Abstract: A method for separation of actinide elements comprising feeding a solution containing actinide elements such as americium, curium, californium and the like, into a resin column in which a weakly basic primary, secondary or tertiary anion exchange resin obtained by resinifying pyridine, imidazole or alkylamine has been packed, and then feeding an eluent of a mixed solution of nitric acid and alkyl alcohol such as methanol, ethanol, propanol and the like into the resin column to chromatographically separate the actinide elements from each other. This method makes it possible to efficiently separate the actinide elements from each other by a unit operation at ordinary temperature and ordinary pressure while avoiding oxidation operation, and hence makes it possible to avoid generation of secondary wastes and operations difficult in terms of engineering, such as precipitation.

Abstract: A powder compacting apparatus (1) has a top plate (3), a die plate (9) and a bottom plate (12), combined with each other as follows. The top plate (3) holds upper punches (4) in place by punch holders (5), and the die plate (9) firmly grips a die holder (8) holding dies (7). The bottom plate (12) keeps lower punches (11) in place gripped by clamps (32) so that an amount of powder fed into each die (7) is compacted to give a consolidated powder piece. The die plate (9) is composed of an upper part (9a) and a lower part (9b) so that the die holder (8) is sandwiched by and between the parts (9a, 9b), in such a manner that these parts cooperate with each other to secure each die (7) in position.

Abstract: Fluorine or a fluorine compound is subjected to a reaction with a spent oxide fuel to produce fluorides of uranium and plutonium, and the fluorides are recovered using a difference in volatility behavior. The spent oxide fuel is subjected to a reaction with an HF gas, whereby uranium, plutonium and most impurities are converted into solid fluorides having low valences or remained as oxides to inhibit volatilization thereof, and then in an F2 fluorination step, the HF fluorination product is subjected to a reaction with a fluorine gas in two stages: one at a low temperature and the other at a high temperature, whereby a certain amount of gaseous uranium and volatile impurities are separated with plutonium kept in a solid form in the first stage, and mixed fluorides of remaining uranium and plutonium are fluorinated into hexafluorides at the same time in the second stage. By such a reprocessing method, plutonium enrichment can be adjusted, uranium and plutonium can be purified, and steps are simplified as well.

Abstract: A method of simply and quickly determining ?-ray releasing nuclides having long half-life without carrying out a chemical separation is provided. By inputting a data of pulses incident to an ?-ray detector in a computer, obtaining time distribution of the incident pulses by using a very short time measuring timer and plotting, and fitting the linear originated in a random event corresponding to the background and the non-linear originated in the correlated events of parent nuclides-progenies by using the least squares method, the whole generating probability P (t) from the parent nuclide to the progeny thereof is obtained. By subtracting the random events portion from the P(t), the correlated events portion is extracted. The radioactivity per unit can be obtained by dividing the extracted correlated events portion by the measured time, the amount of the supplied sample and the counting efficiency.

Abstract: A light-emitting material is provided allowing a light-emitting body having an excellent low-excitation characteristic and high brightness to be obtained by using a light-emitting material containing a light-emitting base material that emits light through radiative transition of electrons in material atoms, the light-emitting base having nanoparticles added thereto and dispersed therein, the light-emitting material also allowing a reduction in excitation energy and an increase in brightness to be simultaneously achieve, thereby allowing, for a wide range of light-emitting bodies, a reduction in excitation energy and a significant improvement in brightness to be achieved in a simple structure. Also provided is a light-emitting body having the light-emitting material and a light-emitting method.

Abstract: There is provided a method for producing oxide dispersion strengthened ferritic steel tube by fabricating a raw tube by mixed sintering of a metal powder and an oxide powder and producing a tube of the desired shape by repeating cold rolling and heat treatment for a total of three times or more. The method comprises performing each of the intermediate heat treatments during the cold rolling by a two-step heat treatment consisting of a first step heat treatment of 1100° C. or lower and a second step heat treatment of 1100 to 1250° C. and higher than the first step temperature, and performing the final heat treatment at 1100° C. or higher.

Abstract: An elongated, thinner apparatus is provided so as to measure surface contamination in narrow places and piping. An elongated plate-shaped radioactivity detection section (10) provided with a scintillator and a radioactivity measuring section (12) provided with a photomultiplier are provided as independent components and are connected together using a wavelength conversion optical fiber cable (14). A measurement signal (optical signal) of ?-rays detected by the radioactivity detection section is transmitted to the radioactivity measuring section via the optical fiber cable.

