IR 05000443/2011010

Draft Letter from C. Miller, Region I to P. Freeman, Site Vice President, North Region, NextEra Energy Seabrook, LLC, Seabrook Station - NRC Inspection Report 05000443-11-010

ML12174A232
Person / Time
Site:	Seabrook
Issue date:	01/20/2012
From:	Christopher Miller Division of Reactor Safety I
To:	Freeman P NextEra Energy Seabrook
References
FOIA/PA-2012-0119 IR-11-010
Download: ML12174A232 (23)

Text

Inspector and Tech Reviewer Areas to focus on:1. Cover letter messages and request2. Executive Summary (ES) for appropriateness and need3. Length of Open URI section 40A5.24. Summary of POD assumptions issues from TIA in ES and 40A55. What to address in plans is only in cover letter (management integration of inspector observationsto TIA)6. No immediate safety concerns in ES only.7. List of ACRONYMS and Reference list ???????Mr. Paul FreemanSite Vice President, North RegionSeabrook Nuclear Power PlantNextEra Energy Seabrook, LLCc/o Mr. Michael O'KeefeP.O. Box 300Seabrook, NH 03874

SUBJECT: SEABROOK STATION -NRC INSPECTION REPORT 05000443/2011010

Dear Mr. Freeman:

On January 20, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat Seabrook Station. The enclosed inspection report documents the inspection results, whichwere discussed on January 20th with you and other members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.The inspectors reviewed selected procedures and records, observed activities, and interviewedpersonnel. The focus of this inspection was activities surrounding your development actionsrelated to the Alkali-Silica Reaction (ASR) problem occurring in safety related structures andother structures of regulatory importance (covered by the maintenance rule). In particular, wereviewed your Prompt Operability Determinations for certain structures based on best availableinformation. At the beginning of the inspection report period, we noted some areas that stillneeded to be addressed based on available information and NextEra satisfactorily addressedthem with revisions to the documents.On January 20, 2011 a final exit meeting was conducted and lead by Mr. Richard J. Conte,Chief Engineering Branch No. 1 of my staff. During the meeting, my staff summarized thechange in status of the new findings and our plans to issue a Task Interface Agreementbetween Region I and the Office of Nuclear Reactor Regulation simultaneously with this report.The TIA was placed in the public document room (ADAMS Accession No. MLXXXXXXX). Thepurpose of the Task Interface Agreement was for the NRR staff to identify the review criteria inevaluating the operability determination for the "B" Electrical Tunnel affected by ASR (part of theControl Building) in assistance to the Region I staff by addressing questions we had on thematter.Also on January 201h, we focused on and summarized observations on your plans with respectto the unwritten assumptions in your operability determinations. The NRC staff noted that these I determinations listed no assumptions in the applicable sections and that the design basis codeACI 318-1971 was based on empirical data for determining certain parameters that were a partof the design bases. Also Poison's ratio on concrete cores were not being tested or evaluatedand this ratio was in the UFSAR. The assumption of this empirical data was that therelationships were for ASR free concrete. Specific areas for which your plans do not addressunwritten assumptions being made in the prompt operability determinations were list in section40A5.2In consultation with our technical reviewers in headquarters and to address the currentshortcomings on unwritten assumptions for your operability determinations, we have determinethat your plans do not sufficiently provide information related to: 1) condition assessment(extent and characterization); 2) cause of the ASR as it impacts current degradation andoperability; 3) estimate of expansion to date and current expansion rate; 4) interim structuralassessment as it impacts current operability vs. longer term structural assessment; and, longerterm monitoring ensure operability in the near future vs. longer term of the duration of thelicense (1-2 years vs. longer); and, 5) short term mitigation or needed remedial actions. This isin distinction to your overall comprehensive plan for the problem.Accordingly, we request that you provide your plans to address the above issues within 30 daysof the date of this inspection report. We noted that, from the exit meeting of January 20th youhave agreed to this request and to review the report in 15 days and let us know of your plans tohonor our request or identify the need for a management meeting. We further request that,should a management meeting be needed on these issues, it should be conducted within 30days of the date of this report and a final response time will be negotiated at the managementmeeting. If your root cause evaluation scheduled for Feb. 2012 and the associate correctiveaction plan for this significant condition adverse to quality addresses the above, please usethem to respond to our request.Also, the report documents two NRC-identified findings of very low significance (Green) thatwere determined to involve a violation of NRC requirements. Because of the very low safetysignificance, and because they are entered into your corrective action program, the NRC istreating these findings as Non-cited Violations, consistent with Section 2.3.2 of the NRCEnforcement Policy. If you contest any non-cited violations in this report, you should provide aresponse within 30 days of the date of this inspection report, with the basis for your denial, tothe 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 NRCResident Inspector at Seabrook Station. In addition, if you disagree with the cross-cuttingaspect assigned to any finding in this report, you should provide a response within 30 days ofthe date of this inspection report, with the basis for your disagreement, to the RegionalAdministrator, Region I, and the NRC Resident Inspector at Seabrook.

P. Freeman2In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any), will be available electronically for public inspection in theNRC Public Document Room or from the Publicly Available Records component of the NRC'sdocument system (ADAMS). ADAMS is accessible from the NRC Web site athttp://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,Christopher G. Miller, DirectorDivision of Reactor SafetyDocket No.: 50-443License No.: NPF-86

Enclosure:

Inspection Report No. 05000443/201110

w/Attachment:

Supplemental Informationcc w/encl: Distribution via ListServ P. Freeman2In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and itsenclosure, and your response (if any), will be available electronically for public inspection in theNRC Public Document Room or from the Publicly Available Records (PARS) component ofNRC's document system (ADAMS). ADAMS is accessible from the NRC Web site athttp://www.nrc.qov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,Christopher Miller, DirectorDivision of Reactor SafetyDocket No.: 50-443License No.: NPF-86

Enclosure:

Inspection Report No. 05000443/201110

w/Attachment:

