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Ion-driven permeation (IDP) behavior has been investigated for deuterium implanted into pure iron with an incident ion energy of 300 eV to 2.0 keV. A permeation spike was observed in the implantation into virgin or annealed membrane. A good linear relationship was observed between the permeation flux and the incident ion flux. The permeation ratio increased with increasing the temperature. It is deduced that the IDP process through pure iron was controlled by a recombination reaction on both sides of the membrane. The permeation flux decreased as the incident ion energy increased. The higher energy run, on the contrary, enhanced the permeation flux of the following runs with lower energies.

Ion implantation driven permeation (IDP) behavior has been investigated for deuterium implanted with low energy (100–1800 eV) into aluminium-lithium alloys (Al-0.89wt.%Li and Al-2.22wt.Li alloys) to stimulate the behavior of tritium. The experimental results showed that the steady state IDP behavior for AlLi alloys was divided into two temperature regions around 620 K for 0.89wt.%-Li alloy and 700 K for 2.22wt.%-Li alloy. In the high temperature regions, the IDP fluxes through both alloys depended significantly on the temperature, while the IDP fluxes were almost constant in the low temperature regions. This fact would suggest that the change in the above temperature dependencies are attributed to the change of the chemical trap site in alloys, because the temperatures at the turning point of the temperature dependencies coincide with those at the phase transition from α to α + δ phase.

Hydrogen isotope exchange reaction rate in tritium and methane mixed gas, as induced by tritium decay and beta radiation, has been experimentally measured. Initially T2 gas was filled to 40 kPa and 20 kPa of CH4 gas was added. The mixed gas spectrum was analyzed periodically by laser Raman spectrometry. The first order HT and H2 formation rates and T2 and CH4 decay rates by hydrogen isotope exchange reaction were observed between 2.9·10−3 h−1 and 4.8·10−3 h−1. Although the estimated hydrogen isotope exchange reaction rate was 1/20–1/10 slower than the rate of H2+T2 mixed gases, it was nearly equivalent to the ion formation rate by tritium beta radiation. This suggested that isotopic hydrogen radicals formed via ionization would disappear in the presence of methane.

A brief survey on recent advances in the field of tritium technology concerning the most important systems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation system, the tritium storage facility and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevant amounts of tritium that constitute the basis of proposed integral process concepts are described.

The Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed a Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. Under the collaboration program of USDOE and JAERI, an integrated process, the “JAERI Fuel Cleanup System” (JFCU) was designed and fabricated for testing at the Tritium Systems Test Assembly (TSTA) as a major subsystem of the simulated fusion fuel cycle. The JFCU has the capability to continuously process a mixture of hydrogen isotopes and impurities of 15 mol/h. The major components are: a palladium diffuser for hydrogen isotopes purification, a catalytic reactor for oxidation of tritium in impurities, cold traps for separation of tritiated water to produce exhaust containing a very low level of tritium, a ceramic electrolysis cell for decomposition of tritiated water, and zirconium-cobalt beds for hydrogen isotopes storage. In early 1990, the JFCU was installed in the TSTA and preliminary testing without tritium was initiated. Components such as the palladium diffuser, cold traps, catalytic reactor, scroll pump and electrolysis cell were tested independently as well as in the integrated process tests.

Application of laser Raman spectroscopy for fusion fuel gas processing was studied by measuring isotopic methanes exchanged with hydrogen isotopes, which are considered to be a major impurities in the processing. For experimental gases, isotopically equilibrated deuterium and methane were prepared in the presence of solid catalyst. Large Raman scattering peaks of v 1, bands were observed at 2,917 cm−1 for CH4 and at 2,100-2,200 cm−1 for deuterated derivatives of methane C(H,D)4. Under a spectral resolution of 5 cm−1, the v 1 bands of CH3D and CH2D2 were observed as an overlapped peak, the relative absolute Raman intensity ratio of each isotopic methane was obtained as CH4: CH3D+CH2D2: CHD3: CD4=230: 74: 144: 100. On the other hand, the Raman intensity ratio obtained from pure deuterated standard methane was CH4: CH3D: CH2D2: CHD3: CD4=230: 53: 33: 115: 105. It was confirmed that isotopically equilibrated hydrogen isotopes and methane mixed gas would be applicable for an alternative standard gas for fusion...

