Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000443/FIN-2012002-012012Q1SeabrookLicensee-Identified Violation10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that that conditions adverse to quality are promptly identified and corrected. Contrary to the above, NextEra did not assure that a previously identified hold down clamp and missing chaffing material on the B emergency diesel generator (EDG) fuel oil return was promptly corrected. Specifically, as part of an October 2008 extent of condition for a similar condition discovered on the A EDG, the missing clamp and chaffing material was planned to be installed but was not. In November 2011, during a surveillance test of the B EDG, a leak occurred due to vibration induced fretting of the fuel oil return line. Though NextEras engineering and maintenance walkdown procedure provided direction intended to identify the missing clamp and chaffing material, NextEra determined that the direction was ineffective in identifying the condition adverse to quality on the B EDG. This performance deficiency was identified in the corrective action program as Condition Report 1710841. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent an actual loss of safety function or contribute to external event core damage sequences. Since the issue was of very low safety significance and was entered into the corrective action program it is considered a licensee-identified, noncited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000443/FIN-2012002-022012Q1SeabrookLicensee-Identified ViolationTechnical Specification 3.8.1.1, A.C. Sources, requires as a minimum, the following A.C. electrical power sources shall be operable: (a) two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System, and (b) two separate and independent diesel generators. TS 3.8.1.1 Action a. requires that with one of the required offsite AC electrical power sources and one of the required independent diesel generator power sources inoperable, operators must demonstrate the operability of the remaining A.C. source by performing TS surveillance requirement (SR) 4.8.1.1.1a. within 1 hour and at least once per 8 hours thereafter. Contrary to the above on October 31, 2011, for 110 minutes, and from January 10 to 17, 2011, when both the A EDG and one of the required offsite power sources were inoperable, NextEra did not perform TS SR 4.8.1.1.1a. within 1 hour and at least once per 8 hours thereafter. In both instances, the offsite AC source via the RAT was not declared inoperable and the applicable TS action was not entered because NextEra did not recognize the impact of the EDG operation on the fast transfer feature in the TS Bases change process. Specifically, NextEra did not ensure appropriate technical evaluations were performed to review change implications against all normal plant configurations. This finding is of very low safety significance (Green) per IMC 0609 because the issue did not result in a total loss of safety function and did not contribute to both a transient initiator and the likelihood that mitigating functions would be unavailable. Specifically a fast transfer would occur following an opening of the EDG breaker if the bus and RAT were in synchronism. If the bus and RAT were not in synchronism, the RAT breaker would close when residual bus voltage relays actuated. Since the issue is of very low safety significance and was entered into the corrective action program as AR 1718306, the issue is considered a licensee-identified, non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000443/FIN-2011010-012011Q4SeabrookAdequacy of Corrective Actions Associated with Calculation Methods for Alkali-Silica Reaction IssueThe NRC staff noted that the methods used in evaluating structural integrity for the selected buildings were based on the correct design basis code ACI318-1971. However, the mathematical relationships in this code were based on empirical data, from testing of non-degraded concrete, for determining key ratios that are a part of the design bases and used for determining tensile and shear strength or capacity in addition to compressive strength. These strength values were important in the building loading analysis during normal or upset conditions such as for seismic events. More importantly, while some testing for the modulus of elasticity was done, it was not clear if the plans would result in additional testing of concrete cores for this parameter or any independent testing associated with other key design parameters such as Poisson\\\'s Ratio, shear modulus, or bulk modulus. With these parameters known, various strengths or capacities can be determined such as for tensile and shear strength. In addition, the plans that the inspectors reviewed did not address variation in mechanical properties of the concrete in different directions due to ASR cracking nor the effect of the ASR expansion on stresses in the rebar. These parameters were important in order to ensure that the current licensing and design basis was maintained. The licensee representatives agreed to address the assumptions or establish relationships for the current conditions at Seabrook. Accordingly this area is unresolved pending completion of license actions as noted above and further NRC staff review.
05000443/FIN-2011005-012011Q4SeabrookReactor Trip Caused by Inadequate Condensate Pump RestorationA self-revealing finding was identified regarding the improper restoration of a condensate pump that resulted in a reactor trip. NextEra workers aligned the B condensate pump for service following maintenance without first venting air from the pump casing in accordance with the system operating procedure. The finding is greater than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone, and because it adversely affects the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors conducted a Phase 1 SDP screening in accordance with IMC 0609 and determined that the finding is of very low safety significance. The finding has a cross-cutting aspect in the area of human performance because NextEra did not ensure that adequate procedures and work packages were available (H.2.c). Specifically, neither the work package nor tagout used to restore the condensate pump to service vented the pump casing, and as a result, air from the pump entered the condensate-feedwater train causing a reactor trip when the A main feedwater pump tripped on low suction pressure.
