Research and development for uranium enrichment by oxidation-reduction chromatography, begun in 1972 despite the prevailing view that a chemical process would require several centuries of operation to attain fuel-grade uranium enrichment and would thus be impractical, led to the establishment of basic process technology permitting attainment of 3% enrichment within several months of operation by the mid 1980, through advance which brought increased electron exchange and oxidation-reduction reaction rates, effective uranium adsorption band formation and maintenance, and equilibrium plate height reductions based on elucidation of mobile-phase dispersion. Theoretical descriptions of the enrichment factor and the isotopic abundance curve of the adsorption band were developed, and the theoretical and experimental development and verification of an intracolumn mechanism of redox agent self-regeneration led to a new "Super" process version characterized by greater simplicity and efficiency than previously thought possible. A pilot plant with enrichment columns of 1m diameter and 3m height, constructed at Hyuga City in Miyazaki Prefecture, demonstrated the recovery of 3% enriched uranium in April 1988. Subsequent development work has brought further advances in process economy and continuous-operation performance.

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Combination of ultra-low-q discharge with adiabatic compression expands the parametric window of Ohmic ignition significantly. The device of this scheme is, in essence, compact. Ohmic heating is enhanced by adiabatic compression because of further intensification of high current density which is specific to the ultra-low-q discharge. Typical example of device and plasma parameters with compression ratio 5 and safety factor 0.5 is as follows; major radius/minor radius=2m/0.5m, energy confinement time required for Ohmic ignition is less than 0.2s, plasma current density is 50MA/m², and the total plasma current is 1.6MA. Without compression longer energy confinement time and larger total plasma current are necessary. Free expansion of the current channel and/or increase of toroidal field induces the transition of ultra-low-q discharge to tokamak.

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Uranium dioxide fuels irradiated in a commercial BWR during 1 to 4 cycles (burn-up: 628GWd/t) have been annealed in an out-of-pile condition and the release rate of ⁸⁵Kr was measured continuously. The annealing temperature ranged 1, 5001, 800°C. The effects of temperature and burn-up have been studied on burst release during temperature ramp and on diffusional release during isothermal annealing.
The burst release evolved on temperature ramp was mainly due to the rapid release of fission gas from grain boundaries, which had previously accumulated there during base irradiation. With increasing burn-up, the critical temperature for the onset of the burst release became lower and the fractional burst release increased. The number of fission gas atoms per unit area of grain boundary, which was calculated from the quantities of the burst released ⁸⁵Kr, was about 24 times larger than saturation values estimated from theoretical considerations based on the assumption of mechanically equilibrated bubbles for 3 and 4 cycle irradiated fuels. Effective diffusion coefficient of ⁸⁵Kr obtained by the isothermal annealing tests appeared to be independent of burn-up (<30GWd/t).

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Fission gas release in the bump tests was correlated to the deformations of claddings via mathematical product of the number of gas atoms and their residing time on grain boundaries. A positive correlation of the deformations with the product indicated that gas bubble swelling of pellets contributed to the pellet-cladding mechanical interaction (PCMI).
Residual gaps prior to the bump tests turned out to be filled in at the bump terminal level by differential thermal expansion of pellet and cladding. Visible macro cracks existing in the central part of the pellet virtually vanished during the tests due to bubble swelling of the pellet. Together with these observations, quantitative image analysis of pellet porosity showed that the aforementioned PCMI was brought about by combination of differential thermal expansion and bubble swelling.
A model to unify gas release, bubble growth and PCMI simulated well the observed behaviors of fission gas release from bump-tested rods. It was deduced by the model that higher burnup retains a higher potential for PCMI, while power reductions and associated gas releases reduce PCMI.
In the analysis in this paper were used the data of the Risφ Transient Fission Gas Release Project.

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Polyisoprene was irradiated in hydrogen atmosphere by ⁶⁰Co γ-rays. The evolved gases were methane, ethane, ethylene, propane, propylene and butane. With respect to saturated hydrocarbons, the yields increased with increasing pressure of hydrogen. On the contrary, the yields of unsaturated hydrocarbons decreased. The G-value of methane in hydrogen of 1MPa was about four times larger than that in vacuum for irradiation of cis-polyisoprene (Cis) at 160kGy. The gas evolution from trans-isomer (Trans) was about the same as that from Cis. The yields of gases from gloves made of natural rubber were smaller than those from Cis or Trans.

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A leaching model for a radioactive pellet package for nuclear power plant wastes was studied. Calculated results obtained in term of the model were compared with experimental ones, where volumetric averages of diffusion coefficients in the Na₂SO₄ pellet and in the cement-glass were used to solve the diffusion equation in the pellet package. Since Na₂SO₄ pellets dissolved quickly, the pellet part were simulated by saturated Na₂SO₄ solution. The leaching ratios were meaured with different volume ratios of cement-glass in the package to discuss the validity of the model. The cement glass diffusion coefficient and distribution factor, which were required to solve the equation, were also measured. The calculated results agreed with the experimental ones within a 20% error.

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A numerical technique based on the particle-tracking method was developed to simulate radionuclide transport in fractured porous media. Considered transport mechanisms are decay chain, advection in fractures, diffusion into rock matrix from fractures and instantaneous sorption. The proposed method was applied to some test problems and good agreements with analytical solutions and results by the finite element method were obtained. Major advantages of the proposed method are its capability to treat arbitrary number of components in a chain and to consider the effect of matrix diffusion on radionuclide transport in fractures.

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The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas.

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Comprehensive analysis on core support barrel (CSB) movements in the ULJ1N Nuclear Plant is performed through noise and structural analysis techniques. Noise signals are taken from the lower channel outputs of ex-core instrumentation system during the full power reactor operation period. Then they are converted into auto-power spectral densities (APSDs) and coherence functions in the frequency range of 050Hz to obtain the vibratory information of CSB movements.
From APSDs, the three different vibration peaks of CSB are detected around the frequencies of 8, 15 and 20Hz, distinctly. These results are also agreed well with those obtained from the structural analysis by ANSYS version 4.3 computer program, which is the finite element method (FEM). Three different vibration mechanisms of CSB at each resonant peaks are identified as two types of the beam mode vibrations (vig., pendulum motion and torsional motion) and the shell mode vibration, respectively.

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