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US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation 

US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation  April 6-7, 2002, Hotel Hyatt Islandia, San Diego, CA. Design Windows of Candidate Fusion Structural Materials. Akihiko Kimura Institute of Advanced Energy Kyoto University. Contents.

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US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation 

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  1. US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation  April 6-7, 2002, Hotel Hyatt Islandia, San Diego, CA Design Windows of Candidate Fusion Structural Materials Akihiko Kimura Institute of Advanced Energy Kyoto University

  2. Contents IAE, Kyoto University 1. Requirements for Fusion Structural Materials 2. Candidates for FSM 3. Progress in R&D of Candidates 1) Reduced Activation Ferritic steels 2) Vanadium Alloys 3) SiC/SiC Composites 4. Design Windows 5. Summary

  3. Requirements for Fusion Materials (1) IAE, Kyoto University ● 14MeV Neutron Irradiation (DT reaction) 1) Heavy displacement damage (200 dpa) 2) High He and H concentration (0.5 at.%) 3) Low activation 14Mev Neutron Blanket

  4. Requirements for Fusion Materials (2) IAE, Kyoto University ● Blanket System Integration 1) Compatibility with coolant and breeder 2) Weld/Joint performance 3) Complex and synergetic effects Plasma 14Mev Neutron First Wall Neutron Multiplier T-Breeder Coolant First Wall Blanket

  5. Requirements for Fusion Materials (3) IAE, Kyoto University ● Industrial Engineering Basis 1) Large-scale manufacturing 2) Impurity control 3) Cost 0 2 4 6 8 10 (m)

  6. Fe-9Cr-2W reduced activation ferritic steels V-4Cr-4Ti reduced activation vanadium alloys SiC/SiC composites Reduced Activation Materials IAE, Kyoto University ●Safety and Environmental Performance Reduction of radioactive waste Some elements, such as Mo, Ag and Nb are strictly limited to a level less than 5 wt.ppm. ●Low-activation elements: C, Cr, W, V, Ta, Ti, Mn, Si, B, Fe

  7. 2. V-4Cr-4Ti reduced activation vanadium alloys 3. SiC/SiC composites SiC/SiC Valloy ODS steel F/M steel Candidates for Fusion Structural Material (1) IAE, Kyoto University 1. Fe-9Cr-2W reduced activation ferritic steels F/M steel (Water or Gas Cooling Blanket) ODS steel （Gas Cooling Blanket) V –alloy （Self-cooling Liquid Blanket) SiC/SiC （Gas Cooling Blanket) 1100 900 Coolant Temperature / °C 700 500 300 30 40 50 20 Thermal Efficiency (%)

  8. Achievements of the Candidates IAE, Kyoto University Materials R&D Maturity (Data Base) ●F/M Steel ■V-Alloy ○SiC/SiC Environmental Performance Radiation Resistance Operation Temperature (Thermal Efficiency) Cost Performance By A. Kimura April, 2002 Engineering Maturity

  9. 9Cr-2W 9Cr-2W (2) (1) Ferritic Steels R&D (1) IAE, Kyoto University ● Improvement of Impact Properties 300 9Cr-2W:by Kimura 9Cr-1Mo: by Klueh (1) JMTR/363K, 0.006dpa: RT (2) MOTA/663K, 22dpa: RT (3) MOTA/663K, 35dpa: RT (4) MOTA/733K, 24dpa: RT (5) MOTA/683K, 36dpa: RT (a) JMTR/493K, 0.15dpa: RT (b) EBR-II/663K, 26dpa: RT (c) MOTA/638K, 7dpa: 638K (d) HFIR/323K, 5dpa: RT (e) HFIR/673K, 40dpa: 673K 250 9Cr-1Mo 9Cr-1Mo (e) 200 150 Shift in DBTT / K 100 (d) (b) (b) (5) (3) (c) 50 580appm He (4) (a) 120appm He 0 0 10 20 30 40 50 -200 -100 0 100 200 300 400 dpa Irradiation Hardening / MPa 1) Higher resistance to irradiation embrittlement. 2)Saturation of irradiation embrittlement at 10 dpa ( above 100 dpa?)

  10. JLS-1 650C(SA) JLS-2 650C(SA) JLS-3 650C(SA) Ferritic Steels R&D (2) IAE, Kyoto University ● Improvement of High Temperature Strength A significant increase in the creep strength has been obtained by increasing W conc. and/or mechanical alloying method. Irradiation study is in progress. 13Cr-ODS 700C(SA) 700 JLS:by Kohyama ODS:by Ukai 13Cr-ODS 700C(PT) 500 9Cr-ODS 700C(SA) 9Cr-ODS 700C(PT) 300 200 Applied Stress / MPa 100 80 F82H 650C(SA) 60 40 4 3 2 1 10 10 10 10 Rupture Time / hr

  11. 1400 12 1200 10 1000 ODS steel 8 800 6 ODS steel 600 4 400 200 2 0 0 0 200 400 600 800 0 200 400 600 800 Ferritic Steels R&D (3) IAE, Kyoto University ● Improvement of High Temperature Strength ODS steel (JNC) PNC316 20%CW Uniform Elongation(%) Tensile Stress (MPa) PNC-FMS PNC-FMS Temperature（℃） Temperature (℃）

  12. 500 300 100 80 60 Ferritic Steels R&D (4) IAE, Kyoto University ● Improvement of Creep Property ODS steel (JNC) Basic Chemical Composition ・9Cr-Martensitic ODS 9Cr-0.12C-2W-0.2Ti-0.35Y2O3 Advanced radiation resistance due to reduced Cr content ・12Cr-ferritic ODS 12Cr-0.03C-2W-0.3Ti-0.25Y2O3 Advanced corrosion resistance ・ Claddings manufactured by cold-rolling PNC316(650℃) (700℃) (750℃) 700℃ Hoop Stress (MPa) 650℃ PNC-FMS (650℃) Mechanical Properties ・Advantageous at higher temperature over 700 ℃ in creep rupture and tensile strength, comparing with PNC316 ・Maintain good ductility 750℃ PNC-FMS (700℃) Target PNC-FMS (750℃) Irradiation ・Up to 15 dpa at 400 ℃ to 530℃ ・Maintain strength and ductility without α’ 10 100 1000 10000 Time to Rupture (hr) M11: 9Cr-0.12C-2W-0.2Ti-0.35Y2O3

  13. Design Windows of RAFS IAE, Kyoto University 1500 1400 1300 1200 1100 1000 Thermal creep (1% creep strain at y/3 ) Temperature(°C) 900 800 Void swelling (>1%) 700 600 He effect (?) 500 400 300 200 He effect (?) 100 DBTT increase by irradiation 0 0 2 4 6 8 10 12 14 16 18 20 Neutron wall loading MWa/m2

  14. And More for Ferritic Steels IAE, Kyoto University ◎Toward DEMO ●Data base ●Helium and hydrogen effects ●Compatibility issues ◎Toward Power Reactor ● All the above ● RAFS/ODS combination design

  15. Vanadium Alloys R&D (1) IAE, Kyoto University ● Ductility Improvement by Alloying ◎ Marked ductility loss by irradiation ◎ Internal oxidation technique Al, Si, Y addition Al2O3, SiO2, Y2O3 Reduction of O in the matrix of vanadium Heat treatment condition is another concern. by Satoh and Abe

  16. Vanadium Alloys R&D (2) IAE, Kyoto University ● Ductility Improvement by Purification ◎ Oxygen: 80 wt.ppm Nitrogen: 130 wt.ppm by Nagasaka and Muroga V-4Cr-4Ti 14 NIFS-HEAT-1 12 C+N+O =290ppm 10 V-4Cr-4Ti 8 Absorbed energy / J 6 4 C+N+O =475ppm V-4Cr-4Ti 2 USDOE-HEAT 0 -200 -150 -100 -50 0 50 100 150 Test temperature / C

  17. Absorbed Energy, E/J・m-3 Test Temperature, T/oC Vanadium Alloys R&D (3) IAE, Kyoto University ● Ductility Improvement by Processing Final annealing: 1000oC, 1h ◎ Thermo-mechanical treatment lowering DBTT (c) (a) (b) (d) (a) 50%CW+IA(1000oC, 1h)+50%CW (b) HIP(1300oC, 147MPa, 2h)+50%CW (c) 75%CW (d) 50%CW by Satoh and Abe

  18. 500 600 700 750 Vanadium Alloys R&D (4) IAE, Kyoto University ● Improvement of Oxidation Resistance 0.1 ◎ Addition of Al and Y appears to be effective to increase oxidation resistance at 700°C. V-4Cr-4Ti V-4Cr-4Ti-0.5Si V-4Cr-4Ti-0.5Al 0.08 V-4Cr-4Ti-0.5Y 0.06 Weight gain (mg/mm2) 0.04 0.02 0 by Fujiwara and Abe Oxidation temperature (˚C)

  19. Design Windows of V Alloys IAE, Kyoto University 1500 1400 Thermal creep life at an applied stresses 110MPa (2/3 of the fracture stress) 1300 1200 1100 Void swelling (>2%) 1000 Temperature(°C) 900 He embrittlement (?) 800 700 600 500 400 300 200 DBTT increase by low temperature irradiation 100 0 2 4 6 8 10 12 14 16 18 20 Neutron wall loading MWa/m2

  20. And More for Vanadium Alloy IAE, Kyoto University ◎Toward DEMO ●Data base ●Helium and hydrogen effects ●Coating technique ◎Toward Power Reactor ● All the above ● Industry engineering basis (Spin-off to non-nuclear)

  21. SiC/SiC Composites R&D (1) IAE, Kyoto University ● Improvement of Tensile Stress (PIP:Polymer Impregnation and Pyrolysis ) by Kotani and Koyama ◎ Good performance at stoichiometric composition

  22. SiC/SiC Composites R&D (2) IAE, Kyoto University ● Improvement of Fabrication Technique Relatively large products Interfacial structure Matrix density by Yang and Noda

  23. SiC/SiC Composites R&D (3) IAE, Kyoto University ● Improvement of Irradiation Response by Kohyama ◎ Development of new fiber (Type-S Nicalon) improved response of mechanical properties to neutron irradiation. The advanced fiber has an almost stichiometric composition. R&D of stoichiometric matrix SiC and appropriate interfacial structure.

  24. SiC/SiC Composites R&D (4) IAE, Kyoto University ● Good Performance to 10 dpa by Katoh and Koyama ◎ No degradation of strength of stoichiometric Hi-Nicalon Type-S fibers. (?)

  25. Design Windows of SiC/SiC IAE, Kyoto University 1500 Thermal creep 1400 Void swelling (>2%) 1300 1200 1100 1000 Irradiation creep (?) Transmutation effect (?) Temperature(°C) 900 800 700 600 500 Decrease in thermal conductivity 400 300 200 Point defect swelling (>2%) 100 0 2 4 6 8 10 12 14 16 18 20 Neutron wall loading MWa/m2

  26. And More for SiC/SiC IAE, Kyoto University ◎Toward DEMO ●Data base ●Helium and hydrogen effects ●Fracture toughness (Interfacial structure) ◎Toward Power Reactor ● All the above ● Radiation resistance ●Production of large components

  27. Shift to DEMO ITER 1st Plasma DEMO startup Power reactor 1st Plasma DEMO H. Load: 10MWa/m2 Temp.: 550°C Construction and Operation POWER REACTOR H. Load: 20-40MWa/m2 Temp.: 600-800°C System startup ITER H. Load: 0.3MWa/m2 Temp.: <150°C Coolant: H2O CDA &EDA Reconstruction and Operation Testing Generation of power CDA &EDA Construction and Operation Requirements *Heat loading: 20-40MWa/m2 *Transmutation He: 3000-6000atppm *Heat efficiency: >50% 1st Material Selection 2nd Material Selection 3rd Material Selection *ready for ITER SUS316L(N)-ITER Grade (0.06-0.08%N) SUS30467(2%B) for shielding DBTT<383K y > 220MPa 105hr at =100MPa *ready for ITER test blanket module F82H, JLF-1 (Martensitic steels) JLF92H DBTT<383K 105hr at =110MPa *need processing R&D and weld/joint tech. JLF92H/92ODS Composites Reduced Activation ODS steels 92-ODS steels *Engineering bases *need purification O,N,C <10wt.ppm *coating tech. DBTT<383K 105hr at =110MPa V-4Cr-4Ti Alloys HP-V-4Cr-4Ti Alloys ODS-V-4Cr-4Ti-Y,Al *fracture toughness, *matrix coating, *matrix/fiber *processing R&D, *leak rate, *weld/joint R&D *Large components *Irradiation effects, *matrix *fiber, *complex, *boundary issues SiC/SiC Composites LS-SiC/SiC LS-3D-SiC/SiC 100 dpa 200 dpa 400 dpa IFMIF CDA &EDA Construction Testing Testing R&D Roadmaps IAE, Kyoto University

  28. Summary IAE, Kyoto University • Ferritic steel is the first candidate of fusion blanket structural materials. The issue of high thermal efficiency can be solved by application of ODS steel. Combined utilization of ferritic steel and ODS steel is quite promising for power reactor. • Vanadium alloys and SiC/SiC composites are also desirable for fusion power reactors. Extensive improvements have been achieved so far. • R&D of material processing for production of large-scale component is strongly demanded.

  29. The solutions will be discovered by you!! Conclusion IAE, Kyoto University Materials R&D Maturity (Data Base) ●F/M Steel ■V-Alloy ○SiC/SiC Environmental Safety Radiation Resistance Operation Temperature (Thermal Efficiency) Cost Performance By A. Kimura July, 2001 Engineering Maturity

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