ML14302A343: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
 
(10 intermediate revisions by the same user not shown)
Line 1: Line 1:
{{Adams
#REDIRECT [[RS-14-314, Requests for Relief for Alternate Examination Frequency Under ASME Code Case N-729-1 for Reactor Vessel Head Penetration Welds in Accordance with 10 CFR 50.55a(a)(3)(i)]]
| number = ML14302A343
| issue date = 10/27/2014
| title = Braidwood, Unit 1 and Byron Units 1 & 2, Requests for Relief for Alternate Examination Frequency Under ASME Code Case N-729-1 for Reactor Vessel Head Penetration Welds in Accordance with 10 CFR 50.55a(a)(3)(i)
| author name = Gullott D M
| author affiliation = Exelon Generation Co, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000454, 05000455, 05000456
| license number = NPF-037, NPF-066, NPF-072
| contact person =
| case reference number = RS-14-314, TAC ME1066
| document type = Letter
| page count = 18
| project = TAC:ME1066
}}
 
=Text=
{{#Wiki_filter:44300 Winfeld Road'Warrenville. IL 60555A Exeton Generation. 630 2000 Office10 CFR 5055aRS-14-314October 27, 2014U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001Braidwood Station, Unit 1Facility Operating License No. NPF-72NRC Docket No. STN 50-456Byron Station, Units I and 2Facility Operating License Nos. NPF-37 and NPF-66NRC Docket Nos. STN 50-454 and STN 50-455Subject: Requests for Relief for Alternate Examination Frequency Under ASME CodeCase N-729-1 for Reactor Vessel Head Penetration Welds in accordancewith 10 CFR 50.55a(a)(3)(i)References: (1) Letter from Patrick Simpson (EGC) to U.S. NRC, "Requests forRelief for Alternate Examination Frequency Under ASME CodeCase N-729-1 and from Requirements for Limited Examination ofReactor Vessel Head Penetration Welds in accordance with10 CFR 50.55a(a)(3)(i)," dated April 2, 2009, ADAMS AccessionNo. ML091030444(2) Letter from Patrick Simpson (EGC) to U.S. NRC, "Supplement ofRequest for Relief from Requirements for Limited Examination ofReactor Vessel Head Penetration Welds in Accordance with10 CFR 50.55a(a)(3)(i)," dated December 17, 2009, ADAMSAccession No. ML093520172(3) Letter from Stephen Campbell (U.S. NRC) to Charles Pardee (EGC),"Byron Station, Unit No. 2 -Relief Request 13R-16 for ReactorPressure Vessel Head Penetration Examination Frequency (TAC NoME1066)," dated January 28, 2010, ADAMS AccessionNo. ML100210231(4) Letter from J. L. Hansen (EGC) to U.S. NRC, 'Third 10-YearInservice Inspection Interval Requests for Relief for AtlernativeRequirements for the Repair of Reactor Vessel Head Penetrations,"dated April 19, 2011, ADAMS Accession No. ML1 11100620Aol7 October 27, 2014U. S. Nuclear Regulatory CommissionPage 2(5) Letter from Jacob Zimmerman (U. S. NRC) to M. J. Pacilio (EGC),"Braidwood Station, Units I and 2 and Byron Station, Unit Nos. I and2 -Relief Requests 13R09 and 13R-20 Regarding AlternativeRequirements for Repair of Reactor Vessel Head Penetrations (TACNos. ME6071, ME6073, and ME6074)," dated March 29, 2012,ADAMS Accession No. ML120790647(6) Letter from David M. Gullott, (EGC) to U.S. NRC, "Revision to the Third10-Year Inservice Inspection Interval Requests for Relief for the Repair ofReactor Vessel Head Penetrations," dated September 8, 2014, ADAMSAccession No. ML14251A536In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i), ExelonGeneration Company, LLC (EGC), submitted relief request 13R-16 and 13R-17 for ByronStation Unit 2 (Reference 1). In Reference 2, EGC withdrew relief request 13R-17. The reliefrequest (i.e., 13R-16) proposed an alternate examination schedule for volumetric and surfaceexaminations to that required by 10 CFR 50.55a(g)(6)(ii)(D)(5) which modified Code Case N-729-1, Note (8) which requires re-inspection' each refueling outage instead of the Code CaseN-729-1 re-inspection frequency. The relief request also requires re-inspection of Byron, Unit2 Penetration 68 each refueling outage because it was previously repaired. In Reference 3,the NRC provided their authorization to Implement Relief Request 13R-16.Note, while Reference 3 allowed a re-inspection frequency of every other outage for ByronStation, Unit 2 penetrations (with the exception of flawed penetration number 68) for theremainder of the third 10-year Inservice Inspection (ISI) interval, the Safety Evaluationstipulated that the alternate schedule was not authorized should any additional indications ofPrimary Water Stress Corrosion Cracking (PWSCC) be found on the Byron Station Unit 2Reactor Pressure Vessel (RPV) head penetration nozzles or associated J-groove welds. Inthe recent refueling outage of Fall 2014 (B2R18), an indication was discovered in ByronStation, Unit 2, RPV Penetration 6. The apparent cause of the indication was attributed toPWSCC. Therefore, the inspection interval approved for Byron Station, Unit 2 in Reference 3is no longer authorized for the remainder of the third 10-year ISI interval and all Byron Station,Unit 2 penetrations are included as part of this relief request.In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(i), ExelonGeneration Company, LLC (EGC), submitted relief requests (RRs) 13R-20 for ByronStation, Units I and 2, and 13R-09 for Braidwood Station, Units 1 and 2, (i.e.,Reference 4). The RRs proposed an alternative repair technique using weld overlays onthe reactor vessel head penetration housing and J-groove welds, using a Westinghouseembedded flaw repair method. EGC proposed the alternative for indications that may be1 Note, 10 CFR 50.55a(g)(6)(ii)(D)(5) refers to intervals as "re-inspection" intervals while theASME Code Case N-729-1 refers to intervals as "reexamination" intervals. For the purpose ofthis relief request, "re-inspection" and "reexamination" are synonymous.
