41results about "Nuclear elements" patented technology

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.

Techniques are described for producing actinium-225 by arranging a converter material at least partially around a target material, where the target material includes radium-226. The converter material comprises a compound of lithium and hydrogen at least partially enriched with lithium-6 and deuterium, and is configured to convert thermal neutrons to fast neutrons, which are then incident on the radium-226. If the fast neutrons have a kinetic energy of at least 6.4 MeV, they will initiate a (n,2n) reaction that converts radium-226 into radium-225, which then decays into actinium-225. The process can desirably be performed within an environment rich in thermal neutrons, such as within a fission reactor.

The gas cooled reactor nuclear power station with the fuel element capable of being purged comprises a fuel pipe, the fuel pipe is used for introducing the fuel element into the nuclear power station, an air inlet and an air outlet which are oppositely arranged in the vertical direction are formed in the pipe wall of the fuel pipe, the air inlet is located above the air outlet, and the included angle between the extending direction of the fuel pipe and the vertical direction is larger than or equal to a preset value; an outlet of the air supply pipe communicates with the air inlet, and the outlet of the air supply pipe faces the air outlet in the vertical direction; the air outlet pipe extends in the vertical direction, an inlet of the air outlet pipe communicates with the air outlet, the inlet of the air outlet pipe is formed in the upper end portion of the air outlet pipe, and the upper end portion of the air outlet pipe is matched with the pipe wall of the fuel pipe; the dust collector is provided with a dust inlet, a gas outlet and a dust outlet, and the dust inlet is communicated with the outlet of the air outlet pipe. Therefore, the gas cooled reactor nuclear power station capable of purging the fuel element has the advantage that dust on the surface of the fuel element can be removed so that operation can be stable.

A method and system (300) for separating thorium from uranium and its decay products is provided. The method includes combining a nuclear fuel feedstock (102) containing thorium and uranium with a first acid (104, 204) to form a first solution. The first solution is contacted with an ion exchange resin (308) that is selective for thorium or uranium. The thorium or uranium is at least partially removed from the first solution by binding the thorium or uranium to the ion exchange resin (308), thereby forming a second solution (106, 206). The second solution (106, 206) is combined with oxalic acid to precipitate the uranium or thorium from the second solution, forming a precipitate (110, 210). The precipitate (110, 210) is separated from the second solution (106, 206).

A fuel fragment screening device of a nuclear fuel loading and unloading system in a gas-cooled high-temperature reactor, comprising a box body, a spiral screen channel, a fuel ball inlet, a fuel ball outlet, and a waste outlet; the box body is arranged on the inlet line of the baffle, with the fuel ball inlet provided on the top of the box body and the fuel ball outlet and the waste outlet provided on the bottom; inside the box body, between the fuel ball inlet and the fuel ball outlet, the spiral screen channel is provided, and the spiral screen channel has a certain angle and is in an inclined state, with the waste outlet being located below the spiral screen channel.During the operation of the present application, when the fuel balls, fuel ball fragment, and the dust mixture pass through the device, the device can completely separate the intact fuel balls from the fuel ball fragment and the graphite dust from the graphite ball fragment, thus effectively preventing the fragment and dust from entering the baffle and affecting its normal operation, thereby improving the normal operation of the fuel loading and unloading system and increasing the reliability of the system operation.

The present disclosure provides various methods, systems, and devices for producing actinium-225 (Ac-225). In one aspect of the disclosure, a method for producing Ac-225 includes inserting an irradiation target assembly into a commercial nuclear reactor. The irradiation target assembly comprises a radium-226 (Ra-226) material. The method further includes producing electricity using the commercial nuclear reactor by generating a neutron flux in the commercial nuclear reactor. The method further includes irradiating the Ra-226 material with the neutron flux generated in the in the commercial nuclear reactor to produce a radium-225 (Ra-225) material. The Ra-225 material decays to produce an Ac-225 material.

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 utility model relates to the technical field of nuclear reactor thermal hydraulics, and particularly discloses a multi-rod-bundle critical heat flux density experimental device, which comprises a shell and heating rods, the shell comprises a first electrode cylinder, a water inlet section, an experimental section, a water outlet section and a second electrode cylinder which are sequentially arranged from left to right, and all the sections are hermetically connected through flanges; shells of the water inlet section, the experiment section and the water outlet section comprise inner shells and pressure-bearing shells, flow channels are formed in the inner spaces of the inner shells, and interlayers are formed between the inner shells and the pressure-bearing shells; the multiple heating rods are located in the flow channel, and heating rod sealing discs used for supporting and limiting the heating rods are arranged in the water inlet section and the water outlet section correspondingly. The first electrode cylinder and the second electrode cylinder are provided with conductive columns used for heating the heating rod in the axial direction of the shell, and the conductive columns extend out of the shell. The two sides of the experiment section along the axial direction of the shell are provided with outlet pipes which are respectively connected with a temperature sensor and a pressure transmitter. The utility model aims to solve the problem that the existing critical heat flux density experimental device is not suitable for experimental research in a multi-rod-bundle high-pressure environment.

