attachment 1, Guidelines for Developing Open-Reference Examinations, and Appendix B, Written Examination Guidelines. Because more than 20 percent, but less than 40 percent, of the questions reviewed were flawed, Blocks 4 and 5 of Appendix I characterized the finding as having very low safety significance (Green). A review of the cross-cutting aspects was performed and no associated cross-cutting aspect was identified. | |
| 05000400/FIN-2015008-01 | 2015Q4 | Severity level IV | NRC identified | Untimely 10 CFR 50.73 Notification of an Inoperable CIV | An NRC-identified Severity Level IV violation of 10 CFR 50.73 was identified for the licensees failure to provide a written report to the NRC within 60 days after discovery of a condition prohibited by Technical Specification (TS) Limited Condition for Operation (LCO) 3.6.3, "Containment Isolation Valves."The issue was placed in the licensees corrective action program as CR 01958628.The inspectors determined that the failure to provide a written report to the NRC within the time limits specified in regulations was a violation 10 CFR 50.73. The violation was evaluated using Section 6.9 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9.d.9of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV, non-cited violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue involving untimely reports to the NRC was strictly associated with a traditional enforcement violation. | |
| 05000400/FIN-2015004-01 | 2015Q4 | Green | H.5 | Self-revealing | B ESW System Safety-Related Cables Submerged in Water | A self-revealing Green non-cited violation (NCV) of Criterion III, Design Control, Appendix B of 10 CFR Part 50, occurred due to the licensees failure to maintain Class 1E (safety-related) electrical cables in an environment for which they are designed. Specifically, the low-voltage safety-related cables associated with the B Essential Service Water (ESW) system were submerged in water, a condition for which they are not qualified. The licensee took immediate actions to lower the water levels in underground cable vaults where submerged cables were discovered, and conduct pump-downs of the safety-related underground cable vaults on an increased scheduled frequency. The licensee entered this issue into the corrective action program (CAP) as nuclear condition reports (NCRs) 1961933 and1962664, respectively, and implemented actions to pump down the non-conforming vaults. The licensees failure to maintain the low-voltage safety-related cables associated with the B ESW system in an environment for which they were designed was a performance deficiency. The performance deficiency was more than minor because the cables known to be submerged are part of the B ESW system, which is a mitigating system and is associated with the Mitigating Systems Cornerstone. The performance deficiency was related to the equipment reliability attribute and failure to maintain the cables in the environment for which they were designed adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the submergence of the safety-related cables adversely impacted the service life of the cables and could cause the B ESW system to be inoperable in the event a cable failed as a result of continuous submergence. The inspectors used Table 2 of Attachment 4, Initial Characterization of Findings, of Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), to determine that the finding was associated with the Mitigating Systems Cornerstone. Using the guidance provided in Table 3 of Attachment 4, the inspectors transitioned to Appendix A, SDP for Findings At-Power, of IMC 0609. The inspectors used Exhibit 2, Mitigating Systems Screening Questions, of the appendix, to determine that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function on the ESW system. The finding was assigned the cross-cutting aspect of work management, as described in the area of human performance, because the licensee failed to implement an adequate preventive maintenance program to monitor and maintain the sump pumping system associated with the safety-related cable vaults. |
| 05000400/FIN-2015007-01 | 2015Q4 | Green | NRC identified | Failure to Establish As-found Testing on 6.9kV Vacuum Breakers | The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish a periodic as-found testing program of safety-related 6.9kV vacuum breakers in accordance with applicable design document IEEE 308-1971. The licensee entered this issue into their corrective action program as action request 01983086 and initiated a procedure change request to have the procedure changed to verify the as-found capability of the breakers before performing the first scheduled preventative maintenance on the breakers in April 2016. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to establish as-found testing could mask degradation of the circuit breakers and decrease the reliability of the breakers to perform their safety-related function when called upon. The finding was determined to be of very low safety significance (Green), because it was a deficiency affecting the design or qualification of a structure, system, or component, which maintained its operability. The team determined that no finding cross-cutting aspect was applicable because the finding did not reflect current licensee performance. | |