October 27, 2014U. S. Nuclear Regulatory CommissionPage 3encountered in the future, and that may be the result of PWSCC. In Reference 5, theNRC provided their authorization to implement Relief Requests 13R-09 and 13R-20,Revision I as a repair method for degradation identified in Reactor Vessel HeadPenetrations.In Reference 6. EGC submitted a Revision to the Relief Request for 13R-09 and 13R-20 (i.e.,13R-09 and 13R-20 Revision 2) which requested relief from performing surface examinations(i.e., dye penetrant (PT)) every cycle under certain conditions. EGC has reviewed the technicalbasis for requiring re-inspection of all nozzles each outage as required by10 CFR 50.55a(g)(6)(ii)(D)(5) along with the examination results and personnel radiationexposure associated with examinations and determined that it is appropriate to relax therequired re-inspection frequency. Attachment 1 provides Relief Requests Byron Station 13R-27and Braidwood Station 13R-14 which are proposing relaxation of the re-inspection frequency asdefined in 10 CFR 50.55a(g)(6)(li)(D)(5) for Byron Station, Unit 1, Byron Station, Unit 2, andBraidwood Station, Unit 1 in accordance with 10 CFR 50.55a(a)(3)(i). Attachment 2 providesElectric Power Research Institute (EPRI) Report 3002003099, "Materials Reliability Program:Reevaluation of Technical Basis for Inspection of Alloy 600 PWR Reactor Vessel Top HeadNozzles (MRP-395)" which provides a technical basis for the re-inspection frequency changerequested. Attachment 2 is provided based on an Outbound Copyright Release authorized forEGC by EPRI on October 24, 2014.EGC requests approval of this proposed relief request by September 4, 2015, prior to beginningof the Bryon Station refueling outage in Fall 2015 (B1R20).There are no regulatory commitments contained in this submittal.If you have any questions regarding this matter, please contact Jessica Krejcie at (630) 657-2816.Respectfully,David M. GullottManager -LicensingExelon Generation Company, LLCAttachment 1: 10 CFR 50.55a Relief Requests Byron Station 13R-27 andBraidwood Station 13R-1 4, "Requests for Relief for Alternate ExaminationFrequency Under ASME Code Case N-729-1 for Reactor Vessel HeadPenetration Welds in accordance with 10 CFR 50.55a(a)(3)(i)"Attachment 2: EPRI Report 3002003099, "Materials Reliability Program: Reevaluation ofTechnical Basis for Inspection of Alloy 600 PWR Reactor Vessel Top HeadNozzles (MRP-395)," dated September 2014 bcc: Project Manager, NRR -Byron StationIllinois Emergency Management Agency -Division of Nuclear SafetySite Vice President -Byron StationVice President -Licensing and Regulatory AffairsRegulatory Assurance Manager -Braidwood StationRegulatory Assurance Manager -Byron StationDirector, Licensing and Regulatory Affairs -WestManager, Licensing -Byron, Braidwood and LaSalle StationsNuclear Licensing Administrator -Byron and Braidwood StationsCommitment Tracking Coordinator -CanteraExelon Document Control Desk Licensing (Hard Copy)Exelon Document Control Desk Licensing (Electronic Copy)B. CaseyR. McBrideH. DoG. NavratilH. MalikowskiD. AnthonyH. SmithL. DworakowsklG. ContradyG. PaniciE. BlondinG. WilhelmsenJ. BauerE. EnglertG. GerzenC. CoteB. PetersB. YoumanP. KusumawatimurrayA. CreameanS. Ahmed
 
==Attachment==
I10 CFR 50.55a RELIEF REQUESTS Byron Station 13R-27 and Braidwood Station 13R-14Requests for Relief for Alternate Examination Frequency Under ASME Code Case N-729-1 forReactor Vessel Head Penetration Welds In accordance with 10 CFR 50.55a(a)(3)(i)
ISI Program Plan Units I and 2, Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 1 of 13)Requests for Relief for Alternate Examination Frequency Under ASME Code CaseN-729-1 for Reactor Vessel Head Penetration Welds in accordance with10 CFR 50.65a(a)(3N1)1.0 ASME CODE COMPONENT(S) AFFECTEDComponent NumbersDescription:Code Class:Braidwood Station, Unit 1 and Byron Station, Units 1 and 2,Reactor Vessels 1 RC01 R (Unit 1) and 2RC01 R (Unit 2)Alternate Examination Frequency Under ASME Code CaseN-729-1 for Limited Examination of Reactor Vessel HeadPenetration WeldsClass IExamination Category: ASME Code Case N-729-1Code Item:B4.20Component Identification: All reactor vessel closure head penetrationsDrawing Numbers:Various2.0 APPLICABLE CODE EDITION AND ADDENDAAmerican Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASMECode), Section XI, 2001 Edition, through 2003 Addenda. Examinations of the reactorvessel closure head penetrations are performed in accordance with10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-1, withconditions.3.0 APPLICABLE CODE REQUIREMENT10 CFR 50.55a(g)(6)(ii)(D)(5) requires that "If flaws attributed to [Primary Water StressCorrosion Cracking] PWSCC have been identified, whether acceptable or not forcontinued service under Paragraphs -3130 or -3140 of ASME Code Case N-729-1, there-inspection interval must be each refueling outage instead of the re-inspection intervalsrequired by Table 1, Note (8) of ASME Code Case N-729-1 ."
