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05000443/FIN-2018003-012018Q3Severity level IVSelf-revealingPressurizer Safety Valve Outside of Technical Specification LimitsA self-revealing Severity Level IV NCV of Technical Specifications 3.4.2.2, All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 psig +/- 3%, was identified when one of the pressurizer code safety valves failed as-found set point testing. Specifically, it was determined that the safety valve had a high as-found set point pressure after the valve was removed from service during the previous refueling outage in April, 2017 (OR18) and the inoperable condition existed for a period of time longer than the allowed T.S. ACTION time.
05000443/FIN-2018411-022018Q2GreenH.3NRC identifiedSecurity
05000443/FIN-2018411-012018Q2GreenH.8NRC identifiedSecurity
05000443/FIN-2018001-012018Q1GreenLicensee-identifiedLicensee-Identified ViolationThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section2.3.2 of the Enforcement Policy.Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control,requires, in part, that measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.Contrary to the above, from an unknown date until January 10, 2018, NextEra did not have a measure for verifying the adequacy of design of seven safety-related electrical manholes. Specifically, there were no original calculations to support the design of the manholes; and when NextEra staff reconstituted the structural design calculation, the results concluded that four of the seven manholes would not meet the design specification unless the loading demands were reduced from 500 to 200 pounds per square foot.Significance/Severity Level: Green because all four structures remained capable of performing their safety function.Corrective Action References:AR 02243800 and AR 02255652
05000443/FIN-2017004-012017Q4GreenH.1NRC identifiedLicensed Operator Examination IntegrityNot EnsuredThe inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 55.49, Integrity of Examinations and Tests, for the failure of the licensee to ensure that the integrity of the written examinations administered to licensed operators was maintained. During the planning of the biennial written examinations, two written examinations would have exceeded the 50 percent overlap criteria limit of questions administered in the previous four weeks of this examination cycle. This failure resulted in a compromise of examination integrity because it exceeded the NextEra Fleet Procedure TR-AA-220-1004, Licensed Operator Continuing Training Annual Operating and Biennial Written Exams, Revision 2, requirement to repeat less than or equal to 50 percent of the questions used during the exam cycle. However, this compromise did not lead to an actual effect on the equitable and consistent administration of the examination because of detection of this issue by the NRC prior to examination administration. This issue was entered into NextEras Corrective Action Program (CAP) as AR 2239906.The failure of NextEras training staff to maintain the integrity of examinations administered to licensed operations personnel was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the performance deficiency could have become more significant in that allowing licensed operators to return to the control room without valid demonstration of appropriate knowledge on the biennial examinations could be a precursor to a more significant event. Using IMC 0609, Significance Determination Process, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance (Green) because although the finding resulted in a compromise of the integrity of written examination, the equitable and consistent administration of the test was not actually impacted by this compromise. This finding had a cross-cutting aspect in the area of Human Performance, Resources, in that leaders ensure procedures are available and adequate to support nuclear safety. Specifically, NextEra established and implemented a procedure that contained instructions to licensed operator biennial exam writers that were unclear regarding regulatory guidance to limit written examination questions overlap. (H.1)
05000443/FIN-2017004-022017Q4GreenH.3NRC identifiedFailure of Exercise Critique to Identify a RSPSWeaknessThe inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Reglations (10 CFR) 50.47(b)(14) and 10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, Section IV.F.2.g. Specifically, Seabrook did not identify and critique a weakness associated with a risk significant planning standard (RSPS) during their critique following the August 30, 2017, emergency preparedness drill. The weakness involved the licensees declaration of a general emergency (GE) that was based on insufficient information. NextEra entered the issue into the corrective action program (CAP) as AR2242073.The inspectors determined that not identifying an exercise weakness related to a GE classification based on insufficient information during the exercise critique was a performance deficiency that was reasonably within the ability of Seabrook to foresee and prevent. The finding is more than minor because it is associated with the Emergency Response Organization attribute of the Emergency Preparedness Cornerstone and affected the cornerstone objective to ensure 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. Specifically, Seabrook personnel did not identify an exercise weakness associated with a RSPS when the incorrect basis for a GE declaration was used by the Site Emergency Director (SED). The finding was assessed using IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016. This attachment directs inspectors to utilize IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued September 22, 2015, because the finding andthe associated weakness is in the licensees emergency preparedness cornerstone. The inspectors determined the finding was a critique finding, the drill scope was full scale, the planning standard was risk-significant, and the performance opportunity was a success utilizing figure 5.14-1, Significance Determination for Critique Findings, and thus determined this finding was of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of Human Performance, Change Management, in that leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, although recent changes to the sites emergency classification and action level standard scheme were effective on July 2017, the new EAL procedure and training regarding the changes lacked sufficient specificity to ensure the users understood the new scheme with respect to the status of the containment integrity. (H.3
05000443/FIN-2017004-032017Q4GreenH.5Self-revealingInadequate Procedure Implementation Results in a Manual Reactor TripA self-revealing Green finding was identified for inadequate implementation of procedure MA 4.5, Configuration Control, Revision 18. Specifically, maintenance technicians failed to properly implement MA 4.5 while backfilling steam generator instrumentation, and inadvertently left an instrumentation valve partially open instead of fully open. This resulted in slow response of the instrument, and ultimately a high steam generator level, a feedwater isolation signal and a manual reactor trip. NextEra promptly rechecked other similar valves, then performed a root cause evaluation that eventually led to additional technician training and improved configuration controls during such evolutions. This finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and 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 to effectively implement MA 4.5 resulted in a valve being left out of its required position, a subsequent lack of steam generator water level control during low power operations, and ultimately required a manual reactor trip. In accordance with IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green), because the fin ding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. Additionally, the finding has a cross-cutting aspect in the area of Human Performance, Work Management, because the organization did not implement a process of planning, controlling, and executing the work activity such that nuclear safety was the overriding priority. Specifically, NextEra did not ensure that a steam generator backfilling activity was properly executed, which resulted in the slow response of a steam generator level indication, the overfeeding of the steam generator, a feedwater isolation signal, and the ultimate requirement to trip the reactor. (H.5)
05000443/FIN-2017008-022017Q3GreenH.5NRC identifiedFailure to Implement Test Program for Appendix R Emergency Lighting UnitsThe team identified a Green, non-cited violation of Seabrook License Condition 2.F, Fire Protection, because NextEra did not implement the fire protection test program to ensure that the emergency lighting units were in conformance with design requirements. Specifically, NextEra did not implement procedure LS0565.31, 8-Hour Emergency Light Inspections, to verify that the Appendix R emergency lighting units would meet the annual inspection requirements, as well as the 3-year preventive maintenance task for battery replacement and the 8-hour capacity test. Additionally, since the 3-year preventive maintenance task was coded incorrectly, there was no process to ensure that the LS0565.31 would be completed going forward. NextEra entered this issue into the corrective action program as AR 2214652. NextEras planned corrective actions included revising the classification of the emergency lighting unit preventive maintenance task in order to ensure that the task is performed at the appropriate frequency.The team determined that this issue was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, failure to conduct the annual inspection requirements and 3-year preventive maintenance activities could result in the emergency lighting units not meeting the 8-hour battery capacity requirement. The team evaluated this finding using Inspection Manual Chapter 0609, Fire Protection Significance Determination Process. Because safe shutdown conditions could be reached and maintained, this finding screened as having very low safety significance (Green). The team determined this finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the three-year preventive maintenance task to replace the batteries in the emergency lighting units was coded incorrectly in the work management system, which resulted in NextEra not completing the required testing and maintenance on the lighting units to ensure that they would perform their function during safe shutdown operations (H.5).
05000443/FIN-2017008-012017Q3GreenH.1NRC identifiedFailure to Correct Condition Adverse to Fire Protection Associated with Fire Safe ShutdownThe team identified a Green, non-cited violation of Seabrook Station Unit 1 Facility Operating License Condition 2.F, Fire Protection, for failure to implement and maintain in effect all provisions of the approved Fire Protection Program. Specifically, although NextEra identified that procedure OS1200.00 did not properly implement a mitigating action for a fire in the Switchgear Room A as prescribed in the Appendix R Safe Shutdown Analysis Report on August 30, 2010 (Action Requests (ARs) 576775 and 1638123), corrective actions were delayed due to higher priority work and were not timely commensurate with the potential safety significance. NextEra entered the issue into the corrective action program as AR 2214834 and planned to reprioritize the preparation and submittal of a license amendment request to resolve the issue.The issue was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors (fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by failing to correct the condition in a timely manner, NextEra did not ensure that the associated fire safe shutdown procedure implemented actions to mitigate a fire in the Switchgear Room A as analyzed in the Appendix R Safe Shutdown Report. The team performed a Phase 1 screening in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process. The deficiency affected the post-fire safe shutdown category because NextEras fire response procedures were degraded. The finding was screened to very low safety significance (Green) because it was assigned a low degradation rating because the procedural deficiencies could be compensated by operator experience and system familiarity. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Resources, in that, NextEra did not ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically,action to submit a license amendment request to support a deviation from the 10 CFR Part 50, Appendix R, III.G.2 requirements for cable separation had been rescheduled five times due to higher priority licensing work (H.1).
