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{{Adams|number = ML092400410}}
{{Adams
| number = ML093160351
| issue date = 11/12/2009
| title = EA-09-145 Seabrook - Final Significance Determination of White Finding with Assessment Followup, and Notice of Violation, (NRC Inspection Report No. 05000443/2009007)
| author name = Collins S
| author affiliation = NRC/RGN-I
| addressee name = St.Pierre G
| addressee affiliation = NextEra Energy Seabrook, LLC
| docket = 05000443
| license number =
| contact person = DeFrancisco, A, (610) 337-5078
| case reference number = EA-09-145
| document report number = IR-09-007
| document type = Enforcement Action, Letter, Notice of Violation
| page count = 9
}}


{{IR-Nav| site = 05000443 | year = 2009 | report number = 007 }}
{{IR-Nav| site = 05000443 | year = 2009 | report number = 007 }}


=Text=
=Text=
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{{#Wiki_filter:November 12, 2009
[[Issue date::August 28, 2009]]


EA-09-145 Mr. Gene St. Pierre  Site Vice President  NextEra Energy Seabrook, LLC  Seabrook Station c/o Mr. Michael O'Keefe  P.O. Box 300    Seabrook, NH  03874
==SUBJECT:==
FINAL SIGNIFICANCE DETERMINATION OF WHITE FINDING WITH ASSESSMENT FOLLOWUP, AND NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 05000443/2009007, SEABROOK STATION, UNIT NO. 1)


SUBJECT:  SEABROOK STATION, UNIT NO. 1 - NRC INSPECTION REPORT 05000443/2009007; PRELIMINARY WHITE FINDING  On July 16, 2009, the NRC completed an inspection at the Seabrook Station, Unit No.The enclosed report documents the inspection findings discussed during an exit meeting on  July 16, 2009, with Mr. Paul Freeman and other members of your stafThis inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your licensThe inspectors reviewed selected procedures and records, observed activities, and interviewed personneThis report documents one self-revealing finding that has preliminarily been determined to be White, a finding with low to moderate increased importance to safety that may require additional NRC inspectionAs described in Section 1R18 of the attached report the finding is associated with the failure to establish adequate design control measures to modify a cooling water flange on the B emergency diesel generator (EDG), which led to the failure of the diesel during a test on February 25, 200This finding was assessed based on the best available information, using the applicable Significance Determination Process (SDP). The final resolution will be conveyed in separate correspondencFollowing the B EDG failure on February 25, 2009, NextEra investigated the event, evaluated the condition of the EDG and its support systems, and restored the EDG and its cooling system to an operable statuFollowing completion of repairs, NextEra performed extensive maintenance operability and reliability runs on the B EDG, and declared it operable on  March 2, 200This finding does not represent an immediate safety concern because of the corrective actions you have takeG. St. Pierre 2  The finding is an apparent violation of NRC requirements and is being considered for escalated enforcement action in accordance with the Enforcement Policy, which can be found on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/enforcement  In accordance with the NRC Inspection Manual Chapter (IMC) 0609, we intend to complete our evaluation using the best available information and issue our final determination of safety significance within 90 days of the date of this letteThe significance determination process encourages an open dialogue between the NRC staff and the licensee; however, the dialogue should not impact the timeliness of the staff's final determinatioWe understand that you continue to evaluate the results of your risk determination for the B EDG failurWe encourage you to provide the results of your evaluation to us when it is finalized using the process as described beloBefore we make a final decision on this matter, we are providing you with an opportunity  (1) to attend a Regulatory Conference where you can present to the NRC your perspective on the facts and assumptions the NRC used to arrive at the finding and assess its significance, or  (2) submit your position on the finding to the NRC in writinIf you request a Regulatory Conference, it should be held within 30 days of the date of this letter and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effectivIf a Regulatory Conference is held, it will be open for public observatioIf you decide to submit only a written response, such submittal should be sent to the NRC within 30 days of the date of this letteIf you decline to request a Regulatory Conference or submit a written response, you relinquish your right to appeal the final SDP determination, in that by not doing either you fail to meet the appeal process outlined in the Prerequisite and Limitation Sections of Attachment 2 of IMC 060Please contact Art Burritt at 610-337-5069, and in writing, within 10 days from the issue date of this letter to notify the NRC of your intentionIf we have not heard from you within 10 days, we will continue with our significance determination and enforcement decision and you will be advised of the results of our deliberations on this matteBecause the NRC has not made a final determination in this matter, no Notice of Violation is being issued for this inspection finding at this timIn addition, please be advised that the number and characterization of the apparent violation may change as a result of further NRC revieThe final resolution of this finding will be conveyed in separate correspondencThe attached report also documents one licensee-identified finding of very low safety significance (Green) that involved a violation of NRC requirements (Section 4OA7). If you contest this violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission,  ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Seabrook StatioThe information you provide will be considered in accordance with Inspection Manual Chapter 030G. St. Pierre 3  In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
==Dear Mr. St. Pierre:==
This letter provides you the final significance determination of the preliminary White finding sent to you in a letter dated August 28, 2009. That August 28, 2009 letter also transmitted NRC Inspection Report No. 05000443/2009007, which provided details regarding the finding. This current letter also provides the result of our assessment of the current performance of Seabrook Station (Seabrook). The updated assessment of Seabrook supplements, but does not supersede, our mid-cycle assessment letter issued on September 1, 2009.


Sincerely,/RA by James W. Clifford Acting for/        David C. Lew, Director Division of Reactor Projects  Docket Nos: 50-443 License Nos: NPF-86
As described in the August 28, 2009 letter, the finding involved the failure to establish adequate design control measures to modify a cooling water flange on the B emergency diesel generator (EDG), one of two safety-related EDGs relied upon to provide electricity to emergency loads during a loss of off-site power. As a result of the inadequate design, the B EDG developed a leak during a routine test on February 25, 2009, and operators secured the B EDG without completing the test. This finding also involved an apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control.


===Enclosure:===
In the NRC August 28, 2009 letter, the NRC had also provided you with an opportunity to request a Regulatory Conference (RC) or provide a written response to the preliminary significance determination. At the request of NextEra Energy Seabrook, LLC (NextEra), an RC was held on September 30, 2009 in the NRC Region I office in King of Prussia, Pennsylvania, to discuss the root cause evaluation of the finding and the differences between the NRCs and NextEras assessment of the safety significance of the finding. During the RC, NextEra staff did not contest the performance deficiency, the related violation, or the NRC description of the event. NextEra also provided additional information in a letter dated September 22, 2009.
Inspection Report 05000443/2009007 


===w/Attachment:===
During the RC, as well as in a subsequent submittal received by the NRC on October 9, 2009, NextEra staff presented a revised risk analysis model, which supported a different view that the finding was of less safety significance than the NRCs preliminary safety determination of the finding as White.
Supplemental Information  cc w/encl:  M. Nazar, Senior Vice President and Chief Nuclear Officer  A. Khanpour, Vice President, Engineering Support M. Warner, Vice President, Nuclear Plant Support  M. Mashhadi, Senior Attorney, Florida Power & Light Company M. Ross, Managing Attorney, Florida Power & Light Company M. O'Keefe, Manager, Licensing Manager  P. Freeman, Plant General Manager  K. Wright, Manager, Nuclear Training, Seabrook Station S. Colman, FEMA, Region I Office of the Attorney General, Commonwealth of Mass K. Ayotte, Attorney General, State of NH O. Fitch, Deputy Attorney General, State of NH P. Brann, Assistant Attorney General, State of Maine R. Walker, Director, Radiation Control Program, Dept. of Public Health, Commonwealth of MA C. Pope, Director, Homeland Security & Emergency Management, State of NH R. Hughes, Director, Licensing and Performance Improvement  J. Giarrusso, MEMA, Commonwealth of Mass D. O'Dowd, Administrator, Radiological Health Section, DPHS, DHHS, State of NH J. Roy, Director of Operations, Massachusetts Municipal Wholesale Electric Company T. Crimmins, Polestar Applied Technology R. Backus, Esquire, Backus, Meyer and Solomon, NH Town of Exeter, State of New Hampshire Board of Selectmen, Town of Amesbury S. Comley, Executive Director, We the People of the United States R. Shadis, New England Coalition Staff M. Metcalf, Seacoast Anti-Pollution League G. St. Pierre 4  In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