ISI Pýocjrem PlanUnits I and Z Third Interval10 CFR 50.65a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 2 of 13)Byron Station, Units I and 2 and Braidwood Station, Unit I have repaired reactor vesselhead penetrations due to PWSCC and therefore, currently require re-inspection1 eachrefueling outage per 10 CFR 50.55a(g)(6)(ii)(D)(5). Byron Station, Units 1 and 2 andBraidwood Station, Unit 1 relief requests 13R-09 and 13R-20 (Reference 4) requirerepaired penetrations be examined and re-examined in accordance with Code CaseN-729-1 including 10 CFR 50.55a(g)(6)(ii)(D) conditions as described in the referencerelief requests.As an alternative to the re-inspection frequency requirements prescribed in10 CFR 50.55a(g)(6)(ii)(D)(5), frequency of examinations will be conducted inaccordance with ASME Code Case N-729-1 Table I requirements based on ElectricPower Research Institute (EPRI) Report 3002003099, "Materials Reliability Program[(MRP)J: Reevaluation of Technical Basis for Inspection of Alloy 600 PWR ReactorVessel Top Head Nozzles (MRP-395)," dated September, 2014 (Reference 6).4.0 REASON FOR THE REQUESTThe criteria to meet the conditions of 10 CFR 50.55a(g)(6)(ii)(D)(5) requires examinationof closure head penetrations each refueling outage if PWSCC has been detected. SincePWSCC has been detected on Byron Station, Units 1 and 2 and Braidwood Station,Unit 1, examination of the reactor vessel closure head penetrations is required eachrefueling outage.Reference 5 allowed a re-inspection frequency of every other outage for Byron Station,Unit 2 penetrations (with the exception of repaired penetration number 68) for theremainder of the third 10-year Inservice Inspection (ISI) interval. The Safety Evaluationspecified that the alternate schedule was not authorized should any additionalindications of PWSCC be found on the Byron Station, Unit 2 Reactor Pressure Vessel(RPV) head penetration nozzles or associated J-groove welds. In the recent refuelingoutage of Fall 2014 (B2R1 8), an indication was discovered in Byron Station, Unit 2, RPVPenetration 6. The apparent cause of the indication was attributed to PWSCC.Therefore, the inspection interval approved for Byron Station, Unit 2 in Reference 5 is nolonger authorized for the remainder of the third 10-year ISI Interval. This relief requestalso includes all Byron Station, Unit 2 penetrations (inclusive of repaired penetrations).As an alternative to the re-inspection frequency requirements prescribed in10 CFR 50.55a(g)(6)(ii)(D)(5), frequency of examinations will be conducted inaccordance with ASME Code Case N-729-1 Table I requirements. ASME Code CaseN-729-1 Table I Note 8 states:"If flaws have been previously detected that were unacceptable for continued service inaccordance with -3123.3 or that were corrected by a repair/replacement activity of-3132.2 or -3142.3(b), the reexamination frequency is the more frequent of the normalreexamination frequency (before RIY [Re-Inspection Years] =2.25) or every secondNote, 10 CFR 50.55a(g)(6)(ii)(D)(5) refers to intervals as "re-inspection" intervals while theASME Code Case N-729-1 refers to intervals as "reexamination" intervals. For the purpose ofthis relief request, "re-inspection" and "reexamination" are synonymous.