| 05000400/FIN-2015008-02 | 2015Q4 | Green | P.3 | NRC identified | Failure to Follow EPM-410 Procedure | An NRC-identified Green NCV of 10 CFR 50.54(q)(2) was identified, for the licensees failure to follow and maintain, in effect, the Emergency Plan when performing monthly testing of the Technical Support Center (TSC). Specifically, the licensee failed to follow procedural steps when recorded values did not meet acceptance criteria as specified in EPM-410, Communication and Facility Performance Tests. The issue was placed in the licensees corrective action program as CRs 01942073, 01940053. The finding was more than minor because it was associated with the Emergency Response Organization (ERO) Performance attribute and 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 failure to follow procedural steps when recorded values did not meet acceptance criteria resulted in a failure to comply with emergency plan. The finding was assessed for significance in accordance with NRC Manual Chapter 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; results in a Green finding. The inspectors identified a cross-cutting aspect in the Problem Identification and Resolution area because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3). |
| 05000400/FIN-2015004-02 | 2015Q4 | Green | H.10 | Self-revealing | Failure to Adequately Implement Post-Maintenance Testing | A self-revealing Green NCV of Criterion XI, Test Control, Appendix B of 10 CFR Part 50, occurred due to the licensees failure to perform adequate post-maintenance testing on the essential services chilled water (ESCW) chillers. Specifically, on multiple occasions the licensee failed to perform section 7.8, Current Signal Input Resistor Adjustment, of procedure CM-I0014, York Essential Services Chilled Water Chiller Temperature Control Maintenance, following maintenance on the temperature controller associated with the ESCW chillers. The licensee took corrective actions to adjust the current limiters on the ESCW chillers to be at the correct setting of 65 amps. The licensee entered this issue into the CAP as NCRs 1944657 and 1950574. The licensees failure to perform adequate post-maintenance testing on the, ESCW Chiller was a performance deficiency. Specifically, the failure to perform section 7.8 of procedure CM-I0014 to ensure the configuration of the temperature controllers were per design specifications resulted in the current limiters on each unit being out of calibration for extended periods of time. The performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. For the A train, continued omission of section 7.8 following maintenance on its temperature controller, had the potential of the pre-rotation vanes pulling excessive load resulting in a trip of the A train chiller. For the B train, continued omission of section 7.8 following maintenance on its temperature controller had the potential of the pre-rotation vanes not picking up adequate load to meet the cooling demands during accident conditions. The inspectors used Table 2 of Attachment 4, Initial Characterization of Findings, of IMC 0609, Significance Determination Process, to determine that the finding was associated with the Mitigating Systems Cornerstone. Using the guidance provided in Table 3 of Attachment 4, the inspectors transitioned to Appendix A, Significance Determination Process (SDP) for Findings At-Power, of IMC 0609. Using Exhibit 2, Mitigating Systems Screening Questions, of the appendix, the inspectors determined that the finding was of very low safety significance (Green) because while the chillers were in a nonconforming condition, operability of the chillers was maintained. The finding was assigned the cross-cutting aspect of bases for decisions, as described in the area of human performance, because the licensee made the decision not to perform section 7.8 of maintenance procedure CM-I0014 without documenting a reason for omitting the section following maintenance on the ESCW temperature controllers (H.10). |
| 05000400/FIN-2015003-06 | 2015Q3 | Severity level IV | NRC identified | Failure to Report the Loss of Emergency Assessment Capability | The inspectors identified a severity level (SL) IV NCV of 10 CFR 50.72(b)(3)(xiii) for the failure to report to the NRC within 8 hours the major loss of emergency assessment capability of the Technical Support Center (TSC). Specifically, on multiple occasions between January 2015 and July 2015, there were unplanned losses of emergency response facility (ERF) function, which resulted in the loss of emergency assessment capability, which the licensee failed to report the condition within the 8-hour time requirement. Subsequently, the licensee notified the NRC once it was realized a report was required and entered the issue in the CAP as AR 757885. The failure to report the loss of emergency assessment capability in the TSC as required by 10 CFR Part 50.72(b)(3)(xiii) was a performance deficiency. The licensees failure to notify the NRC was determined to impact the regulatory process, which requires evaluation through the traditional enforcement process. Based on the examples provided in Section 6.9 of the Enforcement Policy, dated February 4, 2015, Inaccurate and Incomplete Information or Failure to Make a Required Report, the performance deficiency was determined to be a SL IV violation. Specifically, example d.9 states that a SL IV violation involves a failure to make a report required by 10 CFR 50.72 or 10 CFR 50.73. Because the violation was processed as a traditional enforcement violation, no cross-cutting aspect is assigned. | |
