ML15160A614

LaSalle County, Units 1 and 2 - Attachment 6, File No. 1400187.302, Revision 2, Probability of Failure for LaSalle Unit 2 N1 Nozzle-to-Shell-Welds and Nozzle Blend Radii Regions

ML15160A614
Person / Time
Site:	LaSalle
Issue date:	06/08/2015
From:	Wong W, Wu J Structural Integrity Associates
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15160A620	List: ML15160A612 ML15160A614 RS-15-158, Response to Request for Additional Information Regarding Relief Request I3R-14
References
RS-15-158, TAC MF5654, TAC MF5655 1400187.302, Rev. 2
Download: ML15160A614 (14)

Text

ATTACHMENT 6Structural Integrity Associates, Inc. ReportFile No. 1400187.302, Revision 2Probability of Failure for LaSalle Unit 2 N1Nozzle-to-Shell-Welds and Nozzle Blend Radii Regions(Non-Proprietary)13 pages follow V Structural Integrity Associates, Inc." File No.: 1400187.302!C S Project No.: 1400187CALCULATION PACKAGE Quality Program: Z Nuclear [] CommercialPROJECT NAME:LaSalle N702 Relief Request for 60 YearsCONTRACT NO.:00517760, Rev 4CLIENT: PLANT:Exelon Generation Company LLC LaSalle County Generating Station, Units 1 and 2CALCULATION TITLE:Probability of Failure for LaSalle Unit 2 Ni Nozzle-to-Shell-Welds and Nozzle Blend Radii RegionsNOTE: This document contains vendor proprietary information. Such information has been redacted for public release of thisdocument.Document Affected Project Manager Preparer(s) &Revision Pages Revision Description Approval Checker(s)Signature & Date Signatures & Date0 1 -11 Initial IssueA-I -A-2 Jim Wu8/8/14Wilson Wong8/8/14 Wilson Wong8/8/141 1 -11 Revised Proprietary Jim WuA-1 -A-2 Markings Wilson Wong 2/6/152/6/15Wilson Wong2/6/152 3,5,7 Editorial changes and AQ.r" K/q -Revised Proprietary WiMarkings Wilson WongWilson Wong 5/6/155/6/15Jim Wu5/6/15Page 1 of 11F0306-01 R2 CStructural Integrity Associates, Inc!Table of Contents1.0 INTRODUCTION ..................................................................................................... 32.0 METHODOLOGY .................................................................................................. 33.0 SOFTWARE MODIFICATIONS .......................................................................... 34.0 A SSU M PT IO N S ........................................................................................................ 45.0 D E SIG N IN PU T ....................................................................................................... 56.0 FATIGUE CRACK GROWTH ................................................................................ 57.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS .................................. 68.0 PROBABILISTIC FRACTURE MECHANICS EVALUATION ........................... 79.0 RESULTS OF ANALYSES .................................................................................... 710.0 C O N C LU SIO N S ....................................................................................................... 811.0 R E FE R E N C E S ......................................................................................................... 9APPENDIX A LIST OF SUPPORTING FILES ............................................................. A-1List of TablesTable 1: LaSalle Weld Chemistry ........................................................................................... 11Table 2: Probability of Failure Results Summary .................................................................. 11File No.: 1400187.302Revision: 2Page 2 of 11F0306-01R2 CStructural Integrity Associates, Inc!1.0 INTRODUCTIONStructural Integrity Associates (SIA) is contracted by Exelon to perform a plant specific analysis to requestinspection relief for the current licensing period per ASME Boiler and Pressure Vessel Code Case N-702[1], and to the end of the period of extended operation (60 years of operation) for the LaSalle CountyGenerating Station (LGS) RPV nozzles. LaSalle intends to extend their existing relief request forItis also stated inSection 2 of Reference 2 that "It should be noted that only the recirculation inlet and outlet nozzles needto be checked because the P(FIE)s (Conditional probability of failure from