Y. SAKUMA, H. YAMANISHI, T. UDA and H. HIRABAYASHI National Institute for Fusion Science Furo-cho, Chikusa-ku, Nagoya 464-01, Japan At the National Institute for Fusion Science, we are planning to carry out DD experiments using the Large Helical Device, in the near future. 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.
1.
Introduction
When DD experiments using deuterium are carried out in a nuclear fusion experimental facility, tritium is bound to be generated, though the amount may be slight. Here at the National Institute for Fusion Science, we are planning to carry out DD experiments as a second stage program for the Large Experimental Device Helical (referred to as a “LHD” hereafter). The world’s largest helical device is The under  construction. now completed LHD will have a major of  3.9m  and  utilize radius superconducting coils to generate its magnetic field. The main goal of the project are the demonstration of currentless state plasma operation and the achievement of reactor- relevant plasma parameters. Since the disposal of generated tritium within the Laboratory is our policy, it is necessary to develop a tritium cleanup system. To recover discharged    from the tritium experimental facility, it is usual to employ a system (referred to as “wet system" hereafter) by which the entirety of gas including tritium is by converted to water vapor catalytic oxidation so as to adsorb the vapor by a molecular sieve. However, it is not rational to
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deliberately convert the discharged tritium from the elemental state into water vapor, the latter one affects human body 15,000 times more strongly. Besides, in the wet system the device generally tends to become in size. In view of this, we have developed a tritium cleanup system (referred to as “dry system” hereafter) by which the elemental tritium is fixed directly to hydrogen absorbing alloys which have not so far been developed at any other place. Because of the LHD exhaust composition is unclear, we assumed it as shown in table 1. The device will at most generate 430GBq of tritium a year. We assumed all of the tritium would be exhausted during  normal  operation  (DD experiment) and discharge cleaning operation. This means the device will exhaust 430GBq of tritium during normal operation and will also   exhaust   430GBq of   tritium during discharge cleaning operation. In table 2 we showed design conditions of the exhaust of the LHD.
2. DESIGNING

Table 2 Tritium concentration of the outlet of the cleanup system with the legal regulation level [kBq/Nm³]
Q means one of hydroigen isotopes; H,D,T. The main constituent of the exhaust of the experimental facility for DD experiments is deuterium, but there are also mixed a trace of helium and its tritium compounds and a variety of impurities. It is an easy task to remove elemental hydrogen isotopes, namely, protium, deuterium and tritium (referred to simply as “hydrogen” hereafter), from the mixture of hydrogen and helium if one employs hydrogen absorbing alloys. However, it will no longer be a simple matter if there are mixed water vapor, Methane and the like even in a small amount. Such impurities are actually included in the exhaust of a fusion experimental facility. We have completed a conceptual design for a tritium cleanup system in which a    nonvolatile    getter (decomposer) material is employed to let it adsorb carbon and oxygen generated by the decomposition of these impurities caused by the decomposing      material.       And hydrogen absorbing alloys employed to let it adsorb the elemental hydrogen, where these materials may be stored or disposed of as deemed necessary. We have examined the case where titanium is used as a hydrogen absorbing alloy and Zr3Al2 used     as     a     nonvolatile is decomposing material. For titanium is has a low equilibrium hydrogen pressure and is inexpensive¹）and there is an available report as to the decomposing material's action on methane²). The result of design, of a dry tritium cleanup system as shown in Table 2, for the exhaust system in the LHD revealed that outlet    concentrations    are the 0.0kBq/m³ for water vapor tritium and    0.3kBq/m³      for    elemental tritium. In contrast to the outlet concentrations of the wet system previously designed are 64.9kBq/m³ for water vapor tritium and 1.4kBq/m³ for elemental tritium. So that it is obvious that the dry system is far more advantageous from the viewpoint of safety. The comparison of the dry system and the wet system is shown in Table 3. The flow chart of the system is shown in Fig.1. The flow rates, the temperatures and the tritium concentrations at each part are shown in Table 4. The table (a) is in the case of normal operation and (b) is in the case of discharge cleaning operation.
CONCLUSION
3.
A tritium cleanup system by which the elemental tritium is fixed directly to absorbing alloys have been developed. It could perfectly recover the from the tritium exhaust of the Large Helical Device. 1) W.M.Muller. J.P.Blackledge and G.G.Libwitz, Metal Hydrides, Academic Press, New York and London (1968) H.Heimbach, H.R.Ihle 2) and C.H.Wu, Removal of Nitrogen and Methane from Hydrogen by Metal Getters, Fusion Technology 1984, Vol.1, 421-426 (1984)

flow rates, the temperature and the tritium concentaration at each part of the sys
e numbers are corresponding to the numbers in Fig.1.) 'able 4