Supplemental Informationcc w/encl: Distribution via ListServDistribution w/encl: (via e-mail)W. Dean, RAD. Lew, DRAD. Roberts, DRPD. Ayres, DRPC. Miller, DRSP. Wilson, DRSA. Burritt, DRPL. Cline, DRPA. Turilin, DRPR. Montgomery, DRPW. Raymond, DRP, SRIJ. Johnson, DRP, RIA. Cass, DRP, Resident OAM. Franke, RI, OEDORidsNrrPMSeabrook ResourceRidsNrrDorlLpll-2 ResourceROPreports ResourceL. Pinkham, DRSR. Conte, DRSM. Modes, DRSSUNSI Review Complete: (Reviewer's Initials)DOCUMENT NAME: G:\DRS\Engineering Branch 1\-- MModes\20111213 050442_201110Seabrook IP711158A ASR Follow UP.docxAfter declaring this document "An Official Agency Record" it will be released to the Public.To receive a copy of this document, indicate in the box: "C" = Copy withoutattachment/enclosure "E" = Copy with attachment/enclosure "N" = No copyOFFICE RI/DRSI RI/DRS RI/DRP RI/DRS RI/DRSNAME SChaudhary MModes/ ABurritt/ RConte/ CMiller/DATE 1/ /12 1/ /12 12/ /12 12/ /12 12/ /1201FAIl-E9DCQ Docket No.:License No.:Report No.:Licensee:Facility:Location:Dates:Inspectors:Accompanied by:Approved by:U.S. NUCLEAR REGULATORY COMMISSIONREGION I50-443NPF-8605000443/2011010NextEra Energy Seabrook, LLCSeabrook StationSeabrook, NH 03874September 25 -September 30,November 15-17, 2011 (Illinois)November 28-29, 2011January 20, 2012 (Conference Call)M. Modes, Senior Reactor Inspector, Region IS. Chaudhary, Reactor Inspector, Region IW. Raymond, Senior Resident Inspector, SeabrookAtif Shaikh, Reactor Inspector, Region IIIA. Sheikh, Senior Structural Engineer, NRRG. Thomas, Structural Engineer, NRRRichard J. Conte, ChiefEngineering Branch 1Division of Reactor SafetyEnclosure

SUMMARY OF FINDINGS

IR 05000443/2011010; 9/25/2011 -12/2/2011; Seabrook Station (IP 7111115 and IP7111117).This report covers an inspection by regional inspectors, and resident staff; with assistance fromNRR structural specialists. Two Green findings were identified. The significance of mostfindings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter(IMC) 0609, "Significance Determination Process". The cross-cutting aspects for the findingswere determined using IMC 0310, "Components Within Cross-Cutting Areas." Findings forwhich the Significance Determination Process does not apply may be Green, or be assigned aseverity level after NRC management review. The NRC's program for overseeing the safeoperation of commercial nuclear power reactors is described in NUREG-1649, "ReactorOversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding regarding NextEra's operability determinations forCategory I structures with a reduced concrete modulus of elasticity caused by alkali-silicareaction in the concrete. NextEra did not adequately evaluate with available information theeffects of the reduced concrete modulus with respect to key aspects of structural design asdescribed in the Updated Final Safety Analysis Report (UFSAR). Specifically, NextEra did notinitially fully evaluate the effects of the reduced modulus on the dynamic response of Category Istructures to seismic events relative to global response; the changes in the structural naturalfrequency; and, the effects on attached systems, components and anchors. Further, NextEradid not adequately evaluate the adequacy of shear capacity of electric tunnel concretewalls without shear reinforcement to resist lateral forces during seismic events.The failure to fully evaluate the degraded and nonconforming concrete modulus condition asrequired by procedure EN-AA-203-1 001 was a performance deficiency. The performancedeficiency was associated with the Mitigating Systems cornerstone and was determined to bemore than minor based on a comparison with Appendix E.3.i of IMC 0612 because it adverselyaffected the cornerstone objective to ensure the availability, reliability and capability of systemsthat respond to initiating events in order to prevent core damage. The issue was evaluatedusing IMC 0609, "Significance Determination Process" (SDP), and was determined to be of verylow safety significance (Green). The finding had a cross cutting aspect in the area of problemidentification and resolution, P.1(c), related to ensuring that issues potentially impacting nuclearsafety are thoroughly evaluated. Specifically, NextEra did not thoroughly evaluate conditionsadverse to quality, including evaluating the effects of the reduced concrete modulus for impacton operability of the affected structures. (Section 1 R1 5)Severity Level IV. The inspectors identified a non-cited violation of 10 CFR 50.59(d)(1) becauseNextEra did not provide an evaluation that adequately documented why implementing a designchange to address an identified reduction in the concrete modulus of elasticity for severalCategory I concrete structures, did not present a more than minimal increase in the likelihood ofthe occurrence of a malfunction of a structure, system, or component (SSC) important to safetypreviously evaluated in the UFSAR. Specifically, NextEra issued EC272057 on April 25, 2011,to address reduced concrete modulus of elasticity in the control building electric tunnel and theEnclosure tcontainment enclosure building, but did not complete a 10 CFR 50.59 evaluation prior toimplementing changes to the facility as described in the modification.The failure to evaluate changes to the facility as described in EC272057 was contrary to10 CPR 50.59(d)(1) and was a performance deficiency warranting a significance evaluation inaccordance with the NRC Enforcement Manual for Traditional Enforcement and InspectionManual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "IssueDisposition Screening." The violation was determined to be more than minor in accordance withIMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because theinspector could not reasonably determine that the changes would not have ultimately requiredprior NRC approval. The finding was evaluated using the SDP in accordance with IMC 0609,"Significance Determination Process," and determined to be potentially risk significant due to adesign deficiency confirmed not to result in a loss of operability. In accordance with Section6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as Severity Level IVbecause the resulting changes were evaluated by the SDP as having very low safetysignificance (Green). The finding had a cross cutting aspect in the area of human performance-work practices, H.4(b), because NextEra personnel did not follow procedures. Specifically,NextEra personnel did not follow the requirements of Section 5.2.2 of the 5059 ResourceManual when preparing the 5059 screen for EC272057. (Section 1R17)Executive SummaryThe focus of this inspection was activities surrounding development actions related to the Alkali-Silica Reaction (ASR) problem occurring in safety related structures and other structures ofregulatory importance (covered by the maintenance rule). In particular, NRC staff reviewed thePrompt Operability Determinations for certain structures based on best available information. Atthe beginning of the inspection report period, we noted some areas that still needed to beaddressed based on available information and these issues were satisfactorily addressed withrevisions to the documents.Prior NRC review of this area was documented in the following Inspection Reports05000443/2010004, 2010005, 2011002, 2011003, and 2011007 and the results weresummarized in the Background section of this report.One unresolved items was closed and sufficient information was obtained in order to determinethe performance deficiency -50.59 screening to accept as-is conditions for reduced modulus ofelasticity for certain safety related structures. Another unresolved item was left open -PromptOperability Determinations for certain safety related structure with the ASR problem. Anotherfinding of very low safety significance was determined in the course of the operabilitydetermination review. Both findings were summarized above.The inspectors observed that NextEra plans for the ASR problem were not addressing unwrittenassumptions in the operability determinations. These operability determinations listed noassumptions in the applicable sections. The design basis code ACI 318-1971 was based onempirical data for determining certain parameters that were a part of the design bases.Poisson's ratio on concrete cores were not being tested or evaluated and this ratio was in theUFSAR. The assumption of this empirical data was that the relationships were for ASR freeconcrete. Specific areas for which the plans do not address unwritten assumptions being madein the prompt operability determinations were list in section 40A5.2

