In a heavy water reactor a large amount of heavy water is degraded during operation and cannot be used without reconcentration.
In order to reconcentrate the degraded heavy water, we have developed a new plant and constructed an installation having a capacity of 5t./yr.
The plant is now working in good condition. A bigger plant with a capacity of 200 t./yr is being designed by Showa Denko Co., Ltd.

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Gamma scintillation spectrometry is known as a rapid and nondestructive method for determination of ²³⁵U enrichment. The method of the comparison of the ratio of the relative intensities of 184keV from ²³⁵U to 90 keV from ²³⁸U daughters is more accurate and sensitive for low enrichment than the method of the intensity of 184keV alone.
In this paper, using the low enriched uranium dioxide powder, the relationship between the enrichment and the peak ratio, which includes the correction of the experimental condition, was obtained by considering the uranium disintegration schemes and the data of the two standard samples.
Enrichments of 2.5% and 1.5% enriched uranium samples were determined by this relation. They show good agreements with the values informed by U.S. Atomic Energy Comission, supplier of these materials, within the error of ±1.2%.

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Chemical examinations were being made on the natural uranium UO₂ pellets which were irradiated in GETR for one and two months, respectively. The pellets were contained in a stainless steel inner capsule and an aluminum outer capsule, the activities when they arrived at us were 40 and 120 c, respectively. The determination of the release rate of gaseous fission products (⁸⁵Kr) and burnup calculation were performed at these samples.
The release rate of ⁸⁵Kr were determined by comparing the activities of fission gases which were trapped when the capsules were drilled in a vacuum container with those which were trapped by melting a piece of the pellets. The rate were found to be 0.10.13% which means that gaseous fission products were very well retained in these UO₂ pellets.
The burnup values were determined by using ¹³⁷Cs as an indicator and were 0.06 and 0.13%, respectively. These data agreed very well with calculated values from the activities of Al-Co monitoring wires which were irradiated attached to the capsules. The chemical analysis of various parts of pellets were made and no relocation of ¹³⁷Cs was noticiable.

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Many studies have been reported on diffusion coefficients of Kr and Xe in irradiated U, but the values of coefficient reported have not agreed well.
In the present report, these coefficients of diffusion are considered, using the experimental results obtained and the diffusion equations, and the factors influencing the coefficients are discussed. When the sample was heated under inert gas atmosphere, the appearent diffusion coefficient increased. For highly irradiated samples, diffusion coefficient was smaller than less irradiated samples. When the sample was heated in inert gas atmosphere, the release of gases was suddenly increased at the transformation point of samples, and the gas release by thermal cycle was also increased. It is very probable that the gas bubbles migrate as a result of the surface diffusion of the matrix atoms.

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