| 05000400/FIN-2015003-04 | 2015Q3 | Green | H.6 | Self-revealing | Loss of A ESW Train | A self-revealing green NCV of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Criterion III, Design Control, was identified for failure to implement design control measures that verify adequacy of design. Specifically, EC 83681 involved the installation of a new pump bearing with different wear characteristics but the EC failed to evaluate the impact of the bearing replacement on alignment sensitivity of the pump shaft. The licensee took immediate action to align the Normal Service Water system to provide cooling to the heat loads affected by the loss of the A ESW pump. Failure to incorporate alignment requirements for the pump shaft in the work instructions associated with EC 83681 was a performance deficiency. The performance deficiency was related to the equipment performance attribute of the initiating events cornerstone. The performance deficiency was determined to be more than minor because the performance deficiency 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, the failure of the ESW pump shaft resulted in a loss of service water which ultimately led to the loss of the A train of shutdown cooling for a period of twelve minutes. Inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 4 and Appendix G (June 19, 2012), Shutdown Operations Significance Determination Process. The inspectors determined the finding was associated with the Initiating Event cornerstone and required a detailed risk evaluation because the finding involved a loss of safety function. A detailed risk evaluation was completed by a regional Senior Reactor Analyst (SRA). The regional SRA performed a detailed risk review of the finding. The SRA performed the analysis by increasing the maintenance unavailability for the pump, and evaluating it versus the base case. This method was chosen because the pump was in standby service, and the dominant method of determining there was a failure would have been during testing, or operation under non accident conditions. The additional time for the unnecessary repair was used to adjust the base case maintenance unavailability. Online and shutdown risk were evaluated. The total impact was determined to be low enough for the finding to be GREEN for SDP purposes The finding had a cross-cutting aspect in the Human Performance area of Design Margin (H.6). |
| 05000400/FIN-2015003-03 | 2015Q3 | Green | H.6 | Self-revealing | Failure to Implement Adequate Corrective Actions | A self-revealing green finding was identified for failure to implement adequate corrective actions for the repeated failure of PS-4175, low pressure steam inlet crossover pressure switch in accordance with licensee procedure AD-PI-ALL-0100, Corrective Action Program. Specifically, on multiple occasions the licensee failed to install a pressure switch rated for design conditions on the Main Turbine which led to an unplanned reactivity addition, when PS-4175 failed open. The licensee entered this into their corrective action program (CAP) as action request (AR) 755621 and took immediate actions to reduce power to less than 100 percent. Reactor power reached a maximum value of 100.5 percent. Failure to implement adequate corrective action for the repeated failure of pressure switch PS- 4175 in accordance with licensee procedure AD-PI-ALL-0100 was a performance deficiency. The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, if not for the manual actions taken by the operators to insert control rods, the reactivity addition would have continued and would have ultimately resulted in a reactor trip on high neutron flux. Using IMC 0609, Significance Determination Process Attachment 4, Initial Characterization of Findings, and Appendix A, The SDP for Findings At-Power, (June 19, 2012), the inspectors determined the finding was a contributor as a Transient Initiator to the Initiating Events cornerstone. The inspectors determined the finding was of very low safety significance (Green) because it did not result in a reactor trip and it did not cause the loss of any mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors concluded the finding was associated with the design margins aspect (H.6) of the human performance cross-cutting area since the licensee repeatedly failed to install a pressure switch adequate for the operating conditions. |
| 05000400/FIN-2015003-05 | 2015Q3 | Green | P.3 | NRC identified | Failure to implement EQ Program Requirements | The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality affecting the Environmental Qualification (EQ) Program. Specifically, the licensee failed to enter into the CAP the results of the vendor audit of the EQ program which resulted in the licensee blocking open D10 and D11 on June 16, 2015 while the unit was at 100 percent power. The resident inspectors questioned the main control room (MCR) about the doors being open and the licensee immediately closed D10 and D11. The licensee has entered the violation into their CAP as AR 754721, implemented interim guidance as an operations standing instruction (2015-024) not to open D10 or D11 while in mode 1-4. The opening of the tornado door between the main steam tunnel (MST) and the reactor auxiliary building (RAB) was a performance deficiency. The finding was screened in accordance with NRC IMC 0609.04, Initial Characterization of Findings, dated July 7, 2012. The finding was determined to affect the Initiating Events Cornerstone as the MST to RAB tornado door represented a barrier which left RAB systems and components vulnerable to harsh environment conditions should a high energy line break (HELB) occur during the time the doors were open. SDP screening determined that the finding