To address the above and in conjunction with the NRR technical reviewers, the inspectors notedthat the current plans were not finalized and what was existing did not specifically address theunwritten assumptions in the following areas: 1) condition assessment (extent andcharacterization); 2) cause of the ASR as it impacts current degradation and operability; 3)estimate of expansion to date and current expansion rate; 4) interim structural assessment as itimpacts current operability vs. longer term structural assessment; and, longer term monitoringensure operability in the near future vs. longer term of the duration of the license (1-2 years vs.longer); and, 5) short term mitigation or needed remedial actions.The NRC staff summarized why there is no immediate safety concerns for the existingconditions: walkdowns confirm no significant degradation, no visual evidence of distortion norvisible evidence of rebar corrosion; overall evidence of sufficient stiffness remaining; Noappreciable evidence of cracking where found, in isolated sections of the wall; degradationappears to be localized; We know from best available research that the ASR rate slowlyprogresses and there is some evidence that it has progressed to a plateau but it needs to beconfirmed by testing; parameters obtained for compressive strength and modulus of elasticityindicate robust design to strength of the concrete poured (4K psi concrete used in buildings onlyneeding 3K psi concrete and no significant negative shift on seismic analysis.On January 20, 2011 a final exit meeting was conducted. During the meeting, NRC staffsummarized the change in status of the new findings and plans to issue a Task InterfaceAgreement between Region I and the Office of Nuclear Reactor Regulation simultaneously withthis report. The TIA was placed in the public document room (ADAMS Accession No.MLXXXXXXX). The purpose of the Task Interface Agreement was for the NRR staff to identifythe review criteria in evaluating the operability determination for the "B" Electrical Tunnelaffected by ASTR (part of the Control Building) in assistance to the Region I staff by addressingquestions related to the problem.Enclosure 2

