348 results about "Tritium" patented technology

New and unique improvements of prior known fiber optic sights for firearms with day and night time capabilities are disclosed that comprise (A) a sight base, (B) a fiber optic rod mounted in said base having an angle cut at the distal end which is positioned on the underside of the rod, and (C) an artificial light insert that is positioned in a cavity in the fiber optic rod. The cavity is sealed with a fiber optic rod plug and epoxied using an optically clear epoxy. The sight provides increased light output, co-located day and night sight views with increased illumination during the day, and uses a low power tritium insert for night time use. At night the sight achieves a transition from a bright ring during the day to a small central dot at night while maintaining the same color light for both.

The present invention is a power cell for directly converting ionizing radiation into electrical energy. The invented isotopic electric converter provides an electrical power source that includes an electronegative material layered in a semiconductor, to form a first region that has a high density of conduction electrons, and an electropositive material also layered in the semiconductor material to form a second region with a high density of holes. Said N-layers region and P-layers region are separated by a neutral zone of semiconductor material doped with a radioactive isotope, such as, but not limited to, tritium. No junction is formed between the N and P layers regions. Rather, the potential gradient across the neutral zone is provided by the difference between the work functions of the electronegative and electropositive electrodes. Electrical contacts are affixed to the respective regions of the first and second type conductivity which become the anode and cathode of the cell, respectively. Beta particles emitted by the tritium generate electron-hole pairs within the neutral zone, which are swept away by the potential gradient between the first and second regions, thereby producing an electric current.

To increase total power in a betavoltaic device, it is desirable to have greater radioisotope material and/or semiconductor surface area, rather than greater radioisotope material volume. An example of this invention is a high power density betavoltaic battery. In one example of this invention, tritium is used as a fuel source. In other examples, radioisotopes, such as Nickel-63, Phosphorus-33 or promethium, may be used. The semiconductor used in this invention may include, but is not limited to, Si, GaAs, GaP, GaN, diamond, and SiC. For example (for purposes of illustration/example, only), tritium will be referenced as an exemplary fuel source, and SiC will be referenced as an exemplary semiconductor material. Other variations and examples are also discussed and given.

A nuclear fusion power plant having a spherical blast-chamber filled with a liquid coolant that breeds tritium, absorbs neutrons, and functions as both an acoustical and laser medium. Fuel bubbles up through the sphere's base and is positioned using computer guided piezoelectric transducers that are located outside the blast-chamber. These generate phase-shifted standing-waves that tractor the bubble to the center. Once there, powerful acoustic compression waves are launched. Shortly before these reach the fuel, an intense burst of light is pumped into the sphere, making the liquid laser-active. When the shockwaves arrive, the fuel temperature skyrockets and it radiates brightly. This, photon-burst, seeds outgoing laser cascades that return, greatly amplified, from the sphere's polished innards. Trapped within a reflecting sphere, squeezed on all sides by high-density matter, the fuel cannot cool or disassemble before thorough combustion. The blast's kinetic energy is absorbed piezoelectrically.

The invention discloses a small neutron source adopting a windowless gas target, which belongs to the field of nuclear technology and application. In the invention, an ECR (Electron Cyclotron Resonance) ion source is used for generating a deuterium ion and directly extracted by a high-voltage extraction electrode for bombarding a windowless gas target sealed by a plasma; because the windowless deuterium gas target is sealed by adopting the plasma, the windowless deuterium gas target is allowed to bear high-flow-strength beam current; and because the energy loss of the deuterium ion when penetrating through a plasma window is less, the neutron yield is higher. Compared with a neutron pipe, the neutron source provided by the invention has high allowed beam current strength and high neutron yield. Compared with an accelerator neutron source with large size, complex system and high cost, the neutron source provided by the invention has small size, simple system and low cost. Compared with a neutron source for directly carrying out deuterium-tritium reaction on an ion source to generate neutron, the ion source has simple system, low cost and no tritium radioactive processing and circulating problem. The invention has very wide application prospect.

