Physical feasibility of plasma burning control in tokamak reactors by variable toroidal field ripple is shown by using one-dimensional tokamak transport code. Excessive temperature rise can well be suppressed by strengthening the ripple up to about 1.5% with moderate time constant and stable burn is shown to be achievable. These calculations include the noncircular effect of plasma shape and the effect of poloidal angle dependence of ripple structure. Engineering feasibility is also shown by simple estimate.

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The aim of this study is to determine the void generation noise and verify Nomura's assumption (whiteness of the spectrum) in a BWR by means of in-core channel flow-meters. The difference in signals between the outlet and the inlet flow-meters gives information on the void generation noise in the channel. The void generation noise in the Japan Power Demonstration Reactor (JPDR)-IL was determined by the use of turbine flow-meters mounted in the instrumented fuel assembly (IFA). The auto-power spectral density (APSD) of the void generation noise was estimated through multivariate autoregressive (MAR) model fitting by separating the void generation noise from the contributions of the pressure, the inlet temperature and the inlet flow fluctuations. The results gave experimental evidence for the presence of Nomura's void noise. The results also showed that the fluctuations of coolant expansion rate in the channel are affected by the fluctuations of inlet flow rate as well as by Nomura's void noise, which suggests the role of boiling boundary fluctuations as a noise source of BWR neutron fluctuations.

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Temperature coefficients of reactivity have been measured up to 600°C on cluster-type UO₂ fuel for three kinds of ²³⁵U enrichment and on a hollow cluster of sus-cladding tubes by using a hot He gas loop in a heavy-water-moderated, pressure-tube-type critical assembly. A new experimental method has been developed which accurately eliminates the reactivity disturbance caused by heat leakage in the measurement of an extremely small change in reactivity. The fuel (fuel pellet, cladding and pressure-tube) temperature coefficients of reactivity obtained for the temperature range below 300°C are +1.00±0.04, -3.48±0.13 and -6.36±0.25 in the unit of 10^-5%Δk/k•°C for 0.2%, 0.7% and 1.5% ²³⁵U enrichment, respectively. In the higher temperature region above 300°C, each coefficient shifts to positive side by about 2×10^-5%Δk/k•°C. Temperature coefficient of reactivity for the hollow cluster of sus-cladding tubes (cladding and pressure-tube) has a large constant value with positive sign, + (6.42±0.26) × 10^-5%Δk/k•°C, all through the temperature range. A calculational model to analyze a hot-loop-type measurement of temperature coefficients with use of WIMS-D code was proposed and could be successfully applied to the present measurement.

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Physical models of the EULFCI code are presented concerning fuel-coolant interaction (FCI) related materials movements in a fuel pin cavity and coolant channel.
For the in-cavity dynamics, a model of axially non-uniform mixture between molten fuel and FP gas is employed to take into account of melting of fuel pin cavity wall. In the coolant channel, the ejected fuel is treated as multi-sized particles whose movement are represented by master particles. Though parametric, a fuel fragmentation model is combined with the fuel particle dynamic model. For the coolant dynamics, the cross-sectional changes along the axially direction can be treated, and the marker particle method is applied to trace the interaction (FCI) region boundaries.
Calculated results were compared with the results of the TREAT H2 experiment using the multi-sized particle model. The effects of fuel fragmentation time-constant and sodium condensation were parametrically discussed. It is concluded from the results that the EULFCI code could explain the TREAT H2 experiment in the physically reasonable range of parameters.

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The initial burst and the diffusion of fission produced Xe were investigated on UC₂ irradiated to thermal neutron doses between 2.1×10¹⁴ and 1.3×10¹⁶ fission/cm³. In heating runs with 5 and 10°C/min and below 1, 000°C, the burst revealed three steps at around 200, 400 and 600°C (in 5°C/min heating run), to which activation energies were obtained. It was found that the temperatures and the activation energies for each burst were close to those appearing in recovery processes of fission induced defects. In diffusion process above 1, 000°C, on the other hand, the activation energy of Xe in UC₂ increased with increasing fission dose. In the specimens once heated up to 1, 500°C, however, the diffusion coefficient could be expressed as, D ?? 5×10^-1 exp (-344±50 kJ/RT) (cm²/s), which was independent of the level of the fission dose. An enhancement and a suppression of the diffusion were suggested also in the dicarbide as was done previously for the monocarbide of uranium.

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Stability of high-level liquid waste (HLW) from nuclear fuel reprocessing was studied by using a simulated HLW. Fundamental works disclosed that precipitates formed during aging at ambient temperature or refiuxing the simulated HLW in 2 mol/l HNO₃ solution consist mainly of Mo, Zr and Te contributing significantly to the formation of precipitate. When the simulated HLW was denitrated with formic acid or deacidified with NaOH, fractions of precipitated Mo, Zr and Te increased with pH and amounted to over 85% at pH 0.5, where the fraction of precipitated La was below 0.1%. For further treatment of HLW such as partitioning, denitration of HLW to pH 0.5 might be useful for removing Mo, Zr and Te from the solution without significant contamination with rare earths, Am and Cm.

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