Sincerely,/RA by James W. Clifford Acting for/
G. S To support its views, NextEra staff presented information regarding the Seabrook Station Supplemental Emergency Power System (SEPS), a system comprised of two additional diesel generator sets (DGs), which are used to supply backup power to safety-related electrical loads when a safety-related EDG is out of service, or, if a safety-related EDG fails during a loss of all off-site power event. Specifically, NextEra provided an electrical analysis to demonstrate that the SEPS, via one of two of its DGs, is capable of powering electrical loads in a loss of off-site power event when a safety-related EDG is also out of service or has failed during the event.
David C. Lew, Director Division of Reactor Projects Distribution w/encl: (via e-mail)  S. Collins, RA M. Dapas, DRA D. Lew, DRP J. Clifford, DRP A. Burritt, DRP L. Cline, DRP A. Turilin, DRP D. Holody, ORA W. Raymond, DRP, SRI J. Johnson, DRP, RI  E. Jacobs, DRP, OA L. Trocine, RI OEDO H. Chernoff, NRR R. Nelson, NRR D. Egan, NRR, PM R. Ennis, NRR, Backup N. Valentine, NRR ROPreportResources@nrc.gov Region I Docket Room (with concurrences)
ML092400410 SUNSI Review Complete:    LC              (Reviewer=s Initials)  DOCUMENT NAME:  G:\DRP\BRANCH3\INSPECTION\REPORTS\ISSUED\SEA0907.DOC  After declaring this document AAn Official Agency Record@ it will be released to the PubliTo receive a copy of this document, indicate in the box:  "C" = Copy without attachment/enclosure  "E" = Copy with attachment/enclosure  "N" = No copy  OFFICE RI/DRP  RI/DRP  RI/DRP  RI/DRS  RI/ORA  RI/DRP NAME WRaymond/LC for LCline/LC  ABurritt/LC for CCahill/CC    DHolody/AD for DLew/JWC for DATE 08/ 27/09 08/27/09 08/27 /09 08/27/09  08/27/09 08/27/09      OFFICIAL RECORD COPY 1 Enclosure U. S. NUCLEAR REGULATORY COMMISSION  REGION I Docket No.:  50-443  License No.:  NPF-86 Report No.:  05000443/2009007  Licensee:  NextEra Energy Seabrook, LLC Facility:  Seabrook Station, Unit NLocation:  Seabrook, New Hampshire 03874 Dates:  February 25, 2009 through July 16, 2009 Inspectors:  W. Raymond, Senior Resident Inspector    C. Cahill, Senior Reactor Analyst, DRS    K. Mangan, Senior Reactor Inspector, DRS    R. Moore, Reactor Inspector, DRP J. Heinly, Reactor Inspector, DRP J. Rady, Reactor Inspector, DRS E. Burket, Reactor Inspector, DRS  Approved by:  David C. Lew, Director Division of Reactor Projects 2 Enclosure SUMMARY OF FINDINGS IR 05000443/2009007; 02/25/2009-07/16/2009; Seabrook Station, Unit No. 1; Plant ModificationThe report covered a four-month period of inspection by resident and regional inspectorThe significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP) and the cross-cutting aspect of a finding is determined using IMC 0305, "Operating Reactor Assessment Program."  One apparent violation was identifieFindings for which the SDP does not apply may be Green or be assigned a severity level after NRC management revieThe NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 200Cornerstone:  Mitigating Systems  Preliminary WhitA self-revealing apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control was identified following a review of the identified causes for the failure of the B EDG jacket water cooling system on February 25, 200Specifically, NextEra's failure to adequately control design changes implemented on the B EDG jacket water cooling system in January 2009 led to the failure of the gasket on flange JTR005 in the B EDG jacket water cooling system on February 2The inspectors determined that this finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequenceSpecifically, design modification 08MSE11, intended to address flange JTR005 alignment and change the flange gasket design was inadequate and resulted in inoperability of the B EDIn accordance with IMC 0609, "Significance Determination Process," a Phase 3 risk analysis was performed and determined that the calculated delta CDF for the finding was 2.27E-6, which represents a low to moderate safety significance or White findinThe cause of the finding is related to the corrective action component of the cross-cutting area of problem identification and resolution because NextEra  did not thoroughly evaluate problems in a timely manner such that resolutions address causes (P.1(c)). Specifically, NextEra did not adequately evaluate deficient conditions when addressing B EDG cooling water flange leaks, failed to adequately use readily available internal operating experience, and failed to adequately evaluate and correct the impact of engine vibrations on flange JTR005 integrity, contributing to a subsequent failure of the flange. (1R18)  Other Findings  Violations of very low safety significance (Severity Level IV) that were identified by NextEra, have been reviewed by the inspectorCorrective actions taken or planned by the licensee have been entered into the licensee=s corrective action prograThese violations and the licensee=s corrective action tracking numbers are listed in Section 4OA7 of this reporEnclosure REPORT DETAILS 1. REACTOR SAFETY  Cornerstones:  Mitigating Systems 1R18 Plant Modifications (71111.18)  a. Inspection Scope  On February 25, 2009, the B EDG failed to complete a routine operability test when a leak occurred on the engine from a two bolt flange (joint JTR005) on the right bank (RB) turbocharger at the connection to the jacket water return linNRC Inspection Report 2009002 documented NextEra's immediate response and the NRC's initial review of the evenAs of the end of the inspection documented in that report, NextEra's evaluation of the causes for the failure were still ongoing and the inspectors had identified several issues of concern regarding the adequacy of the repairs and modifications completed during the January 2009 overhaul and the adequacy of corrective actions taken to assess and correct the potential effect of the RB turbo vibrations on EDG operabilitThe NRC opened URI 05000443/2009002-01 to track NextEra's completion of the root cause evaluation for the February 25 event and the NRC's subsequent review of NextEra's completed evaluatioTo close URI 05000443/2009002-01 the inspectors reviewed NextEra actions to monitor B EDG conditions and address identified deficiencies including work completed during the B EDG overhaul conducted between January 29 and February 2, 200The inspectors reviewed NextEra modifications to the B EDG jacket water cooling system piping and gaskets on flanged connections, including the design changes in 00MMOD531, 06MSE037, 08MSE211 and EC14490In particular, the inspectors reviewed the flange gasket change completed per maintenance support evaluation (MSE) 08MSE211, and the repairs conducted per work order (WO) 0821400 to address alignmenThe inspectors also reviewed NextEra actions to address vibrations in the RB turbo during engine operation and the results of the root cause investigation for the February 25, 2009 event, including the evaluations conducted for Action Request AR 19144This inspection did not represent an inspection samplb. Findings  IntroductioA self-revealing apparent violation (AV) of 10 CFR 50, Appendix B, Criterion III, Design Control was identified following a review of the identified causes for the failure of the B EDG jacket water cooling system on February 25, 200Specifically, NextEra's failure to adequately control design changes implemented on the B EDG jacket water cooling system in January 2009 led to the failure of the gasket on flange JTR005 in the B EDG jacket water cooling system on February 2DescriptioOperators shutdown the B EDG during a routine operability test on February 25, 2009, when a leak developed in the RB turbo jacket water cooling line at a 2-bolt flanged connectioThe NextEra investigation of the failure found the bolts for flange JTR005 loose and the gasket material severely damaged and blown out along a part of its circumferencPortions of the flange gasket were compressed 60% versus the 4 Enclosure vendor recommended maximum of 16%. The flange faces had irregularities (bowing and cupped surfaces) and there was a misalignment (gap) between the RB turbo outlet flange and the jacket water coolant pipe flangThe gap ranged from 0.164 to 0.245 inches, and by comparison, the installed gasket material had a nominal thickness of 0.0625 incheNextEra evaluated the apparent cause of the flange failure and repaired the flange under EC144905 and Work Order 118563The repairs included changes to address the flange misalignment, gasket material compression, and positive measures to prevent rotation of the boltThe NextEra Root Cause evaluation identified several factors that contributed to the failure of the B EDG jacket water cooling line at flange JTR00In January 2009 NextEra had implemented design change 08MSE211 to change the flange JTR005 gasket design from a 1/8-inch thick full-face gasket to a 1/16-inch annular configuratioThe design change was implemented per work order WO 0821400, which also conducted maintenance to address flange JTR005 alignmenThe root cause was that the 1/16 inch annular gasket installed under 08MSE211 was an inadequate design for the flange specific conditionThe combination of thinner gasket annular design, cupped surfaces, flexed flange, flange gap and bolt loosening from vibration resulted in gasket compression well below the minimum requireThe gasket vendor specified a bolt pre-load to achieve a 6000 psi compressive force, with a minimum of 3244 psi needed to make the flange connection leak tighNextEra found that most of the gasket surface was at 1000 psi or lesThis resulted in an essentially free floating gasket with no sealing pressure in the area where the gasket faileThus, even though flange JTR005 successfully passed a post work test as part of WO 0821400 on January 31, 2009, the as-built gasket design and flange conditions in combination with vibrations which loosened the bolts, left flange JTR005 in a condition to fail with continued B EDG operatioThe cause of the flange JTR005 leak on February 25 was the inadequate design and design control measures used to change the flange gasket from full face to annular configuratioDesign Change 08MSE211 addressed leakage considerations by stipulating attributes in the gasket design that address compressive loaMSE211 did not address the suitability of the gasket design with adequate consideration of the flange performance historThe gasket design did not adequately consider flange specific conditions (bowing under pre-load, surface irregularities), misalignment (gap) or the effects of vibratioMSE211 and WO 0821400 stipulated that the flange condition was required to be "true and flat," but provided inadequate instructions to the workers on how to achieve the required conditionThe work was assumed to be within the skill of the workeThe work order was also intended to correct flange JTR005 alignment issueNextEra concluded the excessive gap found between the flange faces was likely caused by the welding completed during WO 082140Although 08MSE211 stated, "reweld as required ensuring piping is not pulled," the design control measure was inadequate because no specific guidance was provideSimilarly, although WO 0821400 stated the repair "should eliminate any misalignment issues providing care is taken not to 'pull' the flange in final weld out," the work order provided no guidance on how to verify or measure flange JTR005 alignment after weldinDesign change 08MSE211 and WO 0821400 failed to adequately control the welding process relative to flange alignment; failed to address flange specific irregularities; and failed to address vibration that could impact bolt torque and gasket compressive loaAs a consequence, the B EDG jacket water cooling line was left in a condition to fail at flange JTR005 with continued B EDG operatioThe inspectors determined that this was a performance deficiencEnclosure  The inspectors also determined that the primary contributing cause for the performance deficiency was that NextEra did not adequately use internal operating experience or adequately evaluate deficient conditions when addressing the B EDG cooling water flange issueThe work control, corrective action and engineering records show a documented history of leakage from flange JTR00While preparing and implementing the gasket design change per 08MSE211, NextEra did not adequately research the performance history for flange JTR00Readily available plant operating experience showed that a flexible gasket material installed under 06MSE037 was a proven design providing leak free service for two yearThe flexible gasket design could better tolerate flange surface imperfections, was better for a flange experiencing vibrations, and could better accommodate gaps between flange surfaceHad the performance history been adequately considered, NextEra could have either retained the 06MSE037 proven design, or better prepared the 08MSE211 design change to address flange JTR005 conditionFurther, NextEra did not thoroughly evaluate problems such that resolutions addressed causeSpecifically, during the repairs to flange JTR005 per WO 0821400, on January 29, workers requested the use of a locking mechanism on the flange because the fasteners were found less than the required torque (CR200901470). In an evaluation dated February 5, 2009, NextEra concluded a locking feature would be evaluated if the fasteners were loose in the futurThe flange failed during the next EDG run on February 2The failure to adequately review the request for locking devices or evaluate why they were needed was a missed opportunity to prevent vibration induced loosening of the flange boltLocking wires were added to the flange as part of the subsequent design change and repair activity under EC14490AnalysiThe performance deficiency associated with this finding was that inadequate design  control measures used to correct flange alignment and change the gasket design on the B EDG right bank turbocharger jacket water cooling line resulted in the B EDG cooling water line failure on February 25, 200The Seabrook design control manual requires that the design measures for safety related systems consider the equipment performance history and whether materials are suitable for the application and conditionSpecifically, design change 08MSE211 was inadequate because it did not adequately consider the flange performance history and the suitability of gasket materials and thickness relative to flange specific conditions (cupping and bowing); it did not adequately consider welding stresses during repair and then failed to assure flange alignment was acceptable after welding; and, it did not address the impacts of known vibrations on flange performance and gasket compressive loaThe inspectors determined that this finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequenceSpecifically, design modification 08MSE11, intended to address flange JTR005 alignment and change the gasket design was inadequate and resulted in inoperability of the B EDIn accordance with IMC 0609, "Significance Determination Process," Phase 1 worksheets, a Phase 2 risk analysis was required because the finding represents an actual loss of safety function of a single train for greater than the TS allowed outage time of 14 dayThe Phase 2 risk evaluation was performed in accordance with IMC 0609, Appendix A, Attachment 1, "User Guidance for Significance Determination of Reactor Inspection Findings for At-Power Situations."  The total exposure period for the degraded condition 6 Enclosure was approximately 625 hours (26 days). Using Seabrook's Phase 2 SDP notebook, pre-solved worksheets, and an initiating event likelihood of 3-30 days, the inspector identified that this finding is of potentially substantial safety significance (Yellow). The finding affected sequences in the loss of offsite power (LOOP) and LOOP and Loss of Class 1E 4.16 kV AC Bus A (E5) (LEACA) worksheetFor the LOOP condition, sequences that resulted in a station blackout (SBO) were the dominant contributor to core damage. For the LEACA condition, sequences that involved a stuck open relief valve were the dominant contributor to core damagThe sum of the sequences in the LOOP and LEACA, for the identified exposure period resulted in a YelloIn recognition that the Phase 2 notebook typically yields a conservative result, a NRC Region I Senior Reactor Analyst (SRA) performed a Phase 3 risk assessment of this findinThe SRA used the Seabrook Standardized Plant Analysis Risk (SPAR) model, Revision 3.50, dated July 2, 2009, and Graphical Evaluation Module (GEM), in conjunction with the System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE), Version 7, software to estimate the internal risk contributioIn discussions with the licensee, it was discovered that the SPAR model for Seabrook did not credit instrument air accumulatorInformation on these backup accumulators was included in the SDP notebooAdditionally, Seabrook questioned the modeling of switchgear ventilation and provided design information to support the modeling revision to reflect the design success criteriThe SRA worked with Idaho National Laboratory (INL) and modified the model to correct the instrument air dependencies and modified the ventilation success criteriSpecific changes included: 1. Basic event SWS-FAN-FC-RMCOOL (1E-3) was added to the SWS switchgear cooling fault trees (EPS-DGNA-SWS, EPS-DGNB-SWS, SWS-RMCLA, SWS-RMCLB). This event was ANDed with the existing SWS switchgear ventilation logi. Basic event IAS-TNK-FC-ADV (4.8E-8) was added to the atmospheric steam dump valve (ADV) air supply fault tree (MSS-ADVS-AIR). This event allows operation of the ADVs based on the air supply of the accumulatorThe information provided indicated that these accumulators would facilitate 10 cycles over a period of 10 hourThe following assumptions were used for this assessment:  1. To closely approximate the type of failure exhibited by the B EDG, the SRA used the B EDG failure to run basic event <EPS-EDN-FR-1B > and changed its failure probability to 1.0, representing a 100 percent failure-to-run condition. 2. The exposure time for this condition was 625 hours (546.95 hours, plus 77.75 hours of unavailability during troubleshooting and repair). 3. Based upon the nature of the failure, no additional operator recovery credit was        provide. All remaining events were left at their nominal failure probabilitie. Cutset probability calculation truncation was set at 1E-1Based upon the above assumptions, the Seabrook SPAR model internal contribution to conditional core damage probability (CCDP) was calculated at 1.8E-This low E-6 delta CCDP value represents a low to moderate safety significance (White). The dominant internal event sequences involve a loss of offsite power event with subsequent failure of the A EDG and the supplemental emergency power system (SEPS) resulting in a Station 7 Enclosure Black OuAdditionally, the site fails to recover a diesel generator within four hours and the failure to recover offsite power within four hours. These Phase 3 SPAR model results correlate well to the Phase 2 SDP Notebook dominant core damage sequenceThe Seabrook Probabilistic Safety Assessment (PSA) is a full scope model that includes events such as seismic events, internal fires and internal floodThe PSA summarizes the contribution mainly from a turbine building fire or flooding as representing approximately 31% of the total (internal and external) core damage frequency, or nearly one third of the annualized risFor the given exposure period this equates to an external events delta CCDP of 4.7E-The NRC does not have an external risk model for SeabrooConsequently, the SRA used the licensee's external risk assessment to quantify the external risk contribution for this conditioThe SRA used IMC 0609, Appendix H, "Containment Integrity Significance Determination Process," to determine if this finding was a significant contributor to a large early releasThe Seabrook containment is classified as a pressurized water reactor large-dry containment desigBased upon the dominant sequences involving loss of offsite power and station blackout (SBO) initiating events, per Appendix H, Table 5.2, "Phase 2 Assessment Factors - Type A Findings at Full Power," the failure of the B EDG does not represent a significant challenge to containment integrity early in the postulated core damage sequenceConsequently, this finding does not screen as a significant large early release contributor because the close-in populations can be effectively evacuated far in advance of any postulated release due to core damagAccordingly, the risk significance of this finding is associated with the delta CDF value, per IMC 0609, Appendix H, Figure 5.1, and not delta LERThe Seabrook model used to evaluate the condition was RISKMAN model DBGOOS which was based on SB2006NFor the given assumptions, for a failure of the B EDG to run, over the given exposure period, the licensee calculated CDF was 1.48E-The contribution from internal events was 1.01E-6, and external event contribution was 4.7E-Similar to the NRC internal risk contribution, Seabrook's model illustrates that the largest percentage of internal risk is derived from station blackout eventFor the given assumptions, the licensee and NRC results are in close agreemenAs a result, the calculated total risk significance of this finding is based upon NRC analysiThe calculated risk is the summation of internal and external risk contributions (delta CCDP internal + delta CCDP external (fires and floods) = delta CCDP total) which equates to; 1.8E-6 + 4.7E-7 = 2.27E-6 delta CCDAnnualized, this value of 2.27E-6 delta CDF represents a low to moderate safety significance or White findinThe cause of the finding is related to the corrective action component of the cross-cutting area of problem identification and resolution in that the licensee failed to thoroughly evaluate problems in a timely manner such that resolutions address causes (P.1(c)). Specifically, NextEra failed to adequately evaluate deficient conditions when addressing B EDG cooling water flange leaks, failed to adequately use readily available internal operating experience, and failed to adequately evaluate and correct the impact of engine vibrations on flange JTR005 integritEnforcemenCFR 50, Appendix B, Criterion III, "Design Control," states, in part, that measures shall be established to assure that regulatory requirements and the design basis for systems and components are correctly translated into specifications and 8 Enclosure instructionMeasures shall also be established for the selection and review for suitability of application of materials and parts that are essential to the safety-related functions of the systems and componentThe Seabrook Station Design Control Manual (DCM) was developed pursuant to the above to establish design control measures for safety related components, including the emergency diesel generatorDCM Chapter 2, Section 8.0, describes the Maintenance Support Evaluation (MSE) as the design control measure to implement in support of maintenancWhen preparing the MSE, the DCM requires that the design inputs and interdisciplinary review guidelines on Figures 4-1-1 through 4-1-14 shall be used to prepare and develop the design change and understand the areas impacteDCM Figure 4-1-1, Design Inputs, and Figure 4-1-3, Independent Reviewer Guidelines, requires that the design shall consider mechanical requirements such as stresses and vibration; whether materials are suitable for the application; credible failure modes of connected equipment; and, account for equipment performance historContrary to the above, design change 08MSE211, implemented by Work Order 0821400 on January 29 - 31, 2009, to modify and repair a two bolt flange (joint JTR005)
on the B EDG right bank turbocharger, did not adequately consider: mechanical requirements such as stresses and vibration; whether materials were suitable for the application; credible failure modes of connected equipment; and, account for equipment performance historSpecifically, design change 08MSE211 and WO 0821400 did not adequately address the suitability of materials relative to flange specific conditions (cupping and bowing); did not adequately control welding stresses during repair and did not assure post weld flange alignment was acceptable; did not adequately consider the flange performance history and potential failures; and, did not address the impacts of known vibrations on flange performance and gasket compressioAs a result, the B EDG turbocharger flange JTR005 was left in a condition to fail with continued B EDG operation, and the diesel was declared inoperable during a test on February 25, 2009, when the flange gasket blew out causing a rapid loss of jacket cooling wateThis issue was entered into Seabrook's corrective action program as CR 19144Pending final determination of significance, this finding is identified as an AV (AV 05000443/2009007-01, Inadequate B EDG Design Change). Therefore URI 05000443/2009002-01 was closeOA6 Meetings, Including Exit  Exit Meeting Summary  On July 16, 2009, the resident inspectors presented the inspection results to Mr. Paul Freeman and other members of his staff, who acknowledged the findinNextEra acknowledged that none of the material examined by the inspectors during the inspection was considered proprietary in naturEnclosure 4OA7 Licensee-Identified Violation  The following violation of very low safety significance (Severity Level IV) was identified by NextErIt was a violation of NRC requirements that met the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a non-cited violation (NCV).


