Presentation on theme: "Material Development and Testing for Extreme Reactor Applications S.A. Maloy 1, O. Anderoglu 1, T. Saleh 1, M. Caro 1, K. Woloshun 1, F. Rubio 1, M. Toloczko."— Presentation transcript:
Material Development and Testing for Extreme Reactor Applications S.A. Maloy 1, O. Anderoglu 1, T. Saleh 1, M. Caro 1, K. Woloshun 1, F. Rubio 1, M. Toloczko 2, D. Hoelzer 3, T.S. Byun 3, G.R. Odette 4 1 Los Alamos National Laboratory, Los Alamos, NM 87545, USA 2 Pacific Northwest National Laboratory, Richland, WA 99352, USA 3 Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA 4 University 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 2
Advanced Fuels Campaign Mission & Objectives in the Fuel Cycle Research and Development Program Mission Develop 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. Objectives Development of the fuels/targets that –Increases the efficiency of nuclear energy production –Maximize the utilization of natural resources (Uranium, Thorium) –Minimizes generation of high-level nuclear waste (spent fuel) –Minimize the risk of nuclear proliferation Grand Challenges –Multi-fold increase in fuel burnup over the currently known technologies –Multi-fold decrease in fabrication losses with highly efficient predictable and repeatable processes Once- Through Modified Open Continuous Recycle Advanced Fuels High-burnup LWR fuels Accident Tolerant Fuels for improved Safety High-burnup LWR fuels Accident Tolerant Fuels for improved Safety - Deep-burn fuels or targets after limited used fuel treatment - High burnup fuels in new types of reactors - 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)
4 Scientific Approach to Enabling a Multi-fold Increase in Fuel Burnup over the Currently Known Technologies Ultra-high Burnup Fuels Ultra-high Burnup Fuels Coating Liners Advanced Alloys F/M Steels Advanced Alloys Cr Si Al Increasing content F/M Steels HT-9 F/M Steels HT-9 Advanced F/M Steels, e.g. NF616 ODS Steels Advanced Alloys 200 dpa 300 dpa 400 dpa 500 C 600 C 700 C Reduced embrittlement, swelling, creep Enhancements with Fabrication Complexity Enhancements with Fabrication Complexity Enhancements with Fabrication Complexity Different Reactor options to change requirements LFR, GFR FCCI Radiation Temperature Corrosion Develop an advanced materials immune to fuel, neutrons and coolant interactions under specific reactor environments
Outline LANL CMR Hot Cell Tensile Testing ACO-3 Duct Analysis –Tensile Testing –Charpy Testing –Microstructural Analysis STIP – IV irradiated Materials Accident Tolerant Cladding Materials Advanced Material Development for High Dose Applications –ODS steel Processing –Ion Irradiation Testing –Neutron Irradiation Testing to high dose Development of Lead Corrosion Resistant Materials –Lead Fast Reactors –DELTA loop –Long 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 Corrosion Actinide R&D: Separations, Integrated Safeguards Test Lab, Characterization of Irradiated Materials Information Sciences Fuels Research Lab: R&D on Ceramic Fuels Hot-Cells for Processing of Isotopes Nuclear Fuels R&D: MOX Fuel Fabrication and Testing Ion Beam Materials Laboratory, Structural Materials and Fuels R&D World’s first petaFLOPS supercomputer
Slide 7 Mechanical Testing in CMR Wing 9 Hot Cells Tensile Specimen- dimensions are 4 mm x 16 mm x 0.25 mm thick (gage dims. are 1.2 x 5 x 0.25 mm 3 ) Tested at initial strain rate of 5 x /s. Tested at 25 to 700 º C in ultra high purity argon Shear Punch, 3 pt. bend and compression testing capabilities. 16 mm
Slide 8 Testing was Performed on the ACO-3 Duct, one of the most highly irradiated components in a fast reactor FFTF, Hanford site, WA
9 Analysis of Specimens from ACO-3 Duct Total specimens= 144 Charpy, 57 compact tension, 126 tensile specimens, 500 TEM Charpy, 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 25C Completed Microstructural Analysis using TEM, SANS and Atom Probe Tomography
Stress/Strain Curves – HT-9 Irradiated, Room Temp Tests 10 Decreased elongation in irradiated materials. Increased hardening in lower irradiation temperature materials.
