EPRI Irradiated Materials Testing Programs

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EPRI Irradiated Materials Testing ProgramsBoiling Water Reactor Vessel Integrity ProgramMaterials Reliability ProgramPrimary Systems Corrosion ResearchJean SmithSr. Project ManagerEPRI – NRC Materials MeetingJune 5 – 7, 2013

Contents EPRI Irradiated Materials Testing and Degradation ModelsRoadmap Collaborative Research Programs– Zorita Internals Research Project Additional Zorita Materials Projects– Halden Research Program– IASCC Data Compilation and Analysis Primary Systems Corrosion Research Projects MRP Projects BWRVIP Projects 2013 Electric Power Research Institute, Inc. All rights reserved.2

EPRI Irradiated Materials Testing andDegradation Models Roadmap 2013 Electric Power Research Institute, Inc. All rights reserved.3

Zorita Internals Research ProjectMRP, BWRVIP, and PSCR 2013 Electric Power Research Institute, Inc. All rights reserved.4

Zorita Internals Research Project (ZIRP)Objective Increased understanding of irradiation effects on: Mechanical properties: tensile strength, fracture toughness,crack initiation and growth Microscopic properties: grain boundary chemistry and size,void formation, and hydrogen and helium production Project uses Zorita baffle plate materialIMT Gaps P-AS-14 (high): Fluence Impact on SCC of Stainless Steels(IASCC) P-AS-15 (medium): Void Swelling of Stainless Steels P-AS-38 (medium): Fluence Impact of Stainless SteelMechanical Properties (Fracture Toughness and TensileStrength) 2013 Electric Power Research Institute, Inc. All rights reserved.5

Zorita Internals Research ProjectCurrent Status Radiation analysis complete for all 29 cycles Temperature analysis for 8 to 9 representative cycles to becompleted shortly Segmentation and sample cutting underway Cask arrived at Zorita from Sweden in early May 2013 Electric Power Research Institute, Inc. All rights reserved.6

Zorita Internals Research ProjectCutting Plan – Baffle PlatesType 304 Doses ranging from a few dpa to 58 dpa Thickness 28.6 mm ZIRP uses Pieces B1, B2, and B3Plate A (41.22” wide)Plate B (7.8” wide)7.8”Plate C (7.8” wide)41.22”6.67” 52 dpa 32 dpa 15 dpa 11 dpa 2013 Electric Power Research Institute, Inc. All rights reserved.7

Zorita Internals Research ProjectCutting Plan – Core BarrelType 304 Core barrel weld has been located Weld material to be used by MRP,BWRVIP, U.S. NRC 2013 Electric Power Research Institute, Inc. All rights reserved.8

MRP Projects Using Additional Zorita MaterialsThermal and Irradiation Embrittlement and EnvironmentalEffects Testing of Stainless Steel Welds Objective: Determine the combined effects of irradiation and exposure to elevated temperatureon embrittlement of stainless steel welds and characterization of environmental effect onfracture toughness in irradiated stainless steel welds Project uses Zorita core barrel weld material P-AS-13 (high): Thermal & Irradiation Embrittlement Synergistic Effects on CASS andStainless Steel WeldsCGR Testing of Irradiated SS Weld and HAZ Materials Objective: Develop IASCC CGR data in irradiated stainless steel weld and HAZ materials forcomparison to existing data for base materials Project uses Zorita core barrel weld material P-AS-14 (high): Fluence Impact on SCC of Stainless Steels (IASCC)Determination of IASCC CGR, Initiation Rate, and Void Swellingin Zorita Material after Post-Reactor Irradiation Objective: Evaluate IASCC crack initiation and crack growth rates and degree of void swellingin highly-irradiated (near end-of-life conditions) stainless steel base metal and welds Project uses Zorita baffle plate & core barrel weld material P-AS-14 (high): Fluence Impact on SCC of Stainless Steels (IASCC) P-AS-15 (medium): Void Swelling of Stainless Steels 2013 Electric Power Research Institute, Inc. All rights reserved.9

