[en] A rapid and simple process for decontaminating a heavy water reactor and its assemblies with oxalic deuteride is given, comprising three steps of (1) eluting with oxalic deuteride the nuclear fuels, such as Th, U and Pu fission products and other derived radioactive nuclides cohered to the body and assemblies of a heavy water moderated and cooled reactor, (2) decomposing the spent oxalic acid for the elution with radiation and solidifying the dissolved reactor materials, nuclear fuels and radioactive materials, and (3) removing the solids and ions of heavy water together with filtration alone or in combination with ion exchange. In this case, the cohered nuclear fuel of radioative nuclides produce soluble complex salts with oxalic acid. The eluted substances become insoluble when oxalic acid decomposes. The oxalic deuteride may be obtained by dissolving and vacuum-distillating oxalic anhydride. in embodiments, a heavy water solution of oxalic deuteride having a concentration of 0.01M at 10⁰ - 50⁰C was contacted with the surfaces to be treated. Nuclear fuel (U) and radioactive materials were eluted from the surface of reactor materials within 10 hours with a constant concentration. The obtained solution was irradiated with 3 x 10⁷ r of gammaradiation to decompose oxalic acid and to deposit a major portion of the dissolved reactor materials, fuel materials and radioactive materials. The crystalline particle deposit was filtered. (Iwakiri, K.)

No abstract available

[en] Thermal conductivity and crystallite properties of carbon material, which was developed for use as a core-support material in VHTR, were measured after heat-treatments at given temperatures up to 3040⁰C. The thermal conductivity was constant up to 1200⁰C and then it increased with increase of heat-treatment temperatures. Relationships between apparent crystallite size L sub(c(002)), d sub(0)-spacing d sub((002)) and the thermal conductivity were obtained and analyzed for heat-treated samples. The linear relationship between the thermal conductivity and L sub(c(002))sup(2)/d sub((002)) value was obtained, and it was clarified that the change in two-dimensional crystallite structure should be introduced in order to explain the change in thermal conductivity of carbon material in the process of graphitization. (author)

[en] The changes in dimension and bulk density of carbon blocks for core support in VHTR were measured after heat treatment at high temperatures, and also the dimensional changes with time at 900⁰C and 1000⁰C were measured with and without application of compressive stress, for three domestic carbon materials. By heat-treatment at temperatures up to 3040⁰C, two carbons shrinked and one did not, and the bulk density of two carbons increased one decreased. Dimensional changes with time at 1000⁰C were observed with and without compressive stress 284 kg/cm². The results differed with heat treatment above the baking temperatures. The need was indicated for the development of carbon material with dimensional stability at 1000⁰C for VHTR. (author)

[en] A review of research on graphites for High-Temperature Gas-Cooled Reactor in JAERI is given. At first such problems as (1) dimensional change by irradiation, (2) thermal conductivity and coefficient of thermal expansion, (3) mechanical properties, and (4) high temperature corrosion, were pointed out and discussed associated with the core structural point of view. Secondly, recent results of graphites in JAERI are reviewed corresponding to the above problems and also future programs are described. (auth.)

[en] This paper describes the experimental results and discussions on the changes in dimension, volume and thermal expansion coefficient of five nuclear grade graphites which were made from petroleum coke or gilsonite coke and irradiated in the JMTR at 670 -- 1,260⁰C up to the maximum neutron fluence 3.3 x 10²¹ n/cm² (E > 0.18 MeV). Dimensional or volume shrinkages for all samples were observed after irradiation and their changes depended on the kind of graphite studied. The changes in thermal expansion coefficient were also observed for three of five samples. The relations between the changes in dimension or volume and thermal expansion coefficient were obtained, and it was clarified that both the rates of irradiation shrinkage in dimension and volume of the sample having larger thermal expansion coefficient were smaller than for the sample having smaller thermal expansion coefficient. (author)

[en] In a multi-purpose high temperature gas-cooled reactor, a heat exchanger, fuel elements and heat-insulating material are in severe heat transfer condition due to the circulating helium gas of about 1,000⁰C and pressure of 40 kg/cm². It is thus necessary to acquire ample experimental data on such components. Studies made so far on the fuel elements and graphite in Japan Atomic Energy Research Institute and First Atomic Power Industry Group are described: fuel and its production test, out-of-pile test, irradiation test and FP release; graphite and high temperature irradiation test and post-irradiation test, mechanical strength, high temperature corrosion and selection of graphite. (Mori, K.)

No abstract available

[en] The statistical distribution of the tensile and bending strengths has been studied on the gilsocarbon isotropic graphite (IM2-24) and the petroleum coke fine-grained isotropic graphite (7477PT) and the irradiation effect on the statistical nature of strength has also been studied on the former graphite. As a result, the data of tensile and bending strengths can be well approximated by both the Weibull and the normal distributions. Both the Weibull modulus, m and the normalizing factor, σ₀ in the Weibull statistics of strengths of irradiated graphite increased as compared with those for unirradiated one. The effect of specimen volume on strengths was not large as predicted by the Weibull theory. The experimental value of the bending to tensile strength ratio was larger than the calculated one based on the Weibull statistics. However, results based on the pseud-Weibull statistics, in which only the m value in the Weibull parameters m and σ₀ is taken as the same value in the tensile and bending strengths, was comparatively consistent with the experimental value for 7477PT graphite. In conclusion, the weakest link model is considered to be able to apply approximately but not exactly to the fracture of graphite. (auth.)

[en] A new method is presented of measuring porosity and pore structure of nuclear graphites. Argon content in the pores of graphite is measured by activation analysis using the nuclear reaction ⁴⁰Ar(n, γ) ⁴¹Ar, while composition of the gas released from it is revealed by mass-spectrometry. The porosities of SM1-24 and 7477PT graphites were calculated from the argon content and the volume ratio of argon to total gas released. For SM1-24 containing large quantity of argon, the method is effective for porosity measurement. Argon content in 7477PT is too small for the measurement. The effect of heat treatment in a vacuum on argon content of the graphites was studied in relation of the proportion of open pore volume to total pore volume. Comparison of the results by activation analysis with those by mercury porosimetry for SM1-24 graphite showed about 55% open and 30% closed pores of the total porosity. The rest 15% is possibly crystalline distortions and defects, for which none of the methods tested are effective. (auth.)