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Material Development and Testing for Extreme Reactor Applications

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1 Material Development and Testing for Extreme Reactor Applications
S.A. Maloy1, O. Anderoglu1, T. Saleh1, M. Caro1, K. Woloshun1, F. Rubio1, M. Toloczko2, D. Hoelzer3, T.S. Byun3, G.R. Odette4 1Los Alamos National Laboratory, Los Alamos, NM 87545, USA 2Pacific Northwest National Laboratory, Richland, WA 99352, USA 3Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA 4University of California Santa Barbara, Santa Barbara, CA 93106, USA Funded by Department of Energy-Nuclear Energy

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

3 Advanced Fuels Campaign Mission & Objectives in the Fuel Cycle Research and Development Program
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 Continuous Recycle Modified Open Once- Through Advanced Fuels 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 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 FCCI Corrosion Radiation Temperature Coating Liners Advanced Alloys F/M Steels Advanced Alloys F/M Steels HT-9 Advanced F/M Steels, e.g. NF616 Advanced Alloys ODS Steels Cr Enhancements with Fabrication Complexity Si 200 dpa dpa dpa Al Increasing content Different Reactor options to change requirements LFR, GFR 500 C C C Reduced embrittlement, swelling, creep Enhancements with Fabrication Complexity Enhancements with Fabrication Complexity Develop an advanced materials immune to fuel, neutrons and coolant interactions under specific reactor environments

5 Outline STIP – IV irradiated Materials
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

6 Our Unique Facilities and Infrastructure Providing Science and Technology to NE
Ion Beam Materials Laboratory, Structural Materials and Fuels R&D Actinide R&D: Separations, Integrated Safeguards Test Lab, Characterization of Irradiated Materials Information Sciences World’s first petaFLOPS supercomputer Hot-Cells for Processing of Isotopes Fuels Research Lab: R&D on Ceramic Fuels Actinide Cross-Section Measurements Irradiation of Targets for Isotope Production DELTA loop for LBE Corrosion Nuclear Fuels R&D: MOX Fuel Fabrication and Testing

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 mm3) Tested at initial strain rate of 5 x 10-4/s. Tested at 25 to 700ºC in ultra high purity argon Shear Punch, 3 pt. bend and compression testing capabilities. 16 mm

8 Testing was Performed on the ACO-3 Duct, one of the most highly irradiated components in a fast reactor 1 2 3 5 4 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

10 Stress/Strain Curves – HT-9 Irradiated, Room Temp Tests
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

12 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=450C, 155 dpa Precipitation observed • Dislocations of both a/2<111> and a<100> • Loops of a<100> Void swelling observed (~0.3 %) T=505C, 4 dpa No precipitation or void swelling observed. T=384C, 28 dpa G-phase precipitates and alpha prime observed No 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

14 Summary T °C dpa Cr solub at % (G. Bonny) Cr solub (Phase diagram) α′
G-phase Loops voids L nm d x1021 m-3 X1021 380 20 9 8.1 7.8 72 11.3 9.3 14 0.93 X 410 100 9.1 10.1 2.6 16.2 1.4 - 23 440 155 12.4 9.6 9.5 26.5 1.1 18 0.5 28 0.25 466 92 9.9 14.7 505 2 12.5 18.5 475 10,000 hrs 10.3 15.5

15 He/dpa ratios of 50-60 appm/dpa
STIP- (SINQ Target Irradiation Program) Irradiations Provides an Understanding of the Effects of Helium and Irradiation Dose 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-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 ~570 MeV protons He/dpa ratios of appm/dpa

16 Summary of STIP IV Tensile Testing at Room Temperature
Tirr=247C, dose=15 dpa Tirr=120C, dose=8 dpa Tirr=394C, dose=22 dpa Tirr=247C, dose=15 dpa Tirr=380C, dose=22 dpa 12YWT-Control HT-9-Control 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

17 Requirements for Accident Tolerant Fuels (ATF)

18 Measurements on hydrogen evolution performed in steam
Zirc-4 in N2 containing ~25% water vapor to 1100°C 304 in N2 containing ~25% water vapor to 1100°C Hydrogen Production begins in Zircaloy-4 at ~700C and in 304L at ~1000C Similar testing is underway on all advanced alloys in FY13

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

20 Oxide Dispersion Strengthened Alloys
Strength & 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 Y-YO-Ti-TiO-O

21 Typical Processing Route for ODS Alloys
Ball milling Alloy powder Y2O3 Canning MA powder As extruded bar Working Heat treating Hot consolidation (extrusion)

22 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.

23 ODS strengthened Materials Show Excellent Resistance to Hardening under Ion Irradiation
Room temperature irradiation (1.5dpa) Berkovich indenter, 200nm deep indents, constant displacement HT-9 Martensitic Dose [dpa] HT-9 Ferritic Nanohardness [GPa] Dose ion beam Depth [mm] MA957 MA956 Nanohardness [MPa] Dose [dpa] 4mm 10mm 4mm 1mm Beam Depth [mm]

24 Analysis of highly irradiated MA-957 Tubes Underway at PNNL
Tensile testing of MA 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. 24

25 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

26 Core Materials Research and Development – 5 Year Plan
Qualify HT-9 for high dose clad/duct applications (determine design limitations) Data to dpa on F/M and on Inn. Material FFTF (ACO-3 and MOTA) Specimen Analysis Rev. 6 of AFCI (FCRD) Materials Handbook Re-irradiation of FFTF specimens in BOR-60 Advanced Material Development (improved radiation resistance to >400 dpa) STIP- IV (PSI) Specimen PIE Materials Test Station Irradiations MATRIX-SMI and 2 (Phenix) Specimen PIE Data on Advanced Materials to dpa Innovative Material Development Baseline Mech. Prop. Of Inn. Clad Material Innovative Clad Material Downselect ODS Ferritic Steel Material Development Develop ODS Tubing and Weld specifications for innovative Weld material Produce ODS Tubing Advanced Materials Irradiation in BOR-60 and CEFR Accelerated Aging of Advanced Materials (High Dose Ion Irradiations) Advanced Material Development (improved FCCI resistance to >40 % burnup) Development of Coated and Lined Tubes Lined Tube for ATR irradiation PIE on Lined Irradiated Tube Innovative Coating Material Development Fab. Innovative Coated Tube for ATR irradiation FY’11 FY’12 FY’13 FY’14 FY’15 FY’16 Provides data for NEAMS model development of Cladding

27 All results provided to MBM
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). Analysis of segregation in terms of local strain Compositional variations Bulk diffusion Bulk thermal conductivity All results provided to MBM Material specificity MARMOT model development; mesoscale fission gas diffusion and segregation, thermal conductivity (INL) Role of idealized grain boundaries in mass and thermal transport. 7/20/09

28 LBE as a Nuclear Coolant and Spallation Source Target
28 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 SVBR-75, Russia Hyperion, US

29 Conventional and Innovative Materials Selection
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

30 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 100°C DT 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

31 DELTA Corrosion Test Plan
* Corrosion resistance: Exposure up to 3000h in flowing LBE at 500oC 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. Canister loaded with 48 specimens 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 31

32 Conclusions and Future Work
* 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 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 32

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