Abstract: An apparatus which is capable of detecting the incident direction of a radiation incident from afar and which is made to be small in size and light in weight is provided. The apparatus comprises a radiation detection system 12 having a scintillating optical fiber 10 converting an incident radiation into a light signal, light receiving elements 14a, 14b respectively connected to both ends of said scintillating optical fiber and converting the light signals propagating in two directions along said scintillating optical fiber into electric signals, and a radiation analysis system 16 deriving the incident position of the radiation on the basis of the difference between the arrival times of the electric signals from said light receiving elements. The scintillating optical fiber is arranged in a state of being curved in a circular arc shape, and a large number of flat collimators 18 are arranged on the outer periphery of the scintillating optical fiber in a radial pattern at nearly even intervals.

Abstract: In an oxide dispersion strengthened martensitic steel which comprises, by % by weight, 0.05 to 0.25% C, 8.0 to 12.0% Cr, 0.1 to 4.0% W, 0.1 to 1.0% Ti, 0.1 to 0.5% Y2O3 with the balance being Fe and unavoidable impurities and in which Y2O3 particles are dispersed in the steel, by adjusting the Ti content within the range of 0.1 to 1.0% so that an excess oxygen content Ex.O in steel satisfies [0.22×Ti (% by weight)<Ex.O (% by weight)<0.46×Ti (% by weight)], the oxide particles are finely dispersed and highly densified to thereby obtain an oxide dispersion strengthened martensitic steel excellent in high-temperature strength. It is also possible to reduce the amount of oxygen contamination in steel during the mechanical alloying of raw material powders to provide Ex.O within a predetermined range, by carrying out the mechanical alloying in an Ar atmosphere having a super purity of not less than 99.

Abstract: A reprocessing process of spent nuclear fuels for roughly separating U and U—Pu from FP, TRU and the like in a nitric acid solution of spent nuclear fuels by utilizing phenomenon of cocrystallization of hexavalent U and Pu. For example, spent nuclear fuels are sheared and dissolved in nitric acid, and insoluble residues in the nitric acid solution are removed. Then, a nitric acid concentration in the solution is adjusted and a valence of Pu in the solution is adjusted to tetravalence. The solution is then cooled to crystallize uranyl nitrate hydrate crystals and separated into the crystals and a mother liquor, and the separated crystals are recovered as a U product. Then, a nitric acid concentration in the separated mother liquor is adjusted and a valance of U and Pu in the mother liquor is adjusted to hexavalence, and the mother liquor is cooled to crystallize uranyl-plutonyl nitrate hydrate crystals which are separated and recovered as a U—Pu mixed product.

Abstract: A centrifugal extractor of non-contact journaled construction is provided in which, even under the environments of corrosive gase and mist-like liquid such as nitric acid mist generated in the reprocessing of a spent nuclear fuel, no problem in corrosion or deterioration of parts occurs and high reliability is obtained, and operation is enabled for a long period of time free from maintenance. A rotor 12 housed in a rotor housing 10 is journaled by a main shaft 14 and rotated and driven by a motor 16. The main shaft is surrounded by a drive-portion housing 30, and has a thrust magnetic disk 36 on the upper end and radial magnetic disks 46, 48 and a motor-rotor portion 54 in the circumference thereof.

Abstract: In a reactor vessel thermal load reducing system near the surface level of a coolant, the present invention is characterized in that a heat conductive member is installed not contacting the reactor vessel (1) wall in an area above and below the coolant liquid surface, and the heat conducting member is a guard vessel (2) made of good heat conductivity, or a heat conductive plate (20) made of good heat conductivity.

Abstract: A method for treatment of uranium-containing wastes, which realizes the decontamination of steel-based metallic wastes such as carbon steel or stainless steel containing nuclear fuel materials such as uranium or plutonium, produced from nuclear facilities, simultaneously with the reduction of volume thereof. Spent nuclear fuel (uranium) 6, which is mixed with hull 5 as stainless-based metal, is separated from the stainless-based metal, through the separation of uranium that constitutes the spent nuclear fuel 6 as oxides by melting the stainless-based metal wastes.

Abstract: A method of separation and recovery of elements from radioactive liquid wastes, includes a step of bringing into contact a high-level radioactive liquid waste containing separation target elements including Americium, Curium, Zirconium, Molybdenum, Palladium and rare earth elements with solid absorbent containing organophosphorus compounds so that the separation target elements are absorbed in the solid absorbent, a step of bringing into contact the solid absorbent with an acidic solution containing diethylenetriaminepentaacetic acid so that Americium, Curium, Zirconium, Molybdenum, Palladium and heavy rare earth elements are eluted from the solid absorbent, and a step of bringing into contact the solid absorbent underwent the first elution step with water or dilute nitric acid so that light rare earth elements are eluted from the solid absorbent.

