Theoretical calculations of pressure transient and bubble growth were performed for analyzing the experiment on incipient boiling of superheated potassium. The non-equilibrium bubble growth theory with use of evaporation coefficient C and the assumption of quadratic temperature distribution near the interface was combined with the inertial constraint of the finite liquid system which consists of the spherical motion around the bubble and linear motion of upper column.
The combined theory could successfully explain the experimental result that the bubble pressure rises toward the saturation pressure corresponding to the initial bulk temperature after attaining a minimum. A fairly good agreement with the experiment was obtained for a value of C=0.0050.01 on the transients of pressure as well as bubble radius. Half recovery times of the bubble pressure measured also agreed well with the calculation for C=0.005.

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The over-all core hydrodynamics during the reflood phase is quantitatively discussed. The following model was proposed for the core hydrodynamics : (1) When the liquid mass flux and the steam mass flux exist at the quench front, dispersed flow appears. (2) The void fraction downstream of the quench front gradually decreases with time. (3) Finally, the void fraction approaches the value predicted by the modified Cunningham & Yeh and the modified Lockhart-Martinelli correlations within about ±30% error. (4) Even after the concerning region is quenched, the void fraction can be predicted with the same correla-tions. Both correlations have been recently developed.

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Wide range of parameter surveys are made on the DT fusion tokamak experimental reactor next to JT-60. Various physics and engineering requirements are taken into account, e.g. self-ignition, available maximum toroidal β value, α-particle confinement, total fusion power, neutron wall loading, heat flux to divertor plate, structural restriction on major radius, device size, maximum toroidal magnetic field, poloidal field power supply and so on. Theoretical scaling law for the available maximum toroidal β value deter-mined by ballooning mode instability is used. The toroidal magnetic field on plasma axis can be expressed by the aspect ratio A for a given maximum field at the toroidal field coil conductor. Empirical scaling law for the electron energy confinement and neoclassical heat conductivity for the ion are employed. These confinement times can be expressed by the plasma minor radius a and A through the maximum available Β value and the toroidal field on axis. In the similar way, most of the physics and engineering requirements can be mapped on the α-A diagram. This diagram enables us to make systematic and wide range of parameter surveys of the device. In particular, this offers a clear perspective on the device parameters, which can mitigate the engineering difficulties and can also realize the required plasma performances.

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Synthetic conditions of PuN from carbothermic reduction of PuO₂ has been studied in a mixed 8%H₂+92%N₂ stream at a temperature range of 1, 2701, 680°C. In the course of both reactions of the carbothermic synthesis of PuN from PuO₂ and the hydrogenation of C, the vaporization loss of Pu was observed. It increased with temperature in the tem-perature range of 1, 3501, 450°C, and reached to a constant value 1.3% of total Pu in the temperature range of 1, 4501, 680°C, at which PuN was synthesized at a reaction rate of high enough. The minimum mixing ratio (C/PuO₂, mole ratio) for the formation of high purity PuN depends on temperature. The value is 2.15 for 1, 620°C and 2.35 for 1, 680°C. The oxygen and carbon impurities in the PuN obtained were found to be 0.0950.028 and 0.174.012 W/O, respectively.

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Void swelling in 0.045%Ti-modified 316 stainless steel was investigated by 200-keV C⁺ ion bombardment. Three metallurgical conditions, i.e. solution treated, 20% cold-worked and 20% cold-worked+ aged at 898 K for 1, 000 h, were compared. Solution treated materials showed considerably low swelling, while the other two conditioned materials showed much higher swelling. The swelling for 20% cold-worked+ aged materials was the highest. In solution treated condition, a number of fine-scale TiC precipitates appeared uniformly in the irradiated matrix. The precipitates and also solute Ti and C are considered to sup- press void nucleation by vacancy trapping mechanism. In 20% cold-worked condition, high density dislocation networks were observed being heavily decorated dy TiC precipitates, leading to void swelling increase as a consequence of reduced dislocation sink strength and of depletion of solute Ti and C in matrix. In 20% cold-worked+ aged condition, as a result of solute depletion by precipitation of various type precipitates and partial recovery of cold-worked dislocation structure, an enhancement of void swelling occurred. The 0.045% Ti-modified 316 steel was found to be more swelling resistant than type 316 steel.

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The collective dose due to the background radiation outdoors in Japan was analyzed, and the influence of the time trend of the population movement on the per capita dose was estimated. Only the external dose was considered. It was previously shown that the per capita dose calculated for each year between 1952 and 1980 by the use of the surveyed background levels along with the registered population data has been decreasing continuously. In this work, the latest data in 1981 was added, and male and female statistics were pro-cessed separately. The result showed that the male values of the per capita dose have been always lower than the female values, if the deficit is small. It was also shown pre-viously that the prevailing population movement from countryside toward urban areas (often plains of volcanic ash) would be the cause which have changed the per capita dose, i.e. population-weighted mean dose of the whole country. In this work, it was anticipated that the gross population movement from the South-West toward the North-East Japan would be another cause of the decrease.

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A study is made of the nuclear characteristics of blanket/shield design intended for the D-D tokamak reactor incorporating a cooperative effort by the University of Illinois, Brookhaven National Laboratory and Lawrence Livermore Laboratory. The reactor is characterized by high value of plasma beta (β =0.3) and low value of neutron wall loading (W_n= 0.436 MW/m²). The 1 m-thick blanket is composed mainly of graphite, and the 1 m-thick shield is a combination of B4C, Al and H₂O. Multi-dimensional calculations are carried out to confirm the results of one-dimensional calculation and to assess the problems inherent in the design.
Compared to blankets constituted of other materials, the graphite blanket possesses the characteristics of much smaller residual radioactivity, afterheat and biological hazard poten-tial, which could possibly more than offset the lower nuclear heating that can be provided. The design requirements for the magnet shield are satisfied by the present combination and thicknesses of blanket/shield materials.

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