ISI Program Plan Units I and 2, Third IntevalISI Program PlanUnits I and Z Third Interval10 CFR 50.56a REUEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 3 of 13)refueling outage, and [Note (9)] does not apply. Additionally, repaired areas shall beexamined during the next refueling outage following the repair."Byron Station Units 1 and 2 and Braidwood Station Unit I RIY values range fromapproximately 0.420 to 0.488 per operating cycle. This results In a two cycle RIY ofapproximately 1. This RIY value is well under the 2.25 RIY value specified in Note 8 ofCode Case N-729-1 Table I outlined above. Therefore, this relief request refers to areexamination interval of every second refueling outage (i.e., every other outage) sincethis results in the more frequent of the normal reexamination frequency (beforeRIY = 2.25) or every second refueling outage as indicated in Note 8.As described in Reference 6 Section 1, the original technical basis for the reactor vesselhead inspection requirements defined in ASME Code Case N-729-1 is documented inSection 3 of Reference 7. The technical basis is supported by the Reference 8 safetyassessment report and the safety assessments that it references including Reference 9.Note, the inspection requirements for top head nozzles developed on the basis of theReference 8 safety assessment were published in Reference 7. These requirementswere intended to supersede the inspection requirements of NRC Order EA-03-009, butinstead, Reference 8 and Reference 7 formed the technical basis for the inspectionrequirements of ASME Code Case N-729-1 which replaced the NRC order as thecurrent mandatory inspection requirements document (subject to certain conditions aslisted in 10 CFR 50.55a(g)(6)(ii)(D)). The technical basis was originally developed inpart based on plant experience with non-cold heads (i.e., reactor vessels operating atreactor hot-leg temperature (Thot)) experience with PWSCC.The probabilistic fracture mechanics (PFM) analyses performed in Reference 9 used aMonte Carlo simulation algorithm to determine a probability of failure versus time forPWR vessel top heads for a set of input parameters, including operating temperature,inspection types (visual or volumetric NDE), and inspection intervals. Input into thisalgorithm included an experience-based time-to-leakage correlation based on a Weibullmodel of plant inspections, circumferential cracks, fracture mechanic analyses of variousnozzle configurations containing axial and circumferential cracks, and the MRP-developed statistical crack growth rate model for Alloy 600 (Reference 10). The originaltechnical basis for the N-729-1 inspection requirement (contained in Reference 7),concluded that a ceiling of two cycles (i.e., re-inspection/reexamination every othercycle) was a conservative approach.Since the time of the Reference 9 analysis, indications of PWSCC have been identifiedin Alloy 600 CRDM nozzles in five domestic PWR cold heads (i.e., reactor vesselsoperating at reactor cold-leg temperature (Tcold) including Braidwood Station and ByronStation). Since the time of the Reference 2 analysis, indications of PWSCC have beenidentified in five domestic PWR cold heads. Therefore, it is appropriate to assess theimplication of this new experience on the technical basis of the inspection frequencyrequirements evaluated by the MRP. The Reference 6 report evaluates the adequacy ofthe current inspection requirements, including the frequency of periodic volumetric orsurface examinations for heads operating at Tcold, considering the recent cases ofPWSCC reported in PWR cold heads. In addition, the Reference 6 report re-evaluateswhether the original approach of a re-inspection interval of two 18-month fuel cycles (i.e.,every other cycle) is justified for heads operating at Tcold in which PWSCC has been ISI Program PlanUnits I and Z Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 4 of 13)previously detected. The Reference 6 evaluation considers the following technicalaspects: industry examination history of PWSCC, a deterministic crack growth rateanalysis, a probabilistic Monte Carlo simulation analysis and also an assessment of theimpact of boric acid. The Reference 6 evaluation determined that examinationfrequency per Code Case N-729-1 results in an acceptable level of quality and safety.In addition to the technical aspects described above that have concluded an acceptablelevel of quality and safety exists with reexamination every other outage, examination ofthe Byron Station, Units I and 2 and Braidwood Station, Unit I reactor vessel headpenetrations results in approximately 500-1000 mRem each outage. Also, EGC willcontinue to perform Code Case N-729-1 item number B4.10 visual examinations (VE)each refueling outage. Therefore, since the MRP evaluation has determined thatexamination frequency per Code Case N-729-1 (i.e., reexamination every other outage)results in acceptable level of quality and safety, EGC is requesting NRC approval toperform Code Case N-729-1 item number B4.20 examinations of the Byron andBraidwood reactor vessel closure heads without application of NRC condition10 CFR 50.55a(g)(6)(ii)(D)(5).5.0 PROPOSED ALTERNATIVE AND BASIS FOR USEBasis for UseReexamination of the Byron Station, Units I and 2 and Braidwood Station, Unit I reactorvessel closure head penetrations per the Code Case N-729-1 Table 1 prescribedfrequency will continue to ensure that degradation will be detected early and will notresult in significantly increased probability of leakage or significantly reduce nuclearsafety. Continued inspection per Code Case N-729-1 will reduce personnel exposure(i.e., approximately 500-1000 mRem) each outage an inspection is not required to beperformed and will maintain an acceptable level of quality and safety. Reference 6provides an Industry report of the technical basis for this request. A summary of thereport is provided below.5.1 PWSCC Experience for Alloy 600 Reactor Vessel Closure Head NozzlesLaboratory testing is the principal technique applied to determine relative crack growthrates for Alloy 600 