could have affected equipment used to mitigate a LOCA, could have caused a reactor trip, could have resulted in internal flooding conditions, and could have affected equipment relied upon to transition the plant to a stable shutdown condition and required a detailed risk evaluation. A detailed risk evaluation was performed by a regional SRA in accordance with NRC IMC 0609 Appendix A. The major analysis assumptions included: a twenty hour exposure interval, HELBs postulated in all steam and feedwater piping in the MST, pipe break frequency from EPRI Report 1021086, no recovery credit for door closure, and a bounding CCDP value utilized. The CCDP was estimated using the NRC Shearon Harris SPAR model assuming a reactor trip initiator and bounding assumptions that the postulated RAB harsh environmental and flooding conditions would cause failure of the following equipment: auxiliary feedwater system, alternate seal injection system, RAB essential services chillers, component cooling water pumps, charging and safety injection pumps, and the residual heat removal pumps. The dominant sequence was a reactor trip, success of the reactor protection system, and failure of the reactor coolant pump (RCP) seals leading to an unmitigated RCP seal LOCA. The risk was mitigated by the short exposure period and the probability of steam and feedwater HELBs. The analysis determined that the finding represented an increase in core damage frequency of < 1.0 E-6/year, a GREEN finding of very low safety significance. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution in the Corrective Action component because the licensee did not take appropriate corrective actions to address safety issues in a timely manner. |
| 05000400/FIN-2015003-02 | 2015Q3 | NRC identified | Written NRC Biennial Examinations Did Not Meet Qualitative Standards | The inspectors identified an URI associated with 10 CFR 55.59, "Requalification," based on a preliminary determination that between 20 and 40 percent of the written examination questions administered to licensed operators during the biennial requalification examination were potentially flawed. This item is unresolved pending further inspection to determine whether a performance deficiency exists. The NRC-required biennial written examinations are designed to ensure that licensed operators maintain safe standards of knowledge and ability in order to take appropriate safety-related actions in response to actual abnormal or emergency conditions. As part of the biennial licensed operator training inspection, the inspectors evaluated the content of two NRC-required biennial written examinations (Set 1 Exam 1 SRO and Set 2 Exam 4 SRO) that the licensee developed and administered to licensed operators in 2014. Between 20 percent and 40 percent of items reviewed were determined to potentially contain flaws such as more than one implausible distracter, direct lookup, or low level of difficulty. The standard for determining a question flaw was located within site-specific procedures and further defined within NUREG-1021, Operator Licensing Examination Standards for Power Reactors. Several questions were determined to potentially not contain an appropriate level of difficulty. In many instances there may not have been an acceptable amount of knowledge tested for where to find the answer in an open book exam. Compounding the impact on level of difficulty, answer choices could be eliminated without using nuclear power plant operating knowledge. The licensee entered the issue into their CAP as AR 01940942. Pending further guidance from the Office of Nuclear Reactor Regulation (NRR) on the evaluation of the level of difficulty in establishing whether a performance deficiency exists, this issue is identified as URI 05000400/2015003-02, NRC Biennial Written Examinations May Not Have Met Qualitative Standards. | ||
| 05000400/FIN-2015003-01 | 2015Q3 | Green | H.8 | Self-revealing | Failure to Adequately Implement the Clearance and Tagging Procedure | A self-revealing Green NCV of Technical Specification (TS) 6.8.1, Procedures and Programs, for the licensees inadequate implementation of procedure AD-OP-ALL-0200, Clearance and Tagging, when the licensee failed to establish an appropriate clearance boundary to support filling the cooling tower basin. This resulted in excess of 45,000 gallons of water being spilled in the RAB. The licensee initiated corrective actions to address potential equipment degradation and personnel hazards as a result of the spill. The licensees failure to adequately implement procedure, AD-OP-ALL-0200, Clearance and Tagging, Section 5.5, Step 1 was a performance deficiency. Specifically, CO 310942 did not establish isolation between the cooling tower basin and ISW-276, which was not completely assembled. The performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, if operator action had not stopped the leakage, it potentially would have challenged the operability of safety related equipment on multiple levels of the RAB. Using MC 0609, SDP, Appendix G, Exhibit 3 Mitigating Systems Screening Questions, the finding is a deficiency affecting the qualification of a mitigating SSC, however, the SSC function was restored with operator action, resulting in a GREEN finding of very low safety significance. The finding had a crosscutting aspect of procedure adherence, as described in the area of human performance because the licensee allowed the CO to be lifted in the plant without properly establishing an isolation boundary and isolating the cooling tower basin from ISW-276 while not fully assembled. (H.8). |