Embodiments of the present disclosure are directed to a phototherapy eye device. In an example, the phototherapy eye device includes a number of radioluminescent light sources and an anchor. Each radioluminescent light source includes an interior chamber coated with phosphor material, such as zinc sulfide, and containing a radioisotope material, such as gaseous tritium. The volume, shape, phosphor material, and radioisotope material are selected for emission of light at a particular wavelength and delivering a particular irradiance on the retina (when implanted in an eyeball). The wavelength is in the range of 400 to 600 nm and the irradiance is substantially 10⁹ to 10¹¹ photons per second per cm².

The invention provides a low-temperature displacement chromatography hydrogen isotope separation device and a method. The invention makes the mixed gas of hydrogen isotope gas and helium after being cooled sequentially pass through a cooled main separation column and a cooled product gas collecting column which are filled with granulate palladium-loaded aluminum trioxide (Al2O3/Pd) to obtain the product gas rich in heavy isotope components of deuterium and tritium; a hot helium flow passes through and heats the main separation column so as to make the released gas sequentially pass through the cooled secondary separation column and the product gas collecting column which are filled with granulate palladium-loaded aluminum trioxide (Al2O3/Pd) to obtain the product gas rich in heavy isotopecomponents of deuterium and tritium; after the product gas is collected, middle rich gas flowing out from the secondary separation column is directly fed back into a raw material gas tank; and after the middle rich gas feedback process is accomplished, the main separation column and the secondary separation column are heated, and the gas released by heating is collected via a tail gas collecting column. The separation device has simple structure, reasonable separation process, and low cost for the construction and operation of the device. The invention has higher separation factor for hydrogen isotope separation.

The invention relates to a hybrid reactor cladding for realizing long-term energy amplification by using a fast-thermal coupling mixed energy spectrum, which comprises a first wall, a fast fission region, a non fission neutron multiplication area, a thermal fissile area and a tritium breeding area. The structure of the hybrid reactor cladding is the first wall, the fast fission area, the non fission neutron multiplication area, the thermal fission area and the tritium breeding area in sequence from inside to outside. The cladding adopts the way that helium flows circumferentially and is cooled step by step. Helium enters into the cladding from a side pipeline of the first wall, enters into a fast neutron fission area transversely through the cladding after cooling the first wall, enters into the non fission neutron multiplication area and the thermal fission area from the other side of the cladding after cooling the fast neutron fission area and then enters the side pipeline of the first wall and flows out. Liquid lithium lead in the tritium breeding area enters into the cladding from the upper part of the cladding and flows out from the lower part and brings out heat and tritium relying on the flow. According to the hybrid reactor cladding, the long-term energy amplification, higher power generation efficiency and better safety can be achieved.

The invention belongs to the field of cladding designs of a fusion and fission hybrid reactor, and particularly relates to a Th-U self-sustaining circulating full fused salt fuel hybrid reactor system and an operation method thereof. The Th-U self-sustaining circulating full fused salt fuel hybrid reactor system is characterized in that a fast fission breeder reactor provides an initial easily fission fuel required by starting of a thermal fission reactor, a cladding design of the thermal fission reactor utilizes an arrangement strategy of a seed-cladding to improve the entirety neutron economy of the system, and the purposes of a high energy enlargement factor of the system, tritium breeding and thorium-uranium self-sustaining circulating of the system are realized; <233>U is loaded in an energy generating region, so that the energy generating region has a good neutronics property and is mainly used for realizing the purposes of energy amplification of the system, neutron multiplication and most <233>U breeding of the system; superfluous neutrons enter a tritium producing region and are used for tritium breeding and part of <233>U breeding of the system. According to the hybrid reactor, due to self-sustaining circulating of thorium and uranium, a thorium fuel is converted into <233>U and is gradually burn up in an operation process, and the Th-U self-sustaining circulating full fused salt fuel hybrid reactor system can effectively and stably operate for a long time just by gradually adding the thorium fuel and removing the generated fission product.

