382 results about "Radioactive hydrogen" patented technology

A rugged adjustable, optical alignment system is well suited for use in low light environments or at night. A rear sight includes a U-shaped array of, preferably, first second and third substantially cylindrical, transparent, tritium vials arranged such that the elongated cylindrical side surfaces of the vials are exposed to define an array of elongated illuminated segments. The array is preferably in the shape of a “U” having a first horizontally aligned vial positioned below a sight notch with a second vertical vial positioned above and to the left of the notch. A third vertical vial is positioned to the right of the sighting notch, above the first vial. In use, a front blade or front post also containing a tritium insert is aligned within the rear sight's U-shaped array of illuminated segments, thereby providing an intuitive sight picture.

The invention discloses a liquid metal cooling reactor experimental system capable of realizing critical and subcritical running test. The liquid metal cooling reactor experimental system consists of a safety vessel (1), a main vessel (2), a reactor inner supporting structure (3), a reactor core (4), a central measuring column (5), a main heat exchanger (6), a reloading mechanism (7), a reactor top cover (8), a control bar drive mechanism (9) and a neutron source (10); when the critical running test is performed, the neutron source (10) adopts a californium-252 neutron source or a Be-Am neutron source; when the subcritical running test is performed, the neutron source (10) adopts a spallation neutron source or a deuterium tritium neutron source, an isotopic neutron source and a fuel component in the middle part of the reactor core are replaced by an accelerator neutron source according to the real requirement of the spallation neutron source, and the reactor can have the subcritical running test. The reactor experimental system is naturally and circularly cooled by adopting liquid lead-bismuth or lead, so that the usability of the reactor can be improved, the experimental cost can be reduced, and the multifunctional reactor experimental system characteristics can be achieved.

A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360° and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

The invention concerns radiographic equipment. The equipment includes a source of substantially mono-energetic fast neutrons produced via the deuterium-tritium or deuterium-deuterium fusion reactions, comprising a sealed-tube or similar generator for producing the neutrons. The equipment further includes a source of X-rays or gamma-rays of sufficient energy to substantially penetrate an object to be imaged and a collimating block surrounding the neutron and gamma-ray sources, apart from the provision of one or more slots emitting substantially fan-shaped radiation beams. Further included is a detector array comprising a multiplicity of individual scintillator pixels to receive radiation energy from the sources and convert the received energy into light pulses, the detector array aligned with the fan-shaped beams emitted from the source collimator and collimated to substantially prevent radiation other than that directly transmitted from the sources reaching the array. Conversion means are included for converting the light pulses produced in the scintillators into electrical signals. Conveying means are included for conveying an object between the sources and the detector array. Computing means are included for determining from the electrical signals the attenuation of the neutrons and the X-ray or gamma-ray beams and to generate output representing the mass distribution and composition of the object interposed between the source and detector array. The equipment further includes a display means for displaying images based on the mass distribution and the composition of the object being scanned.

A compact neutron generator has at its outer circumference a toroidal shaped plasma chamber in which a tritium (or other) plasma is generated. A RF antenna is wrapped around the plasma chamber. A plurality of tritium ion beamlets are extracted through spaced extraction apertures of a plasma electrode on the inner surface of the toroidal plasma chamber and directed inwardly toward the center of neutron generator. The beamlets pass through spaced acceleration and focusing electrodes to a neutron generating target at the center of neutron generator. The target is typically made of titanium tubing. Water is flowed through the tubing for cooling. The beam can be pulsed rapidly to achieve ultrashort neutron bursts. The target may be moved rapidly up and down so that the average power deposited on the surface of the target may be kept at a reasonable level. The neutron generator can produce fast neutrons from a T-T reaction which can be used for luggage and cargo interrogation applications. A luggage or cargo inspection system has a pulsed T-T neutron generator or source at the center, surrounded by associated gamma detectors and other components for identifying explosives or other contraband.

This invention discloses a kind of tritium determination instrument and tritium determination method. The tritium determination instrument includes the current collection electrode, tapered guard ring and open stainless steel barrel, there are several holes set in the stainless steel barrel. The test gas can diffuse freely into the stainless steel barrel, determines the concentration of tritium in suction pressure and ordinary pressure. The determination speed of the method to determine the concentration of tritium in this invention is fast, it can reduce the radwaste in the monitor process of determining the concentration of tritium. The structure of the tritium determination instrument is simple, the determination method is reliable, it is easy to apply.