| 05000400/FIN-2015404-02 | 2015Q2 | Green | NRC identified | Security | ||
| 05000400/FIN-2015404-03 | 2015Q2 | Green | NRC identified | Security | ||
| 05000400/FIN-2015404-04 | 2015Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000400/FIN-2015404-01 | 2015Q2 | Green | NRC identified | Security | ||
| 05000400/FIN-2015002-03 | 2015Q2 | Green | H.5 | NRC identified | Inadequate Post Modification Testing of 1CZ-1 and 1CZ-2 HMCP Breakers | An NRC-identified Green NCV of TS 6.8.1, Procedures and Programs, was identified for the licensees failure to perform adequate post modification tests (PMTs) on Motor Circuit Protectors (HMCP) breakers for dampers 1CZ-1 and 1CZ-2 as required by procedure AD EG-ALL-1155, Plant Modification Testing. The licensee entered this issue into the CAP as AR 741781. The licensee took immediate corrective action to manually close 1CZ-1 and 1CZ-2 to isolate the MCR boundary. The licensee also changed the setpoint, and revised the PMT to include the direction reversal. The licensees failure to perform adequate PMTs on HMCP breakers for 1CZ-1 and 1CZ-2 as required by procedure AD-EG-ALL-1155 was a performance deficiency. The performance deficiency was more than minor because it was associated with the Procedure Quality Attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to test the highest instantaneous current the HMCP breakers would be expected to experience. This would be during the dampers direction reversal, during the PMT. Therefore, the HMCP breakers for 1CZ-1 and 1CZ-2 had the potential to trip open during a control room isolation signal (CRIS), causing unfiltered inleakage into the MCR envelope in the event of a radiological emergency. Using IMC 0609.04, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, issued on June 19, 2012; the inspectors concluded that a detailed risk evaluation was required since the finding represented a degradation of the barrier function of the control room against smoke and the radiological barrier function provided for the control room. This conclusion was based upon the potential of the HMCP breakers tripping due to a high instantaneous current, during an event that would cause a CRIS such as a high radiation signal at the normal intake or emergency intakes, or smoke detection at the normal intake. A detailed risk evaluation was performed in accordance with the guidance of NRC IMC 0609 Appendix A. A bounding analysis was performed considering potential demands on MCR habitability due to radiation and smoke effects. The major analysis assumptions included: a 94-day exposure period, recovery credit for manual closure of either 1CZ-1 or 1CZ-2, at power core damage probability radiation impact determined from the NRC SPAR model, fuel pool radiation impact from NUREG-1738, and fire risk from IMC 0609 Appendix F. The dominant sequence was a fire impacting the MCR with smoke, failure of operators to isolate the MCR dampers resulting in loss of the operators leading to loss of core heat removal. The risk of the performance deficiency was mitigated by the low initiating event probabilities and the recovery likelihood of MCR damper closure. The result of the analysis was an increase in core damage frequency of < 1.0E-6/year a GREEN finding of very low safety significance. The finding was assigned to the cross-cutting aspect of Work Management in the Human Performance cross-cutting area because the licensees work management processes failed to develop and implement a PMT that adequately tested the breakers to their designed performance. |
| 05000400/FIN-2015002-02 | 2015Q2 | Green | H.8 | NRC identified | Failure to Adequately Implement the Control Room Area HVAC System Procedure | An NRC-identified Green non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for the licensees inadequate implementation of procedure OP-173, Control Room Area HVAC System. Specifically, the licensee failed to adequately implement OP-173 Section 8.3, Placing the Control Room Area HVAC System into Recirculation Manually, and maintain a positive pressure in the main control room (MCR). The licensee entered this issue into the corrective action program (CAP) as action request (AR) 742947, and restored a positive pressure in the MCR. The licensee also revised the associated procedure OWP-RM-01, Control Room OAI (outside air intake) Radiation Monitors, to ensure appropriate actions are taken for the outside air intake supply when radiation monitors are inoperable. The failure to maintain positive pressure in the MCR in accordance with OP-173 was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Appendix B, since it was associated with the procedure quality attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective and, if left uncorrected, the performance deficiency would have the potential for leading to a more significant safety concern. Specifically, the buildup of carbon dioxide (CO2) would impair operators performance and actions. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4 and Appendix G (June 19, 2012), Shutdown Operations Significance Determination Process. The inspectors determined the finding was associated with the barrier integrity cornerstone and required a detailed risk evaluation because the finding involved control room habitability during both normal and accident conditions. A detailed risk evaluation was completed by a regional SRA using the guidance of NRC IMC 0609 Appendix G and Appendix F, Fire Protection Significance Determination Process. A bounding analysis was performed considering potential demands on MCR habitability due to radiation and smoke effects. The major analysis assumptions included: an eleven day exposure period, recovery credit for MCR door closure, shutdown core damage radiation and fuel pool radiation events were considered. The dominant sequence was a fire impacting the MCR with smoke, failure of operators to isolate the MCR resulting in loss of the operators leading to loss of core heat removal. The risk of the performance deficiency was mitigated by the low initiating event probabilities and the recovery likelihood of MCR door closure. The result of the analysis was an increase in core damage frequency of < 1.0E-6/year, a green finding of very low safety significance. The finding had a cross-cutting aspect of Procedure Adherence, as described in the Human Performance cross-cutting area because the licensee failed to comply with OP-173. |
| 05000400/FIN-2015404-05 | 2015Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000400/FIN-2015002-01 | 2015Q2 | Green | P.2 | NRC identified | Failure to Maintain Emergency Assessment Capability | An NRC-identified Green NCV of 10 CFR 50.54(q)(2) was identified for the licensees failure to maintain adequate equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition as required by 10 CFR 50.47(b)(9). Specifically, the data logger for the onsite primary meteorological tower (MET) periodically provided inaccurate meteorological information to the Emergency Response Facility Information System (ERFIS) displays in the MCR and the Emergency Operations Facility (EOF). The inspectors determined that the failure to maintain emergency assessment capability was a performance deficiency. 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, between March 30 and April 28, 2015, the data logger unit on the onsite primary meteorological tower used for dose assessment and dose projections, malfunctioned at least five times. On these occasions, the 15-minute average MET data read by ERFIS was locked and did not update. During these periods, the dose projection process was challenged to provide adequate and timely estimates of radioactive releases, onsite and offsite dose assessment, as well as projected offsite doses. Equipment or systems necessary for dose projection were not functional for longer than 24 hours from the time of discovery and no compensatory measures were implemented until after the inspectors questioned the licensee. The finding was assessed for significance in accordance with NRC IMC 0609, Appendix B, Emergency Preparedness SDP, Attachment 2, and determined to be a very low safety significance finding (Green). The finding has a cross-cutting aspect of evaluation, as described in the area of problem identification and resolution, because the organization did not thoroughly evaluate or address the causes and extent of conditions commensurate with the safety significance of not having accurate MET data for radioactive material releases to the environment or projected offsite doses. |
| 05000400/FIN-2015001-01 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | TS 6.8, Procedures and Programs, Section 6.8.1.a requires, in part, that written procedures be established, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide (RG) 1.33, Revision 2, February 1978. RG 1.33, Appendix A, Section 8.b.(1).(dd) requires, in part, that procedures be established for safety valve surveillance tests. Contrary to the above, on March 12, 2015, engineers used an inadequate procedure to test main steam safety valves. Specifically, engineering procedure EST-224, Insitu Main Steam Safety Valve Test using Assist Device, did not adequately direct personnel to compensate for the head differential between the pressure gauge and the seat of the main steam safety valves. This resulted in the licensee incorrectly adjusting the setpoint of MSSV MS-46 below TS 3.7.1 limits while operating in Mode 1. The licensee identified this issue after incorrectly declaring the valve operable. However, the licensee was able to restore the setpoint and operability of MS-46 within the TS 3.7.1 Limiting Condition for Operation action time. This violation was determined to be of very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The licensee entered this issue into their CAP as AR #737961. As corrective actions, the licensee revised the procedure and restored the setpoint to within TS 3.7.1 limits and retested MS-46. | |
| 05000400/FIN-2014005-01 | 2014Q4 | Green | H.11 | NRC identified | Failure to Adequately Implement the Equipment Clearance Procedure | The NRC identified a Green non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, for the licensees inadequate implementation of procedure OPS-NGGC-1301, Equipment Clearance, when they failed to identify required compensatory measures for a clearance to support installation of a plant modification. This resulted in an unanalyzed condition with no compensatory measures for internal flooding. The licensee entered this into the corrective action program (CAP) as Action Request (AR) #696331 and AR #726784 and took immediate corrective actions to restore the sump pumps to their design configuration. The licensees failure to adequately implement Procedure, OPS-NGGC-1301, Equipment Clearance, Section 9.8, step 3 was a performance deficiency. Specifically, if an internal flood had occurred in the Diesel Fuel Oil Storage Tank (DFOST) building during this period, it could have resulted in both trains of the safety-related fuel oil transfer