Techniques are described for producing radium-225 from a sample containing radium- 226, and milking the sample for actinium- 225, which results in a higher yield of actinium-225 for a given sample in a given period of time compared with conventional approaches. In particular, returning the sample to the reactor while there is still a significant amount of radium-225 in the sample can, over time, produce more actinium- 225 than the conventional approach of milking the sample until the vast majority of radium-225 has decayed into actinium-225.

The present disclosure relates to nuclear chemical, particularly radiochemical, technologies at different stages of the nuclear fuel cycle, such as the production of purified nuclear materials (uranium, zirconium) or the reprocessing of spent nuclear fuel from nuclear power stations, in which extraction processes and operations for purifying nuclear materials are used. An example method, which includes the partial decomposition of nitric acid during continuous evaporation while a solution containing a reducing agent is fed into the bottom part of an evaporator having a circulating bottoms solution, consists in that the process is carried out such that the solution is kept in the bottom part of the evaporator for more than 2 hours under the addition of an aqueous solution of formaldehyde and formic acid (hereinafter “the mixture”) or a solution of formic acid after the process has been started using the mixture.

Disclosed is a method for dismantling a D2O tank of a heavy water reactor installed on an inner wall of a containment building and surrounded by a concrete structure, the method including cutting and opening a front surface of the concrete structure so as to dismantle the D2O tank, descending the D2O Tank by using an dedicated dismantling apparatus, and taking the D2O tank out of the containment building by using a crane, and a dismantling apparatus.

A method and a system for separating yttrium [90Y] from a mother liquor containing strontium [90Sr]. The method comprises the following steps: S1: adding a mother liquor containing strontium [90Sr] to a resin separation column, performing rinsing using a rinsing solution, collecting a rinsing solution containing strontium [90Sr], then emptying liquid from the resin separation column using air; S2: using an eluent solution to perform elution, and collecting an eluent solution containing yttrium [90Y], a total volume of the eluent solution used being at least 11 times the volume of the resin separation column. By means of said separation method, it is possible to perform separation and recovery of yttrium [90Y] from a mother liquor containing strontium [90Sr]. In addition, it is possible to increase a strontium [90Sr] recovery rate, bringing the strontium [90Sr] recovery rate to 98% or higher, as well as decreasing the content of other metal contaminants after recovery.

Provided is a chemical decontamination method that shortens the decomposition time of a reduction decontamination agent. An oxidization decontamination, a decomposition of an oxidization decontamination agent, and reduction decontamination using an oxalic acid aqueous solution are performed on a target piping of a BWR plant. After that, the oxalic acid is decomposed (S7). That is, a part of the oxalic acid is decomposed by irradiating the oxalic acid aqueous solution with ultraviolet rays upstream of a decomposition device (S8), and Fe3+ in the aqueous solution is converted to Fe2+. Hydrogen peroxide is supplied to the decomposition device (S9). In the decomposition device, the oxalic acid is decomposed by a catalyst and hydrogen peroxide, Fe2+ and hydrogen peroxide react to produce Fe3+ and OH*, and the oxalic acid is decomposed by OH*. A corrosion potential of the aqueous solution flowing out from the decomposition device is measured (S11). A concentration ratio calculation device obtains Fe3+ / Fe2+ (concentration ratio) based on the corrosion potential (S12), and A control device controls the supply amount of hydrogen peroxide to the decomposition device based on Fe3+ / Fe2+ (S14 and S16).).

The application provides a complex reactor core collaborative design system suitable for research reactors, relates to the technical field of reactor core design systems, and aims to improve the reliability of reactor core design, and comprises a model library module, a reactor core scheme design module, a reactor core scheme evaluation module and an interactive interface module. The model library module comprises a reactor model library and an irradiation device model library, and is established based on different research reactors and various types of irradiation test pieces in the reactors. The reactor core scheme design module is used for independently initializing and designing the reactor core, solving the problems of missing fuel assemblies and target piece positions caused by channel opening operation, and solving the regional demand of newly added beryllium aluminum assemblies caused by channel closing operation. The reactor core scheme evaluation module is used for evaluating the reactor core. The application has the advantages of intelligent design, easy operation and reliable design quality.