BWRVIP and PSCR Projects Using AdditionalZorita MaterialsCrack Growth Rate & Fracture Toughness Testingof Weld Materials (BWRVIP) Objective: Develop data to help establish the SCC growth ratedisposition curves and understand the relationship between fracturetoughness (FT) and dose for austenitic stainless steels in the BWRenvironment Project uses Zorita core barrel base, weld, and HAZ materials B-AS-09 (high): Assess the Impact of High Fluence on FractureToughness B-DM-06 (medium): Environmental Effects on Fracture ResistanceK-size Effect Testing (PSCR) Objective: Determine the applicability of using data generated fromsmaller specimens to assess the performance of thicker internalscomponents Project uses Zorita baffle plate material 2013 Electric Power Research Institute, Inc. All rights reserved.10

Halden Research ProjectMRP, BWRVIP, FRP, and PSCR 2013 Electric Power Research Institute, Inc. All rights reserved.11

Halden Research ProgramPlant Aging and Degradation EPRI participates in the Halden program which issupported by members from 18 countries andmore than 100 organizations Halden Material Testing Program– Investigations are performed in representative thermalhydraulic, coolant chemistry, and radiation conditions– On-line instrumentation and control of Coolant temperature Water chemistry (H2 / O2) Specimen stress level Constant load vs cycling– Samples manufactured from materials retrieved frompower reactors 2013 Electric Power Research Institute, Inc. All rights reserved.12

Halden Research Program2012 to 2014 Test Plan HighlightsTaskObjectiveApplicability2.1.a. BWR Crack GrowthRate Study Generate long-term CGR data in irradiated specimens in BWR conditions Determine the benefits of HWC in mitigating cracking in irradiatedmaterials with different fluence levelsBWR2.1.b. PWR Crack GrowthRate Study Generate long-term CGR data in irradiated specimens in PWR conditions Determine the effects of hydrogen concentrations, Li/B ratios, and/or Znadditions and the benefits of Post-Irradiation Annealing (PIA)PWR2.1.c. Irradiation of CW 316SS CTs for Future PWRTests Irradiate CW 316 SS CT specimens to higher (4 to 5 dpa) dose for futurePWR CGR testsPWR2.1.d. Crack Initiation(Integrated Time-to-Failure)Study Evaluate the benefits of HWC in mitigating the initiation of cracksirradiated (12 dpa) 304L SS tensile specimens Test under BWR HWC conditions and PWR conditions (including withhigh Li (3.5 ppm) and with Zn additions)PWR/BWR2.1.e. Irradiated MaterialsCharacterization Obtain detailed information on the irradiation history and mechanical andmicrostructural properties of materials being used in IASCC studiesPWR/BWR2.2 Stress Relaxation Study Establish the technical feasibility of on-line irradiation stress relaxationmeasurement during irradiation Measure the irradiation stress relaxation of materials used in LWRs andbenchmark the irradiation stress relaxation data to irradiation creep testsPWR/BWR2.3 RPV Integrity Study Establish proper correlations between the neutron embrittlement datafrom sub-size specimens with those of standard sizePWR/BWR 2013 Electric Power Research Institute, Inc. All rights reserved.13

Halden Research ProgramIASCC: Crack Growth ProgramObjectives Measure on-line cracking response ofmaterials retrieved from commercialreactor components Compare & quantify CGRs in BWR(280-290 C, O2 and H2) vs. PWR(320-340 C, Li, B, H2) conditions Study cracking response as affectedby–––––Stress intensity (K) levelFluenceMicrostructureMechanical properties (yield strength)Flux 2013 Electric Power Research Institute, Inc. All rights reserved.14

Halden Research ProgramIASCC: Time-to-Failure programObjective Determine effectiveness of HWCin reducing susceptibility to theinitiation of cracks in irradiatedmaterialExperimental Miniature tensile specimensprepared from 304L SS Load (up to 100% of YS)applied by means of bellows On-line monitoring of specimenfailures (LVDTs) Compare number of failuresunder HWC & NWC conditions Study in PWR environment isplanned 2013 Electric Power Research Institute, Inc. All rights reserved.15

