[en] Uranium dicarbide is metastable at a temperature below 1500⁰C so that it gradually changes into diuranium tricarbide with the liberation of carbon, an undesirable phenomenon from the viewpoint of using uranium dicarbide as a nuclear fuel. The present invention discloses a technique for preparing a stable uranium sulfide (preferably monosulfide) is heated at a temperature above 1000⁰C, particularly in the range of 1500 to 1800⁰C in vacuum or in an inert atmosphere to convert them into a stable solid solution. In one example, a mixture of uranium dicarbide and uranium monosulfide (95 mol : 5 mol) was pulverized, formed into a block, and heated to 1800⁰C for 5 hours to give a solid solution. When heated to 1450⁰C for 5 hours to give a solid solution. When heated to 1450⁰C for 100 hours, the solid solution underwent no change. Uranium carbide itself, when heated to 1450⁰C, formed diuranium tricarbide in one hour or less. (Kaichi, S.)

[en] Since nuclear grade graphite is a porous solid, its oxidation is a good example of a heterogeneous reaction between gases and a porous solid. Changes in properties of graphite and carbon materials caused by corrosion reactions with water vapor or air cannot be evaluated simply on the basis of weight loss only, because the manner in which the corrosion proceeds into the interior of the materials is quite different from one grade to another. In the reaction at higher temperatures, oxygen reacts with carbon at the surface of graphite and the grains are removed gradually, leading to what is called decrease in thickness'. In this case, although specimen or component made of graphite becomes thinner or decreases its dimensions, its properties such as mechanical strength and thermal conductivity are unaltered. On the other hand, at lower temperatures there are little dimensional changes found in graphite, but corrosion reaction proceeds into the interior of the material because of the relatively low corrosion rate at the surface. Besides, the binder region is preferentially corroded so that the binding force between the grains decreases, causing the separation of grains to lead to significant losses of strength and thermal conductivity. For this reason, it is essential to pay attention to the corrosion in the interior of the material as well as temperature and atmosphere, when it is used for structural components. This report summarizes the results obtained in the experiments in which several nuclear-grade graphites are corroded with water vapor or air in the chemical reaction control and in-pore diffusion control regimes. (1) Difference in the corrosion behavior among the graphite grades is clarified and the relationship between the total weight loss and penetration depth was examined by introducing a parameter common to all the graphite grades, characteristic corrosion length, L_B. (2) Measurements of the surface area of corroded specimens led us find the relation between corrosion behavior and the total surface area in the case of air oxidation in the chemical reaction control and in-pore diffusion control regimes. (3) Taking advantage of the information obtained in (1) and (2) above, the corrosion resistance, a disadvantage of graphite materials, was improved by introducing a compositionally gradient SiC/C material using the reaction between silicon monoxide and carbon. (author)

[en] Various nuclear grade graphite materials are used in the core of Magnox-type Reactors, advanced gas-cooled reactors (AGR) and high temperature gas-cooled reactors (HTGR) as core-support structural materials of the fuel systems and moderating materials of fast neutrons. The graphite-moderated power reactor, which was built by the Japan Atomic Power Company had generated electric power (16.6x104 kWe) from 1966. The reactor was shut down in 1998 and its decommissioning program is being planned. Moreover, Japan Atomic Energy Research Institute (JAERI) had been constructing an HTGR called High Temperature engineering Test Reactor (HTTR) and criticality was attained in November 1998. For the nuclear graphite disposal, it is necessary to determine especially the treatment of long-lived nuclides, such as ¹⁴C which are generated in the graphite components during reactor operation. In relation to the planning and implementation of the decommissioning of radioactive graphite materials, it is essentially important to obtain the basic data about pore structure of graphite components. As a research, which solves the problem of the ¹⁴C concentration, the cooperative research contract of 'The research on the separation of ¹⁴C of the irradiation graphite' is concluded between JAERI and Japan Nuclear Power Corp. in 1999, and the research for the basic data acquisition has been advanced up to the present. To clarify the formation process of ¹⁴C, it is necessary to examine the environmental condition of graphite components inside the reactor, especially the role of nitrogen, which is one of the air components. ¹⁴C was formed with the weight loss by the oxidation of the graphite, because core graphite of Tokai power plant was used in the carbon dioxide coolant under radiation. To reduce ¹⁴C concentration in the core graphite, several methods such as the collection as ¹⁴CO₂ by leaching to aqueous solution and low temperature can be considered. To find the optimum conditions for ¹⁴C reduction, basic data on oxidation reaction and the structure of graphite materials indispensable. In the present experiment, we measure the air oxidation characteristics in the temperature range 450-800degC in Quality 1 graphite and IG-110 graphite. Changes in pore diameter and pore size distribution due to air oxidation are discussed. (author)