The invention provides a device and method for quickly recovering deuterium and tritium in fusion reactor discharged ash gases. The device is divided into a main path and a branch path according to effects and necessities of various units in treatment processes. The main path only comprises a cryogenic pump and a Pd alloy film separating unit, and 96% or more deuterium and tritium fuels in the discharged ash gases can be recovered by virtue of the two units. The method for quickly recovering deuterium and tritium in the fusion reactor discharged ash gases is short in treatment time, and better meets requirements for quick recovery of the deuterium and tritium fuels; and meanwhile, the recovery process is simplified, which is conducive to reduction of the retention volume of tritium in a system, thereby avoiding impurity gases from causing adverse effects on a Pd alloy film separator. Moreover, a small amount of deuterium and tritium not recovered by the main path are recovered by the branch path, thereby ensuring relatively high recovery efficiency. The device and method for quickly recovering deuterium and tritium in the fusion reactor discharged ash gases have the advantages of short treatment cycle, simple process, high recovery rate, good system stability, long service life and the like, and can meet design requirements of existing fusion reactors.

A fusion fuel capsule is disclosed having a substantially spherical ablator shell. The interior surface of the shell is lined with a nanoporous scaffold layer wetted with either a fully or partially liquid mixture of deuterium and tritium.

The invention belongs to an improved turn-off neutron source and in particular relates to a Penning anion source ceramic neutron tube. The Penning anion source ceramic neutron tube adopts a PIG type anion source to generate anions of deuterium (D) and tritium (T). The anion source adopts a lateral lead-out or eccentric lead-out mode; positive high voltage is added on a target; and under the condition of removing electrons from lead-out beams, the lead-out beams are monatomic anions and the utilization ratio of the beams generating neutrons can reach 100 percent. The neutron yield of the neutron tube with the structure is increased by about 2 orders of magnitude compared with that of the former neutron tube under the conditions of the same high voltage and beams. A tube of phi 50 is taken as an example, the life of the improved Penning anion source ceramic neutron tube is prolonged by more than 10 times, the yield is increased by about 100 times, the stability is not more than 3 percent and the rate of the finished products is more than 90 percent.

Method and apparatus for producing controlled steady state nuclear fusion with isotopes of low atomic numbers being the most useful reactants, such as Deuterium, Tritium and Helium3. The apparatus consists of a high voltage power supply and a high voltage spherical capacitor, constructed in such a way, that the outer shell is the anode and contained centrally within it, a hollow cathode, into which positive ions of the reactant gases can be injected through dielectric tubes and confined electrostatically within the cathode, until such high temperatures are reached, as to allow nuclear fusion to take place. The interior chamber of the cathode forms part of a hermetically sealed fuel circuit running through the capacitor, a turbo molecular pump is also connected in line with the fuel circuit, to drive the reactant gas through the reaction chamber The fusion product, which is mainly high energy Neutrons, Protons and alpha particles, is consequently converted to heat in the dielectric medium contained within the space between the anode and the cathode, this heat can easily be extracted and converted into useful energy using known methods.

A filter cartridge apparatus is provided for a respirator. The apparatus contains a wettable filter medium for use with a respirator in order to filter tritium, the filter medium being wetted when in use. The apparatus is also mountable to a cartridge containing a separate HEPA or other filter that is typically maintained in a dry state. Thus, a single unit containing both wet and dry filter media is provided for mounting to a respirator so that tritium or tritium water vapour and particulate contaminants can be filtered at the same time.