pumps being inoperable. The performance deficiency was more than minor because it is associated with the Human Performance Attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using Manual Chapter 0609, Significance Determination Process, Appendix A, Exhibit 2 Mitigating Systems Screening Question, Section B, and Exhibit 4, the finding was determined to require a detailed risk evaluation because the loss of this equipment during an internal flooding initiating event would degrade two or more trains of a multi-train system that supports a risk significant system or function. A detailed risk evaluation was performed by a regional senior risk analyst in accordance with the guidance of NRC IMC 0609 Appendix A, using the Shearon Harris Standardized Plant Analysis Risk (SPAR) model. The major analysis assumptions included: A 28-hour exposure period, the finding was modelled as a non-recoverable common cause failure to run of the Emergency Diesel Generators (EDG), pipe failures of fire protection piping was assumed to result in EDG inoperability and pipe failure data was taken from Electric Power Research Institute (EPRI) Pipe Failure Frequencies for Internal Flooding PRAs, Revision 1. The dominant sequence was a station blackout with auxiliary feedwater system failure and no recovery of the EDGs or offsite power leading to loss of core heat removal and core damage. The risk was mitigated by the short exposure period and the low probability of pipe ruptures resulting in EDG inoperability. The analysis determined that the finding led to an increase of core damage frequency of <1E-6/year, a Green finding of very low safety significance. The finding had a crosscutting aspect of Challenge the Unknown, as described in the area of Human Performance because the licensee allowed the clearance order (CO) to be hung in the plant without properly evaluating and managing the associated risk through the use of compensatory measures (H.11). |
| 05000400/FIN-2014403-01 | 2014Q4 | Green | H.7 | NRC identified | Security | |
| 05000400/FIN-2014403-02 | 2014Q4 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000400/FIN-2014004-02 | 2014Q3 | Green | NRC identified | Loss of Emergency Planning Sirens | The NRC identified a Green NCV associated with emergency preparedness planning standard 10 CFR 50.47(b)(5), which requires in part, that the means to provide alert and notification and clear instruction to the populace within the plume exposure pathway Emergency Planning Zone (EPZ) have been established. Specifically, on April 3, 2014, the licensee unintentionally initiated a complete loss of sirens while responding to a siren system alarm. The licensee entered this issue into the corrective action program (CAP) as Action Request (AR) #679984. As corrective action, the licensee replaced a failed circuit card and restored functionality of the siren system. The licensees failure to comply with WCP-NGGC-0300, Work Request Initiation, Screening, Prioritization and Classification, was a performance deficiency. Specifically, this failure combined with the circuit card failure caused a complete loss of siren functionality for approximately two hours. This finding was more than minor because if left uncorrected, loss of Alert Notification System function has the potential to lead to a more significant safety concern and is associated with the emergency preparedness cornerstone attribute of Facilities and Equipment (Availability of ANS). This ANS unavailability affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Using Manual Chapter 0609 Appendix B, Emergency Preparedness Significance Determination Process (Section 5.5) Failure to Comply with 10 CFR 50.47(b)(5), the inspectors determined this finding to be of very low safety significance (Green) because the loss of siren function was of short duration and did not reach the Degraded Risk Significant Planning Standard (RSPS) threshold. The finding had a crosscutting aspect of Procedure Adherence, as described in the Human Performance crosscutting area because the EPTs failed to comply with the procedural guidance of WCPNGGC- 0300 (H.8). | |
| 05000400/FIN-2014007-03 | 2014Q3 | Green | NRC identified | Failure to Establish Appropriate Procedural Limitations to Prevent Exceeding TS Limits and Safety Analysis Assumptions | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that applicable regulatory requirements in technical specification (TS) 3.7.1.1 and design basis inputs in accident analyses were translated into procedural guidance. Specifically, the licensee did not follow their inservice test program guidance to account for surveillance test equipment instrument uncertainty when establishing the acceptability of Main Steam Safety Valve lift setpoints required by TS 3.7.1.1. Following identification by the team, the licensee generated nuclear condition report 697100 and performed an evaluation of the remaining available margin to the overpressure limit in the safety analysis, and discovered that, after potential instrument uncertainty was taken into account, the margin remained positive, but was reduced from approximately 19 psig to approximately 6 psig. The licensees failure to assure that applicable regulatory requirements in TS 3.7.1.1 and design basis assumptions in accident analyses were correctly translated into procedural guidance, as required by 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, by not accounting for the measurement and test equipment uncertainties as required by the inservice test program, it could have led to the actual lift setpoints exceeding the inputs used in the design basis safety analyses. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. | |
