The solvent extraction behavior of about sixty chemical elements was extensively surveyed in the alkylamine-nitric acid system, clarifying radiochemically the acid dependence of distribution ratios. The organic phases used were 10% (V/V) Amberlite LA-1 and 5% (V/V) tri-iso-octylamine xylene solution.
The highly extractable chemical species were not so many in number through two extraction systems. In most cases the acid dependence curves of a given element were not only similar each other between the two systems, but also resembled to that in 100% TBP-HNO₃ system. The exceptional high K_d value of technetium at 1 N HNO₃ suggested an application to separating ^99mTc from neutron irradiated ammonium molybdate.

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Fuel feed cases studied consist of natural uranium or 1.0 _??_ enriched uranium blended with recycle of 90% of the plutonium produced in the reactor. The fuel irradiation control factor, A, which characterizes the modified bidirectional fueling, investigated in this paper were 0, 0.3 and 0.5 for the case of 1.0 _??_ enriched uranium feed. The principal aim of the modified bidirectional fuel management was to improve the peak to average power density ratio over the improvement which could be expected from the recycle of plutonium and the uranium feed of higher enrichment.
The best peak to average power density ratios obtained with the modified bidirectional fuel management were:
1.93 at A=0.5 for natural uranium feed blended with 90% plutonium produced in the reactor;
1.70 at A=0.3 for 1.0 _??_ enriched uranium feed blended with 90% plutonium produced in the reactor.
The best numbers obtained with the ordinary bidirectional fuel management were:
2.94 for natural uranium feed blended with 90% plutonium produced in the reactor;
2.53 for 1.0 _??_ enriched uranium feed blended with 90% plutonium produced in the reactor;
3.14 for 1.0 _??_ enriched uranium feed without plutonium recycle.
Therefore, the modified bidirectional management could achieve substantial improvement in the peak to average power density ratio. However, the average burn-up experienced by the fuel showed a slight decrease with the modified bidirectional fuel management. For example, the average burn-up changes were as follows;
9086 MWD/t of fuel for the ordinary bidirectional fuel management with natural uranium feed blended with 90% plutonium produced in the reactor;
8340 MWD/t of fuel for the modified bidirectional fuel management (A=0.5) with natural uranium feed blended with 90% plutonium produced in the reactor;
15167 MWD/t of fuel for the ordinary bidirectional fuel management with 1.0 _??_ enriched uranium feed blended with 90% plutonium produced in the reactor;
14638 MWD/t of fuel for the modified bidirectional fuel management (A=0.3) with 1.0 _??_ enriched uranium feed blended with 90% plutonium produced in the reactor;
11352 MWD/t of fuel for the ordinary bidirectional fuel management with 1.0 _??_ enriched uranium feed which was not blended with plutonium.

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Separation and decontamination factors were determined for three steps of an extraction, scrubbing and stripping, using the distribution ratios, scrubbing percents and stripping percents in TBP and seven amine systems irradiated to the extent of 10⁸r.
By this method it is easily shown how irradiation of extractant changes the decontamination factor in an unit step. The decontamination efficiency of fissionable material in an extraction process using irradiated extractants may be derived from fundamental data on extraction, scrubbing and stripping in the irradiated extractant system.

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A method of numerical calculation of the γ-dose rate from the radioactive cloud was devised under the supposition that the distribution of the cloud is uniform along the flow direction; the relative error due to the supposition is less than 5%.
This method is capable of acquiring more accurate values than by the other calculations under the assumptions like the circle-like cross-wind section of the cloud and/or the neglecting of the build-up phenomena.
Actual calculations were made graphically to save time and effort efficiently. As the result of this method, the γ-dose rate distributions for the fission products cloud and the ⁴¹A cloud were obtained. This method of estimating γ-dose rate distributions was verified by the experiments using ⁴¹A cloud dispersion (at JAERI, in 1961).

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The steady-state space-energy distribution of thermal neutrons in the two-region system has been investigated in the diffusion approximation.
The theory has been made free from the assumption that the neutron spectrum is a superposition of the infinite medium spectra of both regions. Here, the neutron spectrum in each region is represented by the combination of several components, each having a definite spectrum and an associated diffusion length which are to be determined as an eigenmode and a corresponding eigenvalue respectively. Such expanded spectra in both regions are interconnected on the boundary so as to satisfy the physical conditions, and hence to determine the amplitude of each eigenmode.
The formalism is a genaral one and the extension to the case of more than two regions is possible with due labors. Numerical results have been obtained for a case of spherical two-region system using the heavy-gas model for the energy transfer kernel.