05000443/FIN-2017002-012017Q2NRC identifiedSeabrook Station Use and Application of Technical Specifications.An Unresolved Item (URI) was identified because additional NRC review and evaluation is needed to determine whether one or more performance deficiencies and non- compliances exist. The inspectors identified an issue of concern (IOC) broadly related to Sea brooks use and application of TS s limiting conditions for operability (LCO). Specifically, performance deficiencies and non- compliances appear to exist when support systems or subsystems have not met the TS definition of operability and NextEra has not entered the associated supported systems TS LCO and applied the required actions. The industry has sometimes used the term cascading to describe the impact of a support systems inoperability on supported systems. A specific example of this IOC involves an inoperable CWT, which is the seismically qualified portion of Seabrooks ultimate heat sink (UHS). The inspectors have questioned whether an inoperable CWT renders systems that it supports (PCCW, EDG s, and RHR) inoperable. Additional information is needed to determine whether one or more performance deficiencies and TS violations exist. A Task Interface Agreement has been submitted to the NRCs Office of Nuclear Reactor Regulation (NRR) to resolve the IOCs presented below regarding the correct application of Seabrooks TSs and the impact of an inoperable CWT on its supported systems. 13 Description : Technical Specification Use and Application Concern: The Seabrook TS s are based on NUREG -0452, Standard Technical Specifications for Westinghouse Pressurized Water Reactors. Seabrook TS 1.21 defines OPERABLE OPERABILITY as a system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s), and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling (emphasis added) and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support (emphasis added) function(s). TS 3.0.2 states that noncompliance with a specification shall exist when the requirements of the LCO and associated ACTION requirements are not met within the specified time intervals, except as provided in Specification 3.0.5. If the LCO is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. Seabrook TS do not contain an exception to LCO 3.0.2, similar to LCO 3.0.6 in the Improved Standard Technical Specifications (ISTS) for Westinghouse Pressurized Water Reactors (NUREG -1431). The ISTS LCO 3.0.6 states, in part, when a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.15, Safety Function Determination Program (SFDP). Background and Licensing Basis : Seabrook Station receives circulating and SW via two large tunnels that were mined a distance of over 3 miles to the Atlantic Ocean. SW is a safety -related system that provides cooling to the safety -related EDGs, PCCW, RHR, and other systems. The tunnels were lined with reinforced concrete following excavation. However, since the tunnels were not formally, seismically -qualified, a reinforced concrete mechanical draft CWT was constructed onsite as the UHS, to provide cooling water to safety -related systems following a seismic event that blocked more than 95 percent of the tunnel water flow to ensure that the requirements of General Design Criteria (GDC) -2, Design Bases for Protection Against Natural Phenomena, are met. Seabrooks conformance with GDC- 2 is described in the UFSAR Section 3.1.1.2. The design bases safety functions of the Station SW system and the UHS are described in UFSAR Sections 9.2.1.1 and 9.2.5.1, respectively. The PCCW systems conformance with GDC -44, Cooling Water, is described in UFSAR Section 3.1.4.15. Licensing Basis Amendments : On April 7, 1993, by letter NYN -93052 (ML17191A390), the licensee submitted license amendment request (LAR) 93- 02: Service Water System/Ultimate Heat Sink OPERABILITY Requirements (TAC No. M85750). The letter stated that the purpose of the LAR was to propose changes to the Seabrook TSs to redefine the requirements for an OPERABLE SW system and to consolidate the SW requirements with the requirements for the UHS. The letter continued by stating that the Seabrook TS 3/4.7.4 (in existence in 1993) required two OPERABLE SW loops with each loop having three 14 OPERABLE pumps (two (ocean) SW pumps and one cooling tower service water (CTSW) pump) when in Modes 1, 2, 3, and 4. The letter asserted that this requirement was unnecessarily restrictive since the second SW pump in each loop is not required for normal or design basis accident conditions and the associated CTSW pump provides the required redundancy during the postulated design basis event. Specifically, the letter stated, in part, The proposed changes: (1) redefine an OPERABLE SW loop as having one OPERABLE SW pump and one OPERABLE CTSW pump;... The letter continued by stating that the consolidation (of TS LCOs 3.7.4 and 3.7.5) is proposed to reduce the potential for confusion between the specifications and to control station operation in a manner consistent with the station design basis. The inspectors identified that the TS wording changes submitted by the licensee and approved by the staff did change the actions for the SW system that consists of ocean SW and CTSW subsystems and ocean and atmospheric UHS. However, given the inspectors understanding of the application of the TS, as described in the above section titled, TS Use and Application Issue of Concern, the revised TS wording does not appear to be sufficient to relieve Seabrook from entering the applicable supported systems (EDGs and PCCW) LCOs when the associated SW subsystems are rendered inoperable. By letter dated October 5, 1994, the NRC app roved Amendment No. 32 to Facility Operating License NPF -86: Primary Component Cooling Water System Operability Requirements LAR 93- 01 and Service Water System/Ultimate Heat Sink Operability Requirements - LAR 93 -02 (TAC M85491 and M85750). The approval letter (ML011800279) states, in part, that this amendment revises the Appendix A TSs relating to the operability requirements for the SW system and the UHS. The safety evaluation report (SER) states, in part, because the tunnels between the Atlantic Ocean and the pump house are not designed to seismic Category I requirements, a seismic Category I CWT is provided to protect against their failure due to a seismic event. Therefore, to meet the design basis for the SW system , each loop must have an operable SW pump and an operable CTSW pump. In addition, the SER states, in part, that the proposed changes to TS 3/4.7.4 reflect the design basis of the SW system in that with two operable loops, each having one operable SW pump and one operable CTSW pump (given each pump's UHS is operable), the system is capable of performing its safety function for all design basis events given the worst case single active failure, including the failure of either EDG. The staff also concludes that the consolidation of the SW system (TS 3.7.4) and UHS (TS 3.7.5) specifications to one TS LCO (3.7.4) was acceptable and necessary to achieve and maintain clarity, within the specifications, of the overall requirements for system operability. The inspectors noted that the LAR and SER statements do not appear to coincide with the language in the approved Amendment No. 32, in that, the revised TS language identifies that the SW system is comprised of two subsystems with the ocean SW subsystem treated separately from CTSW subsystem. The inspectors also noted the addition of an allowed outage time (AOT) of 24 hours for two inoperable ocean SW pumps, and 72 hours for the CWT or two inoperable CTSW pumps. The inspectors noted that the LAR did not appear to identify or acknowledge that the licensing bases for Seabrook requires the CWT basin and one CTSW pump for the SW system to withstand the effects of natural phenomena such as an earthquake, without the loss of capability to perform their safety functions. Additionally, the LAR did not appear to identify or acknowledge that the licensing bases for Seabrook requires ocean SW to withstand the effects of natural phenomena such as tornadoes, without the loss of capability to perform their safety functions. Although these are low probability e vents, in a deterministic 15 licensing regime, the inspectors determined that consistent with the SER, and as detailed specifically by the licensee in the April 1993 LAR, an operable SW system should include two operable loops, with each having one operable ocean SW pump and one operable CTSW pump (given each pump's UHS is operable), such that the system is capable of performing its safety function for all design basis events, given the worst case single active failure, including the failure of either EDG. Specific Examples of the Concern : During the spring 2017 refueling outage, NextEra submitted a one -time LAR (ML17094A764) dated April 4, 2017, regarding the application of the CWT TS. Subsequently, the inspectors reviewed the records of Seabrooks CWT repair activities and OOS times since 2015 and monitored NextEras outage activities. During the review of historical records, the inspectors identified several examples of what could be interpreted as TS inoperability for PCCW and the ED Gs due to an inoperable CWT (TS 3.7.4.b) in Modes 1, 2, 3, and 4. Also, in Modes 5 and 6 during OR18, potential examples of what could be interpreted as TS inoperability were noted for the EDGs and the two RHR loops due to a non -functional CWT. It is important to note that the issue of concern associated with these examples would be based on a conclusion that the SW system / UHS LCO (3.7.4) provides a cooling water support function for both PCCW and EDG, in accordance with the TS definition (1.21) of OPERABILITY, in that the CWT is a necessary component of an OPERABLE SW / UHS due to its seismic qualification. Since the Seabrook TS do not contain an exception to LCO 3.0.2 similar to ISTS LCO 3.0.6 (NUREG 1431, Revision 4), the inspectors position is that the SSCs supported by the UHS (EDGs, PCCW and RHR) could be interpreted as inoperable due to the inoperable UHS. If it is assumed that an inoperable CWT train, a TS support system train, also renders the associated trains of its supported systems inoperable, the inspectors identified instances in the last 3 years where one or more trains of CWT SW inoperability may have exceeded the most limiting TS Action requirements for the associated supported systems. In these instances, NextEra did not enter the associated TS LCOs, and did not perform the applicable ACTIONS for the supported SSCs. Further, on the occasions that the CWT was inoperable, the supported EDG TS Surveillance Requirement 4.8.1.1.1.f(14) could not be met during the CWT maintenance. The inspectors understand that typically the application of TS Surveillance Requirement 4.0.1 would hold and LCO 3.8.1 would not be met and all applicable ACTIONS for the inoperable EDG(s) would be required to be met within the specified time intervals. Below are two specific examples of the IOC: On June 9 through June 10, 2015 (approximately 24 hours), and on October 13, 2016 (approximately 18 hours), both trains of CTSW were inoperable for CWT basin cleaning and inspection while in Mode 1. For this support system, NextEra entered the TS Action 3.7.4.c that provides an AOT of 72 hours to restore at least one train to OPERABLE status or be in hot shutdown Mode 4 within 6 hours and cold shutdown Mode 5 within the following 30 hours (108 total hours). Upon inoperability of this support system (UHS), NextEra did not declare the supported systems (PCCW and the EDGs) inoperable and enter the associated TS Actions. If determined to be applicable, TS 3.7.3 and TS 3.8.1 would have required being in Mode 3 within 7 and 8 hours, and Mode 5 within 37 and 38 hours total, respectively. On April 19, 2017, with the B EDG already inoperable, the A CWT loop was removed from service to replace portions of its CWT pump discharge piping while the 16 plant was in Mode 6 (refueling) with less than 23 feet of water above the reactor flange. LCO 3.7.4 (SW / UHS) only applies in Modes 1, 2, 3, and 4. Before the transition to Mode 6, the B EDG had been rendered inoperable for planned maintenance and testing while the plant was defueled and with no applicable operational mode. In Modes 5 and 6, LCO 3.8.1.2 requires one OPERABLE EDG and TS 3.0.4 requirements were met for entering Mode 6, in part, because of the operable A EDG. While in Mode 6, both trains of ocean SW were operable to supply cooling water. However, the inspectors have interpreted that Seabrooks current licensing basis requires each EDG to be supported by its train of seismically qualified cooling water. If it is assumed that a seismically qualified source of cooling water was required on April 19, when the A CWT loop was removed from service, its supported system, the A EDG m ay have been rendered inoperable for a period of approximately 10 hours at the same time as the B EDG was inoperable for maintenance. Additionally, the inspectors identified a second potential operability concern associated with the RHR system. Specifically, in Mode 6, LCO 3.9.8.2 requires two OPERABLE independent RHR loops while the water level is less than 23 feet above the top of the reactor vessel flange. With less than the required RHR loops OPERABLE, Action 3.9.8.2 requires immediate initiation of corrective action to return the required loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible. This condition may have existed because the A CWT loop was inoperable, which could be interpreted to have resulted in the A RHR loop being inoperable for approximately 65 hours while the plant was in Mode 6 with less than 23 feet of water above the reactor flange. Issues Requiring Resolution through the T ask Interface Agreement Process : 1. Do the current Seabrook Station (50 -443) license and TSs (TS 3.0.2) require parallel/simultaneous entry into both the support system (e.g., the SW system and UHS, TS 3.7.4) and the supported systems (e.g., Electrical Power Systems, AC Sources (diesel generators), TS 3.8.1 and PCCW System, TS 3.7.3) when the definition of OPERABLE (TS 1.21) is not met for the support system? Although one example is provided, the broader question requiring an answer is whether Seabrook is required to cascade their TS. The Seabrook TS have never included nor have been amended to incorporate the non- cascading provisions of ISTS 3.0.6 or the required, accompanying SFDP. 2. Does the October 5, 1994, License Amendment No. 32 on the SW system/UHS operability requirements give NextEra the latitude to remove the entire CWT from service for 72 hours even though it is needed to support key safety -related systems with much shorter LCOs (i.e., when both trains of those systems are OOS )? 