A multilayer device for producing electricity. The device comprising a betavoltaic source layer for generating beta particles, and at least three semiconductor layers each having a bandgap substantially similar to a band gap of the other layers, the at least three layers comprising a doped top layer, an undoped intermediate layer and a doped bottom layer, wherein the top and the bottom layers are doped with opposite-type dopants, and wherein the top layer is closer to the betavoltaic source layer than the bottom layer.

A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium / tritium (or even just a tritium) plasma is generated in the chamber and D or D / T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.

The invention provides a low temperature adsorption method for purifying impurity-containing gas, and continuously producing ultra-pure gas. The method uses liquid nitrogen as cold source, uses liquid nitrogen evaporator to cool raw material gas to required temperature, and uses low temperature adsorption tower group to adsorb to prepare ultra-pure gas with purity more than 6N. In addition, the method also includes cold energy recovery process of product gas and low temperature nitrogen. The method can be used for continuously producing ultra-pure hydrogen, deuterium, tritium, helium, neon, argon, nitrogen, oxygen, carbon monoxide and methane. The invention radically solves defects of directly putting an adsorption tower (column) into a liquid nitrogen groove, and realizes large-scale continuous production of ultra-pure gas.

The invention provides a tritium-containing exhaust gas discharge monitoring automatic sampling system and belongs to the technical fields of radiation protection and environmental protection. The air inlet tube joint in the sampling system is connected with a flow control valve, solenoid valves, dust filters, a mass flow meter and a pressure meter through a gas tube and then is connected with the gas inlets of pumps, and the gas outlets of the pumps are connected with the joints of an exhaust tube through exhaust tubes; the gas inlet and the gas outlet of a total tritium sampler are connected with a gas collection tube and the exhaust tube through flexible tubes respectively; and the signal input end of an automatic control unit is in external connection with a chimney flow monitoring PLC controller, and the signal output end of the automatic control unit is connected with the total tritium sampler, the flow control valve, the solenoid valves, the mass flow meter, the pressure meter and pumps respectively. The sampling system which allows the sampling precision to be less than 1min has the advantages of intelligent regulation and control, strong practicality, improvement of the fragmentary discharge of a tritium effluent, improvement of the representativeness of a tritium monitoring sample when the concentration difference is large, mitigation of the labor intensity of an effluent monitoring sampling personnel, and improvement of the working efficiency.

A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium / tritium (or even just a tritium) plasma is generated in the chamber and D or D / T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.

The invention discloses a thermal hydraulic characteristic testing device and method for tritium blowing helium of a breeding blanket pebble bed of a fusion reactor. The thermal hydraulic characteristic testing device for the tritium blowing helium of the breeding blanket pebble bed of the fusion reactor is mainly used for researching the flow and heat transfer characteristics of the tritium blowing helium of the solid breeding blanket layer of the fusion reactor between breeding agent pebbles of the pebble bed. The thermal hydraulic characteristic testing method includes that using a non-contact type high-frequency induction power to heat quasi metal breeding agent pebbles to simulate high-power density nuclear heat generated by the breeding agent pebbles in the pebble bed, researching the thermal hydraulic characteristics of the pebble bed under changed tritium blowing helium temperature, speed and pressure, breeding agent pebble arrangement mode, size and power density and the like, measuring the temperature distribution of the breeding agent pebbles and pebble bed wall surface, and obtaining a convection heat transfer correlation equation and a flow resistance relation to provide supports for testing and demonstrating the nuclear heat transmission of the neutron deposition in the pebble bed on the breeding agent pebbles and the tritium blowing helium flow and heat transfer characteristics under laboratory conditions, verifying the feasibility of a solid breeding blanket designing scheme and the correctness of a numerical method and data, and researching and developing Chinese solid experiment blanket modules of ITER.