wrought material. However, plant experience is a source of data thatcan in some cases be used to make estimates of the relative crack growth rate forcomparison with statistical assessments of the laboratory crack growth rate data. PlantPWSCC experience for reactor vessel top head nozzles was assessed for cases inwhich meaningful crack growth rate data could be developed. Plant inspectionexperience for both cold heads and heads operating at temperatures significantly aboveToold (i.e., non-cold heads) was assessed with regard to implied relative crack growthrates. The first case of apparent PWSCC detected at a cold head was for the first in-service volumetric/surface examination and was associated with a weld fabrication flaw(Byron Station, Unit 2). As such, this case was not a good candidate for assessment.The crack growth rates implied by the ultrasonic examination data for the other coldhead cases are consistent with the probabilistic crack growth rate inputs developed onthe basis of the MRP-55 (Reference 10) assessment of laboratory crack growth rate ISI Program PlanUnits I and Z Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 5 of 13)data and used in the original MRP-1 05 (Reference 9) probabilistic assessment, as wellas the current probabilistic assessment documented in Section 4 of the attached report.Furthermore, the cases in which relative crack growth rates could reasonably be inferredfor non-cold heads were also consistent with the crack growth rate inputs of theprobabilistic assessments. Hence, the crack growth rate assumptions of the technicalbasis for the N-729-1 inspection requirements remain valid in light of the CRDM nozzleinspection experience.The findings of the top head examinations performed to date support the adequacy ofthe current inspection requirements (i.e., inspections every other outage), including theRIY = 2.25 interval for periodic volumetric/surface examinations:* Since 2004, no circumferential PWSCC indications located near or above the topof the weld have been detected. These are the types of flaws that could producea nozzle ejection were they to grow to a very large size.Since examinations capable of detecting flaws connected to the outer diameter(OD) surface of the nozzle tube were first applied in the early 2000's, there havebeen no reports of top head nozzle leakage (i.e., through-wall cracking) occurringafter the time that the first in-service volumetric or surface examination wasperformed of all Control Rod Drive Mechanism (CRDM) or Control Element DriveMechanism (CEDM) nozzles in a given head. The only incidence of nozzleleakage since 2004 was detected in 2010 during the first in-service inspection(after about six calendar years of operation) performed of a replacement Alloy600 head procured from a cancelled plant. Thus, this initial examinationexperience is not directly relevant to the adequacy of the re-inspection intervalrequirement. No discernible corrosion was detected of the low-alloy steel headmaterial during the bare- metal visual examinations of this replacement Alloy 600head. It is noted that in late 2011 this first replacement head was replaced with ahead having PWSCC-resistant nozzles.The volumetric or surface examinations performed on cold heads and the repeatvolumetric or surface examinations performed on non-cold heads have beeneffective in detecting the PWSCC degradation reported in its relatively earlystages, with modest numbers of nozzles affected by part-depth cracking, oftenlocated below the weld, where the nozzle tube is inside (not directly a part of) thepressure boundary.Five of the 20 operating cold heads with Alloy 600 nozzles have shownindications of PWSCC. This cracking was part-depth. For one of these fiveheads, the indication was associated with a weld fabrication defect. Hence, plantexperience continues to show a very low probability of nozzle leakage for thecold heads given the examinations being performed.5.2 Deterministic Crack Growth AnalysisDeterministic crack growth evaluation can be applied to assess PWSCC risks for specificcomponents and operating conditions. In general, such deterministic evaluationquantifies the time between a certain initial condition with a known or hypothetical flaw ISI Program PlanUnits I and Z Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 6 of 13)size to some adverse condition such as through-wall growth, with a prescribed stabilitymargin, etc., under a set of assumptions. This time may provide information and optionsfor inspection intervals, mitigation, and repair. The evaluation described further inSection 3 of Reference 6 determined the following:" The current N-729-1 volumetric examination interval (i.e., RIY = 2.25) for reactorpressure vessel head (RPVH) without previous PWSCC detection is adequate toprovide sufficient opportunity for flaw detection prior to significant leakage orejection risk." The examination interval for RPVH operating at cold leg temperatures withpreviously detected PWSCC may be extended from the currently requiredinterval of each refueling outage to every other refueling outage withoutintroducing significant added risk of leakage or ejection. For example, allcalculations assume the existence of a roughly 10% through wall surface crack,among other conservatisms, and nevertheless predict times to leakage between7 and 17 Effective Full Power Years (EFPY) at cold head temperatures." The N-729-1 examination interval of each refueling outage for non-cold Alloy 600heads with previously detected PWSCC is considered effective for limiting risksof leakage and ejection while not being overly conservative. This conclusionholds for operating temperatures bounding for the active fleet of Alloy 600 topheads.5.3 Probabilistic Monte Carlo Simulation AnalysisThe purpose of the probabilistic analysis is to quantify the risk of leakage and ejectionmore precisely through comprehensive simulation of the PWSCC degradation process,including the introduction of a PWSCC initiation model. The probabilistic evaluationreplaces many of the conservatisms of the deterministic evaluation with