05000443/FIN-2011005-022011Q4SeabrookLicensee-Identified Violation10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, NextEra did not assure that the design basis for safety-related service water piping installed per design change record 93DCR003 was correctly translated into procedures and instructions. Specifically, when implementing design change 93DCR003, NextEra did not establish a requirement in plant procedures to track the service life of the plastisol liner or establish measures during preventive maintenance activities to assure the material remained bonded to the pipe. The plastisol liner material was found delaminated and generating foreign material in the service water pipe providing cooling water to the B emergency diesel generator on October 10, 2011. This was identified in the corrective action program as Condition Report 1694951 to initiate review of the service water monitoring program, revise the design change process and take other long-term corrective actions. This finding is of very low safety significance (Green) because it did not represent an actual loss of safety function or contribute to external event core damage sequences.
05000220/FIN-2011004-012011Q3Nine Mile PointInadequate Actions to Prevent Vibration Induced Failure on a Socket Weld for a Vent Line on the lA\' FWP Minimum Flow LineA Green self revealing finding was identified for inadequate implementation of corrective actions regarding vibration induced failures of socket welds. This finding resulted in an August 11, 2011, Nine Mile Unit 2 scram due to a failed socket weld on the vent line for the \'A\' feedwater pump (FWP) minimum flow line. NMPNS did not properly consider the impact of high vibration levels on a vent line attached to the \'A\' FWP mini-flow recirculation line. NMPNS corrective actions included upgrading the socket weld to the requirements outlined in industry operating experience (OE). The inspectors determined that the finding was of very low safety significance (Green) through performance of a Phase 1 SOP in accordance with IMC 0609.04, Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems (MS) and Barrier Integrity Cornerstones. Specifically, the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding has a cross-cutting aspect in the area of problem identification and resolution in that NMPNS did not implement and institutionalize OE through changes to station processes, procedures, equipment and training programs. Specifically in 1998 and again in 2010, NMPNS did not institutionalize external and internal OE to reduce the probability of a socket weld failure.
05000443/FIN-2011004-012011Q3SeabrookInadequate Functionality Assessment for Fire Protection SystemThe inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6. 7.1.a that requires that written procedures be established and implemented, including administrative procedures that define authorities and responsibilities for safe operation. Specifically, NextEra identified a degraded condition in the fire protection system on July 15, 2011, but did not properly or thoroughly evaluate the fire protection system performance as required by NextEra procedure EN-AA-203-1001. As corrective action, NextEra completed an operability evaluation that identified degraded fire protection system performance under certain operating conditions for which NextEra implemented administrative controls that would prevent the degraded performance. The performance deficiency was more than minor because a reasonable doubt of operability existed until further engineering evaluations were completed to demonstrate adequate fire system performance under design basis conditions. The finding affected the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage. The issue was evaluated using Appendix F of IMC 0609, Significance Determination Process (SOP), and was determined to be of very low safety significance (Green) because the finding had minimal impact on fire system performance. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1 (c), because NextEra personnel did not adequately implement the operability determination process to ensure that fire system performance was thoroughly evaluated for operability to assure timely and appropriate corrective actions were completed.
05000443/FIN-2011004-022011Q3SeabrookLicensee-Identified ViolationTS LCO 3.4.6.1, RCS Leakage Detection Systems, requires three leakage detection systems be operable, including a containment sump level monitoring system, a containment atmosphere particulate radiation monitoring system and a containment atmosphere gaseous radiation monitoring system. The TS allow plant operation for up to thirty days with one leakage detection system inoperable, and requires a plant shutdown in 6 hours if more than one leakage detection system is inoperable. Contrary to the above, Seabrook operated for greater than 6 hours on October 5, 2010, December 15, 2010, January 4, 2011 and March 10, 2011, with both particulate and gaseous radiation monitors inoperable. On each occasion, RM6548 was credited for RCS leakage detection for more than 6 but less than 24 hours. The finding affected the Initiating Events cornerstone in that a system used to identify reactor coolant system leakage might not have been functional following a operational basis earthquake. The backup gas monitor remained functional but lacked full qualification, as described in Section 40A3 above. This finding is of very low safety significance (Green) per IMC 0609 because the issue did not result in a total loss of safety function and did not contribute to both a transient initiator and the likelihood that mitigating functions would be unavailable. Since the issue is of very low safety significance and was entered into the corrective action program as AR 1633042, the issue is considered a licenseeidentified, non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000443/FIN-2011003-012011Q2SeabrookInadequate Control of Combustible Materials.The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6.7.1h, which requires that written procedures be established and implemented for the fire protection program. Contrary to TS 6.7.1.f , the inspectors identified combustible materials which were not controlled per fire protection procedure FP 2.2, Revision 12\' Specifically, (i) combustible materials were stored within three feet of an energized sample panel in the primary auxiliary building room PB404, a PRA risk significant area; and, (ii) combustible materials in excess of the permissible amounts were stored in waste process building area W8505. The in$pectors identified materials stored in WB505 in excess of FP 2.2 limits on three occasions. Collectively, the NRC observations indicate a weakness in the programmatic control of combustible materials despite the fact that in each case the combustible materials were promptly removed following identification by the inspector. Seabrook entered this performance deficiency into their corrective action program. The performance deficiency was more than minor because, if left uncorrected, inadequate control of combustibles could affect the Mitigating Systems cornerstone objective to assure external factors (fires) do not impact the availability and reliability of syitems which mitigate events. The inspectors assessed the finding using Appendix F of the Significance Determination Process (SDP) Based on a degradation rating of low, which screens to Green in the fire protection SDP, the finding is of very low safety significance. This finding has a cross-cutting aspect in Human Performance, Work Piactices tH.4(b)l because Seabrook personnel did not follow procedures for the control of transient combustibles.
