14results about "Conversion in nuclear reactor" patented technology

This invention discloses a method for producing radioactive isotopes in the core of an advanced high-temperature reactor using a spherical target. This invention overcomes the limitations of using neutron irradiation channels or fuel-related components of nuclear reactors, such as conditioning rods in heavy water reactors and combustible poison rods in pressurized water reactors, to produce radioactive isotopes. It sets a precedent for the mass production of radioactive isotopes using the core of a nuclear reactor, and has significant strategic implications for exploring high-value applications of commercial advanced high-temperature reactors to reduce operating costs and mass-produce radioactive isotopes, including the strategic isotope tritium.

An irradiation target is formed into a sphere. The spherical irradiation target can be iridium metal (10) containing natural or enriched iridium 191. The radiation source of the present invention can be manufactured by manufacturing a spherical irradiation target, accommodating the spherical irradiation target in a rotating capsule (20), and rotating an axial flow impeller (32) by a downward flow of a reactor primary coolant, whereby the rotating capsule (20) is rotated. This radiation source provides an improved nondestructive inspection image having a high geometric resolution, and has no radiation source anisotropy and also has high target recyclability.

The invention relates to a target transport system (1) for a radionuclide generation system, comprising: - a conduit (5) for transporting irradiation targets (7) through the radionuclide generation system; - a target drive system (9) configured to drive the irradiation targets (7) through the conduit (5); - a counting system (12) configured to count the number of irradiation targets (7) passing through the conduit (5). The counting system (12) comprises at least two distance sensors (15), each distance sensor being configured for measuring a radial distance (d1, d2) between the distance sensor (15) and an irradiation target (7), and being arranged at the same axial position along the conduit (5) and regularly spaced at a constant angular distance along the circumference of the conduit (5).

Disclosed are galvanic cells and methods for treating a molten salt by use of the galvanic cells. The galvanic cell can provide anti-corrosion treatment and improved corrosion resistance via electrochemical and chemical reactions involving a corrosive impurity in a molten salt. Reaction products of the chemical and electrochemical reactions include hydrogen gas that can carry hydrogen isotopes evolved in the molten salt. The system can also include removal of the hydrogen gas from the molten salt and separation and recovery of tritium contained therein.

A radioisotope production capsule is described. Each capsule includes generally an inner container for housing one of a target material and a neutron moderator, an outer container surrounding the inner container for housing the one of the target material and the neutron moderator not housed by the inner container, and cladding for isolating the target material from the neutron moderator. One or more modular capsules are placed in each of a plurality of plug fingers. Each single plug finger loaded with one or more capsules is inserted into a guide thimble of an array of guide thimbles in a fuel assembly.

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

This invention relates to a method for controlling the activity ratio of terbium-160 to terbium-161 in the preparation of terbium-161, comprising: S1, placing a natural Gd target or a high-abundance 160Gd or 158Gd target in a reactor for irradiation, then removing and cooling the target, dissolving the Gd target, analyzing the radioactivity of 161Tb ATb-161 and 159Gd AGd-159 in the solution, and calculating the nuclear reaction rates Øthermal σGd-160 and Øthermal σGd-158; S2: Place the natural Tb target in the reactor for irradiation, then remove and cool it. Dissolve the natural Tb target and analyze the radioactivity of 160Tb in the solution (ATb-160). Calculate the nuclear reaction rate Øthermal σTb-159. S3: Based on the nuclear reaction rate data obtained in S1 and S2, calculate the relationship between the irradiation time of a high-abundance 160Gd (greater than 90%) target (containing 158Gd and 159Tb impurities) and the 160Tb / 161Tb activity ratio generated.

A method for processing spent nuclear fuel includes the steps of: boroxidation treatment of spent nuclear fuel pellets to form spent nuclear fuel powder; dissolving the spent nuclear fuel powder in a soluble acid to form a main process stream; extracting uranium (U), plutonium (Pu), neptunium (Np), and technetium (Tc) from the main process stream to form a first extraction stream; separating the first extraction stream into a U / Pu / Np stream and a U / Tc stream; separating the U / Pu / Np stream into a mixture of Np and U / Pu; separating one or more minor actinides and one or more lanthanides from the main process stream to form a second extraction stream; separating one or more minor actinides from one or more lanthanides in the second extraction stream; and separating one or more precious metals from the main process stream.

