Material Development and Testing for Extreme Reactor ApplicationsS.A. Maloy1, O. Anderoglu1, T. Saleh1, M. Caro1, K. Woloshun1, F. Rubio1, M. Toloczko2, D. Hoelzer3, T.S. Byun3, G.R. Odette41Los Alamos National Laboratory, Los Alamos, NM 87545, USA2Pacific Northwest National Laboratory, Richland, WA 99352, USA3Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA4University of California Santa Barbara, Santa Barbara, CA 93106, USA

Funded by Department of Energy-Nuclear Energy

DOE-NE is Performing Research to close the Nuclear Fuel Cycle Fuel Cycle R&D (FCRD) Program*

Advanced Fuels Campaign Mission & Objectives in the Fuel Cycle Research and Development ProgramMissionDevelop and demonstrate fabrication processes and in-pile (reactor) performance of advanced fuels/targets (including the cladding) to support the different fuel cycle options defined in the NE roadmap.ObjectivesDevelopment of the fuels/targets thatIncreases the efficiency of nuclear energy productionMaximize the utilization of natural resources (Uranium, Thorium)Minimizes generation of high-level nuclear waste (spent fuel) Minimize the risk of nuclear proliferationGrand ChallengesMulti-fold increase in fuel burnup over the currently known technologiesMulti-fold decrease in fabrication losses with highly efficient predictable and repeatable processes Once-ThroughModifiedOpenContinuousRecycleAdvanced FuelsHigh-burnup LWR fuelsAccident Tolerant Fuels for improved Safety Deep-burn fuels or targets after limited used fuel treatment High burnup fuels in new types of reactors Fuels and targets for continuous recycling of TRU in reactors (possibly in fast reactors)

*Scientific Approach to Enabling a Multi-fold Increase in Fuel Burnup over the Currently Known Technologies

Ultra-highBurnupFuelsCoatingLinersAdvanced AlloysF/M SteelsAdvanced AlloysCrSiAlIncreasing contentF/M SteelsHT-9Advanced F/M Steels, e.g. NF616ODS SteelsAdvanced Alloys200 dpa 300 dpa 400 dpa500 C 600 C 700 C Reduced embrittlement, swelling, creepEnhancements with Fabrication ComplexityEnhancements withFabrication ComplexityEnhancements with Fabrication ComplexityDifferent Reactor options to change requirementsLFR, GFRFCCIRadiationTemperatureCorrosionDevelop an advanced materials immune to fuel, neutrons and coolant interactions under specific reactor environments

OutlineLANL CMR Hot Cell Tensile TestingACO-3 Duct AnalysisTensile TestingCharpy TestingMicrostructural AnalysisSTIP IV irradiated MaterialsAccident Tolerant Cladding MaterialsAdvanced Material Development for High Dose ApplicationsODS steel ProcessingIon Irradiation TestingNeutron Irradiation Testing to high doseDevelopment of Lead Corrosion Resistant MaterialsLead Fast ReactorsDELTA loopLong Term corrosion tests (>3,000 hours)Summary and Future Work

Our Unique Facilities and Infrastructure Providing Science and Technology to NE

Actinide Cross-Section Measurements Irradiation of Targets for Isotope Production DELTA loop for LBE CorrosionActinide R&D: Separations, Integrated Safeguards Test Lab, Characterization of Irradiated Materials Information Sciences Fuels Research Lab:R&D on Ceramic FuelsHot-Cells for Processing of IsotopesNuclear Fuels R&D: MOX Fuel Fabrication and Testing Ion Beam Materials Laboratory, Structural Materials and Fuels R&D

Slide *Mechanical Testing in CMR Wing 9 Hot CellsTensile Specimen- dimensions are 4 mm x 16 mm x 0.25 mm thick (gage dims. are 1.2 x 5 x 0.25 mm3)Tested at initial strain rate of 5 x 10-4/s.Tested at 25 to 700C in ultra high purity argonShear Punch, 3 pt. bend and compression testing capabilities.