11 Results from Previous PIE Studies Agree Well With These Measurements AC0-1 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 dpa Previous studies (stars in figure, tested at 25C) show similar dependence of yield stress on irradiation temperature HT-9 Control Irradiation Temperature (C) 0.2% Offset Yield Stress (MPa) 25 C 200 C Tirr
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.
13 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 observed No void swelling observed. T=450C, 155 dpa Precipitation observed Dislocations of both a/2 and a Loops of a Void swelling observed (~0.3 %) T=505C, 4 dpa No precipitation or void swelling observed. Small Angle Neutron Scattering Measurements Obtain accurate measurement of ’ vs. dose and irr. Temperature Measurements completed on 5 specimens from ACO-3 duct
Slide 14 Summary T °C dpa Cr solub at % (G. Bonny) Cr solub at % (Phase diagram) α′α′ G-phaseLoopsvoids L nm d x10 21 m -3 L nm d x10 21 m -3 L nm d X10 21 m -3 L nm d X10 21 m XX XXXXXX XXXXXX ,000 hrs XXXXXX
15 STIP- (SINQ Target Irradiation Program) Irradiations Provides an Understanding of the Effects of Helium and Irradiation Dose ~570 MeV protons He/dpa ratios of appm/dpa Materials 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 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 Temperature Highest Hardening observed at lowest irradiation temperatures 12YWT tests reached limits of testing machine before yielding Significant helium accompanies dpa (up to 1300 appm He) Comparison with Phenix irradiated specimens will help quantify helium effect on mechanical properties 12YWT-Control Tirr=394C, dose=22 dpa Tirr=247C, dose=15 dpa HT-9-Control Tirr=380C, dose=22 dpa Tirr=247C, dose=15 dpa Tirr=120C, dose=8 dpa
Requirements for Accident Tolerant Fuels (ATF)
Measurements on hydrogen evolution performed in steam Hydrogen Production begins in Zircaloy-4 at ~700C and in 304L at ~1000C Similar testing is underway on all advanced alloys in FY13 Zirc-4 in N 2 containing ~25% water vapor to 1100°C 304 in N 2 containing ~25% water vapor to 1100°C
19 Advanced Material Development Activities Characterizing 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 weeks Investigating Possible Future irradiations –Domestic Facilities (MTS (18 dpa/yr)) – Collaborating in ATR irradiations –International collaborations Collaborating 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 processing –Mechanical alloying ODS material processing
Oxide Dispersion Strengthened Alloys Strength & damage resistance derives from a high density Ti- Y-O nano-features (NFs) NFs complex oxides (Ti 2 Y 2 O 7, Y 2 TiO 5 ) 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 Y-YO-Ti-TiO-O
Typical Processing Route for ODS Alloys Ball milling Alloy powder Y2O3Y2O3 Canning MA powder As extruded bar Working Heat treating Hot consolidation (extrusion)
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 Martensitic HT-9 Ferritic Dose Dose [dpa] Nanohardness [MPa] Depth [ m] 1m1m MA957 Room temperature irradiation (1.5dpa) Depth [ m] MA956 ODS Strengthened Materials Show Excellent Resistance to Hardening under Ion Irradiation ion beam 4m4m 4m4m 10 m Beam Berkovich indenter, 200nm deep indents, constant displacement
24 Analysis of highly irradiated MA-957 Tubes Underway at PNNL Tensile testing of MA-957 Pressurized tubes –Irradiation conditions in FFTF –(385°C, dpa) –(412°C, 110 dpa) –( °C, dpa) –( °C, dpa) –(750°C, dpa) –Testing will be performed at PNNL Status –Specimens for testing were machined from pressurized tubes at LANL –Tensile testing and TEM work is underway at PNNL –Analysis of in-reactor creep response is complete. Preliminary analysis of creep data –MA-957 is comparable to HT-9 in creep resistance to up to 550°C. At 600°C, 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 powder New consolidation condition explored for 14YWT –2 cans heat treated to nucleate nanoclusters (1 750ºC & 850ºC) –Cans extruded at 1150ºC Fabrication Extruded bar cut into 3 sections One section rolled parallel to extrusion axis at 1000ºC One section rolled normal to extrusion axis at 1000ºC Total reduction in thickness was 55% (~5.5 mm final thickness of 14YWT) No cracking was observed
