The KIWI-TNT experiment was analyzed with a new two-dimensional excursion code MARS to ascertain the validity of the method used in calculating the energy release in the hypothetical accident analysis on fast reactors. In calculating the total energy release, adopting the nominal temperature coefficient, the present calculation was found to underestimate the total energy release by 9 % in reference to the experimental value of 8.8× 10⁹ J. The calculated power pulse width was 1.1 msec, which is about half that was observed in the experiment. Adopting a temperature coefficient reduced to 0.5× 10^-5Δk/Δθ, the calculated energy release agrees with that of the experimental value, but then the discrepancy in power pulse width became the largest.
Assumption that the expansion occurs in the axial direction only approaches the calculated energy release quite close to the values from experiment, although it still leaves a considerable disagreement in power pulse width.

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Neutron transmission measurements on ⁵⁹Co have been made for neutron energies from 0.8 eV to 3 keV using the time-of-flight spectrometer of a linear accelerator at the Japan Atomic Energy Research Institute. The maximum time resolution was 10 nsec/m. Resonance parameters of the 132-eV resonance of ⁵⁹Co were determined from the neutron transmissions of several samples of different thicknesses by the thick-thin method, using an area analysis method based on the Breit-Wigner single-level formula.
The following results were obtained, E₀= 132.0± 0.5 eV, Γ = 6.0± 0.2 eV, Γ n•5.15± 0.06eVand J=4. The value of the potential scattering radius, obtained from the transmission of the thickest sample, was R'= 5.3± 0.5 fm.

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It has been considered so far that oxygen retained in UC is intrinsic to the carbothermic reduction of UO₂, so that some have regarded this reaction as inherently unsuitable for the preparation of pure UC powder, unless the release of CO gas can be kept within prescribed limits during irradiation, Nevertheless, quantitative considerations on the attainable lower limits of oxygen content in the UC have so far been scarcely reported.
The author has estimated the partial pressure of CO gas in the carbothermic reduction and the solubility of oxygen in UC from thermodynamic considerations.
It has been established as a result that the reaction has a large temperature dependency, and that the reduction is easily promoted under conditions of high temperature and low pressure, only 0.00456 ppm of oxygen being soluble in UC at 1, 550°C under 1 × 10^-5 mmHg.

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A new formalism based on the direct operational method** has been developed for numeri-cally calculating the fast neutron spectrum in an infinite homogeneous medium. By this method the fine structures of nuclear data such as the anisotropy of scattering cross sections in the high energy region can be easily taken into account. Application of this method to light water sys-tem has yielded the conclusion that: (1) the effect of anisotropic scattering in the center-of-mass system becomes significant near the source energy and near scattering resonance peaks, (2) the errors introduced by the G.G.-approximation are also significant under similar circumstances and (3) the contribution from inelastic scattering is negligible (>0.2 %) in the case of fission neu-tron source in light water system.

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The previous paper(1) dealt with the problem of fully developed heat transfer in a turbulent flow with uniform internal heat generation in a concentric annulus under conditions of uniform but different heat fluxes at the two wall surfaces. This sequel presents analyses of the heat transfer problem in the thermal entrance region.
First, a graphical representation is given of the variations of Nusselt number at each wall surface with distance along the passage. Second, these fundamental solutions for uniform inter-nal heat generation and uniform wall heat fluxes are applied to the case of arbitrary variations of internal heat generation and wall heat flux along the axial direction.

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The exposure rates of γ-rays from ⁴¹Ar clouds issuing from the stack of a nuclear reactor were calculated on the basis of Pasquill's cloud diffusion formula, using exact γ-ray attenuation and build-up formulas. The results have agreed well with the experimental values of a ⁴¹Ar release test performed in 1961 at the Tokai Research Establishment, Japan Atomic Energy Research Institute.
The cumulative dose of γ-rays received at a given point from shifting ⁴¹Ar cloud for a long period was also calculated by considerations based on the probability concept. The validity of the method has been confirmed by comparison between the value estimated with statistics of meteorological data and the cumulative value of observed γ-exposure. By this method, the maximum allowable release rate was obtained for the Tokai area.

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With the view to gaining basic information to serve in reactor design and to permit better evaluation of fuel properties, the temperature distribution in graphite matrix hollow fuels containing Th is discussed on the basis of information obtained from equations expressing the temperature as a function of power density, fuel properties and cooling conditions.
The relations found as a result among these factors indicate that high power density can be expected by adopting a combination of high thermal conductivity (>0.06 cal/cm•sec•°C) optimized ratio between inner and outer diameters of hollow fuel (about 1:3), and high heat transfer coefficient of gap between fuel and sheath (by reducing the gap width to less than 0.2 mm). The temperature rise in the blanket is found to be insignificant.

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