best estimates,and incorporating uncertainty to reflect lack of specifics about physical variability in theRPVH PWSCC degradation process. Probabilistic predictions are in the form of eventfrequencies and probabilities. Based upon these predictions, RPVH examinationintervals are recommended to achieve acceptable levels of leakage and ejection risk,both relative to risks predicted with currently accepted examination intervals and withrespect to absolute core damage frequency limits.The probabilistic results support the current inspection requirements (i.e., therequirements contained in ASME code case N-729-1) for Alloy 600 RPVH penetrationnozzles, including for plants operating at Tcold. This probabilistic analysis is a key part ofthe updated technical basis of ASME Code Case N-729-1, superseding that ofMRP-1 05, to include industry experience since 2004 and to replace the technical letterMRP 2011-034 (Reference 11) submitted to the U.S. NRC in December 2011. The keyconclusions of this section, discussed further in Section 4 of Reference 6, are as follows:* The risk of ejection is predicted to be acceptably low (below 5E-5 ejections peryear per RPVH, averaged across the operating lifetime) when periodic UTexaminations are performed per the RIY = 2.25 interval of ASME Code Case 1St Program Plan Units I and 2, Third IntervalISI Program PlanUnits I and Z Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 7 of 13)N-729-1. This is true despite taking no credit for more frequent inspectionsrequired after PWSCC detection by N-729-1 as conditioned by10 CFR 50.55a(g)(6)(ii)(D).* Average penetration leakage frequencies due to cracks initiating in the nozzlematerial are below 0.05 new leaking penetrations per year for all the casesevaluated (including cold and non-cold heads), up to and including inspectionintervals of RIY=2.25.* No leaks have occurred since the onset of complete head inspections, so themodel provides a conservative evaluation of the potential for PWSCC flaws togrow without detection because the predicted Average Leakage Frequency (ALF)values are on the order of 0.02 -0.1 leaks per head per year for non-calibratedinitiation models." Even assuming a plant that is nominally as susceptible to PWSCC as the Alloy600 replacement RPVH calibration case (see Reference 6 Section 4.2.2), theprobabilistic analysis demonstrates that a re-inspection (i.e., UT inspection)interval based on RIY = 2.25 is sufficient to minimize the risk of leakage andejection to acceptable levels. The RIY = 2.25 interval generally equates to aninspection interval of four or five 18-month cycles for a head operating at Tcold." Under various conditions, cases were run to investigate a re-inspection (i.e., UTinspection) interval of every refueling outage versus every other refueling outage.The absolute difference In average ejection frequency between these cases isgenerally small (e.g., less than 2E-6 for all conditions evaluated and less than6E-7 for the conditions that did not assume a most-conservative initiation modelbased on the experience of the Alloy 600 replacement RPVH calibration case)." The model sensitivity cases did not show significant deviation and support therobustness of conclusions drawn from the results.The probabilistic Monte Carlo simulation analysis concluded that a reexamination (i.e.,UT inspection) interval of one or two refueling outages for top heads operating at Tcoldthat have previously detected PWSCC results in an acceptably small effect on ejectionand leakage risks. This conclusion is based on a) the acceptable risks achieved whenno credit is given to reducing inspection intervals below RIY=2.25 once PWSCC isdetected, and b) the comparable benefits achieved when using a one or two cycle re-inspection (i.e., UT inspection) interval. The probabilistic analyses conservativelyassume a high likelihood that many PWSCC flaws are initiated and detected in the headover life.5.4 Assessment of Concern for Boric AcidThe concern for boric acid corrosion of the low-alloy steel head material due to primarycoolant leakage at a through-well PWSCC flaw was considered in Section 5 ofReference 6. The concern for structural integrity of the pressure boundary directly dueto circumferential PWSCC is discussed below.
ISI Program PlanUnits 1 and 2, Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 8 of 13)It is concluded that the current requirements for periodic VE for evidence of pressureboundary leakage (per ASME Code Case N-729-1 Table I item number B4.10 asconditioned by 10 CFR 50.55a(g)(6)(ii)(D)) remain valid to address the concern forpotential boric acid corrosion. For heads with EDY > 8 andlor previously detectedPWSCC (i.e., Byron Station, Units 1 and 2 and Braidwood Station, Unit 1), the VEfrequency of every refueling outage is appropriately conservative. For heads withEDY < 8 (effectively all heads with Alloy 600 nozzles operating in U.S. at Tcold) and nopreviously detected PWSCC (i.e., Braidwood Station Unit 2), the original basis forextending the interval to every third refueling outage or 5 calendar years, whichever isless, remains valid.This approach is supported by the demonstrated low probability of pressure boundaryleakage for heads operating at Tcold and the supplemental requirement for the VT-2visual examination of the head under the insulation through multiple access points inoutages that the VE is not completed. Given the large amounts of boric acid depositsthat necessarily accompany substantial rates of boric acid corrosion, the VT-2requirement is an effective supplement to the periodic VE examinations. It isemphasized that this conclusion is not dependent on the volumetric or surfacereexamination interval for heads operating at Tcold with previously detected PWSCCbeing one rather than two 18-month fuel cycles. Plant experience and analyses showthat the probability of leakage is low for heads operating at Tcold with previouslydetected PWSCC, regardless of the volumetric reexamination interval (i.e., one or two18-month fuel cycles).5.5 Repaired NozzlesThe analyses presented in Sections 3 and 