The invention discloses a method for preparing a hydrogen permeation preventing coating by aluminizing by adopting room temperature fused salt, comprising the following steps of: firstly, carrying out surface treatment of degreasing and oxidation film removing on strain steel matrix with a conventional method; secondly, aluminizing by adopting room temperature fused salt, wherein the room temperature fused salt for aluminizing is a system formed by AlCl3 and organic salt, the organic salt is one of alkyl pyridine halide, alkyl imidazole halide and alkaryl chloration ammonium salt, the mole ratio of the AlCl3 to the organic salt is larger than 1.0 and smaller than or equal to 2.0, the technical parameter for aluminizing by adopting room temperature fused salt is as follows: an aluminum wire as the anode has the temperature of 25-60 DEG C and the current density of 5-30mA / cm<2>, and the plating time is 30-200min; and thirdly, heat treatment to obtain an Fe-Al alloy layer, wherein the technical parameter for heat treatment is as follows: the temperature is 670-750 DEG C, and the time is 30min to 24h; fourthly, oxidation, wherein the technical parameter for oxidation is as follows: the temperature is 670-1050 DEG C, the time is 1-200h, and the oxygen partial pressure is 1.0*10-2 Pa to 2.1*104 Pa. The invention is suitable for preparing tritium permeation preventing coatings on surfaces of special-shaped pieces such as the inner wall of a pipeline, and the like, is suitable for the requirement of large-scale industrialized production, and has a self-healing function.

A process for light water detritiation comprised of water distillation for tritium stripping and enriching, followed by chemical conversion of tritium enriched water to elemental hydrogen, and finally by one or more thermal diffusion columns for final tritium enrichment. The combination of process steps takes advantage of water distillation large throughput capability at low tritium concentration with the simplicity of thermal diffusion for small throughput final tritium enrichment. The water distillation front-end and the thermal diffusion back end processes are compatible with any intermediate chemical conversion process such as electrolysis or water gas shift reaction to convert tritiated water to elemental hydrogen.

Systems, methods, and apparatuses for separating tritium from radioactive waste materials and the water from nuclear reactors. Some embodiments involve the reaction of tritiated hydrogen gases with water in the presence of a catalyst in a catalytic exchange column, yielding a more concentrated and purified tritiated water product. Some embodiments involve the use of a permeation module, similar in some respects to a gas chromatography column, in which a palladium permeation layer is used to separate tritiated hydrogen gas from a mixture of gases.

The present invention relates to amino acids, complexes, and compounds comprising deuterium and tritium isotopes preferably alpha deuterated amino acids, polypeptides, antibodies, derivatives and saccharide-amino acid complexes and conjugates. In some embodiments, the invention relates to methods of using compounds comprising deuterium for imaging biochemical concentrations and distributions in mammalian tissues using nuclear magnetic resonance spectroscopy. In some embodiments, the invention relates to the used of said amino acids derivatives and complexes in boron neutron capture therapy. In some embodiments, the present invention relates to the preparation of amino acids, polypeptides, antibodies, derivatives and saccharide complexes / conjugates comprising heavy hydrogen isotopes. In some embodiments, the invention relates to racemizing amino acids starting from compositions of any optical purity. In further embodiments, the invention relates to the preparation of amino acids and their N-acyl counterparts with deuterium incorporated at the alpha carbon.

An apparatus for generating thermonuclear microexplosions includes a first pulsed high-voltage source configured for transmission of a first high-voltage pulse at a first high current. A second pulsed high voltage source is configured for transmission of a second high-voltage pulse at a potential higher than that of the first high-voltage pulse at a second high current having a magnitude less than that of the first high current. An inner transmission line is in electrical contact with the first pulsed high-voltage source, the inner line having a tapered end. An outer transmission line is disposed over the inner line and in electrical contact with the first pulsed high-voltage source. A deuterium-tritium mixture is disposed between the inner transmission line and the outer transmission line, the deuterium-tritium mixture having a tapered end spaced apart from the tapered end of said inner transmission line.

A system and method for tritium separation systems using a mixed bed catalytic exchange process in a Liquid Phase Catalytic Exchange / Closed Loop Continuous Process (LPCE / CLCP) system, that operates as a low temperature and low pressure continuous balanced process, designed to rapidly, economically and safely extract and isolate isotope specific products without generating unwanted products in the form of new waste streams.