REPORT DETAILS

BackgroundIn June 2009, NextEra conducted walkdowns of structures within the scope of license renewalas part of license renewal application preparations. In June, 2010 the License RenewalApplication (LRA) was received by the agency. In October 2010, the License Renewal Auditresults noted the alkali-silica reaction (ASR) problem and pointed to need for a good number ofrequests for additional information in this area since the issue was newly discovered for the site(noted as a area to address in the GALL, Generic Aging Lessons Learned, Revision 1). In theFall of 2010, NextEra performed an Immediate and Prompt Operability Review (POD) based oncore samples taken in Control Building in August 2010. In November 2010, Inspection Report05000443/2010004 and, in February 2011, Inspection Report 2010005, followed developmentsin this area from an operability viewpoint. In these reports, no findings were noted sincelaboratory results determined compressive strength and modulus of elasticity met UFSARvalues (degradation into reserve strength not design margin). In May 2011, Inspection Report05000443/2011002 identified two noncited violations of very low safety significance in thestructures monitoring area with respect to the maintenance rule (10 CFR 50.65 a(1 ) and b(2)).Also in May 2011, License Renewal Inspection (IP71002) Report 05000443/2011007 had anoverall result: "Except for Structures Monitoring Program, results support a reasonableassurance determination for license renewal."As the year progressed, NextEra continued to identify and characterize the below-gradestructures at Seabrook having experienced groundwater infiltration and a resultant reduction inconcrete material properties. NextEra determined the degraded concrete condition was mostlikely due to distress from ASR in the concrete. ASR is a chemical reaction in concrete overtime between the alkaline cement paste and reactive non-crystalline silica which is found incommon coarse aggregates. The reaction only occurs in the presence of water and forms a gelthat expands, forming micro-cracks that change the strength of concrete system. NextEra is inthe discovery phase of condition assessment along with extent of condition reviews in order tosupport other building operability determinations and a plan to conduct an engineeringevaluation which would appear to constitute a long term operability review (40 year licenseterm) along with mitigation and monitoring measures.The appearance of ASR degradation in safety related concrete structures was the first noted inthe nuclear industry in the United States and could be significant related to preservation ofreserve capacity with building design loads as reflected in the current licensing basis duringnormal operations and aging management over the period of extended plant operation. Whilethe problem was noted as an aging effect in a license renewal topical report, the appearance ofASR reflected a newly discovered aging effect at Seabrook that needs to be managed forlicense renewal.In August 2011, Inspection Report 05000443/2011003 addressed a noncited violation of verylow safety significance related to the untimely Initial and Prompt ODs for results on extent ofcondition review for other buildings affected by ASR. Two unresolved items was also opened,one dealing with a potential inadequate screen in accordance with 10 CFR 50.59 for acceptingthe reduce parameter found on compressive strength and modulus of elasticity for the "B"Electrical Tunnel and the Containment Enclosure Building. The other unresolved dealt with theopen prompt operability determinations associated with the "B" Electrical Tunnel and theEnclosure 3building subjected to an extent of conditions review. In August of 2011 a Task InterfaceAgreement (TIA) was issued between Region I and the Office of Nuclear Reactor Regulation onthe ASR issue. The response to the TIA is publicly available (ADAMS Accession No.MLxxxxxxxxx). The purpose of this inspection was to followup on the unresolved items and towork with the NRR technical reviewers in developing the answers to the questions posed in theTIA.1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1 R15 Operability Determinations and Functionality Assessments (71111.15 -0 samples)a. Inspection ScopeAs a part of the review of an unresolved item (see section 4OA5.2), the inspectors, inconjunction with technical reviewers from the Office of Nuclear Reactor Regulation,reviewed the adequacy of operability determinations for the below grade concrete wallsof seismic Category I buildings affected by alkali-silica reaction. The review focused onthe specific design for the buildings affected by an alkali silica reaction and operabilitywas addressed by two prompt operability determinations, one, to address the controlbuilding and the other to address the following extent of conditions with five otherbuilding/structures: Containment Enclosure Building, Emergency Diesel GeneratorBuildings (fuel oil rooms), Residual Heat Removal Equipment Vaults, EmergencyFeedwater Pump House, and the Radiological Control Access (RCA) tunnel.b. FindingsInadequate Operability DeterminationsIntroduction. In April 2011, NextEra identified a degraded and nonconforming conditionrelated to reduced modulus of elasticity for buildings housing safety related equipment,but did not thoroughly evaluate potential impacts in accordance with the requirements inNextEra procedure EN-AA-203. Specifically, the evaluation did not consider the affect ofthe reduced modulus on: overall stiffness and resulting seismic response, concretematerial property changes affecting natural frequency and therefore the amplitude of theseismic response, and the performance of systems and components attached to theaffected structures during seismic events.Description. In April 2011 NextEra determined that certain below grade concrete wallswere affected by alkali-silica reaction (ASR). The analysis of concrete cores showed areduced concrete modulus of elasticity in the control building / electric tunnel(AR581434581434, containment enclosure building (AR1644074) and three other seismicCategory I buildings (AR1664399). The lowest measured modulus was about 40% lessthan the design value of 3.62E+03 ksi.NextEra completed prompt operability determinations for the affected Category Iconcrete structures (reference ARs 581434, 1644074 and 1664399) as required byNextEra Procedure EN-AA-203-1001, "Operability Determinations/FunctionalEnclosure 4Assessments." In accordance with the procedure EN-AA-203-1001, a POD mustinclude: identification of current licensing basis functions and performance requirementsas listed in the UFSAR; identification of the minimum design basis values necessary tosatisfy the SSC design basis safety functions; and evaluation of the effects of thedegraded condition on the ability of the SSC to meet its specified function andperformance requirements.During the week of September 28, 2011, the inspectors in conjunction with technicalreviewers from the Office of Nuclear Reactor Regulation reviewed NextEra's competedPODs for the ASR-affected Category I concrete structures. It was determined that theevaluations were not complete with respect to available information since NextEra didnot evaluate the degraded condition with respect to key aspects of the structure designas described in UFSAR. Specifically, the initial PODs did not adequately address theeffects of the reduced modulus of elasticity in the following areas:1. As a result of further review of the design bases, the NRC staff determined that thewalls below grade in the Control Building 'B' Electrical Tunnel do not contain shearreinforcement to resist dynamic lateral forces acting on the wall during a design basisearthquake. The design of the wall intentionally depends on the strength of theconcrete alone to