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The behavior of tritium sorption, thermal desorption and photo-desorption was investigated for varying metal surfaces. Abstract An experimental reactor such as ITER is planning various tests using several kinds of blanket designs in addition to demonstrating the physics of burning D-T plasmas. The data of neutron irradiation performance of a blanket is needed for the fusion blanket design. The present status of these studies is briefly described in this report. Abstract Hydroxyl groups on the surface of Li 2 O were studied by using a diffuse reflectance method with Fourier transform infrared absorption spectroscopy at high temperature up to K under controlled D 2 O or D 2 partial pressure.

It was found that hydroxyl groups could exsit on Li 2 O surface up to K under Ar atmosphere. Under D 2 O containing atmosphere, only the sharp peak at cm -1 was observed at K in the O—D stretching vibration region. Below K, multiple peaks of the surface -OD were observed and they showed different behavior with temperature and atmosphere.

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Multiple peaks mean that surface is not homogeneous for D 2 O adsorption. Assignment of the observed peaks to the surface bonding structure was also discussed. Tritium release behavior from each observed surface site was discussed. Peculiar tritium release behavior which was observed during temperature increase was also discussed. Abstract For the development of fusion reactor blanket systems, some of the key issues on the tritium recovery performance of solid and liquid breeder materials were studied.

In the case of solid breeder materials, a special attention was focussed on the effects of irradiation on the tritium recovery performance, and tritium release experiments, luminescence measurements of irradiation defects and modeling studies were systematically performed. For liquid breeder materials, tritium recovery experiments from molten salt and liquid lithium were performed, and the technical feasibility of tritium recovery methods was discussed.

Abstract Tritium is sorbed not only on the surface of the specimen or material used in experiments but also on the surfaces of tubes, joints and so forth, which constitute the experimental apparatus. This could lead to incorrect experimental results and erroneous understanding of observed phenomena. The authors call phenomenon "system effect".


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Therefore, it is indispensable to investigate this effect before the behavior of tritium on a certain material is studied. In this work, the system effect was studied using an experimental piping system. Moreover, numerical simulations of tritium behavior in an aluminum-coated box were carried out using results obtained in the experiments.

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Abstract Hydrogen permeation through membranes of selected metals was studied with low temperature plasma generation apparatus. The permeation was found to increase by applying plus voltages to the membrane. The permeation enhanced by the plus bias is due to the dissociation of neutral hydrogen molecules into atoms on the membrane by incident electrons. Tanase 1 , K.

Kurosawa 1 , M. Kato 1 , M. Hashimoto 1 , T.


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Masuzaki 2 , K. Ishida 2 and K. Nagamine 2, 3.

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Through the production run, no leakage of tritium from the facility was observed. About 60 TBq of tritium gas in a high purity was effectively obtained by the batch processing with the improved gas chromatograph. At the Branch, a tritium handling system THS was installed for removing 3 He of a decay product of tritium. Before the fusion experiments, performances of U-getters, Ti-getters, Pd-filter, and all other parts were tested and the fusion experiments have been carried out. Abstract A kind of gas chromatography and a tritium counting device were developed for the separation of hydrogen isotopes.

The Pd-Pt alloy showed the best separation efficiency among the three materials. The bremsstrahlung X-ray counting device developed for measuring high concentration tritium showed a good linearity between the counting rate and the tritium pressure, the specific counting rate being evaluated as The combined use of these two devices is expected to be applicable to the recovery of tritium from the flow of fuel gas in thermonuclear fusion reactor.

Abstract In Isotope Separation Laboratory at Nagoya University, we have been studying water distillation and thermal diffusion for hydrogen isotope separation. The present paper describes some recent developments of the separative analyses on these thechnologies. The device will generate a small amount of tritium, as a fusion product. In order to remove it from the exhaust, we have designed a tritium cleanup system based on a new concept. This system is mainly composed of a palladium permeater, a decomposer and hydrogen absorbing alloys.

It could perfectly recover the tritium without oxidizing it. Abstract Current status of tritium technology to fabricate laser fusion targets is described. Tritium facilities for laser irradiation, deuterium-tritium DT fuel loading system, fabrication of deuterated-tritiated polystyene shells, measurement of partial pressure of tritium and cryogenic technologies to fabricate a thick solid DT layer are briefly mentioned.

Issues to fabricate cryogenic DT target for the coming upgraded laser system is also discussed. The thermal reactor is cooled by light water and moderated by heavy water.

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