Described herein, are methods for providing a protective coating to a storage container for storing nuclear material, the method comprising depositing nickel particles on at least one surface of the substrate to produce the protective coating, wherein the nickel particles are deposited by cold spraying a composition comprising nickel particles and a carrier gas comprising nitrogen. In one aspect, the carrier gas consists essentially or consists only of nitrogen. The methods do not require pretreatment or modification of the nickel particles prior to cold spraying, which makes the methods described herein economically practical. The coatings produced by the methods described herein possess several advantageous properties including, but not limited to, high adhesion strength to the storage system and low porosity. The coatings produced by the methods described herein are effective against chemical attack such as, for example, CISCC.

Disclosed is a method for dismantling metal equipment of a heat transport system in a containment building at a nuclear power plant, the metal equipment including a heat transport pump, a steam generator, a pressurizer, a condenser, and a D2O tank, including: a first dismantling step of dismantling the heat transport pump; a second dismantling step of dismantling the steam generator after the first dismantling step; a third dismantling step of dismantling the pressurizer after the second dismantling step; and a fourth dismantling step of dismantling the condenser and the D2O tank after the third dismantling step.

Disclosed is a method for dismantling a D20 tank of a heavy water reactor installed on an inner wall of a containment building and surrounded by a concrete structure, the method including cutting and opening a front surface of the concrete structure so as to dismantle the D20 tank, descending the D20 Tank by using an dedicated dismantling apparatus, and taking the D20 tank out of the containment building by using a crane, and a dismantling apparatus.

A flow rate formation method for system decontamination of a heavy-water reactor nuclear power plant is provided. A nuclear power plant includes: a reactor; a first coolant pump disposed on one side of the reactor, a second coolant pump disposed on the other side of the reactor, and first coolant system piping passing through the reactor and connecting the first coolant pump and the second coolant pump; a first shutdown cooling pump disposed on one side of the reactor, and first shutdown cooling system piping branching from the first coolant system piping at a downstream end of the first coolant pump and connected to an upstream end of the first coolant pump through the first shutdown cooling pump; and a second shutdown cooling pump disposed on the other side of the reactor, and second shutdown cooling system piping branching from the first coolant system piping at a downstream end of the second coolant pump and connected to an upstream end of the second coolant pump through the second shutdown cooling pump, wherein the first coolant system piping, the first shutdown cooling system piping, and the second shutdown cooling system piping are decontaminated simultaneously by operating the first coolant pump and the second shutdown cooling pump and stopping the second coolant pump and the first shutdown cooling pump.

Disclosed is a filter replacement device which can replace a filter in a radioactivated structure. The filter replacement device comprises: a body; an articulated manipulator provided at one side of the body and having five degrees of freedom; a tool detachably provided at the end of the manipulator so as to perform all operations for replacing a filter in a target object; and a socket detachably provided on the tool.

Drilling device (10) for machining tubes (66) on tube sheets (64) of heat exchangers in a radioactive environment, comprising a transport device (12) having clamping elements (18, 24), characterized in that a drilling device (16) having clamping fingers (30) is held by the transport device (12), that the clamping elements (18, 24) and the clamping fingers (30) are arranged on a common first side of the drilling device (16) and the transport device (12), that the transport device (12) and the drilling device (16) are connected to a support device (14), that the support device (14) has a support plate (50) on which a support plate (28) of the drilling device (16) rests, that the support plate (50) is connected to the support plate (28) by at least one movable connecting element (62), that the support plate (28) is connected to the support plate (50) in a first position of the connecting element (62) is connected without play,that the support plate (28) has a predeterminable play relative to the carrier plate (59) in a second position of the connecting element (62).

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.

In the medical and pharmaceutical fields, radioisotopes may be valuable in imaging and therapeutic aspects. Various irradiation systems are known for use within a nuclear reactor facility to produce radioisotopes. However, current systems face numerous challenges, such as impeded field of view in a viewport, and inability to accurately locate a target in the system. In some embodiments, an isotope irradiation system for use in a nuclear reactor facility is provided. The isotope irradiation system may include a removable support structure, an interface system, a cartridge loading and retrieval system, and an isotope target transport system. The disclosed isotope irradiation system can address challenges associated with conventional systems.

An apparatus comprises a cell with one or more wall panels, a floor, a ceiling, and one or more openings. The cell contains one or more components comprising a dexterous manipulator with at least one articulating robotic arm and a manipulator carriage. The manipulator carriage is coupled to rails and may traverse the cell length. The cell contains a boom with a proximal end, a distal end, and articulating links comprising two or more segments that fold back onto each other within the cell or extend partially or fully through one of the one or more cell openings. The cell also contains a boom carriage (coupled to the proximal end of the boom), boom rails, and a base. The boom carriage may traverse the boom rails in the cell or remain fixed to the base. One or more end effectors are coupled to the distal end of the boom.

