Doryokuro Kakunenryo Kaihatsu Jigyodan

Kabushiki Kaisha Kobe Seiko Sho

Katsuyuki Ohtsuka

Jin Ohuchi

Isao Kondoh

Yoshiharu Takahashi

Toru Suzuki

Hitoshi Miyazaki

Ryukichi Mizuno

Kazutoshi Shimogori

Hiroshi Satoh

Takashi Okada

Kazuo Kitagawa

Hiroshi Matsumoto

Toshio Fukuzuka

Hideaki Tamai

Eiichi Inada

Isao Kondo

Hiroaki Miyo

Daisaku Hayashi

Koji Yanagida

Minoru Harada

Atsuhiro Ochiai

Takeshi Kawasaki

Fumio Kami Kubo

Kazutoshi Miyata

Makoto Kanai

Hideaki Aikawa

The invention provides a method of recovering and storing radioactive iodine by a freeze vacuum drying process, in which off-gas generated when spent fuel is subjected to shearing and dissolving treatments is scrubbed and, when necessary, is subjected to a precipitation treatment by addition of additives, after which waste liquid containing radioactive iodine is freeze-dried by a freeze vacuum drying process to recover radioactive iodine as iodine compounds. As a result, since the radioactive iodine does not vaporize, release of the radioactive iodine into the environment can be eliminated. In addition, consumption of a collecting agent such as silver zeolite for collecting vaporized radioactive iodine can be reduced. The iodine compounds containing the recovered radioactive iodine is given the same composition as a stable, naturally occurring mineral and is solidified and mineralized as by a high-pressure press. This makes it possible to store long half-life .sup.129 I radioactive iodine safely for an extended period by sealing recovered radioactive iodine in stable minerals for a long period of time.

Method of recovering and storing radioactive iodine by freeze vacuum

A method of separating and purifying a spent solvent generated in a nuclear fuel cycle and containing a phosphate and a higher hydrocarbon. This method comprises treating the spent solvent at a temperature not greater than the freezing point of the higher hydrocarbon but not less than the freezing point of the phosphate to selectively freeze the higher hydrocarbon, and separating a resulting frozen solid mainly composed of the higher hydrocarbon from a remaining solution containing the phosphate in a higher concentration. The remaining solution may further be subjected to low-temperature vacuum distillation to separate the solution into the phosphate and a deterioration product thereof.

Method of separating and purifying spent solvent generated in nuclear

In a method for recovering plutonium and uranium from spent nuclear fuel by solvent extraction having solvent consisting of tri-n-butyl phosphate, dibutyl phosphate and n-dodecane, the improvement comprises separating the n-dodecane from the phosphate by freeze-drying and separating the phosphate from each other and residual impurities by fractional distillation.

Spent fuel treatment method

A high-level radioactive waste liquid, produced in a reprocessing plant, is treated by a freeze-drying step. A low-level radioactive waste liquid containing water, nitric acid and nuclides is obtained from the sublimate of the freeze-drying step. Fission products, actinides, corrosion products (iron, chromium, nickel, etc.), sodium nitrate and sodium hydroxide, not sublimated, are separated as a residue. An alkaline solution such as sodium hydroxide solution is added to the residue to dissolve sodium nitrate and sodium hydroxide. The dissolved nitrate and hydroxide are then separated from the residue containing fission products and the corrosion products.

Method of treating high-level radioactive waste liquid

A method for thermal decomposition treatment of a radioactive waste uses an apparatus comprising a container for holding molten matter of a radioactive waste containing a sodium compound, a pair of electrodes contacting the molten matter, and a power source for applying voltage between the electrodes while changing the polarity thereof every several tens of seconds. In this apparatus, the molten matter can be heated in the container by Joule heat, which is evolved by electric current directly flowed through the molten matter, so that the sodium compound contained in the radioactive waste can be decomposed, vaporized and removed to recover a stabilized radioactive solid as a residue in the container.

Method for thermal decomposition treatment of radioactive waste

A method of treatment of a radioactive liquid waste containing fission products and a thermally decomposable sodium compound which comprises heating the liquid waste to convert the sodium compound into oxides of sodium, converting the oxides into sodium, hydroxide, reacting the sodium hydroxide with an alcohol to form a sodium alcoholate, separating the sodium alcoholate from an impurity residue essentially comprising fission products, decomposing the separated sodium alcoholate to form sodium hydroxide and recovering the sodium hydroxide.

Method of treatment of radioactive liquid waste

A process for the treatment of radioactive sodium is provided which comprises the steps of forming radioactive sodium amalgam by mixing radioactive sodium with mercury; reacting the radioactive sodium amalgam with water to form mercury and radioactive sodium hydroxide; recycling the mercury into the step of forming radioactive sodium amalgam to be mixed with the radioactive sodium; and solidifying the radioactive sodium hydroxide in the presence of a solidifying material to be confined in a stable solidified body.

Process for the treatment of radioactive sodium

An apparatus for melting and treating metal wastes including radioactive wastes, including a melting and treating vessel formed of the upper surface of a truck and split mold, wherein metal wastes are charged into the vessel in a closed system and melted by an electroslag melting method using a non-consumable electrode. This invention also provides a method which includes the steps of charging a slag melt starting agent and slag into the vessel in layers, and passing electricity through the non-consumable electrode to thereby heat the starting agent.

Apparatus and method for melting and treating metal wastes

A method of minimizing corrosion deterioration and hydrogen absorption embrittlement of tantalum encountered in performing acid digestion of organic waste matter put in an acid digestion kettle of tantalum containing high temperature high concentration sulfuric acid includes the addition of nitric acid, nitrous acid, nitrates, nitrites and/or nitrogen oxides to the sulfuric acid bath.

Method of stabilized operation of acid digestion kettle of tantalum

A heat treating apparatus using microwaves for continuously treating a substance containing nuclear materials with consideration of the criticality safety of the nuclear materials. The apparatus has a stationary outer cylinder and a rotatable inner cylinder to form therebetween an annular passage for the continuous heat treatment of the substance. The outer cylinder has inlet and outlet ports for the material and at least one microwave waveguide. The material supplied into the annular passage is heat treated and transferred to the outlet port during rotation of the inner cylinder.

Heat treating apparatus using microwaves

A substance such as radioactive liquid waste is heat treated using microwave by introducing the substance into a vessel made of a material reflective to microwave and containing spherical bodies made mainly of a material transmissive (i.e. transparent) to microwave. Microwave is applied to the substance while imparting movement to the spherical bodies. The substance is heat treated while in contact with the spherical bodies and is separated from the spherical bodies and pulverized by the friction and collision between the moving spherical bodies. An apparatus for carrying out the method is described.

Heat treating method and apparatus using microwave

A method for the heat treatment of a radioactive substance by microwave power comprising preparing a nitrate solution of uranium, thorium, plutonium or a mixture thereof, and applying to the nitrate solution microwave energy sufficient to directly convert the nitrate solution into an oxide powder of uranium, thorium, plutonium or a mixture thereof, respectively. Such oxide powder is suitable for the manufacture of nuclear fuel pellets.

Growing community of inventors

Mito, Japan

Katsuyuki Ohtsuka