IASCC Data Compilation and AnalysisPSCR, BWRVIP, and MRP 2013 Electric Power Research Institute, Inc. All rights reserved.16

EPRI IASCC Data Compilation & AnalysisObjectives and ScopeCompile crack growth rate data on irradiated stainless steelsCIR-fast reactor irradiated materialsHalden-LWR irradiated materials & in-reactor testsBWRVIP and MRP irradiated materialsLiterature data (ANL NUREG reports, JNES reports)Convene an Expert PanelReview, screen and categorize the available datausing consensus screening criteriaPanel includes principle investigators, selectedindustry experts, and vendors.Recommend crack growth models and disposition curvesCrack growth models and disposition curves forirradiated stainless steels in BWR and PWRenvironments 2013 Electric Power Research Institute, Inc. All rights reserved.Final report will be issued after review by PSCR, MRP,and BWRVIP17

EPRI IASCC Data Compilation & AnalysisExpert Panel Status Database includes more than 1600 test segments includingthose under cyclic loading For IASCC crack growth rates only test segments underconstant load or periodic partial unloading were considered Test segments ranked on a scale of 1 (best) to 5 (worst)after examining the raw data from crack length vs. timeplots Only data ranked from 1 to 3 was considered suitable fordevelopment for models for BWR NWC, HWC and PWRenvironments 2013 Electric Power Research Institute, Inc. All rights reserved.18

EPRI IASCC Data Compilation & AnalysisNext Steps Work with the EPRI Expert Panel to produce and document the finalconsensus PWR and BWR IASCC models from the extensive EPRIIASCC database that was compiled and ranked by the Expert Panel inearlier phases of this Project A revised version of the draft report on the IASCC database and bothlow ECP (PWR and HWC) and high ECP (NWC) IASCC models,incorporating all revisions requested by the Expert Panel and EPRIreviewers will be prepared by November 2013 The draft report will be sent for PSCR, MRP and BWRVIP review inDecember 2013 Final report will be published in the second quarter of 2014 aftercomment resolution The report will provide the technical basis for crack growth dispositioncurves for irradiated BWR and PWR stainless steel internals 2013 Electric Power Research Institute, Inc. All rights reserved.19

PSCR: Mechanistic understanding on IASCCObjectives Identify the key factors influencing IASCC initiation andcrack growth. Understand the linkage between irradiatedmicrostructures and IASCC Confirm the processes that lead to occurrence ofIASCCGaps Lack of understanding on IASCC No mitigation strategies available Uncertainty on reliability of components for LTO 2013 Electric Power Research Institute, Inc. All rights reserved.20

PSCR: Mechanistic Studies on IASCCClosely collaborating with US DOE, toperform fundamental researches for betterunderstanding of IASCC mechanisms Identification of Key Factors Affecting IASCC ofAustenitic Alloys in LWR Core Materials Establishing a Cause-and-Effect Relationshipbetween Localized Deformation and IASCC APT and Nano-SIMs Characterization ofProton- and LWR Neutron Irradiated StainlessSteels Investigation of PIA as a potential strategy formitigating IASCC of reactor core structuralcomponentsNishioka et al.J. Nucl. Sci. Techn. 45 (2008) 274. 2013 Electric Power Research Institute, Inc. All rights reserved.21

PSCR: Mechanistic Studies on IASCCIdentification of Key Factors Affecting IASCC Assess the role of solute additions, in particular the roles of C, Mo, Ti, Nb, Cr Niand P, on crack growth (CGR) and crack initiation (CI). Understand the linkage between irradiated microstructures and CGR/CI forsolute addition and commercial alloys, also effects of CW and dose. Determine the predictive capability of crack initiation due to proton irradiation,compared to crack initiation due to neutron irradiation. Investigate the role of localized deformation on the IASCC susceptibility inneutron irradiated materials.BWR NWC1.4RA (%) or %IG501.21400.8300.620Elongation (%)60crack growthIASCC initiation0.410NA0.2005.5 (AS13) 10.2 (AS17) 47.5 (AS22)Dose, dpa (Sample) 2013 Electric Power Research Institute, Inc. All rights reserved.22