05000443/FIN-2011003-042011Q2SeabrookUntimely Operability Determination for Degraded Concrete Structures Housing Safety-Related Equipment.The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6.7.1.a that requires written procedures be established and implemented, including administrative procedures that define authorities and responsibilities for safe operation with respect to operability determinations. Contrary to TS 6.7.1.a, NextEra identified a degraded and nonconforming condition related to reduced modulus of elasticity for buiidings housing safety related equipment on May 27,2011 but did not complete an operability determination until EC250348 was issued on June 28,2011 (AR1664399). The delayed entry of the issue into the corrective action process to assess operability was contrary to Section 4.3 of EN-AA-203-1001 that requires operability assessments be completed in a time frame commensurate with the safety significance of the issue (within 8 hours). Seabrook subsequently completed an evaluation of the concrete issues and determined that the buildings housing safety-related equipment remained operable. Seabrook entered this performance deficiency into their corrective action program. The performance deficiency was more than minor because a reasonable doubt of operability for the affected concrete structure$ existed until further engineering evaluations were completed to demonstrate the structures and systems that they housed would remain functional under design and licensing basis conditions. The finding affected the Mitigating Systems cornerstone Objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage. The issue was evaluated using IMC 0609, Significance Determination Process (SDP), and was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of a barrier function, and was not potentially risk significant for external events. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1(a), because NextEra did not enter identified degraded concrete conditions for several site buildings into the corrective actions process in a timely manner, which would have ensured the shift manager completed timely operability evaluations for the affected structures.
05000443/FIN-2011003-052011Q2SeabrookInadequate Operability Determination for Reduced EDG HX Cooling Water Flow.The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6.7.1.a that requires that written procedures be established and implemented, including administrative procedures that define authorities and responsibilities for safe operation with respect to operability determinations. Contrary to TS 6.7 .1.a, NextEra identified a degraded condition related to seryice water flow to the B emergency diesel generator (EDG) heat exchanger (HX) on June 28,2011 but did not fully evaluate the reduced flow under all plant conditions as required by NextEra procedure EN-AA-203- 1001. Fouling of the heat exchanger tubes was subsequently identified and mitigated. Seabrook also completed an evaluation of the B EDG service water flow issues and determined that the EDG remained operable. Seabrook entered this performance deficiency into their corrective action program. The performance deficiency was more than rninor because a reasonable doubt of operability existed untilfurther engineering evaluations were completed to demonstrate adequate service water flow to the B EDG HX existed and the B EDG remained functional under design and licensing basis conditions. The finding affected the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage. The issue was evaluated using IMC 0609, significance Determination Process (SDP), and was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of a barrier function, and was not potentially risk significant for external events. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1(c), because NextEra personnel did not thoroughly assess EDG operability to assure reduced HX SW flow was acceptable under all operating conditions, or assure appropriate corrective actions were timely completed.
05000443/FIN-2011003-022011Q2Seabrooklnadequate 50.59 Screening for Design Change EC 272057.NextEra issued EC272057, Concrete Modulus of Elasticity Evaluation, on April 24, 2011 to address the results of testing that showed a reduction in the concrete modulus of elasticity in the CEB (AR 1644074). EC272057 also address the reduced modulus in the Control Building/Electric Tunnel CB/ET (AR581434). The lowest measured modulus was 2.16E+03 ksifor the CEB and 2.1E+03 ksi for the CB/ET, both less than the design value of 3.62E+03 ksi. EC272057 was supported by calculations C-S-1-10150 and C-S-1-10156 which reflected the degraded conditions in the design calculations CD-20-CALC and CE-4-CALC for the control and containment enclosure buildings, respectively. NextEra concluded the structures remained operable and used EC272057 to disposition the degraded condition as use-as-is, by incorporating the degraded condition into the design basis. In a safety evaluation screen per 10 CFR 50.59 for EC272Q57, NextEra concluded the change to the facility did not require a complete evaluation per 50.59(cX2) because adequate design margin existed and there was no adverse affect on an UFSAR described design function. The inspectors determined the 50.59 Screen for EC272057 did not correctly address Screen Question 5.a: Does the proposed activity involve a change to an SSC that adversely affects an IJFSAR design function? Using the guidance of the Seabrook 1OCFR5059 Resource Manual and NEI 96-07, Revision 1, the inspectors determined that a 50.59 evaluation is specified for changes that adversely affect design function. ln this situation, the ASR impacted concrete with reduced modulus of elasticity which reduces the flexural capacity of the walls would be an adverse effect. Therefore, NextEra should have evaluated the change to the facility per 10 CFR 50.59(cX2). The item is unresolved pending action by NextEra to complete a full 50.59 evaluation for EC272057 and subsequent NRC review of that evaluation to determine whether the performance deficiency is more than minor.