Core Materials Research and Development – 5 Year Plan 26 FY’16 FY’15 FY’14 FY’11 FY’13FY’12 STIP- IV (PSI) Specimen PIE MATRIX-SMI and 2 (Phenix) Specimen PIE Re-irradiation of FFTF specimens in BOR-60 Materials Test Station Irradiations Provides data for NEAMS model development of Cladding FFTF (ACO-3 and MOTA) Specimen Analysis Advanced Material Development (improved radiation resistance to >400 dpa) Qualify HT-9 for high dose clad/duct applications (determine design limitations) Innovative Material Development Innovative Clad Material Downselect ODS Ferritic Steel Material Development Produce ODS Tubing Advanced Materials Irradiation in BOR-60 and CEFR Advanced Material Development (improved FCCI resistance to >40 % burnup) Development of Coated and Lined Tubes Innovative Coating Material Development Rev. 6 of AFCI (FCRD) Materials Handbook Accelerated Aging of Advanced Materials (High Dose Ion Irradiations) PIE on Lined Irradiated Tube Develop ODS Tubing and Weld specifications for innovative Weld material Baseline Mech. Prop. Of Inn. Clad Material Lined Tube for ATR irradiation Fab. Innovative Coated Tube for ATR irradiation Data to dpa on F/M and on Inn. Material Data on Advanced Materials to dpa
7/20/09 Compositional variations Bulk diffusion Bulk thermal conductivity Role of idealized grain boundaries in mass and thermal transport. All results provided to MBM MARMOT model development; mesoscale fission gas diffusion and segregation, thermal conductivity (INL) Analysis of segregation in terms of local strain Development of a Physics-based Model of Radiation Damage in Nuclear Fuels Motivation/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 Target Opportunity Excellent neutron yield Low neutron reaction cross sections Low melting point Excellent thermal properties Not susceptible to radiation damage Challenge Highly corrosive to steel Liquid Metal Embrittlement Liquid Metal Enhanced Creep XT-ADS, EU MYRRHA Hyperion, US SVBR-75, Russia 28
Conventional and Innovative Materials Selection 29 Ferritic/Martensitic steels and graded composites –12 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 alloys –PM 2000 and MA956 ODS steels and FeSi and FeSi model alloys (LANL) Commercially available Fe-Cr-Al alloys Kanthal Series steels (UCB, Berkeley) –APM, APMT, Alkrothal 14, Alkrothal 720 for high temperature applications Alkrothal 3 (UU, KTH, Sweden) Other possible candidate core materials Proposed for fuel cladding, heat exchangers in ADS/LFRs SCK.CEN, Belgium Collaboration –T91 9% Cr ferritic steel –D9 (DIN ) austenitic SS for in-vessel components –AISI 316 austenitic SS for vessel and in-vessel components
LANL Delta Loop Design and Capabilities 30 Design/Performance : * Up to 2 m/s in 2.54 cm diameter ~ 3 m long test section * Up to 100°C T between heater section and heat exchanger exit * Up to 550°C 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 31 * Corrosion resistance: Exposure up to 3000h in flowing LBE at 500 o C and Oxygen concentration 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 system Canister loaded with 48 specimens
Conclusions and Future Work 32 * Achievements: First specimens retrieved (May 2013) after ~ 900 h LBE exposure Ongoing loading of second canister (2000 h) and preliminary corrosion studies * Future Work: Retrieval of specimens after 2000h and 3000h exposure * Collaborations UCB Post Exposure Studies: Cross-section micro Raman, SEM/EDS/FIB/EBSD, nano-indentation studies * Possible Future International Collaborations: MYRRHA Accelerator Driven System - SCK.CEN, Belgium ELECTRA 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