10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that conditions adverse to quality are promptly identified and correcteIn the case of significant conditions adverse to quality, the measure shall assure that the cause of the condition is determined and corrective action is taken to preclude repetitioThe Florida Power and Light (FPL) Energy Quality Assurance Topical Report (QATR) was written pursuant to the above and states in Section A-6 that FPL implements a corrective action program to promptly identify and correct conditions adverse to qualitProcedure PI-AA-205 requires that significant conditions adverse to quality be resolved through corrective actions to prevent recurrencContrary to the above, NextEra Nuclear Oversight issued a finding on April 9, 2009, (QR 090-017) after determining that past corrective actions for B EDG turbocharger vibration issues were inadequate and have not been effective based on a past and recent history of increased vibration, bolt failures, bolt loosening, turbocharger related coolant piping weld failures, coolant system leaks and a failure in some instances to document these conditions in the condition reporting systeThe failure to resolve long standing and increasing vibration and related issues for the B EDG constituted ineffective corrective actioThe finding was more than minor because the ineffective action to resolve turbocharger vibrations impacted the availability and reliability of a mitigating systeFurther, turbocharger vibration was causal to the B EDG failure on February 25, 2009 (reference Section 1R18 above). 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 specificationWhile increased vibrations were causal to the February 25th B EDG failure, they were not the root cause since the cooling water system would have failed due the inadequate gasket design and irregular flange conditionFurther, the finding identified in QR 09-017 is separate from NRC Violation 20090701 since the inadequate design change resulting in the February 25 B EDG failure occurred during the discrete time period of January 29-31, 2009, whereas the corrective actions for the B EDG turbocharger vibrations have been ongoing for a longer period of time (reference 2001 CR 200107312). The inspectors determined that the Criterion XVI violation was licensee-identifieNextEra entered the issue into the corrective action program as CR 0019437A-1 Attachment SUPPLIMENTAL INFORMATION  KEY POINTS OF CONTACT Licensee personnel R. Arn, Engineering K. Browne, Assistant Operations Manager R. Campo, Plant Engineer P. Freeman, Plant General Manager G. Kim, Risk Analyst K. Kiper, Risk Analyst N. Levesque, Engineering Supervisor E. Metcalf, Operations Manager M. Ossing, Engineering Support Manager M. Palumbo, Plant Engineer R. Plante, Maintenance Supervisor R. Samson, Maintenance Supervisor G. St. Pierre, Site Vice President  LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED  Opened 05000443/2009007-01 AV Inadequate B EDG Design Change  Closed  05000443/2009002-01 URI B EDG Emergency Shutdown During Testing on 2/25/09  LIST OF DOCUMENTS REVIEWED  Miscellaneous Operations Logs - Various MRC Associates Report, Modal Analysis of the Turbocharger on the B EDG, June 1992 Fairbank Morse Engineering Report, Turbocharger Vibration, September 27, 1991 Risk Significance of DG-B Failure February 25, 2009, 5/7/09 and 5/27/2009 Engineering Evaluations EE-09-002, Revision 0, 6/24/09; Revision 1, 7/29/09 B EDG Vibration Monitoring Data System Engineering Notes on B EDG turbocharger vibrations, July 1999 Fairbank Morse/Coltec Industries Engineering Report, Turbocharger Vibration, 9/27/91 ARC Associates Report, Modal Analysis of the Turbocharger Section of the B Diesel Generator, 6/9/92
NextEra contended that crediting this one-of-two DG mode of SEPS operation decreases the safety significance of the event to Green.