PSCR: Mechanistic Studies on IASCCIASCC Initiation Mechanism StudyEstablish direct evidence of a cause-and-effect relationship Site-specific approach to confirm whether an IASCC crack can beinitiated at a pre-characterized site Direct measurement approach to determine the degree of localizeddeformation at the crack sites Measurement of displacements due to plastic and elastic strains Digital image correlation measurement for in-plane displacementConfocal microscopy for out-of-plane displacementsEBSD for elastic stressesDevelopment of IASCC mitigation strategy Focus on removing or reducing the degree of localized deformation–––Post-irradiation annealing to remove the dislocation loopsCold-work to reduce the degree of localized deformationPrecipitation-strengthened alloys to prevent localized deformation 2013 Electric Power Research Institute, Inc. All rights reserved.23

PSCR: Development of Advanced RadiationResistant Materials (ARRM)Objectives Develop the next generation of materials for in-corestructural components and fasteners. Determine the degradation-resistance of the currentcommercial alloys Determine the degradation-resistance of the newadvanced alloysGaps No resistance materials for replacement of RIcomponents No resistance materials for new plants (RI components) 2013 Electric Power Research Institute, Inc. All rights reserved.24

PSCR: Development of Advanced RadiationResistant Materials (ARRM)ARRM Project EPRI and the U.S. Department of Energy (DOE) are initiating a global,collaborative research effort to develop the next generation ofmaterials for in-core structural components and fasteners. The two primary research goals are:– By 2022, to develop and test a degradation-resistant alloy that iswithin current commercial alloy specifications– By 2024, to develop and test a new advanced alloy with superiordegradation resistance 10-year project, estimated 12 -15 M work-scope Project work-scope defined by sponsors and managed by EPRI as wasdone in the CIR and NFIR programs 2013 Electric Power Research Institute, Inc. All rights reserved.25

PSCR: Development of ARRMsFuture Steps EPRI and DOE have published “Critical Issues Reportand Roadmap for Advanced Radiation ResistantMaterials Program”, (EPRI Report 1026482, December2012) Both organizations will initiate a long term ( 10 years)collaborative research program to develop and qualifymore radiation resistant materials based this report We are looking for partners (vendors, researchorganizations, utilities, regulators) to join us in thisimportant effort Organizations of the program will be similar to otherEPRI collaborative programs (CIR and NFIR) 2013 Electric Power Research Institute, Inc. All rights reserved.26

MRP: Gondole Void Swelling Irradiation andTestingObjective (Phase 2) Increase accumulated dose to 30 dpa for virgin samples Provoke swelling on other materials to determine kinetics of swelling Investigate possible existence of threshold temperature for swellingIMT Gap P-AS-15 (medium): Void Swelling of Stainless SteelsPhase 1 Results Virgin PWR internals materials & pre-irradiated materials exposed for 5 cycles ofirradiations Virgin materials accumulated 14 dpa Pre-irradiated materials had up to 85 dpa Type 316: No significant swelling; several cases of densification Type 304: 6 cases of significant swelling Type 347 and NMF 18 CW: Significant swelling on each *Significant swelling/densification defined as change in volume 5% 2013 Electric Power Research Institute, Inc. All rights reserved.27

MRP: Gondole Void Swelling ProjectComparison to Void Swelling Model Cluster Dynamics model developed to characterize void swelling in austeniticstainless steels Model being benchmarked with Gondole data1E-7 dpa/s, 287 C, 20 appm He/dpa1E-7 dpa/s, 330 C, 20 appm He/dpa3E-7 dpa/s, 360 C, 10 appm He/dpa3E-7 dpa/s, 360 C, 10 appm He/dpa Gondole Data Gondole DataBlue line segment: Modeling of in-service irradiationRed line segment: Modeling of test-reactor irradiation (Osiris, 360 C)Black squares: Gondole data (irradiation Phases 1 to 5) 2013 Electric Power Research Institute, Inc. All rights reserved.28