An overview is given of the design, construction and initiation of the operation of the facility, Tritium Process Laboratory, JAERI, to study and develop tritium processes for fusion fuel cycle and safe handling technology.

Uranium isotope separation by redox chromatography is analytically studied. The periodic withdrawal of products and tails and the introduction of natural feed are simulated on the assumption of a square cascade for a uranium adsorption band. The influences on the separative power and the lead time until product withdrawal are investigated by varying the magnitude of the isotope separation factor, uranium band length, tails concentration, and so on. Simulating calculations indicate that using ion-exchange resins to achieve uranium isotope separation requires a very long lead time for the production of highly enriched uranium.

The Tritium Process Laboratory (TPL) was constructed at the Japan Atomic Energy Research Institute (JAERI) in June, 1985. The objectives of this laboratory are the research and development of the fusion reactor fuel processing system and safe handling technology of large amounts of tritium. Total tritium inventory in the facility is 10 g and each experiment will be conducted with up to 1 g. Safety systems have been in operation since 1985 and have been tested without tritium. Tritium removal systems, gloveboxes, vacuum pumping system and tritium storage systems etc. have been proven to work as designed so that tritium can be handled safely in this facility. The various items of experimental apparatus were designed, fabricated, installed and checked early in 1988. First tritium was shipped from Oak Ridge National Laboratory in February, and then the “hot” operation was initiated in March, 1988.

A ZrCo intermetallic compound is proposed as the material for recovery, storage and supply of gaseous tritium. It is a substitute for the uranium that has been extensively used for this purpose in spite of its restricted handling as a nuclear fuel material. Pressure-composition isotherms of a ZrCoH2 system were investigated and this material proved to be applicable and advantageous. Some ZrCo beds were designed, fabricated and tested for actual use. A bed for the handling of large amounts of hydrogen isotopes was developed. The test establishes ZrCo beds as practical components for tritium service.

Reports on remarkable improvements in the previously reported mathematical model for multicomponent separating cascades, which are applicable to the cases where the interstage flows and the stage separation factors are input variables for the calculations. The number of the independent variables is greatly decreased for much more efficient iterative calculations by the multidimensional Newton-Raphson method. Particularly, if the stage separation factors are independent of concentrations of the up and down streams, the improved model presents great decreases both in the computation time needed at each iterative step and in the number of total iterations. Several numerical experiments made for a five-component system of N/sub 2/-O/sub 2/-/sup 41/Ar-/sup 85/Kr-/sup 133/Xe, which are separated by using the porous membrane method, indicate that the total computation time is shortened by almost two orders of magnitude if the improved model is used.

A computer code is developed for analysis of dynamic behavior of a single cryogenic distillation column. A dynamic simulation study is performed for the lead column in a typical column cascade which consists of four columns and two catalytic equilibrators. The atom fraction of protium in the bottom product and the reflux ratio are regarded as the controlled variable and the manipulated variable, respectively. Fluctuation or change of the feed composition is considered as the disturbance and the PI control scheme is introduced for a constant value control. The criteria are derived, which can be used in setting the choice of the two PI controller parameters. Investigated are the effects of the controller parameters on the response of the controlled variable to a step change in the feed composition, and the validity of the criteria is verified. The parameter setting method proposed in the present study has two advantages: it accounts for the strong nonlinearity of the column and it predicts the unstable regi...

The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have an important impact on the operation of the fuel processing system. For some blankets, such an aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system in terms of component size and power consumption. The importance of the blanket interface to a fuel processing experiment, such as TSTA, has been identified. The initial work to define the blanket processing system, which is proposed to be added as part of TSTA, will be discussed here.