4 of Reference 6 (i.e., the Deterministic CrackGrowth Analysis and Probabilistic Monte Carlo Simulation Analysis) do not explicitlymodel repaired nozzles. However, as discussed below and further described in Section6.2 of Reference 6, a reexamination interval of two 18-month cycles (i.e., every otherrefueling outage) in the case of previously detected PWSCC in a head operating atTcold is also justified for the periodic Non-Destructive Examination (NDE) required forindividual nozzles that have been repaired using either of the two main methods thathave historically been used. These repair methods are (1) the embedded flaw repair(EFR) with application of a weld overlay on the outer nozzle and weld surfaces and (2)the "ID temper bead mid-wall repair." Note, Byron Station, Units 1 and 2 and BraidwoodStation, Unit I have used the EFR method when repairs have been required. Currently,per approved NRC safety evaluations in response to relief requests associated withthese repair methods, NDE of each repaired nozzle is performed during each refuelingoutage when all nozzles are examined per the volumetric or surface examinationrequirement of ASME Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D).The below discussion justifies that the NDE specific to repaired areas also be performedevery other refueling outage in cases where an interval of two cycles is justified for thegeneral (i.e., non-repaired nozzles) volumetric or surface examination of N-729-1 perReference 6:* The EFR for a flaw connected to the nozzle outer surface involves applyingPWSCC-resistant weld metal (e.g., Alloy 52) over the OD of the Alloy 600 nozzletube and the wetted surfaces of the J-groove weld, overlapping the vessel ISI Program PlanUnits I and Z Third Interval10 CFR 60.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 9 of 13)cladding and extending to the bottom of the nozzle, to isolate the susceptiblematerial from primary coolant. The large majority of reactor vessel penetrationnozzle PWSCC that has been detected has been located on the nozzle outersurface. Without contact with coolant, further PWSCC-induced growth isprevented. This repair is unlikely to significantly affect the stress state at thenozzle Inner Diameter (ID), and to the extent there is an effect on the stress atthe ID, the squeezing of the nozzle tube by shrinkage of the weld overlay uponcooling would tend to reduce the magnitude of the tensile stress at the nozzle ID.Periodic reexamination (i.e., UT) on the nozzle ID, per the standard N-729-1approach as conditioned by 10 CFR 50.55a(g)(6)(ii)(D), monitors the potential forgrowth of an embedded flaw originally located in the nozzle tube, or checks forgrowth into the nozzle tube of an embedded flaw originally located in the weld.The EFR technique has been applied in over 45 different Instances throughoutthe world, and the flaw being repaired has never come into contact with waterafter repair. These repairs have been in place up to 10 years in some cases.Even in the unlikely case that the embedded flaw were to become wetted, anygrowth due to PWSCC would occur at a significantly reduced rate at Tcoldcompared to heads operating at temperatures similar to reactor hot-legtemperature (e.g., 2.8 times slower at 560&deg;F compared to 600&deg;F for the standardgrowth activation energy of 31 kcal/mole). The standard re-inspection (i.e., UT)on the nozzle ID also addresses the potential for new flaws initiating on thenozzle ID in the same manner as for an unrepaired nozzle.In summary, reexamination every other outage is more than sufficient for periodic NDEafter the embedded flaw repair based on (see Reference 6 references 53 and 54 formore Information):" The coverage of the entire outer nozzle surfaces with PWSCC-resistant material," The benefit of operating at Tcold for PWSCC crack growth rates, and" The favorable plant experience, with over 45 repairs, some remaining in servicefor over 10 years to date.5.6 ConclusionsExelon proposes to perform reactor vessel head penetration examinations per CodeCase N-729-1 as amended by 10 CFR 50.55a(g)(6)(ii)(D) with one exception. Instead ofthe re-inspection frequency described in 10 CFR 50.55a(g)(6)(ii)(D)(5), Exelon proposesto perform reexamination every second refueling outage (i.e., every other refuelingoutage). This examination frequency will be applied to all reactor vessel headpenetration nozzles including those nozzles repaired using the EFR method. Thetechnical basis report In Reference 6 supports a reexamination interval of two 18-monthfuel cycles (i.e., every other refueling outage) for reactor vessel heads with previouslydetected PWSCC operating at Tcold.It is noted that ASME Code Case N-729-1 specified that the re-examination interval betwo fuel cycles or before RIY = 2.25, whichever is sooner, in the case of previouslydetected PWSCC requiring repair; however, as described above, the technical basis forthis prescribed re-examination interval had not yet considered PWSCC in RPVHs ISI Program Plan Units I and 2, Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 10 of 13)operated at Tcold. The revised technical basis as contained in Reference 6 has beenupdated to include experience with Tcold heads and, in the same manner as the currenttechnical basis, has demonstrated that reexamination performed every two cycles ascurrently described in in N-729-1 is a conservative approach. The revised probabilisticand deterministic calculations contained in Reference 6 are consistent with theprobabilistic calculations originally performed in MRP-105 (Reference 9) which showedthe RIY = 2.25 interval results in an acceptably small effect on nuclear safety, regardlessof whether PWSCC has been previously detected. Additionally, EGC will continue toperform Code Case N-729-1, Item Number B4.10 visual examination each refuelingoutage. In summary, performing reexamination of the RPV head penetrations per CodeCase N-729-1 defined frequency provides an acceptable level of quality and safety.Table I provides the NDE requirements that will be applied to nozzles repaired with theEFR method.