MRP: Lithium Effects on IASCC InitiationObjective Determine the effect of lithium (Li) on the rate of IASCC initiationfor comparison to recent data generated by EDF suggestingincreased Li concentration may enhance IASCC initiation rateIMT Gap P-AS-14 (high): Fluence Impact on SCC of Stainless Steels(IASCC)Experimental Design Material: Flux thimble tubes from Ringhals Unit (60 and 100 dpa) Lithium levels 2.0 and 8.0 ppm UCL and o-ring specimens 2013 Electric Power Research Institute, Inc. All rights reserved.29

MRP: Effect of Lithium on SCC InitiationBackgroundC7 800 MPa, PWR, 340 CC4 700 MPa, PWR, 340 CC6 700 MPa,PWR, 340 C (fast loading)C8 400 MPa, PWR, 340 C : not brokenC3 500 MPa, PWR,340 CC9 700 MPa, PWR, 290 CC2 600 MPa, PWR, 340 CC1 500 MPa, PWR high Li, 340 CBehavior Chooz A - 30 dpa ?C12 400 MPa, PWR high Li, 340 CC8 350 MPa, PWR high Li, 340 C : not brokenBehavior Chooz A - 30 dpa - High Li ?900800340 C 290 C700No crackingStress (MPa)600[Li] 3.5 ppm [Li] 2.2 ppm500400300200No crackingNo cracking (SEM to be performed)1000110 2013 Electric Power Research Institute, Inc. All rights reserved.100Time to failure (h)30100010000

MRP: Dynamic Strain Effects on IASCCInitiation RatesObjective Compare existing results from static-loaded tests to testsconducted using dynamic loads representative of PWRtransients to better understand EDF baffle bolt experience andIASCC test observationsIMT Gap P-AS-14 (high): Fluence Impact on SCC of Stainless Steels(IASCC)Participation Project hosted by NUGENIA; led by EdF 13 organizations collaborating to provide materials, in-pile & outof-pile testing, modeling 2013 Electric Power Research Institute, Inc. All rights reserved.31

MRP: Dynamic Strain Effects on IASCC Initiation RatesBackground Few data on the effects of temperature or stress variation on IASCC sensitivity Halden Project showed an association of specimen failures with test interruptions andspecimen re-loading. Data obtained on non-irradiated materials show an effect of dynamic loading on initiation Possible role of dynamic straining in IASCC initiation.By Torril Karlsen - Halden 2013 Electric Power Research Institute, Inc. All rights reserved.32

MRP: Dynamic Strain Effects on IASCC Initiation RatesExperimental Approach1. Numerical analysis of the most severe transients for PWRreactors2. Post-irradiation mechanical testing of stainless steel materials(out-of-pile tests) SA 304 and CW 316 irradiated in experimental or commercial reactors 4 and 20 dpa (typical range for IASCC failures) PWR environment using transients identified in Phase 13. In-flux testing of pre-irradiated materials in a material testreactor Neutron flux with load increments or temperature increment to simulate the shutdown and start-up of PWRTests will also be performed without load or temperature increment for comparison4. Tests on non-irradiated materials Assess tapered specimens with representative CWSensitivity to transients on SCC initiation5. Modeling of the effects of transients at grain scale 2013 Electric Power Research Institute, Inc. All rights reserved.33

BWRVIP : Testing of Unirradiated and IrradiatedX-750 and XM-19 MaterialsObjective Develop additional mechanical property data,SCC susceptibility, and CGR data to determinethe long-term viability of currently installedmaterials. Irradiation of materials to be performed at INLATR NSUFIMT Gaps B-AS-26 (high) High-Strength Alloys B-RR-05 (medium) Alternative High-StrengthMaterials 2013 Electric Power Research Institute, Inc. All righ

Crack Growth Rate & Fracture Toughness Testing of Weld Materials (BWRVIP) Objective: Determine the applicability of using data generated from smaller specimens to assess the performance of thicker internals components Project uses Zorita baffle plate material . K-size Effect Testing (PSCR)

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