The invention discloses a radioactive wastewater treatment technology and the synthesis of a treatment agent of radioactive wastewater. A modified product Y3-8103 is used for chemically reacting with radioactive wastewater, all heavy metal elements and all radioactive elements, namely cesium137, uranium 235, thorium, plutonium, thallium, protium, deuterium, tritium, radon, Sr-90 and radium, listed in the periodic table of elements and wastewater generated by all artificial nuclides. Alpha radioactive rays, beta radioactive rays, gamma radioactive rays and neutron radioactive rays are attenuated in a nationally specified mode. The radioactive wastewater treatment technology and the synthesis of the treatment agent of the radioactive wastewater can also be used for treating radioactive seawater or waste residues generated by wastewater treatment and can be used for treating heavy metal wastewater discharged by a lead-acid battery enterprise and an electronic plant and mine nucleus waste residues. The treatment cost of the radioactive wastewater is different along with different radioactive element concentration contents or different waste residue radioactive element concentration contents and different atomic weights in the wastewater, and the treatment cost of the wastewater and the waste residues ranges from RMB 90 yuan/m<3> to RMB 680 yuan/m<3>, radioactive energy with the alpha, beta and gamma radioactive ray radiant quantity smaller than or equal to 0.01Bq/1-1.10Bq/1 can be discharged to the ambient environment by the wastewater and the waste residues, the emission limit standard allowed by the international is reached, and secondary pollution is avoided. The radioactive wastewater treatment technology and the synthesis of the treatment agent of the radioactive wastewater are suitable for treating wastewater and waste residues of a nuclear power plant and treating wastewater and waste residues in mining with the radioactive elements, and are also suitable for treating wastewater containing heavy metal, namely lead, cadmium, mercury, arsenic, chromium, nickel, copper and manganese.

The invention discloses a tritium isotope microcell and a preparation method thereof. The microcell comprises two parts: a transduction unit and an attached part. A Si3N4 layer of the transduction unit is at the upper surface of a silicon-based PN junction; and a ring electrode in the attached part is at epitaxy of the Si3N4 layer. A titanium deuteride + titanium tritide emitting layer is formed by the following steps that: evaporation of a gold layer and a nickel layer is successively carried out on the upper surface of the Si3N4 layer; and deuterium-tritium mixing gas is used to prepare the emitting layer. The gold layer is arranged at the lower surface of the silicon-based PN junction; and one end of a negative wire is connected on the gold layer and the other end of the negative wire is connected on a negative electrode. One end of a positive wire is connected to the ring electrode; and the other end of the positive wire is connected to a positive electrode. Besides, a ceramic chip is arranged below the gold layer and is arranged at the midpoint of a pedestal; and an outer encapsulated layer is used to cover the external portion of the pedestal. According to the invention, the tritium isotope microcell has a micro body size; when the microcell works, there is no need to provide energy from the external world; and uninterruptedly outputting over six years with current of 40 nA to 1 microamp and a power of 60 nW to 0.1 microwatt. Besides, the provided preparation method has advantages of security and reliability.

The invention relates to a high temperature gas cooled reactor tritium-contained wastewater optimization collection system. The system is composed of a first closed-loop circulation system and a second closed-loop circulation system. The first closed-loop circulation system is formed by sequentially connecting a helium purification regeneration system diaphragm compressor, a helium purification regeneration system electric heater, a helium purification system molecular sieve bed, a helium purification regeneration system water and helium cooler, a helium purification regeneration system gas and water separator and an auxiliary molecular sieve bed. The second closed-loop circulation system is formed by sequentially connecting a helium purification regeneration system diaphragm compressor, a helium purification regeneration system electric heater, an auxiliary molecular sieve bed, a helium purification regeneration system water and helium cooler, and a helium purification regeneration system gas and water separator. By the adoption of the system and technology, high temperature gas cooled reactor tritium-contained wastewater can be effectively collected, the running time of a whole high temperature gas cooled reactor helium purification system is prolonged, efficient running of the helium purification system is kept, and the system and technology have great significance in implementation of industrialization of a high temperature gas cooled reactor technology.

The invention provides a preparation method of a lithium-based ceramic microsphere, which is mainly used for preparing a tritium breeder in a fusion reactor blanket. The preparation method comprises the following steps: mixing a binder, water and ceramic powder to obtain uniform slurry by ball milling; adding a defoamer, heating in water bath and then forming sol; dripping into a medium mutually exclusive with water by a dispersing device; after the slurry is dispersed in the medium, forming micro-spherical gel; and washing, drying and sintering the microsphere so as to obtain the lithium-based ceramic microsphere. The preparation method of the invention has simple raw material, convenient forming and low cost, and the prepared ceramic microsphere has good sphericity and internal pore structure.

