The anisotropic diffusion coefficient has been calculated in a cylindrical cell with use made of the integral transport theory. The previous method⁽¹⁾ of calculating the diffusion coefficient requires much computer time to evaluate the generalized first-flight collision probabilities between two mesh points for a square cell. To circumvent this drawback, we introduce new calculation methods for determining the anisotropic diffusion coefficient in a cylindrical cell. Two methods are proposed-one uses the diffusion theory in the outermost moderator region of a cell; the other adopts the integral transport theory in that region as well. For the latter method, a new boundary condition is introduced for the cell surface, which, in the present problem, supersedes the usual isotropic return boundary condition. Using these two methods, the anisotropic diffusion coefficient can be evaluated with very short computer time. Moreover, an analytic expression is obtained for the special case where a cell is composed of a fuel and a moderator.

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The analytical solution of the isotropic, one-dimensional and time-dependent transport equation is expressed by the superposition of a newly defined Green function of the virgin neutron flow. As a result, the solution of the neutron transport equation is given by the superposition of beams possessing wave fronts. The present study has introduced the possibility of clearly explaining neutron transport behavior by beam processes, thus providing a further insight into the interpretation of the transport equation. This method should prove useful when the continuous mode is superior to discrete modes. The solution for the stationary case is also derived.

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Isotopically pure ²³³U samples, with only 3×10^-3ppm ²³²U content, were prepared by thermal neutron irradiation of thoria and subsequent chemical processing. The ²³³U sample thus obtained was reirradiated with a fission neutron spectrum in the core of the Kyoto University Reactor (KUR), and measurements were made of the fission spectrum average cross section for the ²³³U(n, 2n)²³²U reaction. A value of 4.08±0.30mb was obtained for this cross section, in agreement with the renormalized value of Halperin et al. within the limits of experimental error.
In order to assess the energy dependent cross section from the value of this integral measurement, the ²³³U(n, 2n) cross section was calculated assuming a Maxwellian-type fission spectrum and adopting the energy dependent evaluated cross sections of ENDF/B-III and other authors. The values of the cross section thus determined were found to be about 32 to 91% larger than the measured cross section given above. The result of Pearlstein's calculation of the ²³³U(n, 2n) cross section by the statistical model, again assuming the Maxwellian distribution, is smaller than the measured cross section by about 19%.

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The analytical treatment of a distributed reactor having spatially dependent parameters is presented. Both the modal expansion method and the function space method are applied successfully to obtain the optimal feedback system for the terminal cost problem. The explicitly posed control obtained by means of the function space method is particularly convenient for machine computation. A numerical example is presented to show the characteristics of the synthesized optimal feedback system.

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A new method is presented for synthesizing strontium titanate. The reaction between strontium nitrate and meta-titanic acid or titanium dioxide (anatase or rutile) has been studied with use made of thermo-gravimetry and differential thermal analysis, and the compound produced was identified by X ray diffractometry. A mixture of strontium nitrate and meta-titanic acid is converted to strontium titanate at about 600°C by an apparently endothermic reaction. This process can be applied advantageously to the preparation of sources of radiation and heat that utilize radioactive strontium obtained from nuclear fission products.

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Selective removal of uranium from (Th/U)O₂, by fluorination with fluorine was studied experimentally. The fluorination was performed both in a small boat and in a 2 inch inner diameter fluid-bed reactor.
Fuel particles tend to agglomerate in the reactor due to the large amount of reaction heat and the comparatively low melting point of ThF₄. The fluorinated fuels produced in the fluid-bed reactor were found to be partially agglomerated. Fractional retention of uranium was smaller in the agglomerated parts than in the un-agglomerated, and smaller in the outer layers of the cakes than in the core.
On the other hand, it was also established beyond doubt from the results of the small boat fluorination experiment that heavy agglomeration inhibits the volatilization of uranium in the form of UF₆. Inhibition of the violet exothermic reaction by lowering the fluorine pressure in the early stage of fluorination was found to be a very effective method of obtaining high uranium recovery. It was demonstrated that more than 99% of the uranium could be volatilized within 4 to 5hr at a temperature of 580°C.
The experimental results on the effects of temperature, particle size and fluorine concentration are presented. The variations of reaction rate observed in the course of fluorination are also discussed.

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A procedure for separating ²³⁸Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO₃ solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 10² for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5mg of ²³⁸Pu was obtained with an efficiency exceeding 95%.

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