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The kinetics of β-to-α martensitic transformation in uranium alloys containing 0.25_??_1.5a/_??_ Cr have been studied. Upon two alloys of U-0.5_??_ Cr and U-0.75_??_ Cr, the rate curves and the amount of the martensitic transformation under isothermal conditions have been examined in detail by a dilatometric method. According to the results on these two alloys, it is observed that the isothermal completion of the martensitic transformation is inclined to be difficult in the temperature region of the less extent of the undercooling from M_s. However, if the undercooling from M_s is increased, the completion is actually attained when the contraction followed by Johnson-Mehl's rate curve ceases.
In general, the martensitic transformation becomes more sluggish with increasing chromium content and it is not detected by a dilatometer in U-1.5 _??_ Cr alloy. In U-1.0 _??_ Cr alloy, however, the martensitic transformation completes to about 6070% at about 150°C. From the available information obtained on the kinetics of the martensitic transformation, methods for the determination of M_s temperature, especially by means of observing microstructures produced under isothermal conditions, are discussed in detail.

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Particle packing by mechanical vibration was experimentally studied with spherical metal shot and uranium dioxide powder particles for an idealized and actual cases, respectively. The effect of water content on efficiency of packing was studied.
Particles of one, binary and ternary size components in a cylindrical glass container were packed by controlled sinusoidal vibration (10 c10 kc).
With metal shot, densities of 64.0, 86.1 and 92.4% of theorectical were attained from one, binary and ternary components, respectively. With uranium dioxide maximum attained densities were 60.5, 78.9 and 85.0%, respectively.
An economic mixing of two kinds of uranium dioxide, i.e. coarse fused particles (-3+10 mesh) and high fired (Mitsubishi Swageable -50 mesh) UO₂ showed a density of 83.5%, nearly as high as that attained in ternary packings mentioned above.
By adding slight amount of water, packing density decreased abruptly and this effect was more remarkable with packings of finer particles.

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It is well known that an exchangé reaction between metal and its ion in solution occurs at the surface of the metal.
Using this exchange reaction technique, the removal of ⁶⁰Co from an aqueous solution with cobalt metal has been studied. When 2 ml of 0.1 N or 1 N NaNO₃ solution containing trace amounts of ⁶⁰Co was shaked vigorously with 50 mg of cobalt metal (powder) for about 30 min, up to 95100% of ⁶⁰Co was removed from the solution. The higher normality of HCl or HNO₃ solution containing the tracer of ⁶⁰Co was used, the lower rate of removal of ⁶⁰Co from this solution was obtained. This experiment result may be due to the following reason, i.e., a part of cobalt metal was dissolved in acid solution and it resulted in the increase of the content of cobalt ion in this solution.
It was found that up to 95100% of ⁶⁰Co was rapidly removed from 0.5 N NaNO₃ solution containing 10 μg of nickel per ml using this exchange reaction techniques with cobalt metal. In this experiment, however, about 40% of nickel was also deposited on cobalt metal and the separation of ⁶⁰Co from nickel ion was not satisfactory.

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In 1961, the Japan Nuclear Ship Research Association made a contract with the Science and Technical Agency of Japanese Government, on the trial design of nuclear powered oceanographic and supply ship as one of the latter's projects for the peaceful use of nuclear power. The association organized a comittee to supervise the designing in which the author presided as the chairman.
The following presentation is an outline discription of the design which was completed in July, 1962.
The ship has 6, 350 gross tonnage and is equipped with a water cooled, water moderated pressurized water reactor which has 35 MW thermal output and produces 10, 000 s.h.p. The weight of the trial design was not placed in developing a new type of marine reactor nor in detailing about conventional part of a ship's design.
The main object was to study about the feasibility of a comparatively small nuclear powered vessel with proved-type-reactor with satisfactory safety and also to find out where would the problems, technological and legal lie if we were to construct such a ship in a very near future.
We believe that aim of the project has been attained satisfactorily.

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