| 05000400/FIN-2014004-03 | 2014Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 6.8.1 requires the procedures recommended in RG 1.33 to be established, implemented, and maintained. Regulatory Guide 1.33 requires implementation of an RWP system. Specifically, RWP #1014, Task 4, Valve Maintenance RCB (No HRA Access), required HP to be notified prior to the start of work and for HP to be present and perform surveys when breaching a contaminated system. Contrary to these RWP requirements, on November 22, 2013, two workers entered the containment building and cut out two primary CS valves (CS-761 and 762) without having HP present to perform surveys when breaching a contaminated system. After they exited containment, HP discovered the valves on the ground with removable beta-gamma contamination levels up to 200,000 dpm/100 cm2. This finding was of very low safety significance (Green) because there was no substantial potential for overexposure. This was due to the fact that the external dose rates were low and the contamination levels were not high enough to constitute a substantial potential for overexposure. The inspectors noted that no personnel were contaminated as a result of this event. The licensee entered the event into their corrective action program as AR #648061. | |
| 05000400/FIN-2014007-02 | 2014Q3 | Green | NRC identified | Failure to Establish Appropriate Procedural Limitations Based on Design Requirements of the Emergency Diesel Generators | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that applicable regulatory requirements in technical specification surveillance requirement 4.8.1.1.2.e. were correctly translated into procedural guidance. Specifically, appropriate jacket water (JW) and lube oil (LO) standby temperature limitations, which ensured emergency diesel generator (EDG) capability to meet TS SR 4.8.1.1.2.e. requirements, were not translated into procedures for determining EDG operability. Following identification by the team, the licensee generated nuclear condition report 698245 and established administrative limits to ensure the EDG JW and LO temperatures were not allowed to drop below technically supportable limits. The licensees failure to assure that applicable regulatory requirements in technical specification surveillance requirement SR 4.8.1.1.2.e. were correctly translated into procedural guidance, as required by 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was determined to be more than minor because, it was associated with the mitigating systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability and reliability of the EDGs to respond to a design basis accident at the JW or LO temperature conditions at which they considered the EDGs operable. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. | |
| 05000400/FIN-2014007-04 | 2014Q3 | Green | NRC identified | Inadequate Categorization of Valves in Potential Release Paths During Accidents | The team identified a Green non-cited violation of 10 CFR Part 50.55a, Codes and Standards, for the licensees failure to categorize valves that were subject to a specific maximum leakage amount while in the closed position as Category A, as required by their American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code of record. Specifically, the team determined that the licensee failed to correctly categorize six valves that could allow emergency core cooling system (ECCS) leakage into the refueling water storage tank above the water level during ECCS post-accident recirculation operation. During the inspection period, the licensee generated nuclear condition report 699708, and performed an evaluation of the affected valves that verified the valves ability to meet leakage limits based on other monitoring that was in place. The licensees failure to categorize valves that were subject to a specific maximum leakage amount while in the closed position as Category A, as required by their ASME OM Code of record, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the reliability of the physical design barrier of the leak-tightness of valves in the release paths was not assured since leak testing was not performed due to inaccurate categorization. The team determined the finding to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of the hydrogen igniters in reactor containment. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance. | |
| 05000400/FIN-2014004-01 | 2014Q3 | NRC identified | Potential Impact of Sump Pumps out of Service | The inspectors identified an URI associated with an equipment clearance that inadvertently resulted in all sump pumps in the EDG and DFOST buildings being nonfunctional. This item is unresolved pending review and evaluation of the licensees evaluation to determine the impact of a potential internal flood and if a performance deficiency exists. On June 26, 2014, the licensee placed equipment under clearance to support installation associated with an Engineering Change (EC). This clearance removed all sump pumps in the EDG and DFOST buildings from service. Inspectors identified this issue and informed the licensee, who restored the sump pumps to service. Additional inspection activities are needed to determine the impact of a potential internal flood and if a performance deficiency exists. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000400/2014004-01, Potential Impact of Sump Pumps out of Service. |