resist these dynamic forces and this information was not evaluatedfor the degraded conditions. The same was true for the Diesel Generator Building.The modulus of elasticity of concrete was a function of concrete compressivestrength which is generally higher in the as-cast condition than assumed in thedesign. The concrete used in construction of Seabrook structures was formulated tohave a design strength greater than 3000 psi. However, as stated in the SeabrookUpdated Final Safety Analysis Report, Revision 12, Section 3.8, "While variability inconcrete modulus has no significant effect on structural design, it influencesstructural stiffness and natural frequency, and, subsequently, the amplified responsespectra of the seismic analysis."2. Changes in the modulus of elasticity affect the material concrete properties and,therefore, the natural frequency of the structure, which affects how the buildings areanalyzed in the seismic analysis. NRC reviews determined that the promptoperability determinations addressed the dynamic response of the structures in aqualitative manner noting that the ASR impacted walls are below grade and thestructural loadings would be governed by the ground response spectra assumed inthe original design. The initial evaluations did not sufficiently address the impact ofthe reduced modulus on the structure natural frequencies in a quantitative manner tovalidate that the structure response would remain rigid or that there would be aamplification of the response. Specifically, the initial evaluation did not verify there.would be no amplification of the motions beyond those in a ground response spectraas assumed in the seismic analysis per UFSAR 3.7(B).2.3. The initial POD addressed the effects of the reduced modulus on componentshoused within the structures, such as pipe supports, cable trays and componentsupport anchors. NRC review determined that the initial evaluations addressed theimpacts on the internal components in a qualitative manner, but did not verify theequipment performance would remain bounded by the analysis in the original designas described in UFSAR 3.7 and 3.8. Specifically, the initial evaluation did not verifyEnclosure IL5there would be no amplification of the motions beyond those in a ground responsespectra as assumed in the seismic analysis per UFSAR 3.7.(B).2. Further, the initialevaluation did not evaluate the potential impacts on anchor or wall shear capacitiescaused by ASR induced changes in material properties beyond that allowed for asdescribed in UFSAR 3.8.In response to the NRC-identified issues, NextEra completed' additional evaluations thatdetermined the structures and other affected systems and components remainedfunctional for design basis conditions. On October 14, 2011, NextEra completedCalculation C-S-1-10163, and revisions to the PODs for AR581434581434(CB/ ET) andAR1664399 (CEB and other Category I Structures). The NRC determined thatNextEra's additional analysis and revisions to the PODs adequately addressed theconcerns discussed above. Specifically, the analysis confirmed a minor impact on theoverall response of the structure during a seismic event, a small effect on the structure'snatural frequencies that results in no appreciable amplification of the ground responseduring a seismic event, and no impact on the ability of the equipment anchors to performtheir function due to the quality of the concrete and construction methods used.Analysis. The inspectors determined that not following a self imposed standard, notcompletely analyzing the effects of the reduced modulus of elasticity on Category Istructures based on available information, per procedure EN-AA-203-1001 was aperformance deficiency. Specifically, because the reduced modulus affected thedynamic response of Category I structures to seismic events relative to global response,changes in natural frequency and the effects on attached systems and components,procedure EN-AA-203-1 001 required that these impacts be evaluated as part of theprompt operability determination. This performance deficiency was associated with theMitigating Systems cornerstone and was determined to be more than minor because,based on a comparison with Examples 3.i of Appendix E of IMC 0612, it adverselyaffected the cornerstone objective to ensure the availability, reliability and capability ofsystems that respond to initiating events in order to prevent core damage. Specifically,the effects of the reduced modulus of elasticity on the dynamic response of Category Istructures to seismic loading required further evaluation to demonstrate the structuresand enclosed systems remained functional as described in the licensing and designbases. The issue was evaluated using IMC 0609, "Significance Determination Process"(SDP), and was determined to be of very low safety significance (Green). Specifically,when evaluated under IMC 0609, Attachment 4, the performance deficiency was not adesign or qualification deficiency resulting in an actual loss of safety function, was not aloss of a barrier function, and was not potentially risk significant for external events. Thefinding had a cross cutting aspect in the area of problem identification and resolution,P.1(c), related to ensuring that issues potentially impacting nuclear safety are thoroughlyevaluated. Specifically, NextEra did not thoroughly evaluate conditions adverse toquality, including evaluating the effects of the reduced concrete modulus for impact onoperability of the affected structures.Enforcement. Because this finding does not involve a violation and has very low safetysignificance, it is identified as FIN 05000443/2011-10-01, Incomplete OperabilityDetermination for Degraded Concrete Structures Housing Safety-Related Equipmentbased on available information.Enclosure 61R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications(71111.17- 0 sample)a. Inspection ScopeAs a part of the review of an unresolved item (see section 40A5.1), the inspectorreviewed EC272057, dated April 25, 2011, for adequacy in which the EC was a designchange to address reduced concrete modulus of elasticity in the control building electrictunnel and the containment enclosure building. The review was to determine if only a50.59 screening was acceptable to accept "as-is" conditions for this concrete materialproperty which was degraded from the design bases as reflected in the UFSARapparently due to the ASR problem.b. FindingInadequate 50.59 Screen Evaluation for EC272057Introduction: A Severity Level IV NCV of 10 CFR 50.59(d)(1), "Changes, Tests, andExperiments," was identified because NextEra did not provide an evaluation thatadequately documented why implementing a design change to address an identifiedreduction in the concrete modulus of elasticity for several Category I concrete structures,did not present a more than minimal increase in the likelihood of the occurrence of amalfunction of a structure, system, or component (SSC) important to safety previouslyevaluated in the updated safety analysis report (USAR). Specifically, NextEra issuedEC272057 on April 25, 2011, to address reduced concrete modulus of elasticity in thecontrol building electric tunnel and the containment enclosure building, but did notcomplete a 10 CFR 50.59 evaluation prior to implementing changes to the facility asdescribed in the modification.Description: NextEra determined that certain below grade concrete walls were affectedby alkali-silica reaction (ASR). The analysis of concrete cores taken from ASR affectedareas, showed a reduced concrete modulus of elasticity in the control building / electrictunnel (AR581434581434, containment enclosure building (AR1644074) and four other seismicCategory I buildings (AR1664399). The lowest measured modulus was about 40% lessthan the design value of 3.62E+03 ksi.Enclosure 7On April 25, 2011, NextEra