A fusion driving subcritical cladding of transmutation subordinate actinium series nuclide is formed by a plurality of same independent small modules. The whole shape of each small module is D-shaped. Each small module comprises an inner cladding, an outer cladding and a D-shaped plasma body cavity which is formed in each small module. The inner cladding is in a cylindrical shape. The inner cladding successively comprises an inner shield, an inner reflecting layer, an inner first wall and an inner scraping layer from inside to outside along the radial direction. The whole shape of the outer cladding is also D-shaped. The outer cladding successively comprises an outer scraping layer, an outer first wall, a transmutation area, a tritium breeding blanket, an outer reflecting layer and an outer shielding layer from inside to outside along the radial direction. A plurality of transmutation fuel subassemblies are horizontally arranged in the transmutation area along the first wall. By the adoption of the modularized design, fuel is convenient to load and unload. By the adoption of the horizontal arrangement of the transmutation fuel subassemblies in the transmutation area, the filling rate of the transmutation area is improved to enable the structure of the transmutation cladding to be more compact. Metal alloy fuel is adopted by the fuel in the transmutation fuel subassemblies, and a hard energy spectrum is obtained to benefit metal alloy (MA) to carry out the direct and effective transmutation through fission reactions.

The present invention provides a pulsed neutron generator. In a high vacuum and high pressure ceramic tube, when a trigger pulse is added between the source trigger pole and cathode of vacuum arc ion, the deuterium atom is ionized and the initial plasma is diffused, such that the anode and the cathode of the ion source is conducted, the arc current is formed, and the high density plasma is generated. Under the action of accelerating the high pressure, the deuterium ion beam interacts with the tritium titanium target, the deuterium tritium reaction produces, and the pulsed neutron with the energy of 14 MeV is generated. A circular magnetic field is set up at the lead pole of the vacuum arc ion source, which can focus on the formation of the deuterium ion, increase the deuterium ion beam to the target, improve the pulsed neutron yield; The combination of a dry pump, a molecular pump, an ion pump and an air pump is used to improve the vacuum degree of the system, and reduce the loss of deuterium ion beam in the transmission process. The aluminum film with the thickness of 400 nm is coated on the surface of the tritium titanium target, which prevents the heavy ion from reaching the tritium target, and reduces the damage of the target.

The invention provides a method for quickly manufacturing Li2TiO3 tritium breeding small balls. The method comprises the following steps that (1) Li2TiO3 powder and binder powder are mixed to form mixed powder; (2) a small ball three-dimensional model is established in a computer and is stored in the format of STL; (3) the mixed powder obtained in the step (1) is transferred into 3D printing equipment, the small ball three-dimensional model in the format of STL is introduced into a system of the 3D printing equipment; (4) technological parameters are set and the 3D printing equipment is started to obtain small ball entities; (5) the small ball entities are placed into a cold isostatic press to be compacted; (6) binder removal processing is conducted on the small ball entities; (7) the small ball entities are sintered in a sintering furnace. The Li2TiO3 small balls manufactured through the method are high in degree of sphericity, compactness and crushing strength, and large-scale and mechanical production of the Li2TiO3 tritium breeding small balls can be realized.

The present invention relates to a resonant vacuum arc discharge apparatus for producing nuclear fusion. A resonant high-frequency high-voltage alternating current (AC) power supply is used to efficiently power a fusion tube normally containing deuterium, tritium and / or helium-3 vapor. Metals that can hold large amounts of hydrogen isotopes such as palladium and titanium can be used to increase the target density. The nuclear fusion device can be used for energy production, well logging, uranium mining, neutron activation analysis, isotope production or other applications that require a neutron source.

The invention discloses a high-efficiency tritium removal purification device and a tritium removal method thereof. The high-efficiency tritium removal purification device comprises a cabinet body, a gas circulating pump, a gas storage tank, a vacuum pump, a flow controller, a catalytic air suction device and an adsorption bed, wherein the vacuum pump, the flow controller, the catalytic air suction device and the adsorption bed are respectively arranged inside the cabinet body; the gas circulating pump, the flow controller, the catalytic air suction device and the adsorption bed are successively connected to form a circulation loop, and the vacuum pump is connected into the circulating loop; the catalytic air suction device comprises a catalytic oxidation device and an alloy adsorption device in parallel, the catalytic oxidation device comprises a first electromagnetic valve, a catalytic bed and a second electromagnetic valve which are connected in sequence, and the alloy adsorption device comprises a third electromagnetic valve, a getter bed and a fourth electromagnetic valve which are connected in sequence; the flow controller is connected with the first electromagnetic valve and the third electromagnetic valve at the same time, and the adsorption bed is connected with the second electromagnetic valve and the fourth electromagnetic valve at the same time; a gasket is further arranged inside the catalytic bed. When the concentration of tritium in a tritium system exceeds the standard or the tritium leakage occurs, tritium-containing gas is quickly transferred and treated, so that the purification of the tritium is effectively realized.

