Hitachi, Ltd.

Nippon Engelhard Ltd.

Hideo Yusa

Makoto Kikuchi

Kunio Kamiya

Fumio Kawamura

Koichi Chino

Susumu Horiuchi

Shin Tamata

Kiyomi Funabashi

Fumito Nakajima

Masato Takeuchi

Takashi Nishi

Masami Matsuda

Mikio Hirano

Yoshiyuki Takamura

Masaki Takeshima

Hiroyuki Tsuchiya

Akira Oda

Hideichi Miura

Hiroshi Kitaguchi

Takehiko Kitamori

Shigeru Izumi

Katsumi Ohsumi

Shunsuke Uchida

Naohito Uetake

Tsutomu Baba

Tatsuo Hayashi

Tomohiro Sato

Haruo Fujimori

Tatsuo Izumida

Kazumichi Suzuki

Tetsuo Fukasawa

Akira Sasahira

Masao Kitamura

Yoshiyuki Aoyama

Tadashi Gotou

Yoshihiro Ozawa

Yoshikazu Sugimoto

Tetsuro Adachi

Itaru Komori

Norihiko Sagawa

Toshio Ishidate

Takashi Taniguchi

Yomei Kato

A system for discriminating radiation-contaminated fragments on the basis of a predetermined radioactive concentration includes a detection device for detecting the radiation of the fragments, the detection device having a generally vertically-extending path of transfer of the fragments, at least one radiation detector disposed at one of the inside and outside of asid transfer path, and a transfer device for sequentially transferring the fragments in the transfer path; a convey device for conveying the fragments to said detection device; and a controller for determining the radioactive concentration of the fragments in accordance with the radiation detected by the detection device and for judging whether or not the radioactive concentration of the fragments is the predetermined radioactive concentration. The controller is also operable to control the transfer device so as to adjust the speed of transfer of the fragments in the transfer path. The system further includes a density detecting device for detecting the density of the fragments to be conveyed to the detection device. The controller corrects the above determined radioactive concentration by the density detected by the density detecting device.

System for discriminating radiation-contaminated fragments and apparatus

A process and an apparatus for cleaning nuclear reactor cooling water with cation exchange resin whose ion-exchanging groups have a bonding energy of not more than 300 KJ/mole are disclosed, whereby the radiation exposure of operators in an atomic power plant can be considerably reduced, and the waste ion exchange resin can be readily disposed.

Process and apparatus for cleaning nuclear reactor cooling water

The present invention relates to a method and an apparatus for producing a waste package of radioactive waste containing particles of radioactive waste material of low modulus of elasticity, particles of radioactive waste material of high modulus of elasticity, and a solidifying agent in which the particles of radioactive waste material of low modulus of elasticity and the particles of radioactive waste material of high modulus of elasticity are fixed in an almost uniformly dispersed state. According to this invention, the radioactive waste generated from nuclear power plants can be greatly reduced in volume and also a waste package of radioactive waste with high strength and excellent water resistance can be obtained.

Radioactive waste treatment method

A dissolved substance contained in a solution is oxidized or reduced by adding a particulate semiconductor photocatalyst and a water-soluble electron acceptor or donor for the particulate semiconductor photocatalyst to a solution containing a dissolved substance, and irradiating the particulate semiconductor photocatalyst in the solution with an electromagnetic wave having an energy high enough to excite the particulate semiconductor photocatalyst, or by adding a particulate semiconductor photocatalyst and a water-soluble electron acceptor or donor for the particulate semiconductor photocatalyst to a solution containing a dissolved substance and irradiating the particulate semiconductor photocatalyst in the solution with an electromagnetic wave having an energy high enough to excite the particulate semiconductor photocatalyst and having a wavelength outside the absorption wavelength region of a precursor substance capable of producing an oxidation or reduction reaction-inhibiting substance, whereby (1) ions or complexes are converted to simple substances or to ions or compounds with different valencies, and (2) simple substances, ions, complexes or other compounds are separated and recovered or removed by utilizing its actions and effects.

Process and apparatus for oxidizing or reducing dissolved substance

A method of processing radioactive waste resin by pyrolyzing radioactive waste ion exchange resin generated in a nuclear plant such as a nuclear power station. First, the ion exchange resin is pyrolyzed at a low temperature, and the resulting decomposition gas is separated. Second, the ion exchange resin at a high temperature, and the resulting decomposition gas is separated. Finally, the residue of the ion exchange resin is hot-pressed into a molded article.

Method for processing radioactive waste resin

YThis invention concerns solidification products capable of immobilizing radioactive wastes containing water soluble solid components stably for a long period of time, as well as a process for producing the same, in which radioactive wastes are immobilized with an alkaline earth metal silicate compound and the alkaline earth metal silicate compound intakes the water content in the solidifying agent as the bound water to or a hydrate. High temperature and high humidity condition has been found to be necessary upon curing in order to take the water content as the bound water. This invention can provide solidified wastes with less development of detects such as open pores or cracks.