Condition Reports Root Cause Analysis for CR 191440, 194370 Action Requests 00191440, 00191586, 00191608 CR200901470, CR199917417, CR200901470, CR200107312, CR200304671, CR200509803, CR200210604, CR200412056, CR200505245, CR200509803, CR200800136, CR200801690, CR200809251, CR200809307, CR200901505 A-2 Attachment Design Changes DCR 94-00012, EDG Safety Classification Review, DCN 01 06MSE037, EDG Cooling Water System Gasket Replacement (AFLAS) 00MMOD531, EDG Turbocharger Cooling Water Piping Upgrade, DCN 12 94-064, D.G. Cooling Water System Gasket Replacement, Rev 01 08MSE211, EDG Turbocharger CC Water Piping Optional Gasket Configuration and Bolting Type EC144905, EDG Turbocharger CC Piping Outlet Cover Modification/Gasket Replacement
After considering the information developed during the inspection and the information provided by NextEra at the RC and in writing on September 22, 2009 and October 9, 2009, the NRC has concluded that the finding is appropriately characterized as having low to moderate significance to safety, and is therefore characterized as White. In summary, the NRC concluded that NextEra failed to demonstrate that having only one SEPS diesel available during some events would assure adequate power to supply needed safety equipment during those events. The NRC has also determined that NextEras proposal to use single SEPS success criteria, with predicted low margins to the equipment operating limits and extrapolated data based on engineering judgment, represents a loss of defense-in-depth. Additionally, the NRC is concerned about the uncertainty of operator actions as they respond during a potential overload of the SEPS DG. In conclusion, the NRC finds that your current use of two-out-of-two SEPS engine success criteria is appropriate for your risk assessment; this results in the estimate of the annualized incremental core damage probability is 2.27E-6 for Unit 1, a low to moderate safety significance, and therefore, a White finding. Each of the determinations expressed above is detailed in Enclosure 2.