05000443/FIN-2011003-032011Q2SeabrookOperability Evaluation for Degraded Concrete in ASR Affected Plant Structures.NextEra\\\'s analysis of the CEB samples found that the concrete has acceptable compressive strength and reduced but acceptable modulus of elasticity. To evaluate the effects of the reduced modulus, NextEra assessed the increase in strain for CEB building elements and found that the strain at the most limiting element remained less than the American Concrete Institute ACI-318 design stress limit and thus was acceptable. NextEra evaluated the impact on flexural capacity by reviewing the change in bending moment of structural elements. The reduced modulus causes the concrete to have increased flexure which has the effect of shifting the balance point in how load is transferred between the concrete and the imbedded steel (rebar). The reduced modulus causes a shift toward the reinforced steel in tension. The resultant change in bending moment was evaluated to show that the reduction in capacity was minimal and the stresses on the steel and concrete remain below the design stress limits with margin. NextEra\\\'s evaluation of the condition concluded that a change in the dynamic seismic response of the structure would be minor, and the CEB remains capable of performing its design function. The prompt operability determination for the CEB (ARs 1644074 and 1664399) evaluated how the reduced modulus would affect the structure by analysis of locally impacted sections. The evaluation did not address the effects of reduced modulus on the changes to the natural frequencies of the structure and the global response of the structurelo seismic loads. The inspectors requested further information on the effects of the reduced modulus on stresses and strain in the concrete and rebar system for which NextEra will complete additional analyses. The prompt operability determination (POD) for the Control Building (AR581434) as well as for other ASR impacted structures (AR1664399) evaluated the effects of reduced modulus on portions of the below grade structures and the components housed within them. The evaluations lacked details to explain the effects of the reduced modulus on structuralflexure as related to components attached to the structures, such as pipe supports and cable trays. Similarly, the evaluations lacked details with regard to the structure\\\'s response to seismic events as related to structure rigidity and changes in the natural frequency, and the bases to use the ground response spectra. Further, the evaluations lacked details to explain how the function of support anchor bolts would not be adversely impacted by reduced concrete compressive strength in the CBIET. This item is unresolved pending further NRC review of the above issues, action by NextEra to complete additional analysis of the CEB conditions and subsequent NRC Region I review of that analysis, and the completion of reviews by the NRC Office of Nuclear Reactor Regulation specified in the associated task interface agreement (ADAMS No. ML111610530). The result of these reviews will determine whether there is a performance deficiency associated with this item.
05000443/FIN-2011002-012011Q1SeabrookFailure to Monitor Condition of Control Building per 10CFR50.65(a)(1)lnspectors identified a non-cited violation of 10 CFR 50.65(aX1 ) because NextEra did not adequately monitor the condition of an in-scope structure under the Maintenance Rule (MR). Specifically, NextEra did not evaluate the results of their periodic inspections of the condition of the Control Building (CB) to determine the extent and rate of degradation to the structure. Further, in August 2010 after NextEra identified CB concrete strength degradation that called into question the effectiveness of that structures preventative maintenance program, NextEra did not classify the CB as MR (aX1). NextEra entered the degraded structural concrete issue into its corrective action program to address the extent of condition and establish a mitigation strategy (ARs 57412Q and 581434)Ior all in-scope structures. NextEra also initiated AR 1636419 to complete the evaluation for placing the CB into (a)(1) status. This performance deficiency is more than minor because if left uncorrected, the condition could have resulted in the loss of function for the CB structure due to degrading concrete material properties of structures and systems designed to mitigate design basis events. The finding had very low safety significance because despite degraded concrete conditions and loss of design margin, the CB structure remained operable. The inspectors performed a Phase 1 Significance Determination Process (SDP) screening, in accordance with NRC Inspection Manual Chapter (lMC) 0609, Attachment 4, and determined the issue was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of barrier function, and was not potentially risk significant for external events. This finding had a cross-cutting aspect in the area of problem identification and resolution, evaluation (P.1(c)) because NextEra did not ensure issues adverse to quality potentially impacting nuclear safety were promptly identified and evaluated. Specifically, NextEra did not thoroughly evaluate indications of concrete degradation for the CB to determine the extent and rate of degradation to the structure, and once concrete degradation due to alkali-silica-reaction (ASR) distress was identified, NextEra did not evaluate the issue within the context of the MR program to assure the condition of structures was controlled to maintain design margins.
05000443/FIN-2011002-022011Q1SeabrookFailure to Classify and Monitor the Ocean Transition Structures as In-Scope per 10cFR50.65(bX2)Inspectors identified a non-cited violation of 10 CFR 50.65(bX2) because NextEra did not include certain Seabrook buildings as in-scope structures under the MR program. Specifically, NextEra did not classify the intake transition structure (lTS) and the discharge transition structure (DTS) as in-scope structures in the MR database, and as a result did not include them in the periodic inspections completed under the structures monitoring program per PEG04 from 1995 to 2009. NextEra initiated a MR scoping screening worksheet per procedure NAP 415 and upon consideration of the design basis information concluded both transition structures should be in-scope per 10 CFR 50.65(aX1). The NAP 415 scoping results were accepted by the MR Expert Panel on March 15,2011. NextEra initiated CR 1629504 to enter the issue into the Corrective Action Program (CAP) and determine the extent of condition. The performance deficiency is more than minor because if left uncorrected, given the indications of ASR identified in these concrete structures, not monitoring the ITS and DTS structures for degradation could result in the loss of function of structures supporting systems used to mitigate design basis events, used in the emergency operating procedures, or whose loss could result in a reactor trip. The inspectors performed a Phase 1 Significance Determination Process (SDP) screening, in accordance with NRC Inspection Manual Chapter (lMC) 0609, Attachment 4, and determined the issue was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of barrier function, and was not potentially risk significant for external events. This finding did not have a cross cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance.