16 mm

Slide *Testing was Performed on the ACO-3 Duct, one of the most highly irradiated components in a fast reactorFFTF, Hanford site, WA

*Analysis of Specimens from ACO-3 DuctTotal specimens= 144 Charpy, 57 compact tension, 126 tensile specimens, 500 TEMCharpy, Compact Tension testing and thermal annealing completed at ORNL. Completed tensile testing at LANL from 6 different locations along the duct at 25, 200 and the irradiation temperature.Completed Rate Jump Testing at 25CCompleted Microstructural Analysis using TEM, SANS and Atom Probe Tomography

Stress/Strain Curves HT-9 Irradiated, Room Temp Tests*Decreased elongation in irradiated materials.Increased hardening in lower irradiation temperature materials.

*Results from Previous PIE Studies Agree Well With These MeasurementsAC0-1 Duct and Cladding (HT-9) (total dose = ~ 88 dpa)Maximum dilatation (swelling + precipitation + creep) = 0.5 %Yield stress increase ~300 MPa, Tirr =~360C, Ttest= 25C, dose = 36 dpaPrevious studies (stars in figure, tested at 25C) show similar dependence of yield stress on irradiation temperature

Charpy Impact Testing for ACO-3 Duct Upper shelf energy is a function of irradiation temperature, dose, and specimen orientation, while the effect of irradiation temperature is dominant in the transition temperatures.

*TEM analysis of ACO-3 Duct Material (B.H. Sencer, INL, O. Anderoglu, J. Van den Bosch, LANL)T=384C, 28 dpa G-phase precipitates and alpha prime observedNo void swelling observed.T=450C, 155 dpa Precipitation observed Dislocations of botha/2 and a Loops of a Void swelling observed (~0.3 %)

T=505C, 4 dpaNo precipitation or void swelling observed.

Small Angle Neutron Scattering MeasurementsObtain accurate measurement of vs. dose and irr. TemperatureMeasurements completed on 5 specimens from ACO-3 duct

Slide *Summary

TCdpaCr solubat %(G. Bonny)Cr solubat %(Phase diagram)G-phaseLoopsvoidsLnmdx1021m-3Lnmdx1021m-3LnmdX1021 m-3LnmdX1021 m-33802098.17.87211.39.3 140.93XX4101009.110.192.616.21.4--23-4401559.312.49.69.526.51.1180.5280.25466929.914.7XXXXXX505212.518.5XXXXXX47510,000 hrs10.315.5XXXXXX

*STIP- (SINQ Target Irradiation Program) Irradiations Provides an Understanding of the Effects of Helium and Irradiation Dose~570 MeV protonsHe/dpa ratios of 50-60 appm/dpaMaterials for STIP IV irradiation include the following in tensile and TEM specimens:

Structural: HT-9, EP-823, Mod 9Cr-1Mo, 9Cr-2WVTa, T122, 5Cr-2WVTa, A21N, ODS strengthened F/M steels-12YWT and 14YWT (Fe-12Cr-3W-0.4Ti-0.25 Y203, Fe-14Cr-3W-0.4Ti-0.25 Y203), V-4Cr-4Ti, High purity Ta, single crystal Fe (for modelling studies)

Fuels Matrices: ZrN, NiAl, FeAl, RuAl, MgO, Cubic ZrO2, Fissium

Summary of STIP IV Tensile Testing at Room TemperatureHighest Hardening observed at lowest irradiation temperatures12YWT tests reached limits of testing machine before yieldingSignificant helium accompanies dpa (up to 1300 appm He)Comparison with Phenix irradiated specimens will help quantify helium effect on mechanical properties12YWT-ControlTirr=394C, dose=22 dpaTirr=247C, dose=15 dpaHT-9-ControlTirr=380C, dose=22 dpaTirr=247C, dose=15 dpaTirr=120C, dose=8 dpa

Requirements for Accident Tolerant Fuels (ATF)

Measurements on hydrogen evolution performed in steamHydrogen Production begins in Zircaloy-4 at ~700C and in 304L at ~1000CSimilar testing is underway on all advanced alloys in FY13