Solidified radioactive wastes and process for producing the same

A sodium-sulfur battery comprising molten sodium as an active material for a minus pole and molten sulfur as an active material for a plus pole, the molten sodium and molten sulfur being partitioned from each other by a sodium ion-permeating solid electrolyte such as .beta.-alumina as a boundary is provided with a sodium storage tank communicated with the molten sodium in the battery cell, a draining mechanism for withdrawing the molten sodium, when desired, from the battery cell into the sodium storage tank, and a circulating mechanism for purifying and returning the sodium from the sodium storage tank to the battery cell as the molten sodium, sodium can be discarged urgently from the battery at an accident of direct contact between the sodium and sulfur, with the result of improved safety. A power storage system comprising a plurality of the sodium-sulfur batteries connected to one another in series and further connected in parallel can be provided with common sodium storage tanks and relevant draining mechanisms, purifying and circulating mechanisms in equal potential level rows of the batteries, resulting in improved economy and safety.

Power storage system using sodium-sulfur batteries

In storing of radioactive wastes by drying, pulverization and pelletizing, radioactive waste pellets are solidified with an alkali silicate solution as a filler, a substance having an action to harden the alkali silicate solution, and a substance having an action to absorb the water formed by the hardening reaction of the alkali silicate solution, or a substance having both actions to harden the alkali silicate solution and to absorb the water formed by the hardening reaction. The solidification can be attained with easy operation and prolonged stability at a low cost.

Process for solidifying radioactive waste pellets

A waste gas generated in the reprocessing of used nuclear fuel is at first subjected to removal of explosive, gaseous substances such as hydrocarbons and nitrogen oxides, and materials capable of forming the explosive gaseous substances, such as oxygen, and condensible substances such as carbon dioxide, water and ammonia from the waste gas, and then to cryogenic distillation, thereby separating and recovering Kr-85 from the waste gas. As the separation and recovery of Kr-85 is carried out by cryogenic separation after the removal of the substances having a possibility to explode in a cryogenic distillation apparatus and also the removal of condensible substances having a possibility to clog a piping system of the apparatus, the operation of the apparatus is ensured. It is also disclosed that the oxygen and nitrogen oxides can be completely removed by catalytic hydrogen reduction.

Process for treating waste gas in reprocessing of used nuclear fuel

In the concentration of radioactive combustible waste comprising evaporating and drying a radioactive combustible waste slurry by a thin film drier, the radioactive combustible wastes and an incombustible material, such as sodium sulfate, etc. obtained from a nuclear reactor effluent are mixed together in an amount of not more than 70% by weight, preferably 43 to 70% by weight, of radioactive combustible waste on the basis of a mixture on dry basis, and the mixture is fed to the thin film drier and evaporated and dried therein. A risk of powder explosion and fire is prevented in the drying step thereby.

Process for concentrating radioactive combustible waste

N-.beta.-(aminoethyl)-.gamma.-aminopropyltrimethoxysilane [NH.sub.2 (CH.sub.2).sub.2 NH(CH.sub.2).sub.3 Si(OCH.sub.3).sub.3 ] as a silane coupling agent and SiO.sub.(2-x) (ONa).sub.x/2 (OH).sub.x/2 as colloidal silica are mixed into a radioactive liquid waste containing sodium sulfate as a main component, coming from a boiling water-type, nuclear power plant as an effluent. The resulting mixed radioactive liquid waste is supplied into a vessel provided with a rotating shaft with blades. The rotating shaft is revolved while heating the radioactive liquid waste in the vessel, thereby making the radioactive liquid waste into powder. The resulting powder containing the silane coupling agent and the colloidal silica is shaped into pellets by a pelletizer. The pellets having a low hygroscopicity and a high strength are obtained.

Process for treating radioactive waste

A rotary vane type evaporator has a cylindrical vessel and a rotor rotatable therein. The rotor has vane supports fixed to a rotor shaft and vanes each pivotally mounted at the radially inner end to a vane support and having a radially outer end connected with a shoe extending in the circumferential direction of the vessel. Each vane is provided with a balance weight acting to keep constant, irrespective of the wear of the shoe, the pressing force which urges the shoe against the vessel inner peripheral surface. The rotor vanes are disposed in a plurality of stages arranged in axial direction of the rotor. The rotor is so constructed as to uniformalize the wear of the vane shoes over all the stages, whereby the frequency of the interruption of the evaporator operation for the renewal of vanes is lowered to improve the rate of the operation of the evaporator.

Rotary vane type evaporator

A radioactive waste discharged from a radioactive substance handling equipment is dried and powdered, and the powder is pelletized. The resulting pellets are stored in an inner vessel of a store vessel having a double structure for a predetermined period to attenuate the radioactivity of the pellets. Then, the pellets are taken out from the store vessel and packed into a sealing vessel. A binder is injected into the sealing vessel to effect solidification.

Method and apparatus for treatment of radioactive wastes

Granular ion exchange resins in a desalter provided in a nuclear power plant are taken out from the desalter as used ion exchange resins when their capacities are lowered. The used ion exchange resins in a slurry state are led to a thin film drier. A rotating shaft provided with rotating blades is inserted in the thin film drier. When the slurry of the used ion exchange resins is led thereto, the rotating shaft is driven to rotate, and the side wall of the thin film drier is heated. The used ion exchange resins led to the thin film drier are heated and pulverized to powder by the action of rotating blades. The powder is taken out of the thin film drier to the outside and shaped into pellets.

Growing community of inventors

Katsuta, Japan

Hideo Yusa