3. If Amendment No. 32 allows the flexibility to remove both loops of the CTSW or the mechanical draft CWT for 72 hours without affecting the operability of the supported systems, is the current TS language consistent with this flexibility? 4. Do the current Seabrook Station (50 -443) license and TSs (TSSR 4.8.1.1.1.f(14)) - require Seabrook to be capable of simulating each trains CWT actuation signal while the associated EDG is running at minimum accident loading when the CWT or a train of CTSW is removed from service and is inoperable for the A OT specified in TS 3.7.4 and does TS 4.0.1 need to be applied such that the failure to meet a TSSR, whether such failure is experienced during the performance of the surveillance or between 17 performances of the surveillance, shall be a failure to meet the LCO and would require taking the actions in TS 3.8.1. NextEra Position: Initially, NextEra stated its position in its April 4, 2017, one -time LAR (ML17094A764). Additional discussions with NextEra indicate that it is the licensees position that entry into the support system TS alone is sufficient to comply with Seabrook TS 3.0.2 as written even though the Seabrook TS do not include a provision similar to ISTS 3.0.6. (Note: TS 3.0.2 states that noncompliance with a specification shall exist when the requirements of the LCO and associated Action requirements are not met within the specified time intervals, except as provided in TS 3.0.5. If the LCO is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required.) NextEra has since stated its position in this matter as documented in a position paper that can be found in ADAMS at ML17191A412. Specifically, NextEra asserts that the Seabrook SW system consists of two independent loops, each of which can operate with either a SW pump train or a CTSW pump train. NextEra interprets TS Amendment No. 32, approved in October 1994, as having evaluated the impact of SW TS (3.7.4) AOT for both a single and dual train unavailability of the CWT. NextEra believes that the proposed change and acceptance by the NRC staff recognized that the change was intended to redefine the requirements for both the PCCW and SW system as well as the UHS (i.e., the CWT in this case). NextEra believes that the LAR was proposed to take advantage of what the licensee believes to be a redundancy in the SW and UHS designs to provide enhanced operational flexibility. NextEras reading of the SER for the amendment can be interpreted to have stated that the NRC staff agreed with the risk -based methodology and assumptions used, and that the change in SW system unavailability due to the proposed TS amendment and the resulting increase in the total reactor core damage frequency are insignificantly small. Further, NextEra interprets the amendment to read that the staff found the consolidation of the SW system and UHS into one TS to be acceptable and necessary to achieve and maintain clarity within the specifications of the overall requirement s of system operability. (Note: NextEra remained silent regarding the need to meet the GDC requirements governing the protection against natural events for either UHS during the TS AOT.) NextEra interprets the NRCs regulations to have stated that the S ER associated with Amendment No. 32 is not actually part of the regulated licensing basis. Consequently, NextEra believes that a deterministic judgement that the current Seabrook TS was incorrectly made by the NRC via Amendment No. 32 should not be made. NextEras interpretation is that Seabrooks licensing basis remains as originally approved, notwithstanding the current regulatory approach described in Inspection Manual Chapter ( IMC ) 0326 (but not in any regulation). Therefore, NextEra interprets the c urrent TSs to allow removal of redundant portions of SW for limited time periods as recognition of the low probability for occurrence of a natural phenomenon event. Thus it is NextEras position that any new changes to the language of the TS may provide g reater clarity, but offer no substantial offsetting increase in safety. 18 Current Seabrook Administrative Controls : In accordance with Seabrooks procedure, OPMM, Operations Management Manual, Revision 107, Operations Management issued a Standing Operating Order (SOO 17- 002) to the operating department to address the concern with the use and application of TS. The order was effective on February 27, 2017, and remains effective until future resolution of the issue, and revisions to Seabrooks manuals and programs are completed, as appropriate. The order describes the correct application of TS with respect to a supporting function and its potential effect on support system operability, with the exception of the disputed issue related to the CWT- impacted LCOs. In addition, the SOO directs the operators to carefully review TS in order to determine potential operability concerns with respect to the support and supported systems as they are taken OOS . Additional corrective actions were taken to include training for the licensed operators to reinforce and ensure the correct use and application of TS in the future. Therefore, there is no immediate safety concern with respect to the issue of concern. Unresolved Item : The inspectors have coordinated with N RR through the use of the process described in NRR Office Instruction No. (COM -106), Control of Task Interface Agreements, to review this URI regarding the correct application of Seabrooks TS and the impact of an inoperable CWT on its supported systems. Pending resolution this issue is unresolved. (URI 05000443/2017002 -01, Seabrook Station Use and Application of Technical Specifications).
05000443/FIN-2017403-012017Q1GreenH.12NRC identifiedSecurity
05000443/FIN-2016404-012016Q4GreenLicensee-identifiedLicensee-Identified Violation
05000443/FIN-2016007-022016Q3NRC identifiedPotential Missed Evaluation and Reporting of an Adverse Condition to the NRCIntroduction: The team identified an unresolved item (URI) to further review whether NextEras evaluations associated with two PCCW pump motor failures in 2008 and one in 2015, and the associated conclusions not to report the conditions to the NRC, constituted a violation of NRC regulations. Description: As described in Section 1R21.2.1.3.1 above, the team reviewed two time periods where NextEra concluded that PCCW motor failures were the result of a manufacturing defect, however, these were not reported to the NRC. Specifically, a manufacturing defect was identified in a third-party failure analysis, dated January 21, 2009, following the failure of PCCW motors C and D in 2008. A third PCCW motor (B) failure occurred due to the same manufacturing defect in June 2015. These failures appeared to occur from one common cause. During this inspection, the team questioned whether the reporting requirements of 10 CFR Part 21 (Part 21), Reporting of Defects and Noncompliance, were satisfied, because no report was made to the NRC. In response to this concern, NextEra initiated AR 2153374, and initiated a substantial safety hazard (SSH) evaluation for the PCCW pump motor deviations in accordance with Part 21 and NextEra procedure LI-AA-102-1002, Part 21 Reporting. NextEra subsequently completed the SSH determination, and concluded that the deviation (i.e., the manufacturing defect) constituted a defect that could contain an SSH. They notified the NRC in accordance with 10 CFR 21.21(d)(3)(i) reporting requirements on October 20, 2016, via fax (Event Notification 52310). Subsequent to the onsite inspection, and while evaluating NextEras compliance with Part 21 evaluation and reporting requirements, the NRC noted that 10 CFR 21.2(c) stated, in part, that evaluation of potential defects and appropriate reporting of defects under 10 CFR 50.72 and 50.73 satisfies the evaluation, notification, and reporting obligation to report defects under Part 21. While the NRC recognized that NextEra had not made an NRC notification related to the identified PCCW motor manufacturing defect in accordance with 10 CFR 50.72, 50.73 or Part 21, the team did not review NextEras specific reportability evaluations with respect to 10 CFR 50.72 and 50.73. The team did note that NextEras Part 21 reviews, both in 2009 and 2015 did not specifically perform the evaluation specified in 10 CFR 21.21(a)(1) to determine whether the deviation in a basic component, which, on the basis of an evaluation, could create a substantial safety hazard. Since there appears to be overlapping reporting requirements among 10 CFR 50.72, 50.73 and 21.21, and the team did not specifically review NextEras reportability considerations for 10 CFR 50.72 and 50.73, additional inspection is necessary in order to determine whether there was a violation of any of the three reporting regulations. Accordingly, this issue is being treated as an unresolved item (URI) pending further inspection by the NRC staff to determine whether not evaluating and reporting the manufacturing defect associated with the PCCW motors constituted a more than minor violation of NRC reportability regulations. (URI 05000443/2016007-02, Potential Missed Evaluation and Reporting of an Adverse Condition to the NRC).
05000443/FIN-2016007-012016Q3GreenP.1NRC identifiedInadequate Corrective Actions to Preclude Repetition of a Significant Condition Adverse to QualityThe team identified a finding of very low safety significance, involving a non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XVI, Corrective Action, for not performing corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, in 2008, two of four primary component cooling water (PCCW) pump motors failed within a four month period due to a manufacturing defect. NextEra established but did not perform a corrective action to replace all four motors with re-wound motors, free of the identified manufacturing defect. Subsequently, in 2015, a third motor failure occurred due to the same manufacturing defect. NextEras immediate corrective actions included entering this issue into their corrective action program (AR 2153536), implementing an electrical testing program that would provide an early indication of further degradation of the manufacturing defect until motor replacement, and completing a prompt operability determination to assess current PCCW system operability. This finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team screened the finding for safety significance and determined that a detailed risk evaluation (DRE) was required because the finding involved a partial loss of a support system (PCCW pump B) that would increase the likelihood of an initiating event and impacted mitigating equipment (Item C - Support System Initiators of Exhibit 1). The DRE, performed by a Region I senior reactor analyst (SRA), concluded that the performance deficiency resulted in a change in core damage frequency of high E-7/yr, or very low safety significance (Green). The finding had a cross-cutting aspect in Problem Identification and Resolution (Resolution), because NextEra did not take effective corrective actions to address this issue in a timely manner commensurate with its safety significance. Specifically, NextEra did not perform motor replacements for susceptible installed PCCW motors within a reasonable due date as specified by the 2009 corrective action to preclude repetition (CAPR); and plant procedures, programs and resources were available for the decision-making process to schedule the motor replacement.