event F due to event E) forother nozzles are an order of magnitude lower." SIA concluded that the N I nozzle of Unit 2 is thebounding nozzle since it is the only nozzle violating the condition 4 requirements set forth by BWRVIP-241 [2]. To address the elevated fluence issue of certain nozzles in the belt-line region of the RPV, abounding approach is used to qualify all of the indicated nozzles for both units by analyzing the Unit 2NI nozzle using the fluence level from the N6 nozzle (peak fluence at end of the period of extendedoperation) since N6 nozzle is located in the belt-line region and has the bounding fluence among allindicated nozzles.The intent of this analysis is to confirm that the NI nozzles meet the applicable acceptance criteriaconsidering the elevated fluence level, thus qualifying all nozzles identified above. The evaluationconsists of two parts: Finite Element Model (FEM) Stress Analysis and Probabilistic Fracture Mechanics(PFM) Analysis. This calculation package documents the PFM analysis while a previous calculationpackage [8] documented the stress analysis.2.0 METHODOLOGYThe approach used for this evaluation is consistent with the methodology presented in Reference 3 and5. A Monte Carlo simulation is performed using a variant of the program VIPER [4] with somemodifications as described in the following sections. The VIPER program was developed as part of theprogram in Reference [3] for the Boiling Water Reactor (BWR) reactor pressure vessel (RPV) shell weldinspection recommendations. The software was modified into a separate edition, identified asVIPERNOZ, for use in this evaluation.The detailed description of the methodology incorporated in the VIPER/VIPERNOZ program isdocumented in References [3] and [5].3.0 SOFTWARE MODIFICATIONSSeveral modifications were made to VIPER in order to include the capability to perform the evaluationfor nozzle blend radii. The modifications are:1. Include fatigue crack growth analysis,2. Option to perform stress corrosion crack growth and/or fatigue crack growth,File No.: 1400187.302 Page 3 of 11Revision: 2F0306-01R2 CStructural Integrity Associates, Inc!3. User defined flaw size distribution,4. User defined probability of detection (PoD) curves for inspection,5. User defined event occurrence time,6. User defined distribution for selected random parameters,7. User input number of printout for failed and non-failed vessels,8. The constant for margin term for upper bound values of adjusted reference temperature requiredby Appendix G to 10 CFR Part 50 is a user input,9. Pre-service inspection is eliminated,10. Initial flaw size to include clad thickness is a user option,11. Improvement in data structure for analysis results.The modified software for this project is identified as VIPERNOZ to distinguish from the originalVIPER software in Reference [3]. Note that the VIPERNOZ computer program is the same programused in the BWRVIP- 108NP report that was accepted by the NRC in their SER [5].4.0 ASSUMPTIONSThe following assumptions used in the evaluation are consistent with those listed in References [2] and[5]:3. The flaw size distribution, PVRUF, is assumed to be as shown in Figure 5-4 of Reference [6].5. Lower bound constant upper shelf fracture toughness is set to 200 with a standarddeviation of 30 ksi'Iin for un-irradiated material based on the SER report. For irradiated materialthe VIPERNOZ program will make the necessary adjustment based on fluence and Initial RTNDTinputs using guidance from RG 1.99 [12].6. Standard deviation of the mean Kic is set to 15 percent of the mean value of the Kic per the SERreport [5].7. All chemistry information from NI nozzle-to-shell weld and nozzle blend radii wasconservatively taken from BWRVIP-241 fleet bounding data. [2]8. Peak fluence from the Unit 2 N6 nozzles at the belt line region will conservatively be used forthe NI nozzle-to-shell weld and nozzle blend radii.File No.: 1400187.302 Page 4 of I IRevision: 2F0306-01 R2 CStructural Integrity Associates, Inc!5.0 DESIGN INPUTThe LaSalle plant specific input is described below." Vessel Wall Thickness at the weld = 6.5625" (excluding clad) [7]* Vessel Wall Thickness through the Blend = 