Zirc-4 in N2 containing ~25% water vapor to 1100C 304 in N2 containing ~25% water vapor to 1100C

*Advanced Material Development ActivitiesCharacterizing and Testing MA-957 Irradiated to High Dose (>100 dpa)Obtaining Irradiation data on Advanced Alloys (international collaborations)MATRIX irradiations- Samples to be shipped in early 2013?STIP irradiations Samples from STIP IV to be shipped in next few weeksInvestigating Possible Future irradiationsDomestic Facilities (MTS (18 dpa/yr)) Collaborating in ATR irradiationsInternational collaborationsCollaborating with Terrapower for irradiations in BOR-60 in Russia and DOE-RIAR collaborations for additional irradiations in BOR-60.Initial discussions under way for future irradiation in the CEFR in China.Advanced Material Development Friction stir ODS material processingMechanical alloying ODS material processing

Oxide Dispersion Strengthened AlloysStrength & damage resistance derives from a high density Ti-Y-O nano-features (NFs)NFs complex oxides (Ti2Y2O7, Y2TiO5) and/or their transition phase precursors with high M/O & Ti/Y ratios (APT)MA dissolves Y and O which then precipitate along with Ti during hot consolidation (HIP or extrusion)Oxide dispersion strengthened alloys also have fine grains and high dislocation densities

Typical Processing Route for ODS Alloys

Ion Beam Materials Laboratory (IBML)Fundamental irradiation studies performed at the IBML

Irradiations performed on interfaces characterized to the atomic scale

Post irradiation analysis will investigate the role of interfaces on defect formation and accumulation

Aids in model development and provides initial alloy irradiation results.

Nanohardness [GPa]Dose [dpa]HT-9 MartensiticHT-9 FerriticDoseDose [dpa]Nanohardness [MPa]Depth [mm]1mmMA957Room temperature irradiation (1.5dpa)Depth [mm]MA956ODS Strengthened Materials Show Excellent Resistance to Hardening under Ion Irradiationion beam4mm4mm10mmBeamBerkovich indenter, 200nm deep indents, constant displacement

Chart1

4.7789153753.2214328750.16529784890.72564831780.1800507271.10784934321.1078493432