05000443/FIN-2016010-012016Q3Severity level IIINRC identifiedDeliberate Misconduct of Any Licensee EmployeeThe NRC has also preliminarily determined that you committed an apparent violation of Title 10 of the Code of Federal Regulations (CFR) 50.5(a)(1 ), "Deliberate Misconduct." The NRC's deliberate misconduct rule prohibits employees of any licensee from engaging in deliberate misconduct that causes, or would have caused if not detected, a licensee to be in violation of NRC requirements. Your apparent violation is being considered for escalated enforcement action in accordance with the NRC Enforcement Policy. The current Enforcement Policy is included on the NRC's Web site at httg://www.nrc.gov/about. A copy of your apparent violation is provided as Enclosure 2 to this letter. Since the NRC has not made a final determination in this matter, a Notice of Violation is not being issued at this time. Please be advised that the number and characterization of the apparent violation may change as a result of further NRC review. We believe we have sufficient information to make an enforcement decision regarding the apparent violation and are considering issuing to you an Order prohibiting your involvement in NRC regulated activities for some period not likely to exceed three years. However, before the NRC makes its final enforcement decision, we are providing you an opportunity to provide your perspective on this matter, including the significance, cause, and corrective actions, as well as any other information that you believe the NRC should take into consideration by: (1) requesting a pre-decisional enforcement conference (PEC) to meet with the NRC and provide your views in person; (2) requesting Alternative Dispute Resolution (ADR); (3) responding to the apparent violation in writing; or, (4) accepting the violation as characterized in this letter and notifying us of that decision within 1 0 days.A weapon that had been staged at a security post on August 2, 2015, at the NextEra Energy Seabrook, LLC (NextEra) Seabrook Station (Seabrook) was returned to the armory on August 21, 2015, for routine cleaning. While cleaning the weapon, the armorer found that a foam earplug insert and two pieces of rolled up paper had been stuffed in the barrel. NextEra notified the NRC Senior Resident Inspector at Seabrook, who in turn, informed regional staff and management. The region immediately dispatched security inspectors and investigators and, on August 24, 2015, formally launched a high-priority 01 investigation. 01 interviewed site security personnel who had access to the weapon. During initial interviews on August 24, one security officer (SO) acknowledged to 01 that he had stood watch in the position with the affected rifle on two occasions during the subject period. However, the SO testified that he had not placed the materials in the weapon and that he had no information about how the materials got inside of it. Afterward, the SO commented to some of his coworkers that the 01 interview had made him feel like he had something to do with the tampered weapon. He made several comments to other SOs indicating that he may have been involved in tampering with the rifle. He asked a coworker to contact the 01 agents and ask them to meet him at an off-site location. Although 01 contacted the SO, he declined to meet with them at that time. On multiple occasions between August 27 and September 24, 01 attempted tore-interview the SO, but he declined each request. On October 19, 2015, the SO spoke with 01 and stated that he believed he placed the materials in the weapon. The SO told 01 that he did not know why he did it, and adamantly stated that he was not trying to hurt anyone or to assist anyone with gaining access to the site. He acknowledged that it was reasonable to assume that he didn't come forward about what he had done because he was afraid of being fired. The SO also affirmed that he was not aware of adverse issues with any other weapons or equipment at the site. 01 concluded that the SO deliberately placed the materials in the rifle. APPARENT VIOLATION I 0 CFR 73.55(k)(2) requires licensees to ensure that all firearms, ammunition, and equipment necessary to implement the site security plans and protective strategy are in sufficient supply, are in working condition, and are readily available for use. Contrary to the above, from August 14, 2015- August 22, 201 5, NextEra Energy Seabrook, LLC did not ensure that all firearms necessary to implement the site security plans and protective strategy were in working condition. Specifically, foreign material had been introduced into the barrel of a rifle staged at a security post that was established to implement the site protective strategy. Because of the foreign material, the rifle could not have been ensured to fire properly. APPARENT VIOLATION BEING CONSIDERED FOR ESCALATED ENFORCEMENT 10 CFR 50.5(a)(1) states, in part, that any employee of a licensee may not engage in deliberate misconduct that causes a licensee to be in violation of any regulation issued by the Commission. 10 CFR 73.55(k)(2) requires licensees to ensure that all firearms, ammunition, and equipment necessary to implement the site security plans and protective strategy are in sufficient supply, are in working condition, and are readily available for use. Contrary to the above, on August 14, 2015, while employed by NextEra Energy Seabrook, LLC (NextEra) as a security officer at Seabrook Station, you engaged in deliberate misconduct that caused NextEra to be in violation of an NRC regulation. Specifically, you placed foreign material into the barrel of a rifle that had been staged at a security post. Because of the foreign material, the rifle could not have been ensured to fire properly. As a result, NextEra did not ensure that all firearms necessary to implement the site security plans and protective strategy were in working condition.
05000443/FIN-2016007-032016Q3GreenNRC identifiedFailure to Perform Required ASME InService Testing of Manual Isolation Valves for the Atmospheric Steam Dump Valve Block ValvesThe team identified a finding of very low safety significance, involving a non-cited violation of Seabrook Technical Specification Surveillance Requirement 4.0.5, Surveillance Requirements for In-Service Inspection and Testing of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 Components. Specifically, the manual isolation valves for the atmospheric steam dump valves had an active safety function to close, in order to mitigate the radiological consequences of a steam generator tube rupture (SGTR) accident, but had not been placed in the Seabrook In-Service Test Program and tested, as required by the Technical Specifications and ASME Code. As a result, degraded valve performance could go uncorrected without adequate acceptance criteria to ensure that a SGTR would not result in an unacceptable increase in the consequences of that accident (e.g., a more than minor reduction in the margin between the postulated licensing basis radiological release and the regulatory limits). In response, NextEra entered the issue into their corrective action program (AR 2153195) and performed a preliminary assessment of the valves, which concluded that they were fully operable. This finding was more than minor because it was associated with the System, Structure, or Component (SSC), and Barrier Performance attribute of the Containment Barrier Cornerstone and adversely affected the cornerstone objective of ensuring the reliability of associated risk-important SSCs. The team determined that the finding was of very low safety significance (Green) because it was a deficiency confirmed not to represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding did not have a cross-cutting aspect because it was not considered indicative of current licensee performance.
05000443/FIN-2016405-012016Q3GreenNRC identifiedSecurity
05000443/FIN-2016002-012016Q2GreenH.11Self-revealingAutomatic Initiation of Emergency Feedwater Resulting from Performance of Procedural Steps in a Manner Prohibited by Documented InstructionsA self-revealing Green NCV of 10 CFR, Appendix B, Criterion V, Instructions Procedures, and Drawings, was identified, because NextEra did not ensure that activities affecting quality were accomplished in accordance with documented instructions. Specifically, while implementing a procedure following a plant trip that occurred on March 2, 2016, NextEra staff performed steps of a procedure in a manner that was prohibited by a departmental instruction, leading to an automatic initiation of emergency feedwater (EFW) to maintain adequate steam generator (SG) level. NextEra entered this issue into their corrective action program (CAP) and subsequently initiated a root cause evaluation to determine the factors which contributed to the event. Additionally, NextEra took corrective actions (C/As) to provide additional training and guidance for their staff and to resolve issues with existing procedures, which were determined to have been contributing factors during the event. The inspectors determined that this performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability (loss of FW) and challenge critical safety functions during shutdown as well as power operations. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because NextEra did not ensure that individuals stopped when faced with uncertain conditions. Specifically, the individuals involved did not adequately challenge the basis for a decision to disregard a department instruction.
05000443/FIN-2016002-022016Q2GreenH.8Self-revealingMultiple Letdown Isolations Resulting from an Inadequate Procedure and the Performance of Steps Not Prescribed by Established ProceduresA self-revealing Green NCV of 10 CFR, Appendix B, Criterion V, Instructions Procedures, and Drawings, was identified because NextEra did not ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances and that these activities were accomplished in accordance with these procedures. Specifically, a procedure associated with the testing of safety-related containment isolation functions did not contain sufficient instruction to ensure proper control of plant configuration; thus implementation of this procedure resulted in an inadvertent letdown isolation. Additionally, while attempting to perform this test on a subsequent occasion, individuals performed additional steps not prescribed in the associated procedure; the execution of these additional steps resulted in an additional inadvertent letdown isolation. NextEra entered these issues into their CAP and subsequently performed apparent cause evaluations for the two events, made necessary changes to the associated procedure, and provided coaching to NextEra staff. The inspectors determined that this performance deficiency was more than minor because it was associated with the Procedure Quality and Human Performance attributes of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability (letdown isolation) during power operations. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Procedural Adherence, because NextEra failed to ensure that individuals followed processes and procedures appropriately.