05000336/FIN-2010005-012010Q4MillstoneFailure to Provide an Adequate Procedure for Backwashing Condenser Water BoxesA self-revealing finding (FIN) of very low significance was identified for Dominion\'s failure to provide an adequate procedure for backwashing the Unit 2 condenser water boxes in accordance with procedure MP-05-MMM, Manuals, Procedures, Guidelines, Handbooks and Forms. Specifically, in implementing the procedure, the A circulating water (CW) pump automatically ramped down to zero speed shortly after securing the B CW pump. This resulted in a loss of condenser vacuum, which caused an automatic turbine trip. The turbine trip caused an automatic reactor trip. Dominion entered the issue into their corrective action program (CAP) and revised the operating procedure (OP) 2325D. The finding is more than minor because it was similar to NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Example 4b, in that an inadequate procedure led to a reactor trip. The finding was associated with the Procedure Quality attribute of the Initiating Events cornerstone, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Dominion\'s failure to provide an adequate procedure for backwashing Unit 2 condenser water boxes resulted in the variable frequency drive (VFD) logic securing the only CW pump running in that condenser, and subsequently caused a reactor trip. The finding was of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because Dominion did not provide an accurate and up-tO-date procedure for the backwashing of the Unit 2 water boxes.
05000336/FIN-2010005-032010Q4MillstoneLicensee-Identified ViolationTS 3.7.2.1 states that the TDAFW pump has an allowable outage time of 72 hours. TS 4.0.1 requires that the licensee shall declare the TDAFW pump to be inoperable if the pump fails a surveillance test required by TSs. Contrary to this requirement, the TDAFW pump failed a surveillance test on June 30, 2010, and was inoperable for a period of approximately 54 days, which exceeded the TS allowable outage time. Dominion was not aware of the surveillance test failure until an extent of condition review triggered by another failed surveillance test on August 19, 2010, revealed that the TDAFW pump had failed the earlier test. Upon discovery, Dominion restored operability by repairing 3FWA RV45 and placed the condition in the CAP (CR392003 and CR392155). This finding is of very low safety significance because the TDAFW pump was available to fulfill its safety function during the period of time that it was inoperable.
05000336/FIN-2010005-022010Q4MillstoneFailure to Take Adequate Corrective Actions For a Broken Jacket Water Banjo Bolt on the 3 B EDGThe inspectors identified a Green, NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Dominion did not take adequate corrective action following the identification of a degraded condition. Specifically, maintenance personnel identified a broken jacket water fitting (banjo bolt) on the Unit 3, B emergency diesel generator (EDG), but a condition report (CR) was not initiated. Subsequently, an additional similarly degraded fitting resulted in extended unavailability on the Unit 3, B EDG. In response, Dominion entered the issue into the CAP and replaced the broken jacket water fitting. The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent an actual loss of system safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the Problem Identification and Resolution cross-cutting area, Corrective Action Program component, because Dominion did not ensure that issues potentially impacting nuclear safety were promptly identified, fully evaluated, and that actions were taken to address safety issues in a timely manner, commensurate with their safety significance. Specifically, Dominion did not initiate a CR in September 2009 for a degraded condition on the safety-related Unit 3, B EDG.
05000443/FIN-2010005-012010Q4SeabrookLicensee-Identified ViolationTechnical Specification 6.7.1 and Regulatory Guide 1.33 requires that operating activities be implemented in accordance with written procedures. Seabrook procedure OS1 016.05, Step 4.2.26, requires the operator to place the train B standby service water (SW) pump (SW-P41 D) control switch in normal following cooling tower operations. Contrary to the above, on October 8, 2010, the operator left the control switch for SWP41 D in pull-to-Iock after transferring the train B cooling loop from the tower back to the ocean. The train B SW pump was non-functional for about 1 hour 40 minutes until another operator identified the discrepancy during a control board walkdown. The finding had very low safety significance because it did not involve a loss of safety function or impact the safety function for a time greater than the allowed outage time in Technical Specification 3.7.4. Specifically, SW-P41 D was non-functional but recoverable by operator action from the main control board. The violation was licensee identified and entered into the corrective action program as AR 585992.