issued EC272057, "Concrete Modulus of ElasticityEvaluation," to address the reduced modulus. EC272057 dispositioned the degradedcondition as "use-as-is," and incorporated the degraded condition into the design basis.In a safety evaluation screen for EC272057, NextEra concluded the change did notrequire a complete evaluation per 50.59(c)(2) because adequate design margin existedand there was no adverse affect on an UFSAR described design function.10 CFR 50.59 requires licensees to evaluate whether NRC approval is required forproposed changes to the facility. The Seabrook 5059 Resource Manual defines theprocess for completing 10 CFR 50.59 evaluations for changes, tests and experimentscompleted at Seabrook. It includes a screening process that defines criteria used todetermine whether a full 10 CFR 50.59 evaluation must be performed for eachapplicable change, test or experiment. NextEra screened EC272057 in accordance withthe guidance in the 5059 Resource Manual and concluded that the change did notrequire a full evaluation per 50.59(c)(2) because adequate design margin existed andthere were no adverse affects on the UFSAR described design functions. Theinspectors reviewed EC272057 and determined that NextEra's 50.59 Screen forEC272057 did not correctly address "adverse affects" as described in Section 5.2.2 ofthe 5059 Resource Manual. The concrete modulus of elasticity is a design valuespecified in both the Seabrook UFSAR and the ACI 318 Building Code for the applicableplant structures. The reduced modulus of elasticity caused by the ASR occurring inimpacted concrete walls has the potential to affect the flexural capacity and dynamicresponse of the impacted structures. Therefore, the inspectors determined that therewas sufficient evidence that the reduction in the modulus of elasticity was caused by theASR and it was an "adverse affect" as described in Section 5.2.2 of the 5059 ResourceManual and thus required further evaluation per 50.59(c)(2). The additional evaluationrequired by 10 CFR 50.59 d(1) was needed to at least ensure that 10 CFR 50.59 c(2) (ii)and (iv) criteria were not met and, therefore, there would be a need for a licenseamendment per 10 CFR 50.90. The criterion c(2)(ii) and iv) deal with the changeresulting in more than minimal increase in the likelihood of occurrence or in theconsequences of a malfunction of an SSC important to safety previously evaluated in theUFSAR. In response to the inspectors concerns regarding the adequacy of the 50.59evaluation, NextEra rescinded the design change EC272057 from the design basis onSeptember 22, 2011, and initiated additional evaluations of the ASR affected structures.On October 14, 2011, NextEra issued additional information to support of its engineeringevaluation of the ASR impacted structures, including Calculation C-S-1 -10163, thePrompt Operability Determination for AR581434581434(CB/ ET) Revision 1, and the PromptOperability Determination for AR1664399 (CEB and other Category I Structures)Revision 1. The reduced modulus caused the concrete to have increased flexure, butthe results of NextEra's additional evaluations confirmed that the reduction in capacitywas minimal and the resultant stresses on the steel and concrete caused by the ASRdegradation remained below the design stress limits with margin. Similarly, the affect ofthe reduced modulus also reduced the natural frequency of the structures, but theadditional evaluation again determined that the shift in natural frequency was minimaland remained well above the ground response peak frequency range such that theresponse of the structures remained rigid. Therefore, although the effect of the ASR onthe impacted walls was to reduce the design modulus parameter, the structural integrityEnclosure remained fully intact under all design loads, and the buildings remained operable.NextEra actions continue to review the degraded concrete issue within the correctiveaction program, including the effects on the long term reliability of the structures. SeeSection 40A5 of this report for further NRC reviews of the revised operabilitydeterminations for ASR impacted structures.Analysis The inspectors determined that the failure to evaluate changes to the facilityas described in EC272057 was contrary to 10 CFR 50.59(d)(1) and was a performancedeficiency warranting a significance evaluation in accordance with the NRC EnforcementManual for Traditional Enforcement and Inspection Manual Chapter (IMC) 0612, "PowerReactor Inspection Reports," Appendix B, "Issue Disposition Screening." The violationwas determined to be more than minor in accordance with IMC 0612, "Power ReactorInspection Reports," Appendix B, "Issue Screening," because the inspector could notreasonably determine that the changes would not have ultimately required prior NRCapproval.Violations of 10 CFR 50.59 are dispositioned using the Traditional Enforcement processinstead of the SDP because they are considered to be violations that could potentiallyimpede or impact the regulatory process. However, if possible, the underlying technicalissue is evaluated under the SDP to determine the severity of the violation. In this case,the inspector determined the finding could be evaluated using the SDP in accordancewith IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1 -Initial Screening and Characterization of Findings," Tables 3b and 4a, for the MitigatingSystems Cornerstone. The inspector answered "Yes" to Question 5 under the MitigatingSystems Cornerstone column of the Phase 1 worksheet because the inspectorconcluded that the finding screened as potentially risk significant due to a design orqualification deficiency confirmed not to result in a loss of operability or functionality. Inaccordance with Section 6.1 .d.2 of the NRC Enforcement Policy, this violation iscategorized as Severity Level IV because the resulting changes were evaluated by theSDP as having very low safety significance (Green). Further evaluation determined thatthe structures remained operable despite the degraded modulus condition. Uponremoval of EC272057 from the design basis on September 22, 2011, the issue no longerrequired an evaluation per 10 CFR 50.59(a)(2).NextEra personnel did not complete the 50.59 screen properly because theymisunderstood the guidance in the 50.59 Resource Manual regarding the need to screenin changes in design parameters which impact the design function acceptance criteria(Resource Manual Section 5.2.2). The finding had a cross cutting aspect in the area ofhuman performance -work practices, H.4(b), because NextEra personnel did not followprocedures. Specifically, NextEra personnel did not follow the requirements of Section5.2.2 of the 50.59 Resource Manual when preparing the 50.59 screen for EC272057.Enforcement Title 10 CFR 50.59, "Changes, Tests, and Experiments," Section (d)(1)states, in part, that the licensee shall maintain records of changes in the facility orprocedures, and that the records must include a written evaluation that provides thebases for the determination that the change does not require a license amendmentpursuant to paragraph 10 CFR 50.59(c)(2). Contrary to the above, from April 25 toSeptember 22, 2011, NextEra did not provide an evaluation that adequately documentedwhy the reduced concrete modulus of elasticity in Category I structures did not present aEnclosure I9more than minimal increase in the likelihood of occurrence of a malfunction of a SSCimportant to safety previously evaluated in the USAR. Because this failure to properlyevaluate a proposed change is of very low safety significance and has been entered intothe licensee's Corrective Action Program (CR1647722), this violation is being treated asan NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV05000443/2011010-02, Failure to Properly Complete a 5059 Screen for EC272057).4.