05000443/FIN-2016008-012016Q1GreenNRC identifiedFailure to Complete Operability Determinations for ASR-affected Structures10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, 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. NextEra Nuclear Fleet Administrative Procedure, EN-AA-203-1001, Operability Determinations/Functionality Assessments, identifies the responsibilities and requirements for preparation and approval of Immediate Operability Determinations (IOD) and Prompt Operability Determinations (POD) for establishing the acceptability of continued operation of a plant structure, system, or component that is suspected to be degraded or nonconforming. Per Section 2.0, Terms and Definitions, IODs are performed by the Shift Manager without delay (within 8 hours of discovery), using best available information to make an operability declaration. Upon request of the Shift Manager, a POD is performed as a follow-up to an IOD when additional information is needed to confirm the declaration of operability. Contrary to the above, on two occasions between March 17, 2015, and January 22, 2016, Energy Seabrook, LLC (NextEra) did not accomplish an activity affecting quality in accordance with its procedure. Specifically, NextEra received information from vendors identifying non-conforming conditions adversely impacting two reinforced concrete structures at Seabrook Station, and did not complete an appropriate IOD or initiate a follow-up POD to evaluate the impact of that non-conforming condition on structural performance. In particular, 1) On March 17, 2015, NextEra entered a WJE report, entitled Condition Assessment of the Cracking in the RHR and CS Equipment Vault, into the station document tracking system and added the reports recommendations into the Corrective Action Program under Action Report (AR) 01977456, without completing an appropriate IOD or initiating a POD. The report identified structural loading (a load not considered by ACI 318-71, the design and construction code of record) due to ASR as the cause for the excessive bulk expansion and cracking of the RHR/CS Vault interior and exterior support walls; and 2) On December 2, 2015, NextEra initiated AR 02094762 to track recommendations from SG&H report entitled Evaluation and Design Confirmation of As-Deformed CEB, without completing an appropriate IOD or initiating a POD. The report also identified structural loading due to ASR as the cause for deformation of the Containment Enclosure Building (CEB), a condition not conforming with ACI 318-71. This violation is associated with a Green Significance Determination Process finding.
05000443/FIN-2015004-012015Q4NRC identifiedIssue of Concern Regarding Implementation of the Seabrook Structures Monitoring Program and structural evaluations of the CEB and RHR/CS VaultDuring the week of October 26, 2015, the NRC audit team observed the last planned large-scale specimen testing at FSEL, reviewed test program results and analyses completed, to date, and interviewed NextEra staff and their consultants. Audit team activities and conclusions are documented in an NRC audit report (ADAMS Accession No. ML15307A019) dated December 17, 2015. No significant observations or concerns were identified related to the conduct of testing to appropriate quality assurance criteria. NextEra staff planned to have the test data analyses completed by the end of 2015, in support of submitting to the NRC a license amendment request in 2016, to address an ASR-related non-conforming condition with the current licensing basis. NRC inspectors conducted in-office reviews of the root cause evaluation for the Containment Enclosure Building Local Deformation, Event Dated December 19, 2014, completed per AR 02014325, and the Condition Assessment of Cracking in RHR and CS Equipment Vault, documented in Foreign Print (FP) 100903, dated March 17, 2015. NextEras CEB RCE described two root causes. First, regarding the physical causes of CEB deformation, NextEra staff concluded that internal expansion (strain) produced by ASR in the CEB concrete (in-plane direction of the CEB shell) and ASR expansion in the backfill concrete, coincident with a unique building configuration, resulted in CEB deformation. Second, regarding NRC identification of this issue, NextEra concluded this was not identified by plant staff due to an organizational mindset that viewed conditions such as concrete cracks, water infiltration, and misalignment issues as acceptable and inconsequential. Additionally, the RCE identified that NextEra staff did not perform and document comprehensive evaluations of building conditions that could have potentially revealed more significant underlying conditions, such as localized deformation of the CEB. These NRC-identified performance deficiencies were previously dispositioned as non-cited violation (NCV) 2015002-01 and NCV 2015002-02 in NRC inspection report 05000443/2015002 (ADAMS Accession No. ML15217A256). The inspectors determined NextEra corrective actions to address these problems included: 1) a revision to their design control procedures to require pozzolanic materials like fly ash or slag cement to be added to concrete mixes to prevent ASR in any new concrete structures; and, 2) the implementation of multiple training and program changes to correct the organizational mindset issues and strengthen individual responsibilities and accountability for implementation of the Seabrook Structures Monitoring Program. The inspectors concluded the RCE was reasonably thorough and utilized a cause and effect methodology that was appropriate to the problem statement. However, NextEras RCE, dated December 19, 2014, concluded that the reason that NRC inspectors, and not plant staff, identified the presence of localized ASR-induced deformation in Seabrooks concrete structures was due to an organizational mindset that viewed conditions such as concrete cracks, water infiltration, and misalignment issues as acceptable and inconsequential. The inspectors concluded the RCE was reasonably thorough and utilized a cause and effect methodology that was appropriate to the problem statement. The planned and in-process organizational and program related corrective actions appeared appropriately focused on the identified causes of the problem. However, the inspectors concluded that the corrective actions, taken to date, to implement multiple training, program and oversight changes to correct the organizational mindset issues and strengthen individual responsibilities and accountability for implementation of the Seabrook Structures Monitoring Program have either not been implemented or are not yet effective and thus require additional management attention near-term. The inspectors noted that the CEB RCE referenced the results of a Finite Element Analysis (FEA) model of the CEB. The FEA and results were documented in FP 100985. The inspectors review of FP 100985 identified that the FEA model simulated ASR expansion to assess the impact of expansion induced deformation on the structural performance of the CEB. The FEA evaluated CEB design capacity against assumed loading, based on ACI 318-71 criteria, including simulated loads associated with the as-deformed condition. As documented in FP 100985, the FEA model also simulated the impact of the external structural loading due to ASR expansion of the backfill concrete. Based upon the review of the CEB structural assessment described in FP 100985 and the limited structural analysis of the RHR/CS vault documented in FP100903, the inspectors had multiple follow-up questions regarding the CEB and RHR/CS vault structural assessments and the potential impact of these evaluation results on NextEras open Immediate Operability Determinations and Prompt Operability Determinations (PODs) for these ASR-affected structures (reference AR 01664399, AR 01929460, AR 01977456, AR 02004749, and AR 02044627). The follow-up questions were developed via a collegial review by the Region 1 Senior Reactor Analyst and structural engineers from NRC Offices of Nuclear Reactor Regulation and Region 2, under the auspices of the Seabrook ASR Technical Team (ADAMS Accession No. ML14014A378). The questions were documented and shared with the NextEra staff on December 23, 2015 (ADAMS Accession No. ML15357A326). NextEras responses and the inspectors review are planned for the first quarter of 2016 and will be documented in an NRC inspection report. Following the guidance of IMC 0612, Power Reactor Inspection Reports, the inspectors identified an issue of concern regarding NextEras implementation of their Seabrook Structures Monitoring Program. Specifically, the structural evaluations, performed by contractors and accepted by NextEra staff via the FP process, included discussions that identified the potential to exceed limits in the applicable design code (ACI 318-71) for specific locations in the CEB and RHR/CS vault walls. The evaluations further recommended actions to determine whether this was the case. The inspectors noted that the Seabrook staff screened or reviewed these evaluations without documenting a justification in a revision or update to the open PODs for these structures. The additional information requested on December 23, 2015 is required for the inspectors to determine whether this issue involves a performance deficiency. As a result, the NRC opened an unresolved item (URI). The inspectors identified an issue of concern regarding NextEras implementation of the Seabrook structures monitoring program and acceptance of evaluations via the FP and CAP. Additional inspection is warranted to determine whether a performance deficiency exists related to NextEras disposition of FP 100985 for the CEB condition and FP 100903 for the RHR/CS vault condition. Specifically, further inspection is warranted to determine whether NextEra staff properly implemented the Seabrook structures monitoring program for the acceptance and review of structural evaluations potentially impacting the functionality of the CEB and RHR/CS vault, as currently documented in open PODs.
05000443/FIN-2015004-022015Q4GreenLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.55a (g)(4), Inservice Inspection Requirements, requires, in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, must meet the requirements set forth in Section XI of editions and addenda of the ASME Boiler Pressure and Vessel Code. Section XI of the ASME Boiler Pressure and Vessel Code, 2001 Edition with 2003 Addenda, Table IWF-2500-1, Examination Category F-A Supports, requires VT-3 examination of 100 percent of the ASME Class 1 supports, other than piping supports, every ISI Interval (examination item F1.40). Contrary to this requirement, from initial commercial operation until October 14, 2015, (when NextEra staff completed the initial required VT-3 examinations) NextEra did not perform the required ASME Section XI VT-3 examination of ASME Class 1 supports (i.e. seismic support plates and associated load path components) on the CRDM assemblies of Seabrook Unit 1. NextEra staff entered the issue into their CAP as AR 01991880 and completed the VT-3 examinations during the October 2015 refueling outage. The finding is more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the objective to ensure reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Exhibit 2, Mitigating Systems, because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic initiating event. NextEra completed the required examinations on October 14, 2015.
05000443/FIN-2015002-022015Q2GreenP.2NRC identifiedInadequate Characterization of Prompt Operability Determination of the Containment Enclosure BuildingThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because NextEra did not perform an adequate POD of a safety-related plant structure. Specifically, NextEra did not appropriately categorize the operability of the CEB, a safety-related seismic Category I structure, in accordance with EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 19, after identification of a non-conforming condition affecting the structure. NextEra entered the condition into their CAP (AR 02053991), recharacterized the operability of the CEB as Operable but Degraded, and established compensatory measures to monitor for additional structural deformation by performing routine seismic seal gap measurements. This performance deficiency was considered to be more than minor because it affected the design control attribute of the Barrier Integrity cornerstone and its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inspectors determined that the operational capability of the CEB was affected in that compensatory measures were not identified and established to monitor for any further degradation of the non-conforming condition. The finding was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation systems, or heat removal systems. In addition, the affected structures and components remained capable of performing their safety function. The finding is related to the cross-cutting area of Problem Identification and Resolution Evaluation, because NextEra did not thoroughly evaluate an issue to ensure that resolutions address causes and extent of condition commensurate with their safety significance. Specifically, NextEra did not appropriately characterize the CEB non-conforming condition and establish compensatory measures that were commensurate with the safety significance of the condition (P.2).
05000443/FIN-2015002-012015Q2GreenP.1NRC identifiedInadequate Identification of Structural Deformation and Impacts on Associated EquipmentThe inspectors identified a Green NCV of 10 CFR, Appendix B, Criterion XVI, Corrective Action, because NextEra did not ensure that degraded conditions were identified and entered into the corrective action process. Specifically, the inspectors identified multiple instances of material and equipment degradation resulting from deformation of the containment enclosure building (CEB). NextEra entered the condition into their corrective action program (CAP) (AR 02014325) and initiated a root cause evaluation to evaluate the aggregate cause of the non-conforming condition. Additionally, NextEra initiated immediate and prompt operability determinations (PODs), when appropriate, for each of the individually identified material and equipment degraded conditions. This performance deficiency was considered to be more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern if CEB deformation continued to affect plant safety-related structures, systems, and components (SSCs) without appropriate identification and evaluation by NextEra personnel. The finding was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation systems, or heat removal systems. In addition, the structures and components remained capable of performing their safety function. The finding is related to the cross-cutting area of Problem Identification and Resolution Identification, because NextEra did not implement a CAP with a low threshold for identifying issues. Specifically, NextEra failed to identify multiple instances of material and equipment degradation that would have led to the identification of the CEB non-conforming condition (P.1).