Cryogenic distillation is a promising method for a hydrogen isotope separation system in the fuel cycle of nuclear fusion reactors. However, sufficient experimental information has not yet been obtained to enable us to achieve the cascade operation of cryogenic distillation columns. The present study aims to obtain such information. The experimental apparatus used in the present study is composed of a cryogenic distillation column, gas chromatographs and a helium refrigerator. The cryogenic distillation column can be operated in three modes; the total reflux mode, the total recycle mode and the total recycle mode with a feedback stream. In each experiment, the gas in the condenser and the gas or liquid in the reboiler are sampled and analyzed with gas chromatographs after the column pressure, the pressure drop in the packed section, and other operating conditions reach a steady state. The experimental results are arranged with the pressure drop in the packed section and the height equivalent to a theoretical plate (HETP). As a result of the total reflux experiments, it should be noted that the pre-flooding operation of the distillation column before the total reflux operation improves the separating efficiency of the distillation column and reduces the HETP value. The experimental results also show that the pressure drop increases with vapor velocity. The flooding point or the loading point is not observed with vapor velocity lower than 20 cm/s. As a result of the total recycle experiments, the HETP value is observed to decrease with reflux ratio.

An in-situ tritium recovery experiment was conducted in the JRR-2 reactor on Li 2 O and LiAlO 2 spheres. A technique for obtaining tritium diffusion coefficients from the time dependency of tritium release after incremental temperature changes was developed. The tritium retention in LiA1O 2 was higher than in Li 2 O for the same temperature. The apparent diffusion coefficients for LiAlO 2 at temperatures up to 900°C was one order magnitude higher than the previous, lower temperature TRIO data.

A simulation procedure of multicomponent separating cascades is developed. This procedure is applicable and effective in such cases that the cut of mixture of each stage is prescribed and required to be independent of concentration, and the stage separation factors are given as input variables and very large. It can deal with cascades which have multiple feeds and sidestreams. The stage separation factors are allowed to be given as functions of compositions of down streams and up streams. The exact solutions are found out treating all the basic equations simultaneously by use of multi-dimensional Newton-Raphson method. As an application of this procedure to actual cascade calculations, a computer simulation study is carried out for hydrogen isotope separation system by porous membrane method which consists of two cascades and two catalytic equilibrators. It is proved that the Newton-Raphson iterative calculation proceeds stably under the assumed input and output conditions, and the porous membrane method ...

In situ measurement of tritium release during irradiation provides a more realistic understanding of the tritium release mechanism than that obtained by the post-irradiation annealing method. The tritium release curves reached saturated values after irradiation of a few days at the reactor operation of 10 MW. The saturated values were in good agreement with the predicted values of the tritium generation rate obtained from neutronics calculation. Temperature dependence of tritium release under the temperature of 550°C at the pellet center was interpreted by the decomposition of LiOH(T) at the pellet surface. The ratio of tritium gas and tritiated water (HT/HTO) was strongly affected by the moisture generated from the piping system and the pellet. The retained tritium in Li2O was 0.5–0.9 wppm after the final stage of the irradiation of 760°C and the burnup of 0.24% of total Li.

A ceramic electrolysis cell and a palladium diffuser have been developed in Japan and tested with tritium at the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory, to test their feasibility as possible upgrades for the fuel cleanup system (FCU). The ceramic electrolysis cell, made of stabilized zirconia, was operated at 630/sup 0/C for an extended period with a mixture of 3% T/sub 2/O in He carrier gas in a circulation system with an oxidizing catalyst bed. The palladium diffuser was tested with pure tritium gas, circulated at 280/sup 0/C, to verify the compatibility of the alloy with tritium, since the /sup 3/He produced in the metal could cause degradation. The isotopic effects were also measured for both devices.