H. Kayano

Charpy impact testing using miniature specimens and its application to the study of irradiation behavior of low-activation ferritic steels

Abstract The effectiveness of mini-size Charpy V-notch specimens with a 1.5 or 1.0 mm square cross section in measuring the ductile brittle transition temperature (DBTT) and upper shelf energy (USE) compared with full-size specimens is evaluated for a ferritic steel. It is shown that the data from the mini-size specimens can be used to estimate the DBTT and USE for full-size specimens when the measured absorbed energy-temperature curves are normalized by appropriate parameters. The result is applied to the study of neutron irradiation embrittlement of low-activation ferritic steels.

Irradiation embrittlement of neutron-irradiated low activation ferritic steels

Abstract Effects of neutron irradiation and additions of small amounts of alloying elements on the ductile-brittle transition temperature (DBTT) of three different groups of ferritic steels were investigated by means of the Charpy impact test in order to gain an insight into the development of low-activation ferritic steels suitable for the nuclear fusion reactor. The groups of ferritic steels used in this study were (1) basic 0–5% Cr ferritic steels, (2) low-activation ferritic steels which are FeCrW steels with additions of small amounts of V, Mn, Ta, Ti, Zr, etc. and (3) FeCrMo, Nb or V ferritic steels for comparison. In Fe-0–15% Cr and FeCrMo steels, Fe-3–9% Cr steels showed minimum brittleness and provided good resistance against irradiation embrittlement. Investigations on the effects of additions of trace amounts of alloying elements on the fracture toughness of low-activation ferritic steels made clear the optimum amounts of each alloying element to obtain higher toughness and revealed that the 9Cr-2W-Ta-Ti-B ferritic steel showed the highest toughness. This may result from the refinement of crystal grains and improvement of quenching characteristics caused by the complex effect of Ti and B.

Tensile behavior of helium charged VTiCrSi type alloys

Abstract Helium effect on the mechanical properties of the alloy V5Ti5Cr1SiAl,Y (nominal) was studied, adopting various helium charging methods and helium-to-dpa ratio. The first method was helium ion implantation using a cyclotron accelerator at Tohoku University, where helium and displacement levels were 50 appm and 0.02 dpa, respectively. The second, helium was charged by tritium trick technique and following neutron irradiation in FFTF/MOTA-2A, associated with about 80 appm He and 43 dpa. The third was dynamic helium charging experiment (DHCE) conducted in FFTF/MOTA-2B, where helium was generated within specimens during neutron irradiation by tritium decay, and the helium-to-dpa ratio was adjusted to simulate the fusion reactor condition, that is, 177 appm He and 24 dpa. The effect of helium on tensile properties of the VTiCrSiAl,Y alloy depended on the helium charging methods. The uniform elongation of the alloy was 3.2% and total elongation was 8.3% at DHCE condition, which was the most fusion relevant condition of the methods. It is important that tensile properties of the present alloy could be acceptable for fusion reactor component materials.

High-temperature deformation of modified V-Ti-Cr-Si type alloys

Abstract Modified V-Ti-Cr-Si type alloys containing Al and Y were prepared in order to improve the oxidation resistance for fusion reactor applications. The fabrication procedure and the condition of heat treatment were determined for two alloys referred as V-5Ti-5Cr- and V-25Ti-15Cr-. Mechanical properties were measured using miniature-type tensile specimens at temperatures from 300 to 1273 K. The alloy V-5Ti-5Cr- showed the same level of strength and ductility as a reference alloy of V-15Cr-5Ti and the alloy V-25Ti-15Cr- was twice as strong as the reference alloy at temperatures from 300 to 873 K. Dynamic strain aging and work softening phenomena with large elongation were observed at intermediate and high temperature regions, respectively.

Optical properties in fibers during irradiation in a fission reactor

Abstract The effects of irradiation on the performance of optical fibers with silica core were examined during irradiation in a fission reactor, JMTR. The fibers performed well in terms of their radiation characteristics up to the fast neutron fluence of 1.06 × 10 20 n/cm 2 and the gamma ray dose rate of 4.3 × 10 9 Gy at 460 K. Three kinds of optical radiation induced by the reactor irradiation were observed during irradiation. Origins of these optical radiations were studied. Broad optical radiations centered at 0.45 and 0.73 μm are thought to be Cherenkov radiation. The sharp optical radiation at 1.27 μm was thought to be generated in the silica core by the gamma-ray irradiations in the reactor.

Effects of small changes in alloy composition on the mechanical properties of low activation 9%Cr-2%W steel

Tensile and creep tests were carried out for various 9%Cr-2%W steels with small additions of alloying elements to investigate the effects of additions of B, Mn, Ta, Ti, Zr, Al and Y on the mechanical properties of a low activation martensitic steel whose chemical composition (wppm) is as follows: Fe-0.1C-9Cr-2W-0.04Si-0.5Mn-0.26Ta-0.02Ti-0.003B. An increase in the amount of Ti, Mn and Ta and a partial Zr substitution for Ta caused a remarkable increase in tensile strength at 873 K resulting in an increase in creep rupture time at this temperature. An addition of Al and/or Y gave only small changes in tensile and creep behavior at 873 K. Neutron irradiation (7 × 1022 n/m2) did not cause any remarkable changes in tensile and creep behavior of these steels at 873 K, suggesting that irradiation induced defects were annealed out during tests.

The effect of small additional elements on the precipitation of reduced activation Fe9Cr2W steels

Abstract In order to study effects of small additional elements on precipitation of reduced activation Fe9Cr2W steels were irradiated up to 60 dpa at 693 K, 698 K and 733 K in FFTF. Micro-voids were observed in both materials of Fe9Cr2W with or without boron, the density of micro-voids in the steel with boron is larger than without boron, and the mean size of micro-voids is smaller than that without boron. However void swelling was less than 1%. Many precipitates were found to be M 23 C 6 which consists of mainly Cr. Several precipitates which were Ti rich including Si and W were also observed at grain boundary at 733 K. Several Y 2 O 3 particles was observed in an yttrium containing alloy. No precipitation including Al was observed in an Al containing alloy. Ti addition decreased precipitation of Ta-rich M 6 C in 9Cr and 12Cr steels in this irradiation condition.

Effects of neutron irradiation on hydrogen-induced intergranular fracture in a low activation 9%Cr-2%W steel

Abstract Hydrogen charging changed the fracture mode in tensile tests at room temperature from ductile shear rupture to intergranular cracking, resulting in a considerable reduction of the ductility of a low activation 9%Cr-2%W martensitic steel. The critical hydrogen charging current density required to cause hydrogen-induced intergranular cracking was reduced by neutron irradiation, suggesting that neutron irradiation enhanced hydrogen-induced intergranular cracking. This hydrogen-induced intergranular cracking was not caused by irreversible damage due to hydrogen charging, since it disappeared after aging at room temperature. The recovery rate of the fracture mode from intergranular cracking to ductile rupture during aging at room temperature was reduced by irradiation. A mechanism of irradiation-induced enhancement of hydrogen embrittlement in a low activation 9%Cr-2%W martensitic steel is proposed.

Effects of neutron irradiation on fine structure and strength of sic fibers

Abstract Two types of SiC fibers, amorphous SiC(1000) (heat-treated at 1000°C) and microcrystalline SiC(1300) (heat-treated at 1300°C), were neutron-irradiated at 423 K to 8 × 10 23 and 2 × 10 24 n/m 2 , and at 923 K to 9 × 10 24 n/m 2 in JMTR ( E > 1 MeV ). In the 423 K irradiation, density increased and X-ray radial distribution function (RDF) indicated that the atomic arrangement of SiC(1000) scarcely changed and that of SiC(1300) was disordered. Tensile strength and Youngs modulus gradually increased with neutron fluences. In the 923 K irradiation, density increased and RDF indicated that the atomic arrangement of SiC(1300) scarcely changed and that of SiC(1000) was ordered. Youngs moduli of SiC(1000) and SiC(1300) greatly increased. There were differences in density, tensile strength and Youngs modulus between SiC(1000) and SiC(1300) before the irradiation, but their values became almost the same by the irradiation at 423 K to 2 × 10 24 n/m 2 and 923 K to 9 × 10 24 n/m 2 , respectively.

Designation of alloy composition of reduced-activation martensitic steel

An alloy composition of reduced-activation martensitic steel for fusion reactor is designed on the basis of the experimental results of postirradiation microstructure, mechanical properties, such as creep, fracture toughness and tensile properties, hydrogen effects and corrosion. At present, a desired composition of the steel is 0.1C-0.05Si-0.5Mn-9Cr-2W-0.25V-0.02Ti-0.05Ta- < 0.002S- < 0.002P by weight percent. Effects of the other minor elements such as Al, Zr and B are also inspected. An addition of 0.05 wt% Ta increases the high temperature strength but reduces the fracture toughness. Susceptibility to hydrogen-induced cracking is reduced by an addition of 0.03 wt% Al, though it results in a severe degradation of the fracture toughness. An addition of 30 wppm B together with the addition of 0.02 wt% Ti increases the fracture toughness. Void nucleation at grain boundaries, however, is enhanced by the B addition under the FFTF irradiation at 638 K in 10 dpa.