05000443/FIN-2010003-012010Q2SeabrookInadequate instructions to install test equipment caused the A EDG to be inoperableA self-revealing non-cited violation of Technical Specification 6.7.1, Procedures and Programs, was identified related to the failure of the A EDG during a maintenance run per EC145293 on April 15, 2010. Specifically, NextEra did not provide adequate work instructions to control temporary test equipment attached to the EDG. This led to the failure of the jacket water cooling system that required operators to shutdown the engine, resulting in unplanned unavailability for the A EDG. The leak was promptly repaired and the EDG restored to a functional status on April 17, 2010. The issue was entered into the corrective action program as condition report 221321. The finding is more than minor because it is associated with the work control 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 inadequate work instructions intended to flow balance the A EDG coolant system during an instrumented run, resulted in unplanned extended unavailability of the A EDG. The inspectors performed a Phase 1 Significance Determination Process (SDP) screening, in accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, and determined the issue was of very low safety significance because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, and was not potentially risk significant for external events. The finding had a cross-cutting aspect in the area of human performance - resources (H.2.c) because the work instructions were not adequate to assure temporary test equipment was properly installed.
05000443/FIN-2010003-022010Q2SeabrookLicensee-Identified ViolationTechnical Specification (TS) limiting condition of operation (LCO) 3.6.5.1 requires that both containment enclosure emergency air cleanup systems (CEEACS) be operable and in support of this LCO, surveillance requirement (SR) 4.6.5.1.dA requires CEEACS be capable of producing a negative pressure greater than or equal to 0.25 inches of water in the containment enclosure building within 4 minutes of a start signal. Contrary to the above, NextEra did not ensure that controls needed to maintain the CEEACS operable were in place. Specifically, for approximately five hours on March 15, 2010, and approximately four hours on March 17, 2010, charging pump room door P307 was propped open to support planned maintenance. As a result, both CEEACS systems were inoperable due to the inability to meet TS 4.6.5.1.dA. This violation had very low safety significance for the reasons discussed in Section 40A3 above. This condition was identified in NextEra\'s corrective action program as AR 218893.
05000443/FIN-2010002-012010Q1SeabrookLicensee-Identified ViolationThe following violation of NRC requirements was identified by NextEra. It was determined to have very low significance (Green) and to meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a non-cited violation. License condition F of the Seabrook license requires in part that NextEra implement and maintain in effect all provisions of the approved fire protection program as described in the Seabrook\'s Fire Protection of Safe Shutdown Capability Report. The Seabrook Appendix R Report Sections 3.2. 7.2.B.3 and 3.2.7.3.B.3, stated that the containment must remain habitable to allow operator access to manually operate reactor coolant and residual heat removal system valves needed to place. the plant in cold shutdown within 72 hours. Contrary to the above, NextEra did not ensure that systems needed to maintain the containment habitable during certain postulated fire scenarios remained operable. Specifically, the method described in the Appendix R Report for maintaining containment habitable following a postulated fire in Fire Areas CB-F-1A-A and CB-F-1 BA, did not support placing the plant in cold shutdown. For a postulated fire in these areas in order to restore cooling water to containment air coolers, the report directed operators to start a containment air compressor without cooling water. This would cause the compressor cylinders to rapidly overheat causing Significant damage to the compressor that would prevent restoration of containment air cooling. This issue is also discussed in Section 40A3 above. This finding is of very low safety significance (Green) because per IMC 0609, Appendix F, Attachment 1, Section 1.3.1, a finding that only affects the ability to achieve and maintain cold shutdown screens to green with no further analysis
05000361/FIN-2009007-012009Q4San OnofreFailure to Adequately Store and Preserve Materials for Used in Safety-Related ConcreteThe inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XIII, Handling, Storage and Shipping, for the failure of contractor personnel to establish measures to ensure adequate controls for the storage and preservation of material, associated with the admixture and fly ash, to be used in the production of safety-related concrete. Specifically, on December 10, 2009, contractor personnel failed to properly control key materials from being exposed to the elements which could damage or deteriorate the material and adversely impact the properties of safety-related concrete. This finding was entered into the licensees corrective action program as Nuclear Notification NN 200703527. The finding is greater than minor because use of incorrect material, or material whose properties may have been altered due to improper storage, if left uncorrected, would have the potential to lead to a more significant safety concern. The finding is associated with the design control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide release caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not represent an actual open pathway in the physical integrity of reactor containment and because the concrete for the containment opening had not yet been batched or placed into the containment structure. The finding has a crosscutting aspect in the area of human performance associated with work practices since the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c)) (Section 4OA5.2)
05000361/FIN-2009007-022009Q4San OnofreIncorrect Mixing and Batching Associated with ConcreteThe inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, for the failure of contractor personnel to follow procedures to ensure proper mixing and batching of safety-related concrete. Specifically, on December 19, 2009, contractor personnel failed to ensure each batch contained the specified proportion of hydration controlling admixture. This finding was entered into the licensees corrective action program as Nuclear Notification NN 200715236. The finding is greater than minor because the failure to follow procedures for mixing containment concrete, if left uncorrected, would have the potential to lead to a more significant safety concern. The finding is associated with the design control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide release caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because the finding did not represent an actual open pathway in the physical integrity of reactor containment and because the batch of the concrete in question met the desired design strength as verified by testing. The finding has a crosscutting aspect in the area of human performance associated with work practices since the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c)) (Section 4OA5.2)