The application belongs to the technical field of target welding and sealing, and aims to solve the technical problems of complicated operation, harsh environment requirement, low production efficiency and insufficient safety in the existing target packaging process. A medical radioactive isotope irradiation target welding and sealing device and method are disclosed. The device comprises a box-type closed cavity and a front door thereof. The inner wall of the box-type closed cavity and the front door are provided with a metal shielding layer and a neutron and gamma ray shielding layer. The bottom of the box-type closed cavity is provided with a target placing base for fixing and placing the target. The device has a vacuum system, an inert gas filling system, an automatic control system, a semi-automatic process control system and a pressure recovery and safety release system. The method uses the device to perform medical radioactive isotope irradiation target welding and sealing processing. The application can automatically complete the two key processes of frit sealing of the quartz target bottle and welding of the metal shell in a controlled inert atmosphere and a closed shielding environment, and realizes the automation and integration of the preparation process.

According to an aspect, there is provided a medical isotope element for producing medical isotopes during fuel burnup in a nuclear reactor. The medical isotope element includes at least one isotope target precursor, wherein the isotope target precursor converts to a medical isotope when exposed to radiation. In some embodiments, the medical isotope element is part of a fuel bundle. In some embodiments, the element is part of an assembly free of fissile material. According to an aspect, there is provided a method of producing medical isotopes during fuel burnup in a heavy water nuclear reactor. The method includes providing a fuel bundle or the assembly free of fissile material in the heavy water nuclear reactor, irradiating the fuel bundle or the assembly, removing the fuel bundle or the assembly from the heavy water reactor, and harvesting the medical isotope from the fuel bundle or the assembly.

A method for Mo-99 irradiation production and energy spectrum optimization based on reactor activation is proposed. This method establishes a three-dimensional high-fidelity neutronics and burnup coupling model of the reactor and irradiation channels. The full neutron energy range is divided into 238 energy regions, and perturbations are applied to the flux of each individual energy region. This yields the importance indicators and curves of Mo-99 production on flux changes in each energy region, thereby identifying positively and negatively important performance regions. For negatively important performance regions, filter nuclides are screened from the absorption cross-section characteristics of natural isotopes, and burnup calculations are performed at different doping levels to obtain production gains. The isotope scheme is then converted into a natural abundance material scheme to determine the optimal filter material and loading method. This invention, through integrated energy spectrum optimization—refined energy spectrum analysis, filter nuclide screening, determination of the optimal energy spectrum control scheme, and impurity assessment—can increase Mo-99 production without adding extra post-processing steps and reduce the risk of radioactive impurity introduction. It is applicable to the industrial and regional supply of Mo-99 for various reactor platforms, including commercial and research reactors.

Device for reducing spent nuclear fuel, comprising a container for receiving nuclear fuel, wherein the device is configured to transmute long-lived radioisotopes of the nuclear fuel into radioisotopes having a significantly shorter half-life than the initial radioisotopes by interaction with fast neutrons.

Device which is configured to hold a nuclear fuel and has one or more neutron sources which are configured to emit free neutrons which can interact with isotopes of the nuclear fuel, wherein the one or more neutron sources are configured such that the neutron flux strength of the generated neutrons is more than 10 12 neutrons per second and the one or more neutron sources have a power supply by means of which the neutron flux strength can be controlled.

A containment system for an integral molten salt reactor includes a sealed containment structure defining a containment volume. The containment structure includes a fission product adsorbing / absorbing material. The containment system further includes an integral molten salt reactor housed fully within the containment volume of the sealed containment structure. The integral molten salt reactor may permit circulation of a fuel salt therein being heated by fission reactions, and allow for the export of heat from said fuel salt.