Abstract: Provided is a method of treating radioactive liquid waste which reduces the amount of radioactive waste to be generated and is capable of removing a radioactive nuclide from radioactive liquid waste to the extent that the concentration thereof is less than or equal to the measurement lower limit using a simple apparatus configuration. A filtration device is connected to a colloid removal device by a connection pipe. An adsorption tower positioned at the highest stream of an adsorption device is connected to the colloid removal device by a connection pipe. The colloid removal device includes an electrostatic filter. Respective adsorption towers in the adsorption device are sequentially connected by a pipe. A discharge pipe is connected to the adsorption tower positioned at the lowest stream of the adsorption device. Radioactive liquid waste, containing particles having a particle diameter of 1 ?m or greater, negatively charged colloids, and a radioactive nuclide, is supplied to the filtration device.

Abstract: Provided is a method of treating radioactive liquid waste which reduces the amount of radioactive waste to be generated and is capable of removing a radioactive nuclide from radioactive liquid waste to the extent that the concentration thereof is less than or equal to the measurement lower limit using a simple apparatus configuration. A filtration device is connected to a colloid removal device by a connection pipe. An adsorption tower positioned at the highest stream of an adsorption device is connected to the colloid removal device by a connection pipe. The colloid removal device includes an electrostatic filter. Respective adsorption towers in the adsorption device are sequentially connected by a pipe. A discharge pipe is connected to the adsorption tower positioned at the lowest stream of the adsorption device. Radioactive liquid waste, containing particles having a particle diameter of 1 ?m or greater, negatively charged colloids, and a radioactive nuclide, is supplied to the filtration device.

Abstract: Disclosed is a method for treating a radioactive organic waste, the radioactive organic waste including a cation exchange resin adsorbing radionuclide ions, the method including the step of bringing the radioactive organic waste into contact with an organic acid salt aqueous solution containing an organic acid salt and whereby desorbing the radionuclide ions from the cation exchange resin, in which the organic acid salt contained in the organic acid salt aqueous solution includes a cation that is more readily adsorbable by the cation exchange resin than hydrogen ion is. This enables reduction in concentration of a radioactive substance in the radioactive organic waste and reduction in amount of a high-dose radioactive waste.

Abstract: Disclosed is a method for treating a radioactive organic waste, the radioactive organic waste including a cation exchange resin adsorbing radionuclide ions, the method including the step of bringing the radioactive organic waste into contact with an organic acid salt aqueous solution containing an organic acid salt and whereby desorbing the radionuclide ions from the cation exchange resin, in which the organic acid salt contained in the organic acid salt aqueous solution includes a cation that is more readily adsorbable by the cation exchange resin than hydrogen ion is. This enables reduction in concentration of a radioactive substance in the radioactive organic waste and reduction in amount of a high-dose radioactive waste.

Abstract: A method and system of refining natural gas that improves the quality of liquefied natural gas and enables separation and recovery of hydrocarbons other than methane. The method of refining natural gas containing methane; any other hydrocarbon selected from the group consisting of ethane, ethylene, propane, propylene, n-butane, isobutane, 1-butene, n-pentane, and isopentane; carbon dioxide; and hydrogen sulfide, includes adjusting a pressure and temperature of the natural gas so that the methane is in the gas phase, the other hydrocarbon in the liquid phase, and the carbon dioxide and the hydrogen sulfide in the solid phase, respectively; separating the natural gas, of which the pressure and temperature has been adjusted, into a gas containing the methane and a suspension liquid; and separating the separated suspension liquid into a liquid containing the other hydrocarbon and a solid containing the carbon dioxide and the hydrogen sulfide.

Abstract: Most part of an amount of uranium contained in the spent nuclear fuel is removed by making fluorine or a fluorochemical act on the spent nuclear fuel to convert the uranium into UF6, and the uranium is purified through a simple method of distilling the UF6 together with a absorbent. After removing the most part of the amount of uranium, the remaining nuclear fuel material is dissolved and then transferred to an extraction process to recover plutonium. By doing so, a small sized dry process can be employed as a uranium purification process. Since the nuclear fuel material is dissolved and extracted after removing most part of an amount of uranium, a volume of processing solution can be reduced and the machine installation scale can be made small. Accordingly, the reprocessing facility can be extremely downsized.

Abstract: A high-polymer neutron shielding material which scarcely reduces the hydrogen number density when exposed to a high temperature of 150 to 200° C. for a long time period A heat-setting type epoxy resin is employed. The base resin is selected from various epoxy resins such as bisphenol A type epoxy resin. The hardener is selected from alicyclic polyamine, polyamide amine, aromatic polyamine, acid anhydride, and so on. These materials are mixed and hardened at a temperature higher than the room temperature. To give a flame resistance to the hardened resin, a fire retardant such as magnesium hydroxide is added to the mixture. To improve the neutron shielding performance of the hardened resin, a neutron absorbing material is added to the mixture. Further, to increase the moderating performance, hydrogenated bisphenol A epoxy resin is used as the base resin or metal hydride or hydrogen absorbing alloy is added.

Abstract: A high-polymer neutron shielding material which scarcely reduces the hydrogen number density when exposed to a high temperature of 150 to 200° C.

Abstract: Most part of an amount of uranium contained in the spent nuclear fuel is removed by making fluorine or a fluorochemical act on the spent nuclear fuel to convert the uranium into UF6, and the uranium is purified through a simple method of distilling the UF6 together with a absorbent. After removing the most part of the amount of uranium, the remaining nuclear fuel material is dissolved and then transferred to an extraction process to recover plutonium. By doing so, a small sized dry process can be employed as a uranium purification process. Since the nuclear fuel material is dissolved and extracted after removing most part of an amount of uranium, a volume of processing solution can be reduced and the machine installation scale can be made small. Accordingly, the reprocessing facility can be extremely downsized.