05000443/FIN-2009005-012009Q4SeabrookLicensee-Identified ViolationThe following violation of NRC requirements was identified by NextEra, was determined to have very low significance (Green) and to meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a non-cited violation. 10 CFR 50, Appendix B, Criterion III, \"Design Control,\" requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, NextEra did not assure that the design basis for safety-related buried cables was correctly translated into specifications, drawings, procedures, and instructions. Specifically, NextEra did not maintain safety-related underground cables in an environment for which they were designed. The cables were found submerged. This was identified in the corrective action program as 211808 to initiate review of the current manhole and cable monitoring programs, and to initiate long-term corrective actions. This finding is of very low safety significance (Green) because it did not represent an actual loss of safety function or contribute to external event core damage sequences
05000443/FIN-2009004-012009Q3SeabrookFailure to verify that ultimate heat sink isolation valves do not leak in excess of design basis assumptions.The NRC identified a non-cited violation of 10 CFR 50 Appendix-B Criteria III, Design Control, for the failure to verify that service water (SW) isolation valve leakage was within design assumptions for ultimate heat sink (UHS) water inventory. Specifically, the NextEra had not verified by analysis or test that the American Society of Mechanical Engineers (ASME) Class 3 boundary isolation valves, for the safety-related SW piping, provided an adequate leak tight boundary to ensure that the design minimum volume of water would remain in the UHS at the end of a seven-day period with no make-up. Following the identification, NextEra placed the issue into the corrective action program and performed an assessment, which concluded there was reasonable assurance the UHS cooling tower could perform its safety function. The finding was more than minor because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, during a loss of normal ocean water cooling, a leak on the non-safety SW piping could result in a significant loss of inventory from the UHS over a seven-day period. In addition, this finding adversely affected the reliability objective of the protection against external events attribute under the Mitigating Systems Cornerstone. The inspectors determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability or functionality. This finding did not have a cross-cutting aspect because it was not representative of current licensee performance. When NextEra modified the valve seats in the early 1990\'s, they did not verify the modified design by either analysis or test. The valves in question have not been reworked or internally inspected since they were modified. Therefore, the inspectors concluded that this was not reflective of current performance
05000443/FIN-2009004-022009Q3SeabrookLicensee-Identified ViolationThe following violation of very low safety significance (Severity Level IV) was identified by NextEra and was a violation of NRC requirements that meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an non cited violation. Technical Specification 6.7.1 and Regulatory Guide 1.33 requires that maintenance activities be implemented in accordance with written procedures. The Seabrook work control procedures MA 4.5 (Configuration Control) and MA 4.3 (Temporary Modifications and Temporary Alterations) were written per the above. MA 4.5 allows changes to an electrical circuit during maintenance troubleshooting activities, but requires implementation of the modification process per MA 4.3 once troubleshooting is complete. On August 3, 2009, the licensee identified that the environmental qualification (EQ) configuration of the D MSIV control circuits had been degraded on July 30, 2009 while recovering from a faulty test circuit discovered during the quarterly MSIV stroke test in which the valve became partially closed. While troubleshooting the MSIV test circuit, the configuration was changed by disconnecting the connector as allowed under MA 4.5. However, plant workers did not implement a temporary modification per MA4.3 as required after completing the troubleshooting. Had the modification process been entered, the impact of the change on environmental qualification would have been addressed. By not following MA 4.5/4.3, the licensee failed to maintain an electrical circuit configuration that preserved environmental qualification. Upon discovery, the licensee declared the MSIV inoperable and entered TS 3.7.1.5. The connector was reattached restoring the EQ qualification of the electrical circuit, and a slide link was opened to maintain the slow close valve in the retract position. The impact of the circuit modification was acceptably addressed in engineering assessment TAR for AR 202762 and modification EC145071. The finding had very low safety significance because it did not involve a loss of safety function or impact the safety function for a time greater than the allowed outage time in the technical specifications. An engineering assessment determined that the MSIV remained operable but degraded from July 30 to August 3 despite the loss of environmental qualification. The violation was licensee identified and entered into the corrective action program as AR 202762
05000443/FIN-2008005-012008Q4SeabrookLicensee-Identified ViolationTS 6.7.1.a requires that written procedures be established and implemented per Regulatory guide 1.33. Procedure OS1036.01 was written pursuant to the above and requires that steam admission valve MS-V-394 be closed with the control switch in the auto position to align the A emergency feedwater (EFW) pump 37A for standby operation during plant operations at power. Contrary to the above, plant operators did not maintain EFW pump 37A operable during plant operation at full power on October 6, 2008, when the control switch for MS-V-394 was inadvertently moved to open during control board activities. The inadvertent operation of MS-V-394 caused steam generator blowdown to isolate. The operators entered the action statement for TS 3.7.1.2, restored EFW pump 37A to the required standby alignment, and restored steam generator blowdown. The finding was more than minor because the incorrect operation of EFW controls resulted in the unplanned inoperability of the A EFW system and impacted steam generator blowdown during plant operations. The finding had very low safety significance because it did not involve a loss of safety function or impact the safety function for a time greater than the allowed outage time in the TS. The inspectors determined that the violation was licensee-identified. The issue was entered into FPLEs CAP as CR 08-13779