05000443/FIN-2015301-012015Q1Severity level IVLicensee-identifiedLicensee-Identified ViolationAccording to 10 CFR 55.49 licensees are required, in part, to not engage in any activity that compromises the integrity of any application, test, or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. Contrary to this requirement, on February 21, 2105, the licensee identified a vulnerability in the protection of the NRC exam simulator scenario files that were being stored on the simulator workstation computer. Although the files were password protected, the licensee determined that exam security could be bypassed due to an error in file designation. This issue was entered into the licensees corrective action program (AR 02027235). This violation is subject to traditional enforcement because of the potential impact upon the regulatory process of issuing licenses to applicants who could have had access to the scenarios prior to the exam administration. This issue meets the criteria for a Severity Level IV violation because it involved a nonwillful compromise of an examination required by 10 CFR Part 55.
05000443/FIN-2014005-032014Q4GreenNRC identifiedFailure to Periodically Calibrate REM-500 Neutron Survey InstrumentsThe inspectors identified a Green NCV of TS 6.7.1.a, Procedures and Programs, because NextEra failed to conduct appropriate periodic calibration of neutron survey instruments. Specifically, since 1996, NextEra assumed that an operability check of certain neutron survey instruments using an internal alpha check source would provide a calibration equivalent to that performed to a traceable neutron source of a known neutron flux, contrary to the periodic calibration frequency requirements specified in the Seabrook Station Radiation Protection Manual. NextEras immediate corrective actions included capturing this issue in its CAP (AR 01969397), calibrating all of the neutron survey instruments in question, and revising the neutron survey instrument operating procedure to require annual calibrations. This performance deficiency was determined to be more than minor because it adversely affected the Occupational Radiation Safety Cornerstone to ensure the adequate protection of the worker from radiation exposure. Additionally, it was similar to example 6.b in IMC 0612, Appendix E, Examples of Minor Issues, which states that the performance deficiency is more than minor if a radiation protection instrument was not calibrated properly, and when recalibrated the as-found condition of the instrument was not within acceptance criteria for calibration and the accuracy was non-conservative. The issue was evaluated in accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," and determined to be of very low safety significance (Green) since it was not an as low as is reasonably achievable (ALARA) issue and did not involve an overexposure or a potential overexposure and it did not affect any significant neutron exposures of plant personnel. The inspectors determined there was no cross-cutting aspect associated with this finding since it was not representative of current NextEra performance. Specifically, in accordance with IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and were not considered representative of present performance.
05000443/FIN-2014005-012014Q4GreenH.8NRC identifiedFailure to Identify and Evaluate Class 1 Structural Conditions Adverse to QualityThe inspectors identified an NCV of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, of very low safety significance because NextEra staff did not promptly identify nine visual indications of structural problems representing conditions adverse to quality. These problems were observed by NextEra staff during a maintenance rule (MR) walkdown of the Fuel Storage Building (FSB) on November 20, 2014, and documented in walkdown notes as conditions warranting entry into the corrective action program (CAP). However these problems were not entered into the CAP to identify them as conditions adverse to quality until questioned by the inspectors. NextEra staff took corrective actions to enter the issues into their CAP in AR10206192, AR02016238, AR02016225 and AR020168863 and initiated AR02014116 for not promptly identifying these problems. This performance deficiency was considered to be more than minor because it is associated with the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and affected the attribute of design control structural integrity. Specifically, the inspectors determined the finding was more than minor because four of the conditions exceeded American Concrete Institute (ACI) 349.3R-96 "Tier II structural criteria, which indicated they require further technical evaluation and analysis to validate the existing conditions or repair to preserve structural function. This issue was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At- Power, Exhibit 3, Barrier Integrity Screening Questions, and screened as very low safety significance (Green) because the observed FSB degradation did not adversely impact structural or radiological barrier functions of the building. This finding is related to the crosscutting area of Human Performance - Procedure Adherence because individuals did not follow CAP process, procedures, and work instructions (H.8).
05000443/FIN-2014005-022014Q4GreenH.6NRC identifiedFailure to Identify and Evaluate FSB Deviation from Design BasisThe inspectors identified a violation of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, of very low safety significance because NextEra did not promptly identify a condition adverse to quality in December 2013 that involved a deviation from expected settling assumptions in the Seabrook Station design basis for the FSB. FSB elevation measurements were received by NextEra staff in December 2010 and in December 2013 indicating that settling at some locations of the FSB was occurring. NextEra staff did not enter this condition, a condition adverse to quality, into their CAP until December 8, 2014, in response to questions from the inspectors. NextEra initiated AR02011698 to enter this issue in the CAP and AR02014116 to address their staff not entering this issue previously into the CAP. This performance deficiency was considered to be more than minor because it is associated with the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and adversely affected the attribute of design control structural integrity. Specifically, the inspectors concluded that the structural integrity of the FSB was potentially adversely affected because measured settling of the structure deviated from assumed design basis values. Also, this condition exceeded the ACI 349.3R-96 Tier II structural criteria of the Structures Monitoring Program and requires a structural evaluation. This issue was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, and screened as very low safety significance (Green) because the observed degradation does not adversely impact structural or radiological barrier functions for the FSB. This finding is related to the cross-cutting area of Human Performance - Design Margins. The organization did not maintain the FSB within design margins and did not utilize the systematic and rigorous corrective action process.
05000443/FIN-2014403-012014Q4NRC identifiedSecurity
05000443/FIN-2014403-022014Q4NRC identifiedSecurity
05000443/FIN-2014403-032014Q4NRC identifiedSecurity
05000443/FIN-2014007-012014Q3GreenNRC identifiedAlternate Safe Shutdown Areas Affected by Smoke from Cable Spreading Room FireThe team identified a finding of very low safety significance, involving a non-cited violation of Seabrook Unit 1 Operating License Condition 2.F for failure to implement and maintain all aspects of the approved Fire Protection Program. Specifically, NextEra failed to ensure that intake air to the A and B remote shutdown panel areas was not contaminated from products of combustion resulting from a cable spreading room fire. NextEra promptly entered this issue into its corrective action program as condition reports AR 01977233 and AR 01982946. NextEra initiated compensatory measures in the form of four-hour roving fire watches. Long term corrective actions include determining options to eliminate the potential for smoke migration from a cable spreading room fire to the A and B essential switchgear rooms. This finding was more than minor because it was associated with the Protection Against External Factors (e.g., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 1, Step 1.6, a Senior Reactor Analyst examined NextEras probabilistic risk analysis based risk evaluation for the issue and determined this finding resulted in an increase in core damage frequency in the mid E-7 range (Green) or very low safety significance. This finding did not have a cross-cutting aspect because it was determined to be a legacy issue and was considered to be not indicative of current licensee performance.
05000443/FIN-2014007-022014Q3GreenNRC identifiedInadequate Alternative Shutdown ProceduresThe team identified a finding of very low safety significance, involving a non-cited violation of Seabrook Unit 1 Operating License Condition 2.F for failure to implement and maintain all aspects of the approved Fire Protection Program. Specifically, NextEra's alternative safe shutdown operating procedures did not adequately establish decay heat removal and could have challenged the performance goals of alternative shutdown, as required by NextEra's safe shutdown analysis and regulatory requirements. NextEra promptly entered this issue into its corrective action program as condition report AR 01976944 and initiated an operating standing order as a compensatory measure. This finding was more than minor because it was associated with the Protection Against External Factors (e.g., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, a Phase 1 evaluation screened this finding as very low safety significance (Green) because it was assigned a low degradation rating. The team determined this issue had a low degradation rating because the procedural deficiencies could be compensated by operator experience and system familiarity. This finding did not have a cross cutting aspect because it was determined to be a legacy issue and was considered to be not indicative of current licensee performance.
05000443/FIN-2014003-032014Q2GreenLicensee-identifiedLicensee-Identified ViolationTechnical Specification (TS) Surveillance Requirement 4.3.1.2 requires verification of the response time of each reactor trip function every 18 months. During the 18 month surveillance testing of the RCP UV channels conducted on April 6, 2014, three of the four RCP UV relays exceeded their allowable maximum response time, resulting in their associated UV reactor trip channels exceeding the limit of 1.5 seconds. NextEra determined that the three channels were inoperable. TS 3.3.1, Reactor Trip System Instrumentation, requires four channels of RCP UV instrumentation to be operable in Mode 1. With three RCP UV channels inoperable in Mode 1, the plant is required to initiate a shutdown within one hour in accordance with TS 3.0.3. NextEra determined that this condition existed from the time the relays were last calibrated in OR15 (September 20, 2012) until the plant entered OR16 (April 1, 2014). Contrary to TS 3.0.3, Seabrook station operated in Mode 1 with three of four RCP UV channels inoperable for approximately 17 months without taking the required TS actions. NextEra entered this issue into the CAP as AR 01964167 and performed a detailed analysis of the impact of the increased channel response time. NextEra, in consultation with Westinghouse, determined that the safety function of the RCP UV trip channel (prevention of departure from nucleate boiling) was maintained throughout the period of inoperability. NextEra planned to develop a maintenance procedure to allow for on-line re-calibration of the RCP UV relays. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations because the deficiency did not affect a single RPS trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators.