10.4626" (excluding clad) [Path 3, 8]" Vessel Wall Thickness through the Blend = 11.2653" (excluding clad) [Path 1, 8]* Vessel Inner Radius = 126.6875" (excluding clad) [7]* Vessel Clad Thickness at Blend = 0.1875" [7]" Vessel Clad Thickness at Weld = 0.1875" [7]* Vessel Operating Temperature = 528°F [9]* Vessel Hydro Testing Temperature = 100'F [9]* Operating Pressure = 1050 psig [9]* Pressure during Bounding Transient = 1180 psig [9]" End of Life Fluence (54 EFPY/60 years) for N6 Forging at Unit 2= 5.36 x1017 n/cm2 [Table 7-9, 11]* Mean Initial RTndt / standard deviation at Blend Radius" Mean Initial RTndt / standard deviation at the RPV WeldThe weld chemistry is taken from Reference 2 and presented in Table 1.All random variables are summarized in Table 2 of Reference [12]. Most of the input is obtained fromReference [3], except standard deviation for %Cu and %Ni for nozzle blend radii and nozzle-to-shellweld. For nozzle blend radii, these inputs are equal to 0.04407 (Calculated based on Figure 3-1 and 3-26.0 FATIGUE CRACK GROWTHThe fatigue data for SA-533 Grade B Class I and SA-508 Class 2 in a reactor water environment arereported in Reference [13] for weld metal testing at R = 0.2 and 0.7. To produce a fatigue crack growthlaw and distribution for the VIPERNOZ software, the data for R= 0.7 was fitted into a form of ParisLaw. The R= 0.7 fatigue crack growth law was chosen for conservatism. The curve fit results of themean fatigue crack growth law is presented with the Paris Law shown as follows:File No.: 1400187.302Revision: 2Page 5 of 11F0306-01R2 CStructural Integrity Associates, Inc.da= 3.817

• 10_9(AK)2.927 (1)dnwhere a = crack depth, inn = cycleAK = Kmax -Kmm, ksi-in05A comparison to the ASME Section XI [10] fatigue crack growth law in a reactor water environment isdocumented in Reference [13]. It shows a reasonable comparison where the ASME Section XI law ismore conservative on growth rate at high AK.Using the rank ordered residual plot, it was shown that a Weibull distribution was more representativefor the data. The Weibull residual plot with the linear curve fit of the data is shown below:y = -0.3712 + 4.15x (2)where y = ln(ln(1/(1-F))x = ln((da/dn)actuai/(da/dn)mean)F = cumulative probability distributionPer 10CFR 50.55a, the NRC have placed additional, more limiting requirements on the Section XIfatigue crack growth (FCG) in Appendix A for negative R ratios. Since Reference 14 has concluded thatthe main contributing factor of crack growth is SCC and that fatigue crack growth is negligible, theeffect of FCG due to negative R ratios need not be addressed.7.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGSThe stress analyses for the nozzle-to-shell weld and the nozzle blend radius for the Unit 2 NI nozzle arepresented in Reference [8]. The stress analyses were performed for unit pressure and bounding normaland upset thermal transients (Loss of Feedwater Pumps/Isolation Valves Close) for the NI nozzle. Theazimuthal locations evaluated were 0' and 900, which also represent the symmetric un-modeled 1800and 2700 locations of the nozzle. Two through-wall sections were selected. One is at the location of theweld between the RPV and nozzle and the other is at the blend radius location of the nozzle.The bounding load cases analyzed for the NI nozzle include:1. Unit pressure2. Turbine Generator Trip-SCRAM (TGT-SCRAM)3. Loss of Feedwater Pumps/Isolation Valves CloseFor the thermal transients, the through-wall stress profiles that produce the largest stress ranges forthermal fatigue crack growth are presented and used in the evaluation.The number of thermal cycles for the TGT-SCRAM transient are considered to be the total number ofcycles for all normal and upset conditions that involve temperature/ pressure changes in region B of thereactor vessel (754 cycles per Reference 9 for 40 years of operation and approximately 1131 cycles forFile No.: 1400187.302 Page 6 of 11Revision: 2F0306-01R2 CStructural Integrity Associates, Inc!60 years of operation, or 189 cycles for each block of 10 years of operation) for conservatism.Specifically, transients considered were: Design Test (130 cycles), Start Up (117 Cycles), Loss