A diffusion based process for tritium removal from water by tritium transfer from water to an elemental hydrogen stream, followed by a membrane diffusion cascade for tritium stripping and enrichment, and final tritium enrichment by one or more thermal diffusion columns. The combination of process steps takes advantage of membrane diffusion's large throughput capability at low tritium concentration with the simplicity of thermal diffusion for small throughput final tritium enrichment. The membrane diffusion stages use supported or unsupported microporous or hydrogen permeable metal membranes (such as Pd / Ag alloy). The diffusion process is compatible with any front-end process to transfer tritium from tritiated water to elemental hydrogen. The process may be designed and operated at low pressure, with small gas inventory, and no inherent overpressure hazard.

The invention discloses nanostructured lithium orthosilicate ceramic spheres used for tritium propagation and a preparation method thereof. The nanostructured lithium orthosilicate ceramic spheres are made by preparing precursor powder with uniform particle size through a solvothermal method, obtaining a lithium ceramic sphere blank with a uniform microstructure through wet moulding, and carrying out two-step sintering. The prepared lithium orthosilicate ceramics is high in purity, good in sphericility, small in hole size and uniform in hole dispersion, has a crystal grain size in nanometer order and allows the radiation resistance, mechanical properties and tritium release properties of tritium propagation ceramics to be improved at the same time.

A patterning method forms a pattern including a lyophilic region and a lyophobic region. The method includes treating the surface of an object by exposing an atmosphere containing at least one gas selected from the group consisting of hydrogen, deuterium, deuterated hydrogen, and tritium; partially exposing the treated surface to light to form an exposed region and an unexposed region; and applying a liquid onto one of the exposed region and the unexposed region.

Compounds having the general formula II, III, or IVwherein R1 can be a radical of an optionally substituted C-4 to C-7 N-containing heterocycle or a radical of an optionally substituted cyclic or acyclic tertiary amine or isoindoline-1,3-dione: R2,3,4,5,6 can each independently be any one or combinations of the following moieties, cyano, nitro, acyl, alkyl, amido, azido, isothiocyanate, isocyanate, optionally substituted anilino, halogens, ethers, sulfonamides, thioacyl, nitro, aromatic, heterocyclic, olefinic, acetylene, deuterium, or tritium; Y can be either CH, CH2, O, S, OCH2, N—R, N—Ar, C—R, C—Ar; Z can be either H, O, S, S—R or NR. R groups can be either H, aryls, alkyls, or cycloalkyls; “n” can be 1 to 5 carbons in length and stereoisomers, functional analogs, and pharmaceutically acceptable salts thereof and wherein the moiety bridging R1 and N can be a substituted alkylene, optionally substituted alkenylene or optionally substituted alkynylene and where the alkylene group can include an inserted C3-C5 cycloalkyl group, aromatic, and hetercocyclic group; wherein X′ is halogen, or C1-C4 haloalyl; wherein the Rx is a C1-C5 straight chain or branched chain alkyl or a C1-C4 straight chain or branched chain haloalkyl.

The invention provides a method and a system for measuring a gas tritium based on a multi-wire proportional chamber. The method comprises the following steps that: A, an exhaust device discharges air in the multi-wire proportional chamber, and a gas pressure measuring device monitors the air pressure in the multi-wire proportional chamber; B, the exhaust device extracts an air sample containing gas tritium, the air sample is mixed with a working gas in a certain ratio through a gas mixing device, and then the mixed is output to a gas purifying device; C, the mixed gas is purified by the gas purifying device and is output to the multi-wire proportional chamber through an electromagnetic valve; and D, the anode wire of the multi-wire proportional chamber is connected with a positive high pressure, a beta-ray generated from the decay of the gas tritium entering the multi-wire proportional chamber is electrically separated from the working gas, the avalanche amplification is carried out under the action of a high voltage electric field, a pulse signal is output, and the pulse signal is obtained and identified by a data obtaining system so as to realize the measurement of the gas tritium. According to the method, the defects of an existing gas tritium measuring method, such as working efficiency, precision and sensitivity in the measurement, are overcome, the measurement of the environment gas tritium is accurately realized, and active demands on the gas tritium measurement of an existing tritium environment are met.