05000443/FIN-2014202-012014Q2GreenNRC identifiedSecurity
05000443/FIN-2014003-022014Q2GreenH.5Self-revealingUnexpected Main Generator Breaker Pole Closure Results in Reactor TripThe inspectors identified a self-revealing finding of very low safety significance (Green), because NextEra did not ensure that adequate procedural guidance existed in ON1046.12, Operation of the Main Generator Breaker to limit the likelihood of events that upset plant stability. Specifically, Seabrook station experienced an automatic reactor trip from approximately 15 percent reactor power on April 1, 2014 when two of four reactor coolant pumps (RCPs) tripped on low bus voltage. The cause of the reactor trip was directly attributable to the main generator breaker inadvertently closing and actuating the main generator multi-function protective relay. NextEra entered the event into their CAP, and conducted a root cause evaluation to determine the root and contributing causes, extent of condition and extent of cause, and to identify corrective actions to prevent recurrence. NextEra initiated actions to revise ON1046.12 to add controls regarding the potential risk associated with placing the main generator breaker control in local, conducted briefings with Maintenance groups involved in the event, and evaluated the adequacy of other Operations procedures that place equipment in a configuration where protective features are bypassed or defeated. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone, and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was evaluated under IMC 0609, Attachment 4, Phase 1 Initial Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it did not result in both a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of Human Performance - Work Management, because NextEra did not ensure that a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority was implemented. Specifically, ON1046.12, Operation of the Main Generator Breaker did not contain adequate procedural guidance regarding the impacts of positioning the Main Generator Selector Switch to local, take mitigating actions, and minimize time spent at increased risk configurations (H.5).
05000443/FIN-2014003-012014Q2GreenH.8NRC identifiedInadequate Technical Evaluation of Safety-Related StructuresThe inspectors identified a finding of very low safety significance (Green) because NextEra did not perform adequate evaluations of safety-related residual heat removal (RHR) vaults. Specifically, additional technical evaluation and analysis was not adequately conducted on the safety-related A and B RHR concrete vaults when it was determined that they exceeded the quantitative limits specified in NextEra procedures. NextEra entered the failure to perform adequate technical evaluations on concrete structures exceeding American Concrete Institute (ACI) Tier II quantitative requirements into the CAP (action request (AR) 01929460), and planned to perform a formal technical evaluation of the A and B RHR vault conditions in accordance with their structural monitoring program procedure and the ACI 349.3R-96 standard. The performance deficiency was considered to be more than minor because it affected the protection against external factors attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors concluded that the reliability of the structures was affected in that they exceeded the specified Tier II limits without the performance of further technical evaluations. The issue was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined to be of very low safety significance (Green) because it did not represent an actual loss of function of at least a single train for greater than its Technical Specification Allowed Outage Time or two separate safety systems out-of-service for greater than it Technical Specification Allowed Outage Time. This finding is related to the cross-cutting area of Human Performance Procedure Adherence, because NextEra did not follow processes, procedures, and work instructions. Specifically, NextEra personnel did not perform an adequate technical evaluation of two safety-related concrete structures that exceeded the quantitative criteria requiring such an evaluation (H.8).
05000443/FIN-2014008-012014Q1Severity level IVNRC identifiedSecurity
05000443/FIN-2014002-012014Q1GreenP.2NRC identifiedScaffolding Installed with Insufficient Separation to Safety Related EquipmentThe inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, Procedures, because NextEra did not ensure adequate separation was maintained between temporary scaffolding and safety-related equipment. Specifically, six instances of scaffolding installed in the plant were identified with less than the minimum standoff distance to safety-related equipment specified in NextEra procedures and no corresponding engineering evaluation to support these deviations. NextEra entered this NCV into their CAP as AR 01933827 and assessed the six deviations for any impact on the associated safety-related systems. This performance deficiency was considered more than minor because it affected the protection against external factors attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, NextEra did not evaluate scaffolding installations when insufficient separation to safety-related equipment existed after procedural requirements were revised to a more restrictive value. Additionally, it was similar to example 4.a in IMC 0612, Appendix E, Examples of Minor Issues, which states that the issue of failing to appropriately evaluate scaffold installation as required by procedures is more than minor if the licensee routinely failed to perform engineering evaluations. The issue was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power and determined to be of very low safety significance (Green), because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because NextEra personnel did not perform an adequate extent of condition review after revision of their erection of scaffold procedure. This performance deficiency directly contributed to multiple instances of scaffold members erected within two inches of safety-related equipment without an engineering evaluation (P.2).
05000443/FIN-2013005-012013Q4GreenLicensee-identifiedLicensee-Identified Violation10 CFR Part 50.65, paragraph a(4), Requirement for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, states, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. NextEra procedure WM 10.1. On-Line Maintenance, Section 3.3.1, requires that an evaluation of the risk impact of planned maintenance tasks be performed. Contrary to the above, on September 24, 2012, NextEra failed to adequately assess and manage the impact to plant risk during a planned maintenance activity. Specifically, NextEra identified during internal reviews that they had failed to recognize an elevated online maintenance risk level (Yellow) during the performance of the 1-EDE-B-1-B Battery Service Test due to incorrect coding in NextEras PRAX risk model program. The inspectors determined NextEras failure to assess and manage risk during the period when the Battery Service Test was reasonably within NextEras ability to foresee and correct, and was identified as a performance deficiency. This performance deficiency is more than minor, and considered a finding, because it is associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this finding represents a violation of 10 CFR Part 50.65 Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Section a(4), the inspectors used IMC 0609, Appendix K, Flowchart 1 Assessment of Risk Deficit, to analyze the finding. The regional Senior Reactor Analyst determined the incremental core damage probability (ICDP) for the surveillance period (~5-10 minutes) to be several orders of magnitude below the 1E-6 threshold due to the short duration of the systems unavailability. As this finding is not related to Risk Management Actions only, and the ICDP Risk Deficit is not >1E-6, the inspectors determined that the finding is of very low safety significance (Green). The issue was entered into NextEras CAP as AR 1906782.
05000443/FIN-2013004-012013Q3GreenH.14NRC identifiedInadequate Operability Determination Regarding Service Water Leakage and Associated TS ViolationThe inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.4, because NextEra did not follow the requirements of station procedure EN-AA-203-1001, Operability Determinations/ Functionality Assessments. Specifically, NextEra did not properly evaluate and document an adequate basis for operability, when relevant information was available that would have challenged the reasonable expectation of operability threshold for a service water (SW) through-wall leak that degraded incrementally from weepage on August 7, 2013, to a significantly larger leak on August 28, 2013. NextEra completed a temporary non-code repair of the flaw with the installation of a weldolet on September 1, 2013, following NRC review and approval of a relief request. Additionally, under the corrective action process, NextEra completed apparent cause evaluations for the piping flaw, as well as engineering decision-making during the non-destructive examinations and evaluations, and are currently evaluating the fundamental issue of decision-making regarding TS operability and TS compliance. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the prompt operability determination incorrectly concluded the B cooling tower (CT) SW header and the B SW (ocean) pumps were operable, but degraded, versus inoperable. IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 4, External Events Screening Questions, were used to assess this issue and a detailed risk evaluation was completed. The inspectors assumed that functionality of the SW system, based upon the as-found wall thinning, would only be challenged when aligned to the cooling tower basin when the SW piping is subjected to a higher overall sytem pressure. This system configuration is used to mitigate a seismic event following the loss of the normal SW intake structure. Based on low probability of SW piping system failure due to a seismic event and the overall low likelihood of a seismic event of a magnitude sufficient to cause structure, system, and component (SSC) damage, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the decision making component because NextEra failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action. Specifically, NextEra personnel had not considered relevant information in the form of UT data and actual leak propagation to conclude that they no longer had reasonable assurance of operability and did not declare the B header of ocean and CT SW systems inoperable.
05000443/FIN-2013003-012013Q2NRC identifiedTech Spec 6.10.1 Violation - Contractor Electrician Entered High Radiation Area Without Receiving Health Physics BriefingNRC Letter, dated June 1, 2012 (ML12153A155), documented an NRC Office of Investigation (OI) review to determine whether a contractor electrician deliberately entered a high radiation area (HRA) without first receiving a health physics (HP) briefing on the current radiological conditions in accordance with site procedures required by NextEras operating license (NRC Investigation Report Number 1-2011-038). The NRC concluded that the contractor electrician, who had been assigned to conduct work within an HRA, deliberately entered the HRA without first receiving the HP briefing on the current radiological conditions. That issue was being treated as an NCV. In order to facilitate entering this issue into the NRCs Plant Issues Matrix and assessment process, this issue was identified as NCV 05000443/2013003-01, Tech Spec 6.10.1 Violation - Contractor Electrician Entered High Radiation Area Without Receiving Health Physics Briefing.
05000443/FIN-2013008-042013Q2GreenH.7NRC identifiedPrimary Component Cooling Water System Unavailable Following a Seismic EventThe team identified a finding of very low safety significance involving an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that NextEra did not verify the design basis for the primary component cooling water (PCCW) system had been translated into specifications and procedures. Specifically, the team found that NextEra had produced engineering evaluations and maintenance procedures that allowed a limited amount of leakage past the B train PCCW isolation valves. The team noted NextEra used these documents to conclude that a 2.5 gallons per minute (gpm) leak rate identified in April 2011 and a 4 gpm leak identified in October 2012 on B train valves were acceptable. The team reviewed the design and licensing basis of the B train and determined the system did not have a safety related refill capability and, therefore, was required to be leak tight. The team determined that, with leakage past the isolation valves, water would need to be added to the system and concluded that following certain design basis events a safety related refill system would not be available resulting in loss of the PCCW system. Following identification of the issue NextEra entered it into their corrective action program and evaluated the operability of system, concluding the PCCW system was operable based on recent valve-leakage testing results. The team review of the evaluation determined it to be reasonable. The finding is more than minor because it is associated with the protection against external factors (seismic event) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding involved the loss or degradation of equipment designed to mitigate a seismic initiating event and resulted in a DRE in accordance with IMC 0609, Appendix A, Exhibit 4. Based upon the DRE, the finding was determined to be of very low safety significance. The team determined that this finding has a cross-cutting aspect in the area of Human Performance, Resources, because NextEra did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, engineering evaluations and maintenance procedures associated with PCCW isolation valves did not align with the design and licensing basis requirements for a leak tight system.
