83results about "Nuclear elements" patented technology

Embodiments of a ventilated storage system and method are provided for enabling space-efficient storage of casks containing hazardous nuclear materials. Each of a plurality of dry storage casks has a ventilated overpack with air inlets at or near the bottom and air outlets at or near the top so that ambient air flows into the air inlets, through the overpack, and out of the air outlets for venting and cooling the hazardous nuclear material. A storage pad on which the casks sit each have a plurality of air flow channels that provide air to the air inlets situated at the bottoms of the casks. The storage pad enables the casks to be placed closer together in a more space-efficient triangular array.

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.

The embodiment of the invention relates to the technical field of nuclear reactor testing, in particular to a loop type irradiation test device for a reactor, which is used for placing an irradiation sample in a reactor core of the reactor for an irradiation test. The device comprises a shell, the shell comprises a body part, an upper connecting part and a lower connecting part, the upper connecting part and the lower connecting part are connected with the body part, the upper connecting part is used for being installed on a reactor top cover of a reactor, and the lower connecting part is used for being installed in a reactor core vacancy of the reactor; the inner shell is arranged in the outer shell, and the irradiation sample is arranged in the inner shell; a cooling flow path is formed in the inner shell, and a coolant outside the reactor can enter the cooling flow path to cool the irradiation sample and then return to the outside of the reactor. According to the loop type irradiation test device provided by the embodiment of the invention, a large amount of heat does not enter the reactor basically, and the influence on the reactor is small; the cooling effect of the coolant on the irradiation sample can be controlled, it can be guaranteed that the temperature of the coolant outlet is not too high, and the design difficulty of equipment outside the device is lowered.

The invention provides a complex reactor core collaborative design system suitable for a research reactor, relates to the technical field of reactor core design systems, aims to improve the reliability of reactor core design, and comprises a model library module, a model database module, a model database module and a model database module, the model library comprises a reactor model library and an irradiation device model library; the interactive interface module is used for realizing interactive operation; the reactor core scheme design module is used for carrying out autonomous initialization design on a reactor core, and solving the problem of fuel assembly and target position loss caused by pore channel opening operation and the area demand of a newly added beryllium aluminum assembly caused by pore channel closing operation; and the reactor core scheme evaluation module is used for evaluating the reactor core. The method has the advantages of being intelligent in design, easy to operate and reliable in design quality.

A method of removing cesium vapor from a cover gas stream in a nuclear reactor includes the steps of oxidizing the cesium vapor in the cover gas stream to yield cesium oxide particles and removing the cesium oxide particles using a particle filter. The method yields a filtered cover gas having zero to about 2% of the cesium vapor content of the initial cover gas stream, representing a reduction of at least about 98 percent.

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 invention relates to the technical field of isotopes, in particular to a method and system for separating yttrium [90Y] in mother liquor containing strontium [90Sr]. The invention provides a method for separating yttrium [90Y] in mother liquor containing strontium [90Sr], which comprises the following steps: S1, adding the mother liquor containing strontium [90Sr] into a resin separation column, leaching with leacheate, collecting the leacheate containing strontium [90Sr], and emptying the liquid in the resin separation column with air; s2, eluting by using an eluent, and collecting the eluent containing yttrium [90Y]; wherein the total volume of the used leacheate is at least 11 times of the volume of the resin separation column. By means of the separation method, separation and recovery of yttrium [90Y] in the strontium [90Sr] mother liquor can be achieved, the recovery rate of strontium [90Sr] can be increased, the recovery rate of strontium [90Sr] can reach 98% or above, and the content of other metal impurities after recovery can be reduced.

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.

The present invention relates to a method and an apparatus for manufacturing lithium ceramic pebbles for tritium breeding materials and, more specifically, to a method and an apparatus for manufacturing lithium ceramic pebbles by using a fluid system. According to an embodiment of the present invention, core-shell pebbles are manufactured using a fluid system, thereby obtaining pebbles that supplement the shortcomings, such as unstable thermal stability and low mechanical strength and lithium density, of existing pebble raw materials. Additionally, according to an embodiment of the present invention, the production of pebbles through a fluid system can achieve low manufacturing process costs and construct mass production systems, leading to potential applications in nuclear fusion blankets in the future.

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 invention relates to a process for recovering one or more metals selected from technetium, ruthenium, palladium, and rhenium from an aqueous solution. The process comprises precipitating the metal(s) in the solution by adding a reagent to the solution, then collecting the precipitate thus formed. The process is characterized in that the reagent comprises—or is composed of—a mixture of 1,10-phenanthroline and a nickel(II) salt selected from nickel(II) nitrate and nickel(II) sulfate. Applications include: recovering all or part of the technetium, ruthenium, and / or palladium present in aqueous solutions of fission products prior to vitrification of the fission products; and recovering all or part of the ruthenium, palladium, and / or rhenium present in solutions obtained from the nitric acid treatment of industrial and / or municipal waste.

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.

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,

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.

A unique integrated system for recycling spent nuclear fuel (SNF) from a reactor includes a stripping tank to separate fuel pellets from nuclear fuel rods by oxidation to produce a uranium compound. A fluorination tank is connected to the stripping tank. A fluorinating agent is injected into the fluorination tank and reacts with the uranium compound to convert the uranium compound into UF6. An electrolytic extraction tank is connected to the fluorination tank to remove plutonium and actinides by electrolytic extraction. (Fig. 8)

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.

A radioactive nuclear waste storage system includes a cask comprising a hermetically sealed internal cavity configured for holding the waste such as spent nuclear fuel submerged in an inventory of water. One or more pressure surge capacitors disposed inside the cask include a vacuum cavity evacuated to sub-atmospheric conditions prior to storage of fuel in the cask. At least one rupture disk seals a vacuum chamber inside each capacitor. Each rupture disk is designed and constructed to burst at a predetermined burst pressure level occurring inside the cask external to the capacitor. This allows excess cask pressure occurring during a high pressure excursion resulting from abnormal operating conditions to bleed into capacitor, thereby returning the pressure inside the cask to acceptable levels. In one embodiment, the capacitors are located in peripheral regions of the cask cavity adjacent to the circumferential wall of the cask body.

The invention relates to the technical field of fuel element reactor irradiation tests, and provides a fuel element transient irradiation device which comprises a heat insulation pipe, a pressure pipe and a