OTHER ACTIVITIES

40A5 Other Activities.1 (Closed) Unresolved Item 05000443/2011003-02, 50.59 Evaluation for AcceptingReduced Modulus of Elasticity in Certain Safety-Related Structures Affected by ASRa. Inspection ScopeThe inspector reviewed EC272057, dated April 25, 2011, for adequacy in which the ECaddressed reduced concrete modulus of elasticity in the control building electric tunneland the containment enclosure building. The review was to determine if only a 50.59screening was acceptable to accept "as-is" conditions for this concrete material propertywhich was degraded from the design bases as reflected in the UFSAR.b. Observations/FindingsThis issue was closed since the inspectors identified one Severity Level IV NCV, asdescribed in Section 1 R1 7 of this report..2 (Open) Unresolved Item 05000443/2011003-03, Open Operability Determinations forSafety-Related Structures Affected by ASRa. Inspection ScopeThe scope of this review was to update NextEra actions. to date. The inspectorsreviewed the prompt operability determination for the control building, and for the extentof conditions review for other buildings affected by the alkali silica reaction. Theinspectors utilized site records and interviews to develop the design basis for the safetyrelated structures in greater detail than summarized in section 3.8 of the Updated FinalSafety Analysis Report. Additionally, this review was assessed progress in thedevelopment of a plan and schedule to address inspection activities, in-situ andlaboratory testing to address the alkali-silica-reaction degradation with specific focus onthe Control Building as a test case for review.With respect to laboratory conditions the inspector verified: 1) organized and cleanworking area during both sample preparation (measurements and cutting) andcompression testing; 2) adequate lighting available at all times; 3) ambient roomtemperature (- 680F) observed during preparation and testing; and 4) core sampleswere adequately stored and labeled in individual bags. Particular care was taken toensure only one core was handled at any given time so as not to confuse cores duringmeasurements, cutting, and testing. With respect to equipment calibration, the inspectorEnclosure 10verified: 1) caliper (model 500-505-10, serial #0014816) calibration document andcalibration sticker on the caliper; and, 2) compression machine (model CM5000-D, serial#11005) calibration document and calibration sticker on the machine. With respect totest technician qualifications, the inspector also verified qualification records (Level 2qualification for concrete testing up to date). The inspector also reviewed the AltranCommercial Grade Dedication Plan 10-0076 Part 05 -In Field Check List -the check listwas in hand during the preparation and testing.During the week of November 28, 2011, the inspector continued to review historicaldocumentation from the construction phase of the plant, correlation between theconcrete strength value determined by the recent core samples and the original strengthvalues determined at the time of concrete placement. The licensee's projected plan andschedule for further studies and assessment of the ASR problem was discussed andreviewed with cognizant engineering and management personnel. The inspectorsreviewed the licensee's procedures for administration and control of engineering andtesting service vendors and contactors. Additionally, the inspector reviewed the resultsand documentation of IWL inspection of the containment, and the results of thelicensee's efforts in inspection and mapping of 'crazed' cracking in containment andenclosure building wall and the adequacy and validity of the documented results.b. Observations/FindinqsThe inspectors identified one finding which was addressed in section 1 R1 5 of this report.In summary of the finding, the operability determinations did not fully evaluate designinformation that was available on the potential effects of the reduced modulus ofelasticity in the following areas: overall stiffness and resulting seismic response,concrete material property changes affecting natural frequency and therefore theamplitude of the seismic response, and the performance of systems and componentsattached to the affected structures during seismic events. Procedure EN-AA-203-1001required that these impacts be evaluated as part of the prompt operability determinationbased on available information.The unresolved item was kept open because the operabilitiy determinations have notbeen finalized. Several other observations were made in order to update this item. Thereviewers noted that NextEra had engaged knowledgeable vendors, appropriateconsultants, and recognized experts for testing, analysis, and evaluation of the effects ofalkali-silica-reaction, on the serviceability and safety of the affected structures. Althoughthe NextEra plan has not been refined before the NRC staff may be able to determinethat if it meets the rigor expected of an Appendix B Quality Assurance program, apreliminary schedule of actions generally consistent with the proposed plans from thecontractors was underdevelopment. It was noted that some of the elements of an agingmanagement program had been developed and a final one was under developmentputting for putting in place, including monitoring and trending. Additional core sampleswere planned, the operability determination was to be updated as new information,analysis, and assessments became available on an as needed basis. The extent ofcondition was being comprehensively determined along with building initial assessmentsthrough the work of its consultants and contractors. NextEra was also developing andscheduling actions to determine where the alkali-silica-reaction was on the alkaliEnclosure 11depletion curve (relates to extent of potential future degradation), although the tests maytake up to two years to complete, and provide reliable data.On October 14, 2011, NextEra issued additional information in support of its engineeringevaluation of the alkali-silica-reaction impacted structures, including Calculation C-S-1-10163, the Prompt Operability Determination for AR581434581434(for the Control Building andElectrical Tunnel), Revision 1, and the Prompt Operability Determination for AR1664399(other Category I Structures), Revision 1. In C-S-1-10163, the fundamental frequencywas evaluated using the measured modulus of elasticity determined in concrete coresamples taken from the building walls. The calculation evaluated the impact of thereduced modulus (compared to the design value) on the wall stiffness with respect to theground response spectra for the Seabrook site. The building response frequency wascalculated using the principles and equations of engineering mechanics for a uniformlyloaded fixed-fixed beam model (a simple span fixed at both ends during a seismicevent). The effect of the reduced modulus was similarly evaluated to assess the impacton the natural frequency of the structures. The seismic analysis, for Seabrook describedin Updated Final Safety Analysis Report Section 3.7(B).2, was used in the design ofCategory I Structures, systems and components at Seabrook.The inspectors completed a detailed review of C-S-1-10163 and verified that thecalculation inputs were supported by plant data and the design references cited in thecalculation. No inadequacies were identified. The results of C-S-1-10163 supportedNextEra's conclusions in the revised prompt operability determinations, AR 581434581434andAR1 664399.Also, during the week of November 14, 2011, a Region III inspector reviewed laboratorytesting for compressive strength on fifteen concrete core samples taken from the controlbuilding in the October 2011 time frame. The testing was conducted at a laboratory inNorthbrook, Illinois. The scope of this review was as noted above. For the testing theweek of November 14, 2011, all 15 core samples were compression tested.Photographs were taken for all core samples prior to loading for compression test andafter fracture. Three cores had small length samples cut from them during the cuttingphase to be used by Seabrook for further petrography in the near future. Samplepreparation (capping) was done in accordance with ASTM C617. Compression testingwas done in accordance with ASTM C39. No concerns were noted with respect toquality control during all aspects of compression testing.Other observations were made during the week of November 14, 2011. Multiplelaboratory engineers, licensee engineer, and Altran engineer were involved in makingcall on fracture patterns. All but one of the obtained compressive strengths were fairlyconsistent with previous lab's results (2010, 2011 data). Core sample L5-C exhibitedhighest compressive strength of 6610 PSI whereas the previous lab's data identifiesstrength at 3950 PSI. This core compressive strength value was the only apparentoutlier amongst the data set. All 15 destroyed cores are to be shipped back to Seabrooklater today including the cut samples to be used for petrography.During the week of November 28, 2011, the IWL Examination Report for the PrimaryContainment in October 2011, recorded information related to concrete conditions toensure no unacceptable surface conditions for cracking (greater than 40 mils) and reportEnclosure 12on other conditions such spalling and discoloration conditions. Related to this reviewthe inspector noted in AR 01641413 an evaluation of crazed cracking on the exteriorsurface of the primary containment at azimuth 3150 and elevation (-)30 feet, 00 inches.While several factors were identified by NextEra in support of structural integrity of thisstructure, it was noted the continued evaluation would be done in accordance with theextent of condition review per AR 574120574120which identified the loss of concrete strengthdue to alkali-silica reaction (ASR) in other buildings noted herein. The inspector notedthat no specific cause for the crazed cracking was identified which could be due ASR orother mechanisms. Further the inspector questioned the reliance on the April 2008 10CFR 50 Appendix J, Type A Test at 49.6 psig that showed no evidence of crackinggreater than 40 mils without the use of a before and after crack mapping effort. Nounacceptable conditions were found and the extent of condition review noted above is apart of this open unresolved item.During this inspection, it was determined that NextEra and its contractor wereconducting a remodeling effort on the Containment Enclosure Building (an extent ofcondition building reviewed in AR 01664399) using current data from core sample andin-situ reviews such as crack mapping etc. The purpose of this remodeling was toconduct a seismic reanalysis to demonstrate the effects of the reduced modulus onstructural response because of the unique design of the building and because of theneed to address a global response vs. a localized response. The completion of thisreview was not expected until early 2012. NRC review determined that the initialevaluation for the CEB did not address the response of the entire structure to seismicloading comparable to the methods described in UFSAR 3.8 and how the inducedseismic stresses would shift between the concrete and the steel in adjoining sections ofthe structure. In response, NextEra pointed out that they began development of ananalytical model to reanalyze the CEB using the as-measured elastic modulus (40%reduced) applied to that ASR-impacted sections. The results of this analysis will befurther reviewed as a part of an Engineering Evaluation scheduled to be completed byMarch 2012.Notwithstanding the acceptable revised operability determination based on availableinformation, the inspectors noted that these determinations listed no assumptions in theapplicable sections. The inspectors also noted that design basis code ACI 318-1971was based on empirical data for determining certain parameters that are a part of thedesign bases. Also Poison's ratio on concrete cores are not being tested or evaluatedand this ratio was in the UFSAR. The assumption of this empirical data was that therelationships were for ASR free concrete. Further, based on a review of theimplementation schedule from contractor-submitted plans that do not have NextEraapproval as of Nov. 29th, the inspectors noted that the plans do not address the followingwhich are directly related to addressing the unwritten assumptions being made in theprompt operability determinations:1. The plans do not appear to test concrete cores for the following key designparameters from the design basis code ACI 318-1971, as for tensile and shearstrength, rebar bond strength and Poison's ratio.2. The plans do not appear to address nondestructive testing to assess the currentprogression of the ASR expansion rate before the destructive tests of the concretecores.Enclosure 133. There was a apparent lack a clear framework for concrete core sampling in thebuildings to ensure how representative the core sampling addressing the need forrandom core sampling in distinction to smart sampling on worst caseconditions/doing a bounding calculation along addressing the impact of too muchcore boring and re-grouting on the building structural integrity.4. The plans do not appear to address potential effects of other degradations from anaggressive groundwater environment along with the presence of ASR.In summary at the close of the inspection, NextEra continued to work on: their plans andimplementing schedule, building initial assessments along with evaluation results foradditional core sampling; identifying the in-situ and out-of-situ testing (concrete coresamples and nondestructive testing of concrete core samples) for the structure areasaffected by alkali-silica-reaction; need to address key design parameters for thebuildings, such as compressive strength, tensile strength, bond strength (between rebarand the concrete), modulus of elasticity and Poisson's Ratio in terms of how alkali-silica-reaction has affected the non-alkali-silica-reaction functional relationship between theseparameters per the design code ACI 318-1971; and, remodeling efforts on theContainment Enclosure Building.Overall, this area remains open pending further project and test plan development byNextEra and further NRC staff review of the final operability determinations on or aboutMarch 2012.4OA6 Meetings, Including ExitOn September 30 and December 2, 2011, the inspectors presented the interim results ofthis inspection to Mr. P. Freeman, Site Vice President, and Seabrook Station staff. Theinspectors also confirmed with NextEra that no proprietary information was retained byinspectors during the course of the inspection.On January 20, 2011 a final exit meeting was conducted and lead by Mr. Richard J.Conte, Chief Engineering Branch No. 1. Others involved in this conference are noted onthe list of contacts. During the meeting, the NRC staff's final disposition of theunresolved items and new findings were summarized. Other comments and questionswere communicated to NextEra Management with respect to the ASR problem in safetyrelated structures.Enclosure 14ATTACHMENT:

SUPPLEMENTARY INFORMATION

Enclosure

A-1ATTACHMENTSUPPLEMENTARY INFORMATIONKEY POINTS OF CONTACT

Licensee Personnel

B. Brown, Supervisor, Civil Engineering

V. Brown, Senior Licensing Analyst

K. Browne, Plant General Manager

J. Esteves, Plant Engineering

P. Freeman, Site Vice President

P. Gurney, Reactor Engineering Supervisor

M. Collins, Manager, Design Engineering

M. O'Keefe, Licensing ManagerKey Manager Participants for Teleconference of January 12, 2012NRC Staff

C. Miller, Director Division of Reactor Safety (DRS), Region I

R. Conte, Chief Engineering Branch No. 1, DRS, Region I

A. Burritt, Chief Reactor Projects Branch No. 3, Region ILIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATEDOpened:05000443/2011-010-0105000443/2011-010-02Closed:05000443/2011-003-02Updated05000443/2011-003-03Documents Reviewed:FIN Inadequate Operability Determination for DegradedConcrete Structures Housing Safety-Related EquipmentNCV Failure to Properly Complete a 50.59 Screen forEC272057URI Review of 50.59 screening to accept-as-is reduced valuesfor concrete properties in safety related structures.NCV Prompt Operability Determination for Safety RelatedStructures affected by ASR.Attachment

AA-2LIST OF ACRONYMSAR ????IMC Inspection Manual ChapterNCV Non-Cited ViolationNRC U.S. Nuclear Regulatory CommissionNRR Nuclear ????TS Technical SpecificationURI Unresolved ItemAttachment