05000443/FIN-2013008-012013Q2GreenNRC identifiedFailure to Verify Adequate Fault Protection for Safety Related Equipment from NON-SAFETY Related Load FaultThe team identified a finding of very low safety significance involving a non-cited violation (NCV) of the 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, NextEra did not appropriately select and review, for suitability of application, a safety related over-current protection device for a safety related power panel (EDE-PP01B). Specifically, NextEra did not consider the effects of the current-limiter function of safety related inverters, which supplied the safety related power panel and would limit fault current at the over-current protection device. As a result, the safety related over-current protective devices would not have prevented a postulated fault of a non-safety related load, supplied from the safety related power panel, from causing a momentary loss of voltage to the power panel and all associated safety related loads. In response, NextEra entered the issue into their corrective action program and performed a preliminary analysis that determined an existing non-safety related fuse would provide adequate over-current protection. NextEra credited the use of this fuse as an interim compensatory measure in their operability assessment in order to conclude the system was operable. The team determined the analysis and associated assessment were reasonable. The finding was more than minor because it was similar to Example 3.j of NRC IMC 0612, Appendix E, and was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance because the issue was a design or qualification deficiency that did not result in inoperability of the system. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current performance.
05000443/FIN-2013008-022013Q2GreenNRC identifiedCondensate Storage Tank Water Level Above Limits of Seismic QualificationThe team identified a finding of very low safety significance involving an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that NextEra did not assure the seismic design requirements for the condensate storage tank (CST) were translated into specifications and procedures. Specifically, the team found that NextEras seismic design calculations for the CST was based, in part, on a maximum tank level. The maximum tank level was used to ensure that the floating cover inside the CST would not strike the top of the tank. NextEra engineers had concluded that this impact could cause a failure of the CST or cover during a seismic event. However, the team identified that the high level alarm and operating procedure limits for the tank were above the level credited in the calculation. Additionally, the team determined that NextEra routinely operated the CST tank above the maximum level assumed in the calculation. Following identification NextEra entered the issue into their corrective action program and proceduralized a lower maximum allowable water level for the CST to prevent a seismically induced impact of the floating cover on the tank. The finding is more than minor because it is associated with the protection against external factors (seismic event) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The finding involved the loss or degradation of equipment so a detailed risk evaluation (DRE) was performed. Based upon the DRE, the finding was determined to be of very low safety significance. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current performance.
05000443/FIN-2013008-032013Q2GreenH.13NRC identifiedFailure to Perform Preventative Maintenance on the Supplemental Emergency Power SystemThe team identified a finding of very low safety significance, in that NextEra did not perform preventative maintenance (PM) on supplemental emergency power system (SEPS) electrical components as required by the approved engineering design modification for SEPS. As a result, the system\'s reliability to respond to a loss of off-site power event had not been maintained at a high confidence level, as assumed in NextEra\'s design and probabilistic risk analyses. In response, NextEra entered the issue into their corrective action program, evaluated the effect on equipment reliability for the never-performed PMs, and implemented an accelerated schedule to complete the missed PM tasks. The finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. In addition, the finding was associated with the Procedure Quality and Equipment Performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. The team determined the finding was of very low safety significance because it was not a design or qualification deficiency and did not result in the loss of the SEPS system or train function. This finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because the most recent quarterly system health report (4th quarter 2012) had stated that SEPS PMs had not been scheduled or performed, these reports had been reviewed by NextEra management, however actions were not taken to develop electrical PMs on the SEPS.
05000443/FIN-2013002-022013Q1GreenH.14Self-revealingFailure to Evaluate Service Water Cooling Tower LevelA self-revealing NCV of technical specification (TS) 3.7.4 Service Water System/Ultimate Heat Sink, resulted from operators failure to follow procedures to evaluate a faulty SW cooling tower basin level instrument. Specifically, because NextEra personnel did not properly follow their Conduct of Operations procedure and the Operations Management Manual, an inaccurate level gage was used to determine SW cooling tower basin level. This resulted in the SW cooling tower basin level dropping and remaining below its TS minimum value for approximately 17 days. NextEras immediate corrective actions included conducting a fast fill of the cooling tower basin via the fire protection system to restore operability on December 7, 2012, and entering the issue into their CAP as CR 1830734. Planned corrective actions included implementing a process for operations department oral boards to focus on standards applications, fundamentals, and use of situational questions. This performance deficiency is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the SW cooling tower basin level was below its TS minimum level of 42.15 feet for 17 days. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations (IMC 0609A). The inspectors determined that the finding was of very low safety significance (Green) because the deficiency did not affect the design or qualification of the SW system and it did not represent a loss of system safety function. Although the finding did involve the degradation of equipment specifically designed to mitigate a seismic initiating event, the SW cooling tower had sufficient margin available to satisfy its design basis requirements and safety function. This finding has a crosscutting aspect in the area of Human Performance, Decision Making, because NextEra did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed, rather than a requirement to demonstrate that it is unsafe in order to disapprove the action. Specifically, NextEra failed to properly evaluate which SW cooling tower level gage was inoperable and thus relied on an inoperable indication for SW cooling tower level.
05000443/FIN-2013002-012013Q1GreenH.12Self-revealingLoss of DC Control Power to Switchyard #2A self-revealing finding of very low safety significance was identified for failure to follow procedures associated with switchyard maintenance activities on January 24, 2013. Specifically, in preparation for the planned maintenance on switchyard battery (SYB) #3, operators incorrectly performed NextEra procedure ON1048.07, Switchyard Battery Operation, which led to a loss of power on switchyard system (SYS) #2, disabled the SYS#2 breaker automatic closure feature, and increased the risk of a loss of offsite power. Corrective action was subsequently taken to secure the maintenance on SYB#3, and return it and the battery charger to service to supply loads to both Switchyard System #1 (SYS#1) and SYS#2. NextEra entered this issue into their corrective action program (CAP) as condition report (CR) 1841980. This performance deficiency is more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone, and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. Specifically, not properly performing NextEra procedure ON1048.07 resulted in the loss of the SYS#2 breaker automatic closure feature, thereby increasing the risk of an initiating event due to a loss of off-site power. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations (IMC 0609A). The inspectors determined that the finding was of very low safety significance (Green) because the deficiency did not cause a reactor trip, and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, because NextEra personnel did not utilize human error prevention techniques commensurate with the risk of the assigned task, such that work activities were performed safely. Specifically, NextEra personnel did not verify that the switchyard battery charger switch manipulation would result in the appropriate system response.
05000443/FIN-2012005-012012Q4GreenH.6Self-revealingFailure to Correct a Condition Adverse to Quality for the L-5 Fici ConnectionA self revealing, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified because the high pressure Swagelok fitting for the L-5 fixed in-core detection instrument failed and caused an unisolable reactor coolant leak. Specifically, NextEra did not implement timely and effective corrective actions to address a degraded Swagelok fitting associated with the L5 in-core instrument connection that was identified as a condition adverse to quality in 2006. The inspectors determined that not taking timely and effective corrective action to correct a condition adverse to quality was a performance deficiency. The inadequate corrective actions led to the failure of the L-5 fixed in-core detection instrument Swagelok fitting on October 21, 2012. The inspectors determined that this issue was within NextEras ability to foresee and correct, because this fitting was identified as leaking during previous operating cycles, was assigned additional monitoring and the adverse trend of increased leakage at L-5 at low pressures continued from the time it was identified in 2006. NextEra entered this into their corrective action program as AR 01815351 and implemented immediate corrective actions to cut the connection for the L-5 instrument, as well as two others showing signs of leakage, and capped the tubes prior to recommencing start-up. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally it is similar to example 4.d of Inspection Manual (IMC) 0612, Appendix E, because this was a failure to implement a corrective action that did have a safety impact, because the fitting failed and caused a 4 gpm non-isolable leak from the reactor coolant system. The inspectors evaluated the finding using IMC 0609, Attachment A, because the operational impact occurred after the residual heat removal pump was secured for start-up. The inspectors determined that the finding was of very low safety significance (Green) because the deficiency would not result in exceeding the small loss of coolant accident (LOCA) leak rate and would not have affected other systems used to mitigate a LOCA. This finding has a cross-cutting aspect in the area of Human Performance, Resources, because actions were not taken to maintain long term plant safety by minimization of long standing equipment issues. Specifically, NextEra did not manage the ongoing degradation of the L-5 in-core instrument connection fitting connection while long term corrective actions were implemented.
05000443/FIN-2012005-022012Q4GreenH.8Self-revealingFailure to Adequately Implement Procedure Led to Reactor Coolant System Leakage from Pressurizer Safety Valve FlangeA self-revealing, non-cited violation of technical specification 6.7.1, Procedures and Programs, was identified after the control room received a high discharge temperature alarm for pressurizer relief valve RC-V-116 while pressurizing the reactor coolant system during start-up preparations on October 21, 2012. Specifically, NextEra personnel did not properly implement maintenance procedure MS0519.17, Crosby Pressurizer Mechanical Safety Valve Removal and Installation. This led to the reactor coolant system leakage past the RC-V-116 flange gasket that caused the high discharge temperature alarm. The inspectors determined that not properly implementing procedure MS0519.17 was a performance deficiency that was within NextEras ability to foresee and correct. NextEra entered this into their corrective action program as AR1815307 and implemented immediate corrective actions to retorque the bolts and replace the gasket on RC-V-116. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, because NextEra personnel did not properly implement procedure MS0519.17, eight bolts on the inlet flange of pressurizer RC-V-116 were not adequately torqued. This resulted in reactor coolant system leakage during preparations for reactor start-up on October 21, 2012, and required NextEra operators to return the plant to cold shutdown. Additionally, this was similar to more-than-minor example 2.e in IMC 0612, Appendix E, because the procedure non-compliance resulted in a negative safety consequence in that it impacted the ability of the flange to perform its function to prevent reactor coolant system leakage. The inspectors evaluated the finding using IMC 0609, Attachment A, because the operational impact occurred after the residual heat removal pump was secured for start-up. The inspectors determined that the finding was of very low safety significance (Green) because the deficiency would not result in exceeding the small loss of coolant accident (LOCA) leak rate and would not have affected other systems used to mitigate a LOCA. This finding has a cross-cutting aspect in the area of Human Performance, work practices, because personnel did not follow the procedures. Specifically, when tensioning the bolts on the pressurizer relief valve RC-V-116 inlet flange, NextEra personnel did not verify there was a gap for eight of the twelve bolts on the inlet flange of the valve as required by maintenance procedure MS0519.17
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