Shiro Jitsukawa

Interactions between fusion materials R&D and other technologies

The importance of interactions between fusion materials research and development (R&D) and other technologies is emphasized to make attractive and realistic fusion technology integration activities. The focuses are on: (1) materials design and processing, (2) safety issues relating to materials and (3) material performance evaluation methodologies, including 14 MeV neutron source utilization for fusion material R&D. As typical examples, material design activities on reduced activation ferritic steels, vanadium alloys and SiC/SiC composite materials are provided. The safety assessment of reactor systems and reactor design code consideration including prediction methodologies of materials performance are also discussed.

Fracture toughness and tensile behavior of ferritic-martensitic steels irradiated at low temperatures

Abstract Disk compact tension and sheet tensile specimens of the ferritic-martensitic steels F82H and Sandvik HT-9 were irradiated in the High Flux Isotope Reactor (HFIR) at 90°C and 250°C to neutron doses of 1.5–2.5 dpa. For both steels, radiation hardening was accompanied by a reduction in strain hardening capacity (SHC). When combined with other literature data it is apparent that severe loss of SHC occurs in F82H for irradiation temperatures below ∼400°C and in HT-9 for irradiation temperatures below ∼250°C. For both alloys, severe loss of SHC does not correlate with brittle behavior during fracture toughness testing.

Recent progress in reduced activation ferritic steels R&D in Japan

The Japanese reduced activation ferritic steels (RAFSs) R&D road map towards DEMO is shown. The important steps include high-dose irradiation in fission reactors such as the high flux isotope reactor at Oak Ridge National Laboratory, irradiation tests with 14 MeV neutrons in the International Fusion Materials Irradiation Facility and application to ITER test blanket modules to provide an adequate database of RAFSs for the design of DEMO. The current status of RAFS development is also introduced. The major properties of concern are well-known, and process technologies are mostly ready for fusion application. RAFSs are now certainly ready to proceed to the next stage. A materials database is already in hand, and further progress is anticipated with the design of the ITER test blanket. Oxide dispersion strengthening steels are quite promising for high temperature operation of the blanket system, with potential improvements in radiation resistance and in corrosion resistance.

Response of reduced activation ferritic steels to high-fluence ion-irradiation

Abstract Effects of high-fluence irradiation in fusion-relevant helium production condition on defect cluster formation and swelling of reduced activation ferritic/martensitic steels (RAFs), JLF-1 (Fe–9Cr–2W–V–Ta) and F82H (Fe–8Cr–2W–V–Ta), have been investigated. Dual-ion (nickel plus helium ions) irradiation using electrostatic accelerators was adopted to simulate fusion neutron environment. The irradiation has been carried out up to a damage level of 100 displacement per atom (dpa) at around 723 K, at the HIT facility in the University of Tokyo. Thin foils for transmission electron microscopy (TEM) were prepared with a focused ion beam (FIB) microsampling system. The system enabled not only the broad cross-sectional TEM observation, but also the detailed study of irradiated microstructure, since unfavorable effects of ferromagnetism of a ferritic steel specimen were completely suppressed with this system by sampling a small volume in interests from the irradiated material.

Stress-strain relations of irradiated stainless steels below 673 K☆

Abstract Most specimens, irrespective of thermo-mechanical treatment, exhibited proof stress levels of above 800 MPa and uniform elongations below 1% after irradiation in the High Flux Isotope Reactor (HFIR). Only the solution annealed specimens irradiated at a low temperature of 328 K showed uniform elongations larger than 5% and proof stresses smaller than 800 MPa. Irradiation in the High Flux Reactor (HFR) caused more hardening than did irradiation in the HFIR. Ductility loss and change in work hardening characteristics by HFR irradiation were evaluated from reduction of area values. Residual ductility was revealed to be larger than 0.5 in natural strain, and the irradiation was estimated to have a small effect on work hardening characteristics and on fracture stress. The ductility of the irradiated alloys was found to be about 58% of that for the unirradiated alloys, as has been previously reported for irradiation in the HFIR. It was also demonstrated that true stress-strain relations, except for the fracture conditions, could be represented by Swifts type constitutive equation.

Radiation damage of HFIR-irradiated candidate stainless steels for fusion applications

Data of the swelling behavior and tensile properties were summarized on two candidate alloys; Japanese Primary Candidate Alloy (JPCA) and Type 316 stainless steel (J316) irradiated in the High-Flux Isotope Reactor (HFIR) at temperatures ranging from 573 to 873 K to a maximum dose of 57 dpa (4200 appm He). Little temperature dependence of small swelling less than about 1% was observed in both solution-annealed (SA) and cold-worked (CW) JPCA and J316 at irradiation temperatures < 673 K up to nearly 56 dpa. At 773 K, the swelling increased rapidly in SA although CW alloys still had good swelling resistance. Tensile properties below 673 K were different from those obtained above 773 K; in the low temperature region, no substantial change of the work-hardening character was induced by irradiation, while in the high-temperature region, the apparent work-hardening exponent n and/or strength coefficient A became large. Through the observation of microstructural change, this is possibly due to precipitation. Substantial decrease in ductility was obtained in the high-temperature region with dose beyond 50 dpa.

Highly thermal conductive, sintered SiC fiber-reinforced 3D-SiC/SiC composites: experiments and finite-element analysis of the thermal diffusivity/conductivity

Chemical vapor infiltrated (CVI) and polymer impregnated and pyrolized (PIP) SiC/SiC composites were fabricated by using highly thermal conductive, sintered SiC fiber. These composites had relatively high thermal diffusivity/conductivity values, which were two or three times larger than those reinforced with Hi-Nicalon or Hi-Nicalon Type S fibers. The improvement of thermal diffusivity by using this fiber was more noticeable for PIP composites than for CVI composites. A 2D finite element thermal analysis supported the experimental results, and revealed that highly thermal conductive SiC fiber was much effective for PIP composites and that the increase of fiber volume worsened CVI composite thermal diffusivity if low thermal conductive SiC fibers were used.

Radiation effects at fusion reactor He : dpa ratios: Overview of US/Japan spectrally tailored experiments☆

Abstract Spectrally tailored experiments in the Oak Ridge reactors ORR and HFIR are being used to determine the properties of austenitic stainless steels in the relatively unexplored radiation damage regime represented by the proposed ITER first wall and blanket design. These collaborative US/Japan experiments reproduce the temperature (60 to 400°C), damage rate (~1 MW y/m2), neutron fluences (10 to 30 dpa) and helium generation rates (10 to 15 appm/dpa) typical of the ITER environment. The status of these experiments is described. Following irradiation in the ORR to ~8 dpa, data have been obtained on the tensile, swelling, and irradiation creep properties for a variety of austenitic stainless steels. These data are being utilized in the development of design equations. The experiments have been re-encapsulated in assemblies designed for insertion into the new RB∗ irradiation facilities when the HFIR restarts.

Microstructural study of irradiated isotopically tailored F82H steel

Abstract The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250–400 °C to 2.8–51 dpa in HFIR has been examined using isotopes of 54 Fe or 10 B . Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 °C to 2.8 dpa. Helium atoms also influenced them at around 300 °C, and the effect of helium atoms was enhanced at 400 °C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of α′-precipitates on dislocation loops.

Tensile properties of austenitic stainless steels and their weld joints after irradiation by the ORR-spectrally-tailoring experiment☆

Abstract Tensile specimens of the Japanese heat of PCA (JPCA) and type 316 stainless steels were irradiated in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) to a peak dose of 7.4 dpa and a peak helium level of 105 appm in the temperature range between 328 and 673 K. Specimens of type 316 steel with weld joints produced by tungsten inert gas (TIG) and electron beam (EB) welding techniques were also included. Irradiation caused both increases in flow stress and decreases in elongation. Weld joint specimens exhibited both lower strength and elongation after irradiation. The reduction of area (RA) for the TIG weld joint specimens decreased by a factor of 5 compared to unirradiated base metal specimens, however, they still fractured in a ductile mode. The EB weld joints maintained RA levels similar to that of the unirradiated base metal specimens. Post-radiation ductilities of weld joints and base metal specimens of these steels should be adequate for their application to next generation fusion experimental devices, such as the International Tokamak Experimental Reactor (ITER).