Since early 1980's, JAERI and DOE (LANL) have collaborated on developments of tritium technology for fusion reactors. About ten years collaboration, specially under the Annex III and IV agreements, has produced many remarkable outcomes on tritium fuel processing and tritium safety handling technologies.
Under the Annex III agreement, the two "process-ready" components for tritium fuel processing such as a palladium diffuser and an electrolysis cell were tested. Through tritium tests, both components were verified to be feasible and attractive for various applications in fusion fuel processes. Based on the results obtained in the component tests, an integrated process has been developed and designed by JAERI for full scale demonstration of the plasma exhaust processing. The process loop will be tested using tritium under the Annex IV agreement.
In the first three years of joint operation of TSTA under Annex IV, a number of the integrated loop runs and non-loop tests of the subsystems were performed using 100 g-level tritium to demonstrate a tritium fuel processing cycle. The engineering database and operating experience of TSTA obtained in the runs and tests would be reflected in design and fabrication of tritium fuel processing systems for next-step fusion devices such as ITER and FER.

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Performance of compact intense 14-MeV neutron source based on latest experimental data of Ultra-Low-q discharge is studied. Neutron wall loading of 1.4 MW/m² is attained with compact device with following parameters ; major radius and minor radius are 2 and 0.45 m, respectively, toroidal field is 10 T, the plasma current is 10 MA, and the edge q value is 0.5. Required energy confinement time for achieving these specifications is 55 ms and this is attainable if the present database can be extrapolated to the high current regime. Without use of anomalous effect that is proper to the Ultra-Low-q discharge, energy confinement time longer than 1 s is necessary to obtain sufficient amount of neutron wall loading.

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A subcriticality measurement method that employs an intrinsic neutron source is proposed based on the one-point subcritical reactor model and the method is demonstrated through laboratory experiment and analysis.
The test fuel assemblies were LWR-type 5×5, 8×8 and 14×14 arrays. Each of these assemblies was submerged in water, and multiplied neutrons originally emitted in spontaneous fissions of ²³⁸U in the fuel rods were measured around the assembly. A cadmium box was placed around the assembly and then removed, and the neutron countrate ratio was measured. This ratio was converted into a k-effective value with the aid of calculated parameters, and compared with the value obtained through a Monte Carlo calculation. Fuel enrichment ranged 15 w/o, and the k-effective value was in the range of 0.250.85 in the present study.
Almost all the experimental k-effective values agreed with those calculated within ±0.03 Δk for the uniformly fueled and source-distributed multiplying system in water.

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Buoyancy and Marangoni convection in melted zone generated by the melting and resolidification process of stainless steel under high power CO2 laser irradiation was analyzed numerically. Governing equations with vorticity and stream function were solved for the flow field in the melted zone. Latent heat of fusion for melting and resolidification was treated with temperature recovery method. Development of the melted zone with time and the effect of the convection on heat transport process were discussed. Temperature in melted zone became more uniform and the depth of melted zone became thinner in depth and wider radially, as the convection was taken into account, in comparison with the cases in which it was neglected. Marangoni convection generated a high velocity convection with some small value of the temperature coefficient of surface tension. Comparison of the analytical results with the experimental results for several seconds irradiation under atmospheric condition showed that, the velocity of the surface of melted zone was almost in agreement with the results assuming the buoyancy convection, and the net evaporation coefficient was estimated to be much smaller in the experiment under atmospheric condition than that under vacuum condition.

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A three-dimensional analysis method for sloshing behavior of fast breeder reactor (FBRs) is developed. The method treats the coolant in a reactor vessel as a potential flow with moving liquid surfaces. The Laplace equation of a velocity potential is solved by a boundary element method with its boundary condition described by a Bernoulli equation.
The vibration test results of a rectangular water pool are calculated by the method. Then, the method is applied to analysis of sloshing behavior of uni- and multi-vessel type FBRs. The latter consists of vessels for the core, heat exchangers and pumps. These vessels are connected by piping. In the case of the uni-vessel type FBR, heat exchangers and pumps are placed in the reactor vessel. The characteristics of sloshing behavior of both the reactors are presented.

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Five neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No. 3). Two of them are for thermal neutrons and the other three are for cold ones. The characteristic wavelength of the thermal neutron guide tubes is 2 A, and those of the cold neutron guide tubes are 4 and 6 A. The longest guide tube is 59.9 m long and the total length of guide tubes is 232.1 m.
The beam sizes are 2 cm×20 cm for the thermal neutron beams and 2 cm×12 cm for the cold neutron beams. A curved part of the neutron guide is assembled by a polygonal approximation with use of 85 cm long straight units. The neutron mirrors of these units are made of natural Ni deposited borosilicate glasses. The Ni layer is about 2, 000 A in thickness.
The mean fabrication error of guide tube units is 4 μm. The mean installation errors are 8 μm for the positional abutment error and 5×10^-6 rad for the angular error. The neutron losses by these errors will be about 5%, and the neutron fluxes at the exits of the neutron guides are estimated to be about 2×10⁸ n/cm²•s.

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In the next century, the "fission breeder" concept will not be practical to solve the global energy problems, including environmental and North-South problems. As a new measure, a simple rational Th molten salt breeding fuel cycle system, named "Thorium Molten-Salt Nuclear Energy Synergetics [THORIMS-NES]", which composed of simple power stations and fissile producers, is proposed. This is effective to establish the essential improvement in issues of resources, safety, power-size flexibility, anti-nuclear proliferation and terrorism, radiowaste, economy, etc. securing the simple operation, maintenance, chemical processing, and rational breeding fuel cycle. As examples, 155 MWe fuel self-sustaining power station "FUJI-II", 7 MWe pilot-plant "miniFUJI-II", 1 GeV-300 mA proton Accelerator Molten-Salt Breeder "AMSB", and their combined fuel cycle system are explained.

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