4.0452871253.375458750.57278289681.04119803780.39108396810.3910839681

3.5558346253.264539751.24893289320.61803630120.28354281030.2835428103

4.3223693.71982351.67093078960.64979399340.57739711770.5773971177

4.1208201255.0590088751.2912662210.22178171430.17603743850.1760374385

4.346354254.9006038751.18774901290.35789755230.25347808230.2534780823

4.347194.8474258751.74796529960.2595859310.17194793150.1719479315

4.1410294.8995778751.54138000950.21828409320.1213393780.121339378

4.129148754.84502651.27840576690.35789755230.26572152810.2657215281

4.1648851254.9593858751.43412658150.2595859310.20275072460.2027507246

4.5989791255.157163251.58416471280.21828409320.18709712680.1870971268

4.4243595.0356948751.42425030450.19784817570.21309094350.2130909435

3.3452975.4089331.66434235620.20154230990.17958750550.1795875055

3.2859074.2491081.53166118910.55346310140.10112826630.1011282663

3.3308623754.4928841.61842426670.1800507270.21726795250.2172679525

4.43609971431.68319290870.25307473980.2530747398

7.81.5575418033

8.42.0250109123

1.5389851698

1.7648671394

1.8064559225

1.9790808262

1.7137132436

1.9371958748

1.573094765

1.8227616767

1.9555950517

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4.0552501331

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15.2273040154

21.6381339161

30.7978558901

36.9936150904

33.2077364359

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2.0667436558

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0.0025430423

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HT-9 Ferritc RT

HT-9 Martensitic RT

dose

Sheet6

HT-9 HT HE RTHT-9 HT RT PHT-9 HT P 300CHT-9 HT P 550CHT-9 HE RTHT-9 RT PHT-9 P 300CHT-9 P 550C

percent hard increase=312836200724175

2735738237237231000273573823

797

273273273

Sheet6

37

1224

817

35

HT-9 HT

HT-9

Dose P summary

HT-9 1dpa Ferritic

HT-9 1 dpa Martensitic

anealeing temperature

HT-9 Martensitic He 0.2 dpa RT

MIBML datqa

Dose P

====== H (10) into Layer1=======

SRIM-2006.02

==========================================

Ion andTarget VACANCY production

See SRIM Outputs\TDATA.txt for calc.details

==========================================

See file :SRIM Outputs\TDATA.txt for calculationdata

Ion = HEnergy = 900 keV

============0RIAL ======================================

Layer 1 : Layer 1

Layer Width100000A ;

Layer # 1-Density = 8.465E22 atoms/cm3 = 7.785 g/cm3Take the charge in red box and divide by charge of 1 proton 1.602E-19 C/p then divide by area gives fluence

Layer # 1-Fe = 88 Atomic Percent= 88.7 Mass Percent

Layer # 1-Cr = 12 Atomic Percent= 11.2 Mass PercentThis assumes that the current is right. I couldn't get hold of Christiana. Current x Time = charge

====================================================================

Total Ionscalculated =5123

Total Target Vacancies0

Total Target Displacements = 34 /Ion

Total Target ReplacementCollisions =2 /Ion

634740948813982000036235955056179800007402247191011240000

!!!! NOTE :2nd Column below is numberof Primary Knock-Ons !!!!

(PKO are number of TargetAtoms Recoiling fromtheIon. )Area(cm2):0.125mmx 5mm2mm x 2mm

==========================================================Currenthours

Table Units are >>>> Vacancies / Angstrom / Ion> Vacancies / Angstrom / Ion> Vacancies / Angstrom / Ion

*Analysis of highly irradiated MA-957 Tubes Underway at PNNLTensile testing of MA-957 Pressurized tubesIrradiation conditions in FFTF(385C, 18-43 dpa)(412C, 110 dpa)(500-550C, 18-113 dpa)(600-670C, 34-110 dpa)(750C, 33-120 dpa)Testing will be performed at PNNLStatusSpecimens for testing were machined from pressurized tubes at LANLTensile testing and TEM work is underway at PNNLAnalysis of in-reactor creep response is complete.Preliminary analysis of creep dataMA-957 is comparable to HT-9 in creep resistance to up to 550C. At 600C, MA-957 creep resistance remains high while HT-9 creep resistance begins to rapidly decline.

New ODS 14YWT heat produced with low N and C powderNew consolidation condition explored for 14YWT2 cans heat treated to nucleate nanoclusters (1 h @ 750C & 850C)Cans extruded at 1150CFabricationExtruded bar cut into 3 sectionsOne section rolled parallel to extrusion axis at 1000COne section rolled normal to extrusion axis at 1000CTotal reduction in thickness was 55% (~5.5 mm final thickness of 14YWT) No cracking was observed

Core Materials Research and Development 5 Year Plan*FY16FY15FY14FY11FY13FY12STIP- IV (PSI) Specimen PIE MATRIX-SMI and 2 (Phenix) Specimen PIERe-irradiation of FFTF specimens in BOR-60 Materials Test Station Irradiations Provides data for NEAMS model development of CladdingFFTF (ACO-3 and MOTA) Specimen AnalysisAdvanced Material Development (improved radiation resistance to >400 dpa) Qualify HT-9 for high dose clad/duct applications (determine design limitations) Innovative Material DevelopmentInnovative Clad Material DownselectODS Ferritic Steel Material DevelopmentProduce ODS TubingAdvanced Materials Irradiation in BOR-60 and CEFRAdvanced Material Development (improved FCCI resistance to >40 % burnup) Development of Coated and Lined TubesInnovative Coating Material DevelopmentRev. 6 of AFCI (FCRD) Materials HandbookAccelerated Aging of Advanced Materials (High Dose Ion Irradiations)PIE on Lined Irradiated TubeDevelop ODS Tubing and Weld specifications for innovative Weld materialBaseline Mech. Prop. Of Inn. Clad MaterialLined Tube for ATR irradiationFab. Innovative Coated Tube for ATR irradiationData to 250-300 dpa on F/M and 100-150 on Inn. MaterialData on Advanced Materials to 80-100 dpa

7/20/09Compositional variations

Bulk diffusion

Bulk thermal conductivityRole of idealized grain boundaries in mass and thermal transport.