Abstract: A method of separating and recovering useful rare FP contained in spent nuclear fuels (platinum group element (Ru, Rh, Pd), Ag, Tc, Se, Te) selectively and at high recovery percentage is provided. Nitric acid solution to be processed containing useful rare FP in spent nuclear fuels is electrolytically reduced by using Pd2+ or Fe2+ as a catalyst and rare FP are collectively deposited on an electrode and then deposits on the electrode are collectively dissolved by electrolytic oxidation. Then, the deposit-dissolved solution is electrolytically reduced at low current density, medium current density and high current density, successively, whereby Ag.Pd group, Se.Te group and Ru.Rh.Tc group are separately deposited and recovered, group by group. A cooperation system for nuclear power generation and fuel cell power generation can be provided by utilizing the recovered rare FP as electrode materials and production and purification catalysts for hydrogen fuel in fuel cell.

Abstract: This invention is provided for improvement of corrosion-resistant property of a crucible and for promotion of safety in a pyrochemical reprocessing method for the spent nuclear fuel. The spent nuclear fuel is dissolved in a molten salt placed in the crucible. In a pyrochemical reprocessing method, the nuclear fuel is deposited, and the crucible (2) is heated by induction heating. Cooling media (5, 6) are supplied to cool down, and a molten salt layer (7) is maintained by keeping balance between the heating and the cooling, and a solidified salt layer (8) is formed on inner wall surface of the crucible.

Abstract: An apparatus for non-destructively inspecting an arbitrary heat exchanger tube among a group of heat exchanger tubes. The apparatus includes a radiation detector inserted in a heat exchanger tube to be inspected, at least one radiation source inserted in a plurality of heat exchanger tubes surrounding the heat exchanger tube to be inspected, and a CT processing unit. A cross section of the heat exchanger tube to be inspected is imaged by the CT processing. Also, by setting at least one radiation source in an inner portion of the heat exchanger tube, on the inner side of a group of heat exchanger tubes or on the outer side of the group of heat exchanger tubes, and by setting at least one radiation detector carrying a collimator on the outer side of the group of heat exchanger tubes, a cross section of the group of heat exchanger tubes can be imaged by the CT processing.

Abstract: A centrifugal extractor has a rotor supported rotatably in a housing 10, and an aqueous phase liquid and an organic phase liquid are supplied to the outer circumference of the rotor and mixed between the housing and the rotor. A mixed phrase liquid is sucked into the rotor and is separated into two phases in a centrifugal force field generated in the inside of the rotor, and the separated aqueous phase liquid and organic phase liquid are discharged to an aqueous phase collector 50 and an organic phase collector 44, respectively. A cavity portion is provided in the center of the rotor, and a neutron absorption body 60 is disposed in the cavity portion. Preferably, the rotor has a lower rotating and supporting mechanism (such as a sliding bearing 66), and a neutron absorption material is sealed into the lower rotating and supporting mechanism. By such a centrifugal extractor, criticality safety and durability can be enhanced, even when the extractor is designed to be larger-size and larger-capacity.

Abstract: A radioactive dust monitor comprising a dust collecting electrode serving also as a radiation detecting surface, an ionization line extending in air in a spaced relation with the dust collecting electrode and a radiation detection part. The radiation detection part is provided with a scintillator disposed close to said dust collecting electrode and a photoelectric conversion part converting light emitted from the scintillator into an electric signal. Dust in air is collected by the dust collecting electrode due to a corona discharge which is generated by applying a negative high voltage to the ionization line and by applying a positive high voltage to the dust collecting electrode, and light emitted from the scintillator due to radioactive rays radiated from the collected radioactive dust is detected by the photoelectric conversion part.

Abstract: An intermediate heat exchanger-incorporated type steam generator having an intermediate heat exchanger tube 20 and a steam generating heat exchanger tube 22 disposed separately in a vessel 10 storing a secondary coolant 14 therein. A pump mechanism is made of an electromagnetic pump mechanism formed by an electromagnetic driving coil 24 provided on an outer circumference of the vessel and a magnetic core 26 attached to an inner cylinder 12 disposed in the vessel. At least one porous plate or slotted plate 18 is preferably disposed between the intermediate heat exchanger tube and steam generating heat exchanger tube.