In the field of medicine and pharmacy, radioisotopes may be valuable for imaging and treatment. Various irradiation systems have been known to be used inside nuclear reactor facilities to create radioisotopes. However, current systems face challenges such as obstructed view in the viewport, and an inability to positively locate targets in the system. In some embodiments, an isotope irradiation system (150) is provided for use in a nuclear reactor facility. The isotope irradiation system may include a removeable support structure (140), an interfacing system (152), a cartridge load and retrieval system (160), and an isotope target transportation system (182). The disclosed isotope irradiation system may address the challenges associated with conventional systems.

The application discloses a research reactor overall design system and method, which realizes the whole-chain optimization of the research reactor overall design from demand input to parameterization and from single-discipline calculation to multi-specialty cooperation through a closed-loop mechanism of demand traction, two-way coupling of the reactor core and irradiation application, cross-disciplinary iterative optimization and comprehensive performance evaluation. Compared with the prior art, the application can significantly improve the efficiency and reliability of the research reactor overall design, ensures that the design result meets the scientific research and application tasks, the safety and economic requirements, and thus guarantees the multifunctional adaptability and global optimization of the research reactor in the early design stage.

The present invention relates to a method for cutting a nuclear reactor and inserting cut materials, the method comprising the steps of: preparing cut materials by cutting a nuclear reactor through a plurality of processes; classifying the cut materials into a plurality of stack shapes; and inserting the cut materials into a packaging container according to the stack shapes, wherein the packaging container includes an external container, an internal container and a cover, and the inserting step includes the steps of: preparing the external container; accommodating the internal container in the external container; and mounting the cover on the upper portion of the internal container.

Title: A Process for Treating Plutonium Retained in the Waste Organic Phase of the PUREX Process The present invention relates to a process for treating plutonium retained in a waste organic phase of a PUREX process. The waste organic phase of the PUREX process comprises an organic solvent and plutonium, the process comprises: contacting the waste organic phase of the PUREX process with an aqueous phase re-extraction solution containing 2,6-pyridinedicarboxylic acid to perform re-extraction, to obtain a plutonium re-extraction product; the plutonium in the re-extraction solution may be loaded onto a column for adsorption, transformation and desorption by an anion exchange column to recover the plutonium in the re-extraction solution.The method can effectively elute and recover plutonium in the waste organic phase, especially can effectively elute plutonium in the waste organic phase which has been stored for a long time. Figure for abstract: Fig.1.

Techniques are described for producing actinium-225 by arranging a converter material at least partially around a target material, where the target material includes radium-226. The converter material comprises a compound of lithium and hydrogen at least partially enriched with lithium-6 and deuterium, and is configured to convert thermal neutrons to fast neutrons, which are then incident on the radium-226. If the fast neutrons have a kinetic energy of at least 6.4MeV, they will initiate a (n,2n) reaction that converts radium-226 into radium -225, which then decays into actinium-225. The process can desirably be performed within an environment rich in thermal neutrons, such as within a fission reactor.

The present invention provides a simulated spent fuel storage tank for nuclear power station including a simulated storage tank, simulated spent fuel assemblies received in the simulated storage tank, and an electric heater, wherein the simulated spent fuel assemblies are electrically connected to the electric heater via cables. The simulated spent fuel storage tank for nuclear power station of the present invention can be used for thermal property identification and operation training test. The arrangement state and the thermal power of the spent fuel assemblies can be precisely simulated according to the actual loading and storage quantity and the heat release quantity of the spent fuel assemblies, and the thermal property identification of the spent fuel storage tank is achieved. The simulated spent fuel storage tank for nuclear power station of the present invention can also be used to simulate vacuum drying, automatic welding, and cutting and retrieval operations in a real heating environment, to simulate a real spent fuel assembly load in order to perform storage tank transfer and storage operations, such that the skill level of operators can be improved, thereby preventing the occurrence of incidents caused by human factors.

Method for repairing a steam generator tube of a nuclear reactor, and corresponding steam generator. The tube (5, 115) is integral with an outer casing (3) of the steam generator (1) and includes an outer end (7) connected to an outer conduit (9, 118) for the circulation of a secondary heat transfer fluid, the tube (5, 115) having an inner passage (15) delimited by an inner surface (17), the steam generator (1) including an inner conduit (13, 117) housed inside the outer casing (3) and a thermal sleeve (19, 119) having a proximal end (21) connected to the inner conduit (13, 117) and a distal end (23) engaged in the inner passage (15) of the tube (5, 115), such that the inner passage (15) is in fluidic communication with the inner conduit (13, 117) through the thermal sleeve (19, 119),the distal end (23) extending against or in the immediate vicinity of an annular area (35) of the inner surface (17) of the tubing (5, 115). The method includes a step of introducing a protective sleeve (45) into the inner passage (15), the protective sleeve (45) being pressed against the inner surface (17) and covering the annular area (35). Figure for the abstract: 4,