The invention discloses a tritium gas flow generator system with constant concentration and a method thereof. The tritium gas flow generator system comprises a closed loop circuit in which a tritium gas flow generator, a monitor for to-be-detected tritium gas and a tritium gas flow circulation pump are installed sequentially. The closed loop circuit is also provided with a temperature control unit matched with the tritium gas flow generator, an inlet end and an outlet end of the temperature control unit are both provided with a temperature sensor. An inlet end and an outlet end of the tritium gas circulation pump are both provided with a pressure sensor. A hygrometer is arranged between the temperature control unit and the monitor for to-be-detected tritium gas, and a nitrogen reservoir in communication with the tritium gas flow generator is also included for supplying the nitrogen reservoir with dry nitrogen. The tritium gas flow generator system achieves value traceability of a gas-borne tritium monitor, realizes accurate calibration of the gas-borne tritium monitor, and achieves gas-borne tritium monitoring of large-scale nuclear power devices, military isotope production reactors, military isotope content test sites, and military isotope packaging use sites.

The invention discloses a preparation method of a styrene-divinylbenzene copolymer hydrophobic catalyst support, which is characterized by comprising the following steps: adding deionized water, organic matter polyvinyl alcohol-1799, inorganic salt and sodium dodecyl benzene sulfonate into a reaction vessel, heating to 25-40 DEG C while stirring, adding a mixture of styrene, divinylbenzene, benzoperoxide, toluene, n-heptane and dichloroethane, heating to 70-90 DEG C, carrying out suspension polymerization reaction for 7-9 hours while stirring, and filtering; and washing the reaction solid with deionized water and anhydrous ethanol, extracting with acetone in a Soxhlet extractor for 12-48 hours, washing the solid with methanol and deionized water, drying, and screening to obtain the styrene-divinylbenzene copolymer hydrophobic catalyst support. The Pt/SDB hydrophobic catalyst prepared by using the product as the support has favorable use effect when being used for heavy water tritium extraction and wastewater detritiation.

The invention discloses a displacement chromatography hydrogen isotope separation device which comprises a rack as well as a cooling part, a separation column, an air supply part, an air collection part, a lifting part and a driving part. By adopting the displacement chromatography hydrogen isotope separation device disclosed by the invention, different adsorption functions that hydrogen and hydrogen isotope are replaced and separated by using a separation material can be achieved, the lifting part is driven by the driving part to precisely control the lifting position of the separation column, and then the separation column can be descended or lifted off the cooling part, so that adsorption or desorption of hydrogen isotope by the separation material under different temperature conditions can be achieved, a high retention rate of deuterium-tritium isotope in the separation material can be ensured, and a deuterium-tritium isotope gas obtained through final separation and selection is relatively high in purity; compared with the prior art, the device needs no parts for introducing deuterium or separating deuterium, meanwhile deuterium and tritium do not need to be replaced by introducing hydrogen after the separation material is in saturation adsorption of hydrogen, and in addition, the device is simple in structure, low in energy consumption rate, simple in process and low in cost.

The invention discloses a passive feedback structure suitable for controlling plasmas of a future fusion reactor rapidly. The passive feedback structure comprises a plurality of passive feedback structure independent blocks, the inner sides of the passive feedback structure independent blocks and the outer sides of the passive feedback structure independent blocks are provided with detachable cladding assemblies and permanent tritium breeding claddings respectively, and the passive feedback structure independent blocks are arranged evenly along the 360-degree angle in the large ring direction to form a structure with the overall appearance in the shape of a drum. The passive feedback structure independent blocks are composed of conductive copper sheets, stainless steel supporting base materials, cooling channels and side rib plates. The passive feedback structure can achieve rapid control over the plasmas and avoid instability of vertical displacement of the plasmas while meeting the requirements of the structure that a future fusion reactor increases a tritium breeding cladding, a divertor window and the like under the condition that space is limited, meanwhile, play the role of the supporting tritium breeding cladding structure, meet the requirements for long-term high-temperature resistance and irradiation, and guarantee stable operation of the plasmas of the future fusion reactor device.