Materials Challenge for Clean Nuclear Fusion Energy Development of Low Activation Structural Materials T. Muroga National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan Symposium on Materials Challenges for Clean Energy June 21-25, 2009, NIST, USA 1
Roadmap of Fusion Energy Development for Japan 2020 2030 2040 ITER D-T Plasma Performance DEMO Designing Construction Operation 50dpa 100-150dpa Power Generation Materials Performance under Irradiation IFMIF Operation Fission reactors and charged particle irradiation IFMIF : International Fusion Materials Test Facility 2
Nuclear Fusion as a Clean Energy Source Nuclear fusion is a clean energy source from the CO 2 emission standpoints One of the very limited candidates of major energy source in the future From radiological standpoint, nuclear fusion is a clean energy source relative to nuclear fission No high level radioactive waste is produced Neutron-induced radioactivity of the components (low to middle level waste) and tritium are the only radiological issues Neutron-Induced radioactivity is an issue for : (1) Maintenance of in-vessel components (lifetime of maintenance tools) (2) Waste disposal or reuse (3) Potential hazard 3
Structural Materials for Fusion Blankets Blanket is the key component of fusion reactors Energy conversion (neutron to heat) Tritium production (Li + n T) The structural materials need to have high temperature strength, resistance to neutron irradiation and compatibility with coolants. Neutrons Fusion DEMO Reactor (FFHR) Reactor Core D+T He+14.1 MeV n Blanket 4
The Concept of Low Activation Materials Low Activation Materials is defined as : The materials in which neutron irradiation-induced radioactivity decrease to a level where economical waste disposal or recycling in the next fusion plant (possible zero emission) is feasible in <50 years of cooling. Use of Low Activation Materials can increase largely the attractiveness of fusion system as a clean energy source 5
Element Restriction for Low Activation There is strong element restriction for low activation materials Ni, Cu, Al, Mo, Nb cannot be used as alloying elements Al, Ag, Nb Mo should be strictly controlled to ~ppm levels None of the commercially-established materials can be used for the low activation structural components no limit design dependent <0.1% Classification of the elements by the restriction for low activation 6
Radioactivity of Fusion Candidate Materials Three candidate materials are under development for fusion application as low activation materials Reduced activation Ferritic/Martensitic steel Vanadium alloys SiC/SiC composites The candidate materials satisfy the recycling limit Contact Radioactivity( Dose (Sv/h) /h) 10 5 10 4 10 3 10 2 10 1 10 0 10-1 10-2 10-3 10-4 10-5 FFHR -Li Blanket First Wal Reduced Neutron 1.5MW/m 2 Activation 30 years foperation Ferritics F82H) V- 4Cr- 4Ti NIFS-Heat Pure SiC/SiC SS -316 316SS SS316LN -IG Low activation ferriticf82h Vanadium alloynifs -HEAT -2 SiC (no impurity Pure V-4Cr-4Ti 10-2 10-1 10 0 10 1 10 2 10 3 10 4 Cooling Time after Shutdown (Years) Full-remote recycling Full-hands-on recycling Contact dose of fusion structural materials after shutdown of a reactor 7
Low Activation would also be the Future Key Phrase for Nuclear Fission Systems Disposal of Light Water Reactor in-vessel components The activity can be reduced by three orders using low activation materials Low Activation Concrete is already investigated in nuclear fission community Reduction of Eu and Co levels for satisfying clearance level Contactdose (Sv/h) 10 7 10 6 10 5 10 4 10 3 10 2 10 1 10 0 10-1 10-2 10-3 10-4 10-5 V-alloy SS316 10-4 10-3 10-2 10-1 10 0 10 1 10 2 10 3 10 4 10 5 10 6 Cooling time after shut-down (years) Radioactivity after 60 y operation of PWR Relative Radioactivity Clearance Zone Low Activation Concrete Activation by 152 Eu and 60 Co Cooling Time (y) Conventional Concrete Radioactivity of concrete after 20 y operation 8
Operation Temperature and Efficiency 50 *Fusion AGR DREAM* Conversion efficiency / % 40 30 20 FBR PWR SSTR* BWR RAF steel Water HTGR ARIES-1* FFHR* SiC ceramic V alloy He Flibe, Li, Li-Pb 200 300 400 500 600 700 800 900 1000 Coolant temperature / o C High operation temperature is essential for high conversion efficiency Plant efficiency is influenced also by other factors (e.g. net working rate) 9
History of Vanadium Alloys for Fusion Reactors Vanadium alloys were once candidates of FBR core materials in 1970s V-Nb-Cr, V-Nb-Mo, V-Cr-Zr (JP-NIMS), V- Ti-Cr, V-Cr-Fe (US), V-Ti-Si (Germany) The programs were concluded because of insufficient compatibility with Na Vanadium alloys attracted attention for fusion reactors from 1980s Low activation property -- potentiality for recycling High temperature strength, high thermal stress factor V-XCr-YTi 950 1hr The effects of Cr+Ti (wt%) on Ductile-Brittle Transition Temperature of V-Cr-Ti alloys Fusion materials programs identified V-4Cr-4Ti as the leading candidate Industrial infrastructure of vanadium alloys are still quite limited 10
Impurity Reduction is the Key Issue for V-Alloys Vanadium is a reactive metal, whose properties are influenced heavily by C, N, O impurities C, N, O impurities need to be controlled Contamination from the environment may be an issue Vanadium alloys must satisfy low activation criteria Nb, Mo, Al, Ag need to be controlled to 1~10 ppm Cross contamination by sharing facilities for other purposes may be an issue Elongation (%) The effect of O, N and C levels on Elongation NIFS has promoted a program for development of high purity V-4Cr-4Ti 11
Commercial Production Process of Metal Vanadium Ores O Al C N Nb, Mo (Cross Contamination) 12
Improvement of the Process Intermediates Conventional process Improved process C N O C N O Ammonium Vanadate 12 41 12 41 (Calcination) V 2 O 5 100-150 20-90 44 10 10-20 44 Al 20-80 10 10-20 10 (Themic reduction) V-Al-O 180 300-400 800-5000 60 60-120 800-5000 (EB refining) Metal V 300 400-710 50-150 100 60-160 40-100 13
Production of Purified Metal Vanadium Metal Vanadium with <100wppm O and <200ppm N were produced by improving the production process 500 400 Laboratory Scale US-DOE Program (900, 2200 kg) Conventional (50 kg) Improved (25 kg) Oxygen / wppm 300 200 Large ingots by US-DOE Conventional large ingots of commercial vanadium in Japan 100 This study 0 0 100 200 300 400 500 600 Nitrogen / wppm O and N levels of metal V produced 125 kg ingots of purified V 14
Alloying Technology Alloying to V-4Cr-4Ti was carried out by EB and VAR methods Special cares were taken to avoid contamination with O and N from atmosphere and with Nb and Mo by cross-contamination EB melting from V, Ti, Cr flakes 1st VAR Ingot scaling EB-welding Hot Forging canning 160 kg ingot 2 nd VAR Ingot N and O Nb, Mo (Cross Contamination) 15
Resulting Purity of the Alloys Purification of metal vanadium resulted in low O level relative to US alloys. US alloy had high Nb level because the infrastructure was common to Nb production system This is a valuable lesson for future manufacturing of low activation V-alloys Chemical compositions of V-4Cr-4Ti ingots produced by NIFS and USDOE Programs ID C O N B Na Mg Al Si V Cr Mn Fe Ni Cu NIFS-HEAT-1 56 181 103 7 17 <1 119 280 Bal. 4.12 <1 80 13 4 NIFS-HEAT-2 69 148 122 5 <1 <1 59 270 Bal. 4.02 <1 49 7 2 US832665 170 330 100 3.7 0.01 0.17 355 785 Bal. 3.25 0.21 205 9.6 0.84 US832864 37 357 130 <2 <1 193 273 Bal. 3.8 228 <50 ID As Zr Nb P S Ca Co Ag Sn Sb Ti W Mo Ta NIFS-HEAT-1 1 <10 1.4 16 9 3 2 <0.05 <1 <1 4.13 <1 23 58 NIFS-HEAT-2 <1 2.5 0.8 7 3 12 0.7 <0.05 <1 <1 3.98 <1 24 13 US832665 1.4 <46 60 33 16.5 <0.26 0.30 0.078 0.24 0.17 4.05 25 315 <19 US832864 106 <30 4 3.8 <50 16
V-4Cr-4Ti Products (mm) 26t 6.6t 4.0t 1.9t 1.0t 0.5t 0.25t 2d 8d Plates, Sheets, Wires and Rods Thin pipes Creep tubes W coating by plasma spraying Laser weld joint Various high purity products were manufactured 17
V-4Cr-4Ti Tubing by Drawing O impurity before fabrication ~400 ppm O impurity before fabrication ~130 ppm Quality of tubes increased by reducing O levels 18
Welding Impurity Control is the Key Issue Welding can results in contamination with C, N, O by two ways (a) Contamination of the weld joints with C, N, O from atmosphere during the welding (b) In V-4Cr-4Ti, large fraction of C, N, O are stored in Ti-CON precipitates. Welding results in contamination of matrix with C, N, O (a) is avoidable by highly controlling the welding atmosphere (b) can be avoided only by reducing the original level of C, N, O in the material Base Metal Weld Metal 19
TIG Welding with Ultra-High Purity Wire Ultra-high purity wires were used for TIG welding for the purpose of reducing impurity level of weld joints With the decrease in the oxygen level in the weld joint, DBTT decreased Absorbed Energy (J) 15 10 5 0 NH1/HP -145 C US / HP -90 C NH1 / NH1-85 C -200-100 0 100 Test temperature (C) US / US 47 C Plate / Wire DBTT Absorbed energy of Charpy tests for weld joints 300 200 100 NH1 / HP Weld joint Alloy Name O level (appm) US-Large Heat (US) 310 NIFS-HEAT-1 (NH1) 180 High Purity Wire (HP) 36 US / HP NH1 / NH1 US / US DBTT = +60 K / 100 wppm oxygen 0 00 50 100 150 200 250 300 350 400 Oxygen in weld metal / wppm DBTT of weld joints as a function of oxygen level 20
Radiation Effects at Low Temperature Irradiated Vanadium alloys showed plastic instability Interaction of dislocations with radiation-induced defect clusters (black dot images by TEM) is thought to be the key process Atom Probe Elemental Mapping showed the black dots are decorated with Ti, O and TiO. Elemental mapping of radiation-induced defect clusters by Atom Probe 10nm 35nm (Rice 1998) 11nm Ti O Change of tensile property and typical microstructure after deformation in V- 4Cr-4Ti irradiated at low temperature (Nita 2003) TiO 21
Enhanced Uniform Elongation in V-Cr-Ti-Si,Al,Y Uniform elongation after irradiation decreased to ~ 0 after irradiation at <400ºC However, the uniform elongation recovered by addition of Si, Al, Y Si, Al and Y can scavenge impurity O in V-4Cr-4Ti matrix and suppress the formation of the decorated defect clusters Uniform elongation of V- 4Cr-4Ti after neutron irradiation (Abe, Satou 2003) 22
Nano Structural Materials for Fusion Nano oxide dispersion strengthened (ODS) materials are under development for fusion as advanced options : Enhanced high temperature strength Improved radiation damage resistance because of high density of defect sinks ODS low activation steel : Fe-9Cr-W Collaboration with FBR and SCWR ODS vanadium alloys ODS Cu alloys : Heat sink materials W, Be, C CuCrZr to ODS-Cu Steels ITER First Wall Tensile Stress (MPa) 1400 1200 1000 800 600 400 200 0 0 300 500 700 900 1100 400nm Ferritic steel Stress (MPa) 9CrODS steel Temperature (K) 800 600 400 200 0 T Test = 298 K ODS V-alloys Strain (%) 12 10 8 6 4 2 T Test = 1073 K 23 Uniform Elongation (%) 9Cr Ferritic Steel vs. 9Cr ODS steel V - 1.0 vol% Y 2 O 3-0.7 vol% YN V NIFS-HEAT-1-4.4 wt% Cr - 4.1 wt% Ti 20%
Summary Materials development is the key to successful development of fusion energy. Use of low activation structural materials can largely increase the attractiveness of fusion reactors as a clean energy source. High temperature operation is essential for high efficiency energy conversion Vanadium alloys are attractive candidate low activation structural materials for fusion reactors because of low activation and high temperature strength. Control of both potentially-radioactive impurities (Nb, Mo, Al, Ag) and interstitial impurities (C, N, O) is essential for vanadium alloy development. Nano Structural Materials are longer-term advanced options to enhance the operation temperature window and radiation resistance of structural components. 24