No abstract available

[en] For the evaluation of oxidation characteristic for the trial production of SiC/C compositionally gradient carbon and SiC/C compositionally gradient carbon fiber reinforced carbon (CFRC) materials, which are composed of a surface SiC coating layer, an intermediate SiC/C layer and substrate material, the oxidation test in air at 800degC for up to 100 h was performed compared with each other three kinds version, i.e., substrate material, one with intermediate SiC/C layer and one coated by SiC layer. It was shown that SiC/C compositionally gradient material exhibited the best oxidation resistance among each substrate series, and the conception of SiC/C compositionally gradient material could be applied to carbon and CFRC materials as well as graphite material. (author)

[en] For the developed compositionally gradient graphite material composed of surface SiC coating layer, middle SiC/C layer and graphite matrix, the thermal cyclic oxidation test was performed together with two kinds of the SiC coated graphite materials in air environment. It was made clear that the developed material exhibited high performance under severe thermal cyclic condition independent of the morphology of middle SiC/C layers and had the longer time or the more cycle margins from crack initiation to failure for surface SiC coating layer compared with the SiC coated graphite materials. (author)

[en] Oxidation reactions of candidate graphites for core support post with atmospheric air were studied in a temperature range between 550₀C and 1000₀C. The reaction rates, temperature dependence of the rates and distribution of bulk density in the oxidized graphites were measured and the characters obtained were compared between the brand of graphites. On the basis of the experimental results, dimension and strength of the post after corrosion with air, which might be introduced in rupture accident of primary coolant tube, were discussed. In the case of IG-11 graphite, it was proved that the strength of post is still sufficient even 100 hours after the beginning of the accident and that, however, it is necessary to insert more deeply the post against graphite blocks. (author)

[en] Effect of irradiation on the changes in properties such as length, volume, bulk density, thermal conductivity, thermal expansivity, electrical resistivity, and Young's modulus were studied for two kinds of nuclear graphites which were thermally oxidized by water vapor prior to irradiation. The samples were irradiated at 800-1020⁰C to the neutron fluence 6.6x10²⁴ n/m² (E > 29 fJ). The experimental results have shown that the irradiation-induced percentage changes of the properties were the same for both thermally oxidized and unoxidized samples. (orig.)

[en] To evaluate the influence of surface SiC coating and intermediate SiC/C layers on the mechanical strength of graphite matrix of the developed compositionally gradient graphite materials, which are composed of a surface SiC coating layer, an intermediate SiC/C layer and a graphite matrix, bending and compressive tests were carried out for SiC/C compositionally gradient graphite materials, SiC coated graphite materials, graphite materials with the intermediate SiC/C and virgin graphite materials at room temperature. The results of the tests were discussed in comparison with the specific surface areas which is a measure for the surface condition of samples. For the graphite with the intermediate SiC/C layer, the strength was independent of the Si content in the intermediate SiC/C layer and almost the same as that for the graphite substrate, although the specific surface area increased with an increase in the Si content in the intermediate SiC/C layer. The strength of the surface SiC coated specimens, i.e., SiC/C compositionally gradient graphite materials and SiC coated graphite materials, was about twice as that of the respective substrate, i.e., the graphite with the intermediate SiC/C layer and the virgin graphite. (author)

[en] The effect of hydrogen on the rate of reaction with water vapor was studied for four kinds of graphites including IG-11 which is a candidate material for use in the core of the Experimental High Temperature Gas-cooled Reactor. The experiment was carried out at 800 - 1000sup( o)C in the helium flow which contained 0.65 - 1.0 % water vapor and 0 - 1.5 % hydrogen. The rate of reaction with water vapor decreased by addition of hydrogen and the effect of hydrogen was much larger for highly purified graphite than for contaminated one. Based on the experimental results, the effect of hydrogen in the reactor condition was estimated and it was shown to be negligible small at such impurity level as the coolant of the reactor. (author)