A. Kimura

Irradiation hardening of reduced activation martensitic steels

Abstract Irradiation response on the tensile properties of 9Crue5f82W steels has been investigated following FFTF/MOTA irradiations at temperatures between 646 and 873 K up to doses between 10 and 59 dpa. The largest irradiation hardening accompanied by the largest decrease in the elongation is observed for the specimens irradiated at 646 K at doses between 10 and 15 dpa. The irradiation hardening appears to saturate at a dose of around 10 dpa at the irradiation temperature. No hardening but softening was observed in the specimens irradiated at above 703 K to doses of 40 and 59 dpa. Microstructural observation by transmission electron microscope (TEM) revealed that the dislocation loops with the a〈100〉 type Burgers vector and small precipitates which were identified to be M6C type carbides existed after the irradiation at below 703 K. As for the void formation, the average size of voids increased with increasing irradiation temperature from 646 to 703 K. No voids were observed above 703 K. Irradiation softening was attributed to the enhanced recovery of martensitic structure under the irradiation. Post-irradiation annealing resulted in hardening by the annealing at 673 K and softening by the annealing at 873 K.

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 Feue5f8Crue5f8W steels with additions of small amounts of V, Mn, Ta, Ti, Zr, etc. and (3) Feue5f8Crue5f8Mo, Nb or V ferritic steels for comparison. In Fe-0–15% Cr and Feue5f8Crue5f8Mo 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.

Void swelling of Japanese candidate martensitic steels under FFTF/MOTA irradiation

Abstract Microstructural observations of six Japanese candidate 7–9% Cr reduced activation martensitic steels were carried out after heavy neutron irradiation in order to investigate the void swelling behavior of each steel. Neutron irradiations were performed in the FFTF/MOTA up to 67 dpa at temperatures between 638 and 873 K. Transmission electron microscope observations revealed that voids were formed in all the steels irradiated to 67 dpa at 703 K, and the highest void swelling was observed in JLM-1 which was added with 30 wt.ppm of boron (0.74%), and the minimum void swelling was observed in F82H steel (0.12%). The 9% Cr martensitic steels showed the peak of void swelling at temperatures around 700 K, where void swelling gradually increased with increasing irradiation fluence to 30 dpa and increased rapidly above it. It is considered that the incubation period of void swelling of 9% Cr martensitic steels (JLM series) is about 30 dpa. JLM-1 showed the highest void swelling rate (0.045%/dpa at most). The addition of 30 wt.ppm of boron enhanced void swelling, while it was suppressed by the addition of 100 wt.ppm Ti in the 9% Cr martensitic steel. The JLF-3 steel (7.03% Cr) and F82H (7.65% Cr) showed less void swelling than JLF-I (9.04% Cr). The alloying effects on the swelling behavior of the steels were interpreted in terms of the difference in the precipitation morphology of carbides.

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 Feue5f89Crue5f82W 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 Feue5f89Crue5f82W 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.

Impact properties of intermetallic compounds

Abstract The impact fracture energies of γ-TiAl and Co 3 Ti intermetallic compounds were measured at temperatures between 293 and 1273 K using miniaturized U-notch specimens. For comparison, the impact fracture energy of a ferritic steel was also measured. The fracture energy of γ-TiAl gradually increases with increasing test temperature and the fracture mode changes from cleavage fracture at 293 K to intergranular fracture at elevated temperatures. The fracture energy of γ-TiAl is almost constant at around 1073 K and appears not to vary with its tensile properties, since the yield stress drastically decreases, while the tensile elongation drastically increases at that temperature. The increase in the fracture energy of γ-TiAl with temperature is considered to be as a result of the susceptibility to hydrogen-induced cleavage fracture decreasing at higher temperatures. A Co 3 Ti (Co-23.1at.%Ti) intermetallic compound breaks in a completely ductile manner when impact tested, while it undergoes severe intergranular embrittlement when deformed at a slow strain rate.

Neutron irradiation effects on the microstructure of low-activation ferritic alloys

Abstract Microstructures of low-activation ferritic alloys, such as 2.25% Cr-2% W, 7% Cr-2% W, 9% Cr-2% W and 12% Cr-2% W alloys, were observed after FFTF irradiation at 698 K to a dose of 36 dpa. Martensite in 7% Cr-2% W, 9% Cr-2% W and 12% Cr-2% W alloys and bainite in 2.25% Cr-2% W alloy were fairly stable after the irradiation. Microvoids were observed in the martensite in each alloy but not in bainite and δ-ferrite in 12% Cr-2% W alloys. An addition of 0.02% Ti to 9% Cr-2% W alloy considerably reduced the void density. Spherical (Ta, W) and Ti-rich precipitates were observed in the Ti-added 9% Cr-2% W alloy. Precipitates observed in 9% Cr-2% W and 7% Cr-2% W alloys are mainly Cr-rich M 23 C 6 (Ta, W) and Ta(W)-rich M 6 C and Fe-rich Laves phase. In 2.25% Cr-2% W alloy, high density of fine (Ta, W)-rich M 2 C type precipitates as well as M 6 C were observed. Spherical small α Cr-rich particles were observed in both martensite and α-ferrite in 12% Cr-2% W alloys. Correlation between postirradiation microstructure and irradiation hardening is shown and discussed for these alloys.

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.

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.