05000443/FIN-2008003-032008Q2SeabrookFailure to control a high radiation area as a locked high radiation areaA self-revealing non cited violation of Technical Specification 6.11.2 was identified. Specifically, on May 1, 2008, FPLE failed to identify and control an existing high radiation area with dose rates greater than 1000 millirems per hour in the reactor containment building. A worker was exposed to higher than expected radiation levels of approximately 2,270 mrems per hour. The worker received a dose of 4 millirem. The finding is more than minor because it is associated with the occupational radiation safety cornerstone attribute of exposure control and affected the cornerstone objective, because not controlling the locked high radiation areas could increase personal exposure. The finding was determined to be of very low safety significance (Green) using the SDP assessment because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. FPLE entered the issue into the corrective action program as a Condition Report 200806982. This finding had a cross-cutting aspect in the area of human performance, work control, because FPLE did not appropriately plan work by incorporating job site conditions that impact radiological safety. Specifically, FPLE did not adequately assess changing area dose rates caused by operating activities, and thus did not adequately plan a work task with due consideration of the actual radiological conditions at the job site (H.3(a)). (2OS1)
05000443/FIN-2008003-022008Q2SeabrookInadequate corrective actions to prevent recurrence of mispositioned stow-operated valves caused inadvertent drain of 2000 gallons from RCSA self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified because FPLE did not implement corrective actions to prevent recurrence of mispositioned valves caused by difficult to operate stow-operator reach rods. Specifically, on April 20, 2008, a mispositioned (partially open), stow-operated filter drain valve, CS-V-1190, resulted in the inadvertent draining of 2000 gallons of water from the reactor cavity while operators placed the reactor letdown system into service. The drain valve was partially open because it was difficult to operate when positioned with its stow-operator. The mispositioning of a stow-operated valve in a safety system was a repeat occurrence of a similar event in October 2007. This finding was more than minor because it was associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of configuration control in the charging system unintentionally drained 2000 gallons from the reactor cavity, which affected the shutdown critical safety function of maintaining adequate reactor inventory. The finding was determined to be of very low safety significance (Green) using the SDP Phase 1 assessment, since the finding did not result in a loss of control of shutdown operations and adequate mitigation capabilities remained available. FPLE entered this issue into the corrective action program as Condition Report 08-06260. The finding has a cross-cutting aspect in the area of problem identification and resolution because FPL Energy did not take appropriate corrective actions to address safety issues in a timely manner commensurate with their safety significance and complexity (P.1.d). Specifically FPL Energy did not take adequate corrective actions to assure the correct positioning of stow-operated safety system valves and thereby prevent recurrence of a significant condition adverse to quality. (Section 1R20
05000443/FIN-2008003-012008Q2SeabrookFailure to follow tagging procedure caused inadvertent drain of 200 gallons from RCSA self-revealing non-cited violation of Technical Specification 6.7.1.a was identified for the failure to implement written procedures governing safety-related activities. Specifically, on April 20, 2008, FPL Energy Seabrook (FPLE) failed to implement tagging and configuration control procedures, resulting in the loss of configuration control during shutdown operations when flow was established through a partially disassembled charging system valve. This resulted in a 200 gallon leak of reactor cavity water onto the floor of the Primary Auxiliary Building (PAB). The letdown flow path was established while work was in progress on valve CS-V-299. A clearance boundary was modified with the incorrect assumption that CS-V-299 was intact. This finding was more than minor because it was associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of configuration control in the charging system unintentionally drained 200 gallons from the reactor cavity, which affected the shutdown critical safety function of maintaining adequate reactor inventory, and caused an uncontrolled leak of radioactively contaminated water to a work area. The finding was determined to be of very low safety significance (Green) using the SDP Appendix G assessment, since the finding did not result in a loss of control of shutdown operations and adequate mitigation capabilities remained available. FPLE entered this issue into the corrective action program as Condition Report 08-06270. The finding has a cross-cutting aspect in the area of human performance, work control, since FPL Energy did not plan and coordinate work activities consistent with nuclear safety (H.3(b)). Specifically, FPLE revised a clearance tagging boundary without verifying the status of affected work activities in accordance with site procedures. (Section 1R20
05000443/FIN-2008003-042008Q2SeabrookLicensee-Identified Violation10 CFR 50.9 requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 8, 2007, FPLE submitted a NRC Form 398 application for an individual=s senior reactor operator license that was not complete and accurate in all material respects. Specifically, the application indicated the individual met the requirement for three years of responsible power plant experience; however, this was inaccurate because the individual had less than the three years of responsible power plant experience. This information was material to the NRC because the NRC used the information submitted on the 368 to allow the applicant to take the initial license exam, and ultimately, issue the individual an SRO license. The traditional enforcement process was used to disposition the violation because it impacted the NRC=s ability to perform its regulatory function. The finding was more than minor because it was a non-willful compromise of an application required by 10 CFR Part 55 that contributed to an individual being granted a SRO license. The violation was licensee identified via an internal audit and entered into their corrective action program (CR 08-01388). FPLE performed a root cause evaluation and informed the NRC. The finding was of very low safety significance because the licensed individual properly performed licensed duties and because the NRC would most likely have granted a waiver of experience requirements, based on the applicant=s work history, had a waiver been requested. (05000443/200800304, Inaccurate Information on Initial Operator License Application, EA-08-164)
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Jerrica-20Johnson-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"