ofFeedwater Heater (80 Cycles), SCRAM (180 Cycles), Shut Down Vessel Flooding (111 Cycles), Unbolt(123 Cycles), Loss of Feedwater Pump/Isolation Valves Close (10 Cycles), and Natural Circulation StartUp (3 Cycles).The number of thermal cycles for the Loss of Feedwater Pump/Isolation Valves Close transient is 10cycles for 40 years of operation per Reference 9. However, there are three internal cycles within themain transient, the last of which occurs after an indefinite time and can be bounded by the TGT-SCRAMtransient. Therefore, only the first two internal cycles are considered for the Loss of FeedwaterPump/Isolation Valves Close transient, which amounts to 20 cycles for 40 years of operation (10 cyclesx 2 internal cycles) and 30 cycles for 60 years of operation.8.0 PROBABILISTIC FRACTURE MECHANICS EVALUATIONThe probabilistic evaluation is performed for the case of 25% inspection for the extended operatingperiod (with zero inspection coverage conservatively assumed for the initial 40 years of operation).For the nozzle blend radius region, a nozzle blend radius crack model [15] was used in the probabilisticfracture mechanics evaluation for the reliability of the in-service inspection program. For this locationand crack model, the applicable stress is the stress perpendicular to any path cut along the nozzlelongitudinal axis (nozzle hoop stress).For the nozzle-to-vessel shell weld, either a circumferential or an axial crack could be initiated due toeither component fabrication (i.e. considering only welding process) or stress corrosion cracking. FromReference [3], it is shown that the probability of failure for a circumferential crack is much less than anaxial crack, due to the difference in the stress (hoop versus axial) and the influence function of the crackmodel. Therefore, the probabilistic fracture mechanics evaluation for the nozzle and vessel shell weldwould concentrate on the axial crack. An axial elliptical crack model with a crack aspect ratio of a/l =0.2 is used in the evaluation for the nozzle-to-vessel shell weld. The inspection PoD curve is the userinput of Figure 42 of Reference [12], with an inspection interval every 10 years. The calculation ofstress intensity factor is at the deepest point of the crack.The analyses are performed using VIPERNOZ, a modified version of the program VIPER, [4], with themodifications as described in Section 3.0. The number of simulations is 5 million.9.0 RESULTS OF ANALYSESThe reliability evaluation is presented using plant specific inspection coverage. The probabilities offailure (PoF) from the limiting Low Temperature Overpressure (LTOP) events and Normal OperatingConditions are summarized in Table 2. The in-service inspection of 25% inspection for the extendedoperating term (with zero inspection coverage for the initial 40 years of operation) is used at both thenozzle blend radius as well as the nozzle-to-shell weld.File No.: 1400187.302 Page 7 of 11Revision: 2F0306-01R2 rStructural Integrity Associates, Inc!10.0 CONCLUSIONSThe probability of failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii in thelimiting NI nozzle at LaSalle Unit 2 is below the criteria of 5 x 10-6 per year [17]. The LaSalleN1nozzles still meet the acceptable failure probability considering 60 year thermal cycles and theelevatedfluence level of the N6 nozzles. Therefore, N1, N2, N3, N5, N6, N7, N8, N9, N16, and N18 nozzles atLaSalle Units 1 and 2 still qualify for reduced inspection using ASME Code Case N-702 to the end ofthe period of extended operation (60 years of operation).File No.: 1400187.302Revision: 2Page 8 of IIF0306-01R2 CStructural Integrity Associates, Inc!11.0 REFERENCES1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle InnerRadius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.2. BWRVIP-241: BWR Vessel Internal Project, Probabilistic Fracture Mechanics Evaluation for the BoilingWater Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA. 1021005. EPRIPROPRIETARY INFORMATION.3. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05),"Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.4. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), StructuralIntegrity Associates.5. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basisfor the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel ShellWelds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007.6. B WR VIP-108NP: B WR Vessel and Internals Project, Technical Basis for the Reduction of InspectionRequirements for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii.EPRI, Palo Alto, CA: 2007. 10161237. GE Drawing, "Recirculation Outlet Nozzle NI," LaSalle II MPL# B13-D003, SI File No.1400187.202.8. SI Calculation Package, "Finite Element Model Development and Thermal Mechanical StressAnalyses for the Unit 2 NI Nozzle," Revision 0, SI File Number 1400187.301.9. Thermal Cycle Diagramsa. General Electric Drawing Number 158B8136, Sheet 1, Revision 6, "Reactor Vessel NozzleThermal Cycles," SI File No. 1400187.207b. General Electric Drawing Number 73 1E776, Sheets 1 and 2, Revision 3, "Reactor VesselThermal Cycles," LaSalle Unit 1, SI File No. 1400187.205c. General Electric Drawing Number 761E581, Sheets I and 2, Revision 1, "Reactor VesselThermal Cycles," LaSalle Unit 2, SI File No. 1400187.206.10. ASME Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of NuclearPower Plant Components, 2007 Edition with 2008 Addenda.11. EXL-LSA-001-R-003, "LaSalle County Generating Station Unit 2 Reactor Pressure Vessel FluenceEvaluation at End of Cycle 15 with Projections to 32 and 54 EFPY," Revision 0, SI File Number1400187.209.12. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May1988..13. Bamford, W. H., "Application of corrosion fatigue crack growth rate data to integrity analyses ofnuclear reactor vessels," Journal of Engineering Materials and Technology, Vol. 101, 1979.File No.: 1400187.302 Page 9 of I IRevision: 2F0306-01 R2 CStructural Integrity Associates, Inc!14. EPRI Memo 2012-138, "BWRVIP Support of ASME Code Case N-702 Inservice Inspection Relief,"From Chuck Wirtz to All BWRVIP Committee Members, August 31, 2012.15. ASME publication, "Fracture Mechanics Analysis of JAERI Model Pressure Vessel Test," S.A.Delvin and P C. Ricardella, 78-PVP-91..16. BWRVIP-173-A: "Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials," EPRI,Palo Alto, CA, 2011, 1022835, SI File Number BWRVIP-173-A. EPRI PROPRIETARYINFORMATION.17. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule(10 CFR 50.61), NUREG-1806, Vol. 1, August 2007.File No.: 1400187.302Revision: 2Page 10 of I IF0306-01R2 CStructural Integrity Associates, Inc!Table 1: LaSalle Weld ChemistryMean Chemistry%Cu I %Niý1 Nozzle-to-shell-weldIm No IIn no 0INI Nozzle Forging Blend RadiusNote: %Cu and %Ni were obtained from Reference 2, Table 5-1.Table 2: Probability of Failure Results SummaryPoF per year from PoF per year from Maximum PoF perLTOP events for 25% Normal Operating year [171In-Service Inspection Condition for 25% In-for period of Extended Service Inspection forOperation (Zero period of Extendedinspection for initial Operation (Zero40 years)* inspection for initial40 years)Nozzle Blend Radii 1.4 x 10-9 4.2 x 10-7 5.OE-6Nozzle-to-shell-weld <<2.0 x 10-10 3.3 x 10-9 5.OE-6*Note: Values include 1 x 10' probability of LTOP event occurrence.File No.: 1400187.302Revision: 2Page 11 of 11F0306-01R2 CStructural Integrity Associates, Inc.!APPENDIX ALIST OF SUPPORTING FILESFile No.: 1400187.302Revision: 2Page A- I of A-2F0306-OIRI CStructural Integrity Associates, Inc!File Name DescriptionLCNS_Blend_pl.INP VIPERNOZ input file for Path I at nozzle blend radii.LCNS _Blendcp3.1NP VIPERNOZ input file for Path 3 at nozzle blend radii.LCNS _Weldp2.1NP VIPERNOZ input file for Path 2 at nozzle-to-shell-weld.LCNS _Weld-p4.INP VIPERNOZ input file for Path 4 at nozzle-to-shell-weld.LCNS _Blendpl.OUT VIPERNOZ output file for Path 1 at nozzle blend radii.LCNS _Blend-p3.OUT VIPERNOZ output file for Path 3 at nozzle blend radii.LCNS _Weldp2.OUT VIPERNOZ output file for Path 2 at nozzle-to-shell-weld.LCNS _Weld-p4.OUT VIPERNOZ output file for Path 4 at nozzle-to-shell-weld.VIPERNOZv2.EXE VIPERNOZ executable programISPCTPOD.EXE VIPERNOZ probability of detection curve input fileFLWDSTRB.EXE VIPERNOZ flaw size distribution curve input fileFile No.: 1400187.302Revision: 2Page A-2 of A-2F0306-OIRI