All results provided to MBMMARMOT model development; mesoscale fission gas diffusion and segregation, thermal conductivity (INL)Analysis of segregation in terms of local strainDevelopment of a Physics-based Model of Radiation Damage in Nuclear FuelsMotivation/Approach Develop mechanistic materials models with improved accuracy and predictive power using atomic level simulation techniques for application in meso-scale and/or continuum models (MBM).

LBE as a Nuclear Coolant and Spallation Source TargetOpportunityExcellent neutron yieldLow neutron reaction cross sectionsLow melting pointExcellent thermal propertiesNot susceptible to radiation damage

ChallengeHighly corrosive to steelLiquid Metal EmbrittlementLiquid Metal Enhanced CreepXT-ADS, EUMYRRHAHyperion, USSVBR-75, Russia*

Conventional and Innovative Materials Selection *Ferritic/Martensitic steels and graded composites12 Cr F/M HT-9 steel and Russian EP823 steel (LANL)New HT composite tube integrated to Delta Loop; provided by R. Ballinger (MIT, USA)

Oxide dispersed strengthened (ODS) and model alloysPM 2000 and MA956 ODS steels and FeSi and FeSi model alloys (LANL)

Commercially available Fe-Cr-Al alloysKanthal Series steels (UCB, Berkeley)APM, APMT, Alkrothal 14, Alkrothal 720 for high temperature applicationsAlkrothal 3 (UU, KTH, Sweden)

Other possible candidate core materialsProposed for fuel cladding, heat exchangers in ADS/LFRsSCK.CEN, Belgium CollaborationT91 9% Cr ferritic steel D9 (DIN 1.4970) austenitic SS for in-vessel componentsAISI 316 austenitic SS for vessel and in-vessel components

LANL Delta Loop Design and Capabilities*Design/Performance :* Up to 2 m/s in 2.54 cm diameter ~ 3 m long test section* Up to 100C DT between heater section and heat exchanger exit* Up to 550C operation in the test section* Capable of free convection flow* All 316L construction* Robust/Safe to operate continuously* Gas injection system for oxygen control

DELTA isometric view

DELTA Corrosion Test Plan ** Corrosion resistance: Exposure up to 3000h in flowing LBE at 500oC and Oxygen concentration 10-6-10-8 wt%.* Flow-rate resistance: LBE velocities up to 3.5 m/s

* Grand total of 144 specimens tested Thin test coupons are placed in a cylindrical holder that is lowered into the test section.Gen-4 Module US Technological Impact: Understanding flow velocity and high-temperature effects on LBE steel corrosion properties for exposure times >2000h is critical in the conceptual design of advanced systemCanister loaded with 48 specimens

Conclusions and Future Work** Achievements: First specimens retrieved (May 2013) after ~ 900 h LBE exposureOngoing loading of second canister (2000 h) and preliminary corrosion studies* Future Work: Retrieval of specimens after 2000h and 3000h exposure* CollaborationsUCB Post Exposure Studies: Cross-section micro Raman, SEM/EDS/FIB/EBSD, nano-indentation studies* Possible Future International Collaborations:MYRRHA Accelerator Driven System - SCK.CEN, BelgiumELECTRA European LBE Fast Reactor KTH, Uppsala Univ., Sweden

Publications: Contribution submitted to JOM August 2013 and Materials Selection Milestone LANL Reports LANL Science Highlights PADSTE AOT MST (Oct. 2012) F. Rubio et al., Rio Grande Symp. on Advanced Materials RGSAM 2012

****************Material specificity*