1St Program Plan Units I and 2, Third lnteivalISI Program PlanUnits I and Z Third Interval10 CFR 50.65a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 11 of 13)Table IEmbedded Flaw Repair Methods and Inspection RequirementsRepair Location In Flaw Orientation Repair Repair NDE ISI NDEOriginal In Original Method Note (2) Note (2)Component ComponentVHP Nozzle/Tube ID AxialSeal weld UT or SurfaceCircumferential SurfaceVHP Nozzle/Tube Axial orOD above J-groove Circumferential Note (1) Note (1) Note (1)weldVHP Nozzle/Tube Axial orOD below J-groove Circumferential Seal weld UT or Surface UT or SurfaceweldUT andJ-groove weld Axial Seal weld Surface, UT and Surface,Note (3) Notes (3) and (4)UT and UT and Surface,J-groove weld Circumferential Seal weld Surface, Notes (3) and (4)_ I_ Note (3)Notes:(1) Repair method to be approved separately by NRC.(2) Preservice and Inservice Inspection to be consistent with10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of CodeCase N-729-1 with conditions; or NRC-approved alternatives to thesespecified conditions.(3) UT personnel and procedures qualified in accordance with10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of CodeCase N-729-1 with conditions. Examine the accessible portion of theJ-groove repaired region. The UT plus surface examination coverageequals to 100%.(4) Surface examination of the entire embedded flaw repair (EFR) shall beperformed during each refueling outage. Surface examinations may bediscontinued after the EFR or any localized welded repairs to the EFRsubsequent to initial installation have been in service for two fuel cyclesand the most recent examination satisfies ASME Section III, NB-5350acceptance standards.
1St Program Plan Units I and 2, Third IntervalISI Program PlanUnits I and Z Third Interval10 CFR 50.S6a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 12 of 13)6.0 DURATION OF THE PROPOSED ALTERNATIVEThe duration of the proposed alternative is for the remainder of the Byron Station,Units 1 and 2, Third Inservice Inspection Interval currently scheduled to end in July 15,2016.The duration of the proposed alternative is for the remainder of the Braidwood Station,Unit 1, Third Inservice Inspection Interval currently scheduled to end in July 28, 2018.7.0 PRECEDENTSNone8.0 REFERENCES1. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating PrimaryWater 2077 Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials(MRP-55) Revision 2078 1, EPRI, Palo Alto, CA: 2002. 1006695. [freely available onwww.epri.com]2. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis of PWRReactor 2074 Vessel Top Head Nozzle Cracking (MRP-105), EPRI, Palo Alto, CA:2004. 1007834. [NRC 2075 ADAMS Accession No. ML041680489]3. Tood RV Closure Head Nozzle Inspection Impact Assessment, EPRI, Palo Alto, CA:2011. 2183 MRP 2011-034. [NRC ADAMS Accession No. ML12009A04214. Letter from David M. Gullottl EGC to USNRC, "Revision to the Third 10-YearInservice Inspection Interval Requests for Relief for Alternative Requirements for theRepair of Reactor Vessel Head Penetration," dated September 8, 2014. [NRCADAMS Accession No. ML14251A536]5. Letter from Stephen Campbell (U.S. NRC) to Charles Pardee (EGC), "Byron Station,Unit No. 2 -Relief Request 13R-16 for Reactor Pressure Vessel Head PenetrationExamination Frequency (TAC No ME1066)," dated January 28, 2010, [ADAMSAccession No. ML100210231]6. Materials Reliability Program: Reevaluation of Technical Basis for Inspection of Alloy600 PWR Reactor Vessel Top Head Nozzles (MRP-395). EPRI, Palo Alto, CA: 2014.3002003099 [freely available on www.epr.com]7. Materials Reliability Program (MRP): Inspection Plan for Reactor Vessel ClosureHead Penetrations in U.S. PWR Plants (MRP-1 17). EPRI, Palo Alto, CA: 2004.1007830 [freely available on www.epri.com]8. Material Reliability Program: Reactor Vessel Closure Head Penetration SafetyAssessment for U.S. Pressurized Water Reactor (PWR) Plants (MRP-1 10):Evaluations Supporting the MRP Inspection Plans, EPRI, Palo Alto, CA: 2004,1009807 [ADAMS Accession No. ML04168050619. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis of PWRReactor Vessel Top Head Nozzle Cracking (MRP-105), EPRI, Palo Alto, CA: 2004.1007834. [NRC ADAMS Accession No. ML041680489]
ISI Program Plan Units I and 2, Third Interval10 CFR 50.55a RELIEF REQUESTS BYRON STATION 13R-27 andBRAIDWOOD STATION 13R-14(Page 13 of 13)10. Materials Reliability Program: Crack Growth Rates for Evaluating Primary WaterStress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55)Revision 1, EPRI, Palo Alto, CA: 2002. 1006695. [freely available on www.epri.com]11. Materials Reliability Program: Tcold RV Closure Head Nozzle Inspection ImpactAssessment, EPRI, Palo Alto, CA: 2011. MRP 2011-034. [NRC ADAMS AccessionNo. ML12009A0421
}}

Latest revision as of 02:44, 11 April 2019

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML14302A343&oldid=1432549"