A. Hasegawa

D. Kato

H. Atsumi

H. Kurishita

K. Sato

M. Fukuda

M. Hatakeyama

S. Nogami

T. Muroga

T. Nagasaka

T. Nozaki

T. Toyama

V.Kh. Alimov

Y. Hatano

Y. Nagai

Y. Oya

Y. Ueda

National Institute for Fusion Science (NIFS-Repository)

0 1 2 3 4 5 610-310-210-1100 Depth (m) 673 K 773 K 873 K damage profile  D concentration in W (at.%) W-5Re  Understanding of tritium (T) behavior (diffusion, trapping, desorption, etc.) in neutron-irradiated plasma facing materials (PFMs) is indispensable for evaluation of tritium mass balance and assessing safety margins of fusion reactors. Tungsten (W) has been recognized as a primary candidate of PFMs, and the effects of neutron irradiation on T behaviors have been examined in the Japan–US joint research project TITAN.1–3) In this project, W specimens are irradiated with neutrons in the High Flux Isotope Reactor at Oak Ridge National Laboratory, and the retention of hydrogen isotopes is examined in Idaho National Laboratory using a linear plasma machine called Tritium Plasma Experiment (TPE).1-3) The objective of this work is to get better understating of the mechanisms underlying hydrogen isotope trapping in irradiated W through characterizations of defects in the W specimens prepared in the TITAN project, ion-irradiation experiments, and computer simulation. In FY2012, the effects of Re on deuterium (D) retention was examined because neutron irradiation results in not only formation of radiation defects but also that of transmutation elements such as Re and Os. The irradiation effects of neutrons were simulated by 20 MeV W ions. Specimens used were plates of W–Re alloy, in which the concentration of Re was 4.6 mass%. This specimen is hereafter denoted as W–5Re alloy. For comparison, plates of ITER-grade W and recrystallized W (RX-W) were also used as specimens. These specimens were irradiated with 20 MeV W ions to 0.5 dpa at room temperature and 873 K. Damage profiles were evaluated using the SRIM2008 program. The displacement threshold energy of W was set to 90 eV.The irradiated specimens were exposed to D2 gas at 100 kPa and 673–973 K. Duration of exposure was 36 ks (10 h) at 673 K, and 10.8 ks (3 h) at 773 and 873 K. The concentration of trapped D was examined by nuclear reaction analysis (NRA). Fig. 1 shows depth profiles of D and damage profile in the W-5Re alloy irradiated at room temperature. The damage level took the maximum value at the depth of 1.4 m, and the damaged zone extended up to ~2 m. Deuterium was clearly enriched in the damaged zone due to the trapping effects of radiation damages. The D concentration in the damaged zone decreased with increase in the temperature of exposure to D2 gas. Such reduction in D concentration was ascribed to decrease in occupation probability of traps with increase in temperature. Similar enrichment of D in the damaged zone and reduction of D concentration with increase in temperature were observed for all specimens examined in the present study.  The D concentration at the damage peak is plotted in Fig. 2 as a function of the temperature of exposure to D2 gas. Since no significant difference was observed between ITER-grade W and RX-W, only the results of W–5Re alloy and ITER-grade W are shown. The D concentration in the specimens irradiated at 873 K was lower than the values for the specimens damaged at room temperature. This observation indicated that a part of radiation damages was recovered during the irradiation at 873 K. At the same irradiation temperature, D concentration in W–5Re alloy was lower than that in ITER-grade W. In other words, Re has preferable effects from the viewpoints of T inventory. Further investigation is required to understand the mechanisms underlying the reduction in D retention by Re.                 Fig. 1 Depth profiles of D and damage profile in W–5Re alloy irradiated with W ions at room temperature and exposed to D2 gas at temperatures indicated in the figure.                 Fig. 2 Deuterium concentration at the damage peak in    W–5Re alloy and ITER-grade W.  1) Hatano Y. et al.: Mater. Trans. 54 (2013) 437. 2) Hatano Y. et al.: Nucl. Fusion 53 (2013) 073006. 3) Hatano Y. et al.: J. Nucl. Mater., in press, doi:10.1016/j.jnucmat.2013.01.018. 600 650 700 750 800 850 900 95010-210-1100 ITER-W, 0.5 dpa@RT ITER-W, 0.5 dpa@873K W-5Re, 0.5 dpa@RT W-5Re, 0.5 dpa@873K  D concentration  (at.%)Temperature of exposure to D2 gas () A vanadium alloy is expected as a candidate material of blanket system with liquid Li coolant for fusion reactor application. Since there is not yet enough knowledge about radiation effect and compatibility between liquid Li coolant and structural materials, its engineering design cannot be determined. In order to resolve this problem, it is necessary to exam the compatibility experiment in liquid Li with Li-enclosed capsule in pile, so that it is required to develop the Li-enclosed capsule as an irradiation vehicle in pile.  In this study, we discuss the compatibility experiment in liquid Li with Li-enclosed capsule in the domestic reactor such as JMTR and Joyo, and aim at acquiring the data in order to design the capsule.  Task items in this plan are shown in the following;  Analyses of the tritium production and leakage amount from Li-capsule in irradiation  Evaluation of thermal expansion difference (TED) in Li-fulfilled capsule  Analysis of corrosion resistance of material against Li coolant  Design of Li-enclosed capsule   A cylindrical vessel made by V-5Ti alloy was manufactured in order to design an irradiation capsule enclosed with liquid lithium. A series of rolling work and deep drawing of the V-5Ti alloying were performed in Taiyo Koko Co. Ltd. The cylindrical vessel composed of a pipe with 105mm in length x 22mm OD x wall thickness of 0.5mm. In order to reduce the number of welded joint in the vessel, the deep drawing of pipe products made in V-5Ti was attempted and the vessel manufacture was succeeded. By means of the vessel, specimens of V-4Cr- 4Ti alloys designated as NIFS-Heat-2 were enclosed in loquid lithium in the vessel by electron beam welding in vacuum. In order to perform the electron beam welding, a handy-type container was fabricated. The container was prepared for transporting the vanadium specimens and liquid lithium no to touch the air and moisture. Figure 1 shows a photo that the vanadium alloy specimens were fulfilled with liquid lithium in the vessel. Figure 2 shows a view of the handy-type container for electron beam welding. The manufacturing process of electron beam welding was performed in the SFC Co. Ltd. After the weldment, a visual inspection on the weld point was examined and no crack and leakage could be found. Consequently, it is concluded that the manufacturing process of lithium bonding irradiation capsule and enclosed process of liquid lithium in the vessel can be established. A lithium compatibility test for vanadium alloys at the elevated temperature of 700C will be performed with this capsule in 2013. A microball with 1.2mm diameter necessary for encapsulation of thermal expansion difference monitor (TED) was made on an experimental basis. It was found that the processing technique was not enough in the reial manufacture of microball, and a design of the enclosure plug geometry to take place of microball for endcap encapsulation has been examined. It is expected that component design implementation of TED will be over in 2013 and the settlement of encapsulation approach are aimed in future. A TED performance test in the liquid lithium and the corrosion test are going to be performed soon after the manufacture of TED is succeeded.    Figure 1 : Cylindrical vessels made by V-5Ti alloys with vanadium specimens fulfilled in lithium.  Figure 2 : The handy-type container for electron beam welding. 577§78. Mechanism Underlying Trapping of Hydrogen Isotopes in Neutron-irradiated Plasma Facing MaterialsHatano, Y., Alimov, V.Kh., Nozaki, T. (Univ. Toyama), Kurishita, H., Nagai, Y., Toyama, T., Hatakeyama, M., Hasegawa, A., Nogami, S., Fukuda, M. (Tohoku Univ.), Oya, Y. (Shizuoka Univ.), Sato, K. (Kyoto Univ.),  Ueda, Y. (Osaka Univ.), Atsumi, H. (Kinki Univ.),  Muroga, T., Nagasaka, T., Kato, D.

Mechanism Underlying Trapping of Hydrogen Isotopes in Neutron-irradiated Plasma Facing Materials

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Mechanism Underlying Trapping of Hydrogen Isotopes in Neutron-irradiated Plasma Facing Materials

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