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Ida, Yuma*; Matsuura, Hideaki*; Nagasumi, Satoru*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki
no journal, ,
JAEA and Kyushu University have studied the tritium production method using high temperature gas-cooled reactors (HTGR) for initial fusion reactors. In this method, lithium compounds are loaded into the reactor core and tritium is produced with 
Li(n,
)T reaction. We studied about optimization of lithium loading method, effective tritium containment method and nuclear thermal design of lithium loaded HTGR, and consequently we confirmed the feasibility of the tritium production method. Then, we started preliminary study for lithium irradiation experiment by test reactors. This paper describes evaluation results of tritium production and tritium containment for proposed lithium irradiation capsule.
Ida, Yuma*; Matsuura, Hideaki*; Nagasumi, Satoru*; Koga, Yuki*; Okamoto, Ryo*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo
no journal, ,
Tritium production method using HTGRs (High Temperature Gas-cooled reactors) is studied as the tritium supplying method for initial D-T fusion reactors. In this method, tritium is produced by Li (n,)T reaction. The amount of tritium production and the tritium confinement capability were evaluated in case of the irradiation capsule including the Li compound is installed into the HTGRs in the past. In this study, the tritium confinement capability is evaluated for the irradiation capsule with ZrC layer by performing calculations of the amount of tritium leakage. The calculation results showed that the amount of tritium leakage is decreased to one fifth with the ZrC layer.
850
C) conditionsOkamoto, Ryo*; Matsuura, Hideaki*; Ida, Yuma*; Koga, Yuki*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; Nagasumi, Satoru; et al.
no journal, ,
Currently, many researches to achieve DT nuclear-fusion power generation are under proceeding but the method to provide initial tritium loaded to fusion prototype reactor is not clear. The method of tritium production by using high temperature gas-cooled reactor (HTGR) was proposed. In this method, lithium rods are loaded to the reactor core of HTGR and tritium is produced by Li(n,)T reaction. And the method to reduce the spilled tritium by using the lithium rod with zirconium layer was proposed. In this study, the experiments to evaluate the performance of hydrogen absorption in the zirconium layer were conducted under the temperature condition more than 700C which is the normal operation condition for the very high temperature gas-cooled reactor (VHTR). The experimental result concerning solubility and diffusion factor of hydrogen in the zirconium layer will be presented and discussed.
Ida, Yuma*; Matsuura, Hideaki*; Nagasumi, Satoru; Okamoto, Ryo*; Koga, Yuki*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; et al.
no journal, ,
Large quantity of tritium is demanded for starting up of fusion reactor and engineering test using tritium for fusion blanket system. However, tritium is very rare and kg order of tritium must be produced artificially. Tritium production, by Li(n,)T reaction using the high temperature gas-cooled reactor (HTGR), has been proposed. In this method, loading of Li rods into burnable poison (BP) holes in HTGR is considered. In this paper, the Li rod suited to the demand for the utilization in High Temperature engineering Test Reactor (HTTR) is designed, and tritium production and leakage from Li-rod capsule are evaluated by adjusting the thickness of LiAlO
, alumina, and Zr layers. A scenario of irradiation test supposed to be conducted at HTTR for demonstration of the tritium production and containment performance of the Li rod is presented.
C) conditionsOkamoto, Ryo*; Matsuura, Hideaki*; Ida, Yuma*; Koga, Yuki*; Suganuma, Takuro*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; et al.
no journal, ,
It has been proposed that lithium rods, which are cylindrical lithium compounds, are loaded into a HTGR and tritium for initial fusion reactors is produced by Li(n,)T reaction. In this study, it was discussed that the lithium rods are covered with zirconium layers to prevent the produced tritium leak. The solubility and diffusion coefficient of hydrogen in zirconium were measured and the effectiveness of the zirconium layers on prevention of tritium leakage was estimated with the measured values. As a result, the tritium leakage ratio with the zirconium layers was estimated two orders lower than that without the zirconium layers, and hence it was considered that the zirconium layer is very effective on the prevention of the tritium leakage.
Koga, Yuki*; Matsuura, Hideaki*; Okamoto, Ryo*; Ida, Yuma*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; Nagasumi, Satoru; et al.
no journal, ,
Large quantity of tritium is demanded for starting up of fusion reactor and engineering test using tritium for fusion blanket system. Tritium production, by Li(n, )T reaction using the high temperature gas-cooled reactor (HTGR), has been proposed and the method to produce tritium by loading the lithium rods as burnable poison in the reactor core has been studied. In this presentation, the design of lithium rods to be loaded to High Temperature engineering Test Reactor (HTTR) and its irradiation test plan to demonstrate tritium production are presented.
Okamoto, Ryo*; Matsuura, Hideaki*; Ida, Yuma*; Koga, Yuki*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; Nagasumi, Satoru
no journal, ,
A study on tritium production using a high-temperature gas-cooled reactor has been carried out and it was proposed that zirconium is loaded into the lithium irradiation capsule to confine tritium within the irradiation capsule under high temperature condition. In this study, zirconium loading method was examined by numerical calculations to improve the tritium confinement. As a result, it was found that improvement in the tritium confinement can be expected by loading spherical zirconium into the irradiation capsule.
Ida, Yuma; Obata, Hiroki; Kimura, Yasuhiko; Onozawa, Atsushi
no journal, ,