Drawings Drawing B20466, DG Cooling Water System Detail Drawing 1-NHY-310882, CWD for Pressurizer Pressure Control Valve PCV-456B P&ID 1-NHY-506402, DB - DG B Lube Oil System Control Loop Diagram P&ID 1-NHY-504120, DG - DG Temperature Scanner Logic Diagram P&ID 1-NHY-310008, 4160 Bus E6 One Line Diagram P&ID 1-DG-B20463, Diesel Generator Lube Oil System Train B Detail 1-NHY-310002, Unit Electrical Distribution One Line Diagram, Rev. 40 1-NHY-310010, D1A and DG-1B One Line Diagram Sh.1, Rev. 14 1-NHY-310010, DG-1A and DG-1B One Line Diagram Sh.2, Rev. 4  Work Orders Work Orders (WO) 0821400, 0812472, 0442764, 05131067, 072419, 0805715, 01185637  Procedures PI-AA-205, Condition Identification and Corrective Action PI-AA-01, Corrective Action and Condition Reporting ES0815.002, General Welding Procedure, Rev 00, Chg 21 ES0815.004, Welding of Carbon Steel Materials, Rev 00, Chg 08 ES1807.001, Visual Examination Procedure for Welding, Rev 07, Chg 02 MA-AA-203, Work Order Planning Process, Rev 5 MA-AA-202, Work Order Execution Process, Rev 2 MS0517.03, Flange Maintenance, Rev 9
You have 30 calendar days from the date of this letter to appeal the staffs significance determination for this finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2. An appeal must be sent in writing to the Regional Administrator, Region I, 475 Allendale Rd., King of Prussia, PA 19406.


Manuals FPLE Quality Assurance Topical Report (QATR), Section A-6, "Corrective Action" Design Change Manual (DCM), Revisions 37- 45 DCM Sections 1.0, 2.0 andDCM Figures 4-1-1 through 4-1-14 DCM Figure 4-1-1, Design Inputs DCM Figure 4-1-3, Independent Reviewer Guidelines A-3 Attachment LIST OF ACRONYMS  AR  Action Request CR  Condition Report DCM  Design Control Manual EDG Emergency Diesel Generator LERs  Licensee Event Reports MSE  Maintenance Support Evaluation NCV                Non-Cited Violation NRC  U.S. Nuclear Regulatory Commission NRR  Nuclear Reactor Regulation PARS  Publicly Available Records RB  Right Bank RV                  Reactor Vessel SDP Significance Determination Process TS Technical Specifications UFSAR Updated Final Safety Analysis Report WO                  Work Order
The NRC has also determined that the failure to establish adequate design control measures to modify the cooling water flange on the B EDG is a violation of 10 CFR Appendix B, Criterion III, Design Control, as cited in the enclosed Notice of Violation (Notice) (Enclosure 1). The circumstances surrounding the violation were described in NRC Inspection Report No.
 
05000443/2009007. In accordance with the NRC Enforcement Policy, the Notice is considered an escalated enforcement action because it is associated with a White finding.
 
You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response.
 
As a result of our review of Seabrook performance, including this White finding, we have assessed Seabrook to be in the Regulatory Response column of the NRCs Action Matrix.
 
Therefore, we plan to conduct a supplemental inspection using Inspection Procedure 95001, Inspection for One or Two White Inputs in a Strategic Performance Area, when your staff has notified us of your readiness for this inspection. This inspection procedure is conducted to provide assurance that the root cause and contributing causes of risk significant performance issues are understood, the extent of condition is identified, and the corrective actions are sufficient to prevent recurrence.
 
G. S In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response, will be made available electronically for public inspection in the NRC Public Document Room or from the NRCs document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
 
Sincerely,
/RA/
 
Samuel J. Collins Regional Administrator
 
Docket No.: 50-443 License No.: NPF-86
 
Enclosures:
1. Notice of Violation 2. NRC Basis for Final Significance Determination
 
cc w/encl: Distribution via ListServ
 
ML093160351
"C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No Copy
 
OFFICE
 
RI/ORA
 
RI/DRP
 
RI/DRS
 
RI/DRP
 
RI/DRP
 
NAME
 
ADeFrancisco
*WRaymond CCahill ABurritt
 
JClifford
 
DATE
 
10/30 /09
 
10/29/09 10/ 26 /09 10/ 30 /09
 
11/03/09
 
OFFICE
 
RI/DRS
 
RI/ORA
 
RI/ORA
 
HQ/OE
 
RI/ORA
 
NAME
 
DRoberts
 
KFarrar MB for KF DHolody
**GBowman
 
SCollins
 
DATE 10/29 /09
 
10/28 /09 11/ 04 /09 11/05/09
 
11/08/09
 
G. S Distribution w/encl: (via e-mail)
ADAMS (PARS)
SECY CA OEMAIL OEWEB W Borchardt, EDO B Mallett, DEDR L. Trocine, OEDO R. Zimmerman, OE B. Sosa, OE N. Hilton, OE L Lopez,OE N Hasan, OE M. Crutchley, OE G Bowman, OE E Leeds, NRR Enforcement Coordinators RII, RIII, RIV (C Evans, S. Orth, W. Jones)
C Marco, OGC E Hayden, OPA H Bell, OIG C. McCrary, OI S Titherington-Buda, OCFO M Williams, OCFO DScrenci/NSheehan, RI K Farrar, RI D Holody, RI A DeFrancisco, RI/M. McLaughlin C ODaniell, RI Region I OE Files (with concurrences)
RI ORA Mail Resource RI DRP Mail Resource S. Collins, RA M. Dapas, DRA J. Clifford, DRP D. Lew, DRP M. Gray, DRP A. Burritt, DRP L. Cline, DRP/A. Turilin, DRP W. Raymond, DRP, SRI/J. Johnson, DRP, RI E. Jacobs, DRP, OA RidsNrrODMail Resource RidsNrrPMSeabrook Resource RidsNrrDorpl1-2 Resourse H. Chernoff, NRR R. Nelson, NRR D. Egan, NRR, PM ROPreports@nrc.gov Region I Docket Room (with concurrences)
 
NOTICE OF VIOLATION
 
NextEra Energy Seabrook, LLC
 
Docket No. 50-443 Seabrook Station
 
License No. NPF-86
 
EA-09-145
 
During an inspection documented in NRC Inspection Report No. 05000443/2009007, issued on August 28, 2009, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is set forth below:
 
10 CFR 50 Appendix B Criterion III, Design Control, states, in part, that measures shall be established to assure that regulatory requirements and the design basis for systems and components are correctly translated into specifications and instructions. Measures shall also be established for the selection and review for suitability of application of materials and parts that are essential to the safety-related functions of the systems and components.
 
The Seabrook Station Design Control Manual (DCM), developed pursuant to the above to establish design control measures for safety related components, including the emergency diesel generators (EDG), states, in Chapter 2, Section 8.0, that the Maintenance Support Evaluation (MSE) is the design control measure to be implemented in support of maintenance. When preparing the MSE, the DCM requires that the design inputs and interdisciplinary review guidelines on Figures 4-1-1 through 4-1-14 shall be used to prepare and develop the design change and understand the areas impacted. DCM Figure 4-1-1, Design Inputs, and Figure 4-1-3, Independent Reviewer Guidelines, requires that the design shall consider mechanical requirements such as stresses and vibration; whether materials are suitable for the application; credible failure modes of connected equipment; and, shall account for equipment performance history.
 
Contrary to the above, on January 31, 2009, NextEra Energy Seabrook, LLC, completed Work Order 0821400 on the B EDG without adequately establishing measures to assure that regulatory requirements and the design basis for systems and components were correctly translated into specifications and instructions. Specifically, design change 08MSE211, implemented by the Work Order, to modify and repair a two bolt flange (joint JTR005) on the B EDG right bank turbocharger, did not adequately: (1) control welding stresses during repair, assure post weld flange alignment was acceptable, or address the impacts of known vibrations on flange performance and gasket compression; (2)
address the suitability of gasket materials relative to flange specific conditions (cupping and bowing); and, (3) consider the flange performance history and potential failures to account for equipment performance history and credible failure modes of connected equipment. As a result, the B EDG turbocharger flange JTR005 failed during B EDG operation on February 25, 2009, causing a rapid loss of jacket cooling water and the EDG being declared inoperable.
 
This violation is associated with a White finding.
 
Pursuant to the provisions of 10 CFR 2.201, NextEra Energy Seabrook, LLC, is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional Administrator, Region I, and a copy to the NRC Resident Inspector at the facility that is the
 
Notice of Violation
 
subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation; EA-09-145" and should include: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time. If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.
 
Because your response will be publicly available in the NRC Public Document Room or from the NRCs document system (ADAMS), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be made available without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
 
In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days.
 
Dated this 12th day of November 2009
 
Enclosure 2: Evaluation of the Licensees Positions, NRCs Response, and the NRC Final Significance Determination  
 
On September 30, 2009, the NRC held a Regulatory Conference with representatives of NextEra Energy Seabrook, LLC (NextEra) Seabrook Station, to discuss the significance of a finding involving the failure to establish adequate design control measures to modify a cooling water flange on the B emergency diesel generator (EDG). The inadequate modification led to the development of a leak from the B EDG during a routine test on February 25, 2009, requiring operators to secure the B EDG without completing the test.
 
At the conference, NextEra staff highlighted differences between the NRCs risk assessment of the finding, as provided in NRC Inspection Report No. 05000443/2009007, dated August 28, 2009, and NextEras risk assessment that the licensee developed subsequent to identification of the preliminary White finding. NextEra staff also provided additional information in writing to the NRC on October 9, 2009.
 
SUMMARY OF NEXTERAS POSITION
 
Specifically, NextEra staff presented information regarding the Seabrook Station Supplemental Emergency Power System (SEPS), a system comprised of two additional diesel generator sets (DGs), which are used to supply backup power to safety-related electrical loads when a safety-related EDG is out of service, or, if a safety-related EDG fails during a loss of all off-site power event. Specifically, NextEra provided an electrical analysis to demonstrate that only one of the two SEPS DGs are needed to provide the required electrical loads in a loss of off-site power event when a safety-related EDG fails, or is out of service, during the event, and that crediting this one-of-two DG mode of SEPS operation decreases the safety significance of the event to Green. In support of its position, NextEra stated that: (1) SEPS capacity, based on a one of two SEPS diesel generator success criteria, is supported by detailed analysis and vendor test data; (2) electrical analysis of the SEPS design has identified additional margin in the capability of the SEPS system, enabling a single SEPS DG to supply loss of offsite power loads without tripping; and, (3) operator actions required for successful SEPS operation do not include reduction of SEPS load.
 
NRC EVALUATION OF NEXTERAS POSITION
 
1. NextEra contended that SEPS capacity, based on a one of two SEPS diesel generator success criteria, is supported by detailed analysis and vendor test data.
 
The NRC determined that NextEra did not demonstrate acceptable results using one of two SEPS for some low probability events, such as both SEPS DGs starting and one subsequently failing (failure-to-run scenario). To support its demonstration, NextEra relied on a sensitivity analysis in its evaluation of the one of two SEPS success criteria.
 
NextEras analysis of the SEPS failure modes and reliance on a sensitivity analysis is inconsistent with the standard modeling of equipment failures in the Seabrook Probabilistic Risk Assessment (PRA). Further, the NextEra analysis was performed using generic diesel failure data, and the NRC also concluded that the application of generic class 1E EDG failures, operating within their design, is not representative of the failure rates that would be anticipated in a single SEPS operating at 108% of its design with degraded frequency and voltage. The NRC further determined that NextEras evaluation did not consider the residual heat removal and cooling tower pump test
 
Enclosure 2
 
scenarios impacting loading on the diesels. Regulatory Guide 1.200 states that success criteria analysis as it relates to PRA determines the minimum requirements for each function and ultimately the systems used to perform the functions necessary to prevent core damage given an initiating event. Therefore, the NRC concludes that the only mode of SEPS operation that has been adequately demonstrated by design, factory and on-site testing is based on a two of two SEPS diesel generator success criteria, not a one of two criteria as described in NextEras evaluation.
 
2. NextEra contended that electrical analysis of the SEPS design has identified sufficient margin in the capability of the SEPS system, enabling a single SEPS DG to supply loss of offsite power loads without tripping.
 
The NRC considered that one of two SEPS appears to work by NextEras engineering analysis. However, the engine overload condition combined with narrow margins to operating limits and protection settings, presents uncertainty in engine and support system performance, such that successful single SEPS operation cannot be reasonably assured. Most significantly, the NRC determined that there is very little margin between the calculated electrical current level for the predicted EDG overload condition and the over current setpoint for the DG output breaker. In addition, NextEras evaluation for the single SEPS was based on factory tests of a similar engine and it did not provide sufficient information to validate that the data from the similar engine was applicable to the Seabrook SEPS units. Therefore, the NRC concludes that adequate additional margin is not supported by the electrical or risk analysis.
 
3. NextEra contended that operator actions required for successful SEPS operation do not include reduction of SEPS load.
 
Regarding the role of operators in the risk evaluation, the NRC acknowledges the NextEra analysis of operator actions to successfully assure operation using a single SEPS engine. However, while the path through emergency procedure ECA 0.0 appears to demonstrate success under tightly constrained assumptions, there remains some question as to what operators would do if placed in the postulated scenario. The EDG fault setting report provided by NextEra for the overload condition includes many fault effects that were not addressed by NextEras analysis. The NRC continues to question whether: (1) these faults would prompt operators to take actions that would negatively impact the success of a single SEPS engine run; and, (2) the operators would reduce EDG loading per the procedure or remove needed loads. Therefore, the NRC notes that uncertainty in operator actions combined with the limited guidance provided in the available procedures makes it difficult to assure success on a single SEPS engine.
 
In summary, the NRC concluded that NextEra failed to demonstrate success for one of two SEPS during some events as appropriate. The NRC has also determined that NextEras proposal to use single SEPS success criteria, with predicted low margins to the equipment operating limits and extrapolated data based on engineering judgment, represents a loss in defense-in-depth. Additionally, the NRC determined that uncertainty in operator actions for SEPS load reduction is a concern. In conclusion, the NRC finds that continued use of the two-out-of-two SEPS engines recovery criteria is appropriate in the risk assessment; this results in the estimate of the annualized incremental core damage probability is 2.27E-6 for Unit 1, a low to moderate safety significance, and therefore, a White finding.
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Latest revision as of 08:19, 14 January 2025

EA-09-145 Seabrook - Final Significance Determination of White Finding with Assessment Followup, and Notice of Violation, (NRC Inspection Report No. 05000443/2009007)
ML093160351
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 11/12/2009
From: Collins S
NRC Region 1
To: St.Pierre G
NextEra Energy Seabrook
DeFrancisco, A, (610) 337-5078
References
EA-09-145 IR-09-007
Download: ML093160351 (9)
v • d • e


Text

November 12, 2009

SUBJECT:

FINAL SIGNIFICANCE DETERMINATION OF WHITE FINDING WITH ASSESSMENT FOLLOWUP, AND NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 05000443/2009007, SEABROOK STATION, UNIT NO. 1)

Dear Mr. St. Pierre:

This letter provides you the final significance determination of the preliminary White finding sent to you in a letter dated August 28, 2009. That August 28, 2009 letter also transmitted NRC Inspection Report No. 05000443/2009007, which provided details regarding the finding. This current letter also provides the result of our assessment of the current performance of Seabrook Station (Seabrook). The updated assessment of Seabrook supplements, but does not supersede, our mid-cycle assessment letter issued on September 1, 2009.

As described in the August 28, 2009 letter, the finding involved the failure to establish adequate design control measures to modify a cooling water flange on the B emergency diesel generator (EDG), one of two safety-related EDGs relied upon to provide electricity to emergency loads during a loss of off-site power. As a result of the inadequate design, the B EDG developed a leak during a routine test on February 25, 2009, and operators secured the B EDG without completing the test. This finding also involved an apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control.

In the NRC August 28, 2009 letter, the NRC had also provided you with an opportunity to request a Regulatory Conference (RC) or provide a written response to the preliminary significance determination. At the request of NextEra Energy Seabrook, LLC (NextEra), an RC was held on September 30, 2009 in the NRC Region I office in King of Prussia, Pennsylvania, to discuss the root cause evaluation of the finding and the differences between the NRCs and NextEras assessment of the safety significance of the finding. During the RC, NextEra staff did not contest the performance deficiency, the related violation, or the NRC description of the event. NextEra also provided additional information in a letter dated September 22, 2009.

During the RC, as well as in a subsequent submittal received by the NRC on October 9, 2009, NextEra staff presented a revised risk analysis model, which supported a different view that the finding was of less safety significance than the NRCs preliminary safety determination of the finding as White.

G. S To support its views, NextEra staff presented information regarding the Seabrook Station Supplemental Emergency Power System (SEPS), a system comprised of two additional diesel generator sets (DGs), which are used to supply backup power to safety-related electrical loads when a safety-related EDG is out of service, or, if a safety-related EDG fails during a loss of all off-site power event. Specifically, NextEra provided an electrical analysis to demonstrate that the SEPS, via one of two of its DGs, is capable of powering electrical loads in a loss of off-site power event when a safety-related EDG is also out of service or has failed during the event.

NextEra contended that crediting this one-of-two DG mode of SEPS operation decreases the safety significance of the event to Green.

After considering the information developed during the inspection and the information provided by NextEra at the RC and in writing on September 22, 2009 and October 9, 2009, the NRC has concluded that the finding is appropriately characterized as having low to moderate significance to safety, and is therefore characterized as White. In summary, the NRC concluded that NextEra failed to demonstrate that having only one SEPS diesel available during some events would assure adequate power to supply needed safety equipment during those events. The NRC has also determined that NextEras proposal to use single SEPS success criteria, with predicted low margins to the equipment operating limits and extrapolated data based on engineering judgment, represents a loss of defense-in-depth. Additionally, the NRC is concerned about the uncertainty of operator actions as they respond during a potential overload of the SEPS DG. In conclusion, the NRC finds that your current use of two-out-of-two SEPS engine success criteria is appropriate for your risk assessment; this results in the estimate of the annualized incremental core damage probability is 2.27E-6 for Unit 1, a low to moderate safety significance, and therefore, a White finding. Each of the determinations expressed above is detailed in Enclosure 2.

You have 30 calendar days from the date of this letter to appeal the staffs significance determination for this finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2. An appeal must be sent in writing to the Regional Administrator, Region I, 475 Allendale Rd., King of Prussia, PA 19406.

The NRC has also determined that the failure to establish adequate design control measures to modify the cooling water flange on the B EDG is a violation of 10 CFR Appendix B, Criterion III, Design Control, as cited in the enclosed Notice of Violation (Notice) (Enclosure 1). The circumstances surrounding the violation were described in NRC Inspection Report No.

05000443/2009007. In accordance with the NRC Enforcement Policy, the Notice is considered an escalated enforcement action because it is associated with a White finding.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response.

As a result of our review of Seabrook performance, including this White finding, we have assessed Seabrook to be in the Regulatory Response column of the NRCs Action Matrix.

Therefore, we plan to conduct a supplemental inspection using Inspection Procedure 95001, Inspection for One or Two White Inputs in a Strategic Performance Area, when your staff has notified us of your readiness for this inspection. This inspection procedure is conducted to provide assurance that the root cause and contributing causes of risk significant performance issues are understood, the extent of condition is identified, and the corrective actions are sufficient to prevent recurrence.

G. S In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response, will be made available electronically for public inspection in the NRC Public Document Room or from the NRCs document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.

Sincerely,

/RA/

Samuel J. Collins Regional Administrator

Docket No.: 50-443 License No.: NPF-86

Enclosures:

1. Notice of Violation 2. NRC Basis for Final Significance Determination

cc w/encl: Distribution via ListServ

ML093160351

"C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No Copy

OFFICE

RI/ORA

RI/DRP

RI/DRS

RI/DRP

RI/DRP

NAME

ADeFrancisco

  • WRaymond CCahill ABurritt

JClifford

DATE

10/30 /09

10/29/09 10/ 26 /09 10/ 30 /09

11/03/09

OFFICE

RI/DRS

RI/ORA

RI/ORA

HQ/OE

RI/ORA

NAME

DRoberts

KFarrar MB for KF DHolody

    • GBowman

SCollins

DATE 10/29 /09

10/28 /09 11/ 04 /09 11/05/09

11/08/09

G. S Distribution w/encl: (via e-mail)

ADAMS (PARS)

SECY CA OEMAIL OEWEB W Borchardt, EDO B Mallett, DEDR L. Trocine, OEDO R. Zimmerman, OE B. Sosa, OE N. Hilton, OE L Lopez,OE N Hasan, OE M. Crutchley, OE G Bowman, OE E Leeds, NRR Enforcement Coordinators RII, RIII, RIV (C Evans, S. Orth, W. Jones)

C Marco, OGC E Hayden, OPA H Bell, OIG C. McCrary, OI S Titherington-Buda, OCFO M Williams, OCFO DScrenci/NSheehan, RI K Farrar, RI D Holody, RI A DeFrancisco, RI/M. McLaughlin C ODaniell, RI Region I OE Files (with concurrences)

RI ORA Mail Resource RI DRP Mail Resource S. Collins, RA M. Dapas, DRA J. Clifford, DRP D. Lew, DRP M. Gray, DRP A. Burritt, DRP L. Cline, DRP/A. Turilin, DRP W. Raymond, DRP, SRI/J. Johnson, DRP, RI E. Jacobs, DRP, OA RidsNrrODMail Resource RidsNrrPMSeabrook Resource RidsNrrDorpl1-2 Resourse H. Chernoff, NRR R. Nelson, NRR D. Egan, NRR, PM ROPreports@nrc.gov Region I Docket Room (with concurrences)

NOTICE OF VIOLATION

NextEra Energy Seabrook, LLC

Docket No. 50-443 Seabrook Station

License No. NPF-86

EA-09-145

During an inspection documented in NRC Inspection Report No. 05000443/2009007, issued on August 28, 2009, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is set forth below:

10 CFR 50 Appendix B Criterion III, Design Control, states, in part, that measures shall be established to assure that regulatory requirements and the design basis for systems and components are correctly translated into specifications and instructions. Measures shall also be established for the selection and review for suitability of application of materials and parts that are essential to the safety-related functions of the systems and components.

The Seabrook Station Design Control Manual (DCM), developed pursuant to the above to establish design control measures for safety related components, including the emergency diesel generators (EDG), states, in Chapter 2, Section 8.0, that the Maintenance Support Evaluation (MSE) is the design control measure to be implemented in support of maintenance. When preparing the MSE, the DCM requires that the design inputs and interdisciplinary review guidelines on Figures 4-1-1 through 4-1-14 shall be used to prepare and develop the design change and understand the areas impacted. DCM Figure 4-1-1, Design Inputs, and Figure 4-1-3, Independent Reviewer Guidelines, requires that the design shall consider mechanical requirements such as stresses and vibration; whether materials are suitable for the application; credible failure modes of connected equipment; and, shall account for equipment performance history.

Contrary to the above, on January 31, 2009, NextEra Energy Seabrook, LLC, completed Work Order 0821400 on the B EDG without adequately establishing measures to assure that regulatory requirements and the design basis for systems and components were correctly translated into specifications and instructions. Specifically, design change 08MSE211, implemented by the Work Order, to modify and repair a two bolt flange (joint JTR005) on the B EDG right bank turbocharger, did not adequately: (1) control welding stresses during repair, assure post weld flange alignment was acceptable, or address the impacts of known vibrations on flange performance and gasket compression; (2)

address the suitability of gasket materials relative to flange specific conditions (cupping and bowing); and, (3) consider the flange performance history and potential failures to account for equipment performance history and credible failure modes of connected equipment. As a result, the B EDG turbocharger flange JTR005 failed during B EDG operation on February 25, 2009, causing a rapid loss of jacket cooling water and the EDG being declared inoperable.

This violation is associated with a White finding.

Pursuant to the provisions of 10 CFR 2.201, NextEra Energy Seabrook, LLC, is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional Administrator, Region I, and a copy to the NRC Resident Inspector at the facility that is the

Notice of Violation

subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation; EA-09-145" and should include: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time. If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

Because your response will be publicly available in the NRC Public Document Room or from the NRCs document system (ADAMS), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be made available without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days.

Dated this 12th day of November 2009

Enclosure 2: Evaluation of the Licensees Positions, NRCs Response, and the NRC Final Significance Determination

On September 30, 2009, the NRC held a Regulatory Conference with representatives of NextEra Energy Seabrook, LLC (NextEra) Seabrook Station, to discuss the significance of a finding involving the failure to establish adequate design control measures to modify a cooling water flange on the B emergency diesel generator (EDG). The inadequate modification led to the development of a leak from the B EDG during a routine test on February 25, 2009, requiring operators to secure the B EDG without completing the test.

At the conference, NextEra staff highlighted differences between the NRCs risk assessment of the finding, as provided in NRC Inspection Report No. 05000443/2009007, dated August 28, 2009, and NextEras risk assessment that the licensee developed subsequent to identification of the preliminary White finding. NextEra staff also provided additional information in writing to the NRC on October 9, 2009.

SUMMARY OF NEXTERAS POSITION

Specifically, NextEra staff presented information regarding the Seabrook Station Supplemental Emergency Power System (SEPS), a system comprised of two additional diesel generator sets (DGs), which are used to supply backup power to safety-related electrical loads when a safety-related EDG is out of service, or, if a safety-related EDG fails during a loss of all off-site power event. Specifically, NextEra provided an electrical analysis to demonstrate that only one of the two SEPS DGs are needed to provide the required electrical loads in a loss of off-site power event when a safety-related EDG fails, or is out of service, during the event, and that crediting this one-of-two DG mode of SEPS operation decreases the safety significance of the event to Green. In support of its position, NextEra stated that: (1) SEPS capacity, based on a one of two SEPS diesel generator success criteria, is supported by detailed analysis and vendor test data; (2) electrical analysis of the SEPS design has identified additional margin in the capability of the SEPS system, enabling a single SEPS DG to supply loss of offsite power loads without tripping; and, (3) operator actions required for successful SEPS operation do not include reduction of SEPS load.

NRC EVALUATION OF NEXTERAS POSITION

1. NextEra contended that SEPS capacity, based on a one of two SEPS diesel generator success criteria, is supported by detailed analysis and vendor test data.

The NRC determined that NextEra did not demonstrate acceptable results using one of two SEPS for some low probability events, such as both SEPS DGs starting and one subsequently failing (failure-to-run scenario). To support its demonstration, NextEra relied on a sensitivity analysis in its evaluation of the one of two SEPS success criteria.

NextEras analysis of the SEPS failure modes and reliance on a sensitivity analysis is inconsistent with the standard modeling of equipment failures in the Seabrook Probabilistic Risk Assessment (PRA). Further, the NextEra analysis was performed using generic diesel failure data, and the NRC also concluded that the application of generic class 1E EDG failures, operating within their design, is not representative of the failure rates that would be anticipated in a single SEPS operating at 108% of its design with degraded frequency and voltage. The NRC further determined that NextEras evaluation did not consider the residual heat removal and cooling tower pump test

Enclosure 2

scenarios impacting loading on the diesels. Regulatory Guide 1.200 states that success criteria analysis as it relates to PRA determines the minimum requirements for each function and ultimately the systems used to perform the functions necessary to prevent core damage given an initiating event. Therefore, the NRC concludes that the only mode of SEPS operation that has been adequately demonstrated by design, factory and on-site testing is based on a two of two SEPS diesel generator success criteria, not a one of two criteria as described in NextEras evaluation.

2. NextEra contended that electrical analysis of the SEPS design has identified sufficient margin in the capability of the SEPS system, enabling a single SEPS DG to supply loss of offsite power loads without tripping.

The NRC considered that one of two SEPS appears to work by NextEras engineering analysis. However, the engine overload condition combined with narrow margins to operating limits and protection settings, presents uncertainty in engine and support system performance, such that successful single SEPS operation cannot be reasonably assured. Most significantly, the NRC determined that there is very little margin between the calculated electrical current level for the predicted EDG overload condition and the over current setpoint for the DG output breaker. In addition, NextEras evaluation for the single SEPS was based on factory tests of a similar engine and it did not provide sufficient information to validate that the data from the similar engine was applicable to the Seabrook SEPS units. Therefore, the NRC concludes that adequate additional margin is not supported by the electrical or risk analysis.

3. NextEra contended that operator actions required for successful SEPS operation do not include reduction of SEPS load.

Regarding the role of operators in the risk evaluation, the NRC acknowledges the NextEra analysis of operator actions to successfully assure operation using a single SEPS engine. However, while the path through emergency procedure ECA 0.0 appears to demonstrate success under tightly constrained assumptions, there remains some question as to what operators would do if placed in the postulated scenario. The EDG fault setting report provided by NextEra for the overload condition includes many fault effects that were not addressed by NextEras analysis. The NRC continues to question whether: (1) these faults would prompt operators to take actions that would negatively impact the success of a single SEPS engine run; and, (2) the operators would reduce EDG loading per the procedure or remove needed loads. Therefore, the NRC notes that uncertainty in operator actions combined with the limited guidance provided in the available procedures makes it difficult to assure success on a single SEPS engine.

In summary, the NRC concluded that NextEra failed to demonstrate success for one of two SEPS during some events as appropriate. The NRC has also determined that NextEras proposal to use single SEPS success criteria, with predicted low margins to the equipment operating limits and extrapolated data based on engineering judgment, represents a loss in defense-in-depth. Additionally, the NRC determined that uncertainty in operator actions for SEPS load reduction is a concern. In conclusion, the NRC finds that continued use of the two-out-of-two SEPS engines recovery criteria is appropriate in the risk assessment; this results in the estimate of the annualized incremental core damage probability is 2.27E-6 for Unit 1, a low to moderate safety significance, and therefore, a White finding.

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