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Abstract: Effects of neutron irradiation were examined on microstructural evolution in powder metallurgically processed V–1.6Y (in wt%) alloys comprising essentially fine grains and very fine dispersoids of Y2O3 and YN. The alloys feature a highly pure matrix free from gaseous interstitials such as oxygen and nitrogen. Optical and transmission electron microscopy observations revealed that the martensite of a vanadium bct structure with a tetragonality (c/a) of 1.06 formed heterogeneously in some of the bcc grains of V–1.6Y irradiated at 290 and 600°C to 0.25 and 0.6 displacement per atom, respectively, but not in unirradiated V–1.6Y. The martensite had {101)bct habit planes parallel to {110}bcc in the matrix and a high density of {101)bct microtwins as the lattice invariant shear strain. To the authors’ knowledge, the emergence of martensite is the first to be observed for vanadium or its alloys in neutron irradiated states. The cause of the martensite formation is discussed. [Copyright &y& Elsevier]

Abstract: In order to elucidate the effects of pre-irradiation microstructures and irradiation conditions on radiation embrittlement and radiation-induced ductilization (RIDU), fine-grained Mo–0.2wt%TiC specimens with high and low reduction rates in plastic working, which are designated as MTC-02H and MTC-02L, respectively, were prepared by powder metallurgical methods. The specimens were neutron irradiated to 0.1–0.15dpa with controlled 1-cycle and 4-cycle heating between 573 and 773K, and 473 and 673K, respectively, in JMTR. Vickers microhardness and three-point bending impact tests and TEM microstructural examinations were made. The degree of radiation embrittlement, assessed by DBTT shift due to irradiation, was strongly dependent on the reduction rate and cycle number. The 4-cycle irradiation suppressed the radiation embrittlement compared with the 1-cycle irradiation, and the suppression was much more significant in MTC-02L than in MTC-02H. The observed behavior is discussed in connection with RIDU and microstructural evolution caused by the 4-cycle irradiation. [Copyright &y& Elsevier]

Abstract: In order to improve the resistance to serious radiation embrittlement in refractory metals, a process for microstructural control is presented, with significance of controlling the substance, size, number density and distribution of dispersed particles. The process is applied to vanadium, a metal of group VA, with ultra-fine grains and finely dispersed, thermally stable particles. It is shown that the developed V–(1.6–2.6)%Y alloys have grain sizes between 170 and 340nm and particle (Y2O3 and YN) sizes between 13 and 25nm and exhibit the following noted features: (1) The microstructure is stable against vacuum heating at 1573K for 1h. (2) The vanadium matrix is very low in solute oxygen and nitrogen, which are responsible for environmental embrittlement of vanadium. (3) The alloys exhibit good tensile strength and elongation at low temperatures; approximately 1.5 times more strength and about 70% of the elongation compared with a V–4Cr–4Ti (Nifs-Heat-1) alloy. (4) The developed alloys show good resistance to neutron irradiation at 563K to 0.25dpa and 873K to 0.60dpa. These fabrication processes and noted features are applicable to other group VA metals such as tantalum. [Copyright &y& Elsevier]

Surface modifications of toughened, fine-grained, recrystallized tungsten (TFGR W) materials with 1.1 wt.% TiC and 3.3 wt.% TaC dispersoids due to repetitive ELM-like pulsed (∼0.15 ms) helium plasma irradiation have been investigated by using a magnetized coaxial plasma gun. No surface cracking at the center part of the TFGR W samples exposed to 20 plasma pulses of ∼0.3 MJ m −2 was observed. The suppression of surface crack formation due to the increase of the grain boundary strength by addition of TiC and TaC dispersoids was confirmed in comparison with a pure W material. On the other hand, surface cracks and small pits appeared at the edge part of the TFGR W sample after the pulsed plasma irradiation. Erosion of the TiC and TaC dispersoids due to the pulsed plasma irradiation could cause the small pits on the surface, resulting in the surface crack formation. [ABSTRACT FROM AUTHOR]

Abstract: About one thousands of small size specimens will be irradiated in the High Flux Test Module (HFTM) with a limited irradiation volume of 0.5l in the International Fusion Materials Irradiation Facility (IFMIF). It is necessary to verify that the experimental data of these small specimens mechanical characterization can be safely extrapolated to standard specimen data, enabling a sound dimensioning of DEMO reactor. The program of small specimen test technique (SSTT) in IFMIF/EVEDA (Engineering Validation and Engineering Design Activity) phase for fatigue, fracture toughness and crack growth measurement is summarized, and recent progress and some analysis of the experiments for small size specimens have been shown. [Copyright &y& Elsevier]

Abstract: High heat flux experiments using a hydrogen–helium mixture beam have been carried out on powder metallurgy tungsten (PM-W) and ultra fine grain W–TiC alloy (W–0.5wt%TiC-H2). The energy of is 18keV. Beam flux and heat flux at the beam center is 2.0×1021 atoms/m2s and 7.0MW/m2, respectively. Typical ratio of He/D ion is 0.25. Beam duration is 1.5–3s and interval of beam shot start is 30s. The samples are irradiated up to a fluence of 1022–1024 He/m2 by the repeated irradiation pulses. After the irradiation, surface modification by the irradiation and hydrogen retention, surface composition have been investigated. Surface modification by hydrogen–helium mixture beams is completely different from results of single beam irradiation. In particular, mixture beam irradiation causes remarkably high hydrogen retention. [Copyright &y& Elsevier]

Abstract: In order to study hydrogen blistering and subsequent cracking behavior of pure W, K-doped W, and La2O3-doped W, 1keV (main ion components) ion beams were irradiated at 653K to fluences up to 1×1025 H/m2. Two pre-irradiation heat treatments were done for stress relief (900°C) and recrystallization (1300°C for pure W and 1500°C for K-doped and La2O3-doped W). It was found that blister characteristics and cracking behavior strongly depended on microstructures and dopant materials. For W materials with layered microstructure, blister shapes were mostly spherical-like, while for W materials with recrystallized (or disordered) microstructures, blisters had complicated plateau-like shapes with many cracks. Addition of K or La2O3 dopants increased the number of blisters and exfoliations for both stress relieved and recrystallized W. [Copyright &y& Elsevier]

Abstract: Deuterium retention in Toughened, Fine-Grained Recrystallized W (TFGR W-1.1wt%TiC) was studied, compared to pure W. D implantation was performed to a fluence of 1×1024 m−2 at temperatures of 473–873K, followed by TDS. It was found that D retention in TFGR W is higher than in pure W at all irradiation temperatures. Namely, at 673K, D retention in TFGR W is six times higher than pure W. TDS spectrum of TFGR W irradiated at 573K has a large peak around ∼700K with small shoulder up to ∼1100K. In the case of D+He simultaneous irradiation, D retention is about 30% lower than for pure D. In addition, plasma exposure experiment was also conducted in TEXTOR, followed by NRA. Higher retention in TFGR W-1.1wt%TiC could be attributed to high grain boundary diffusion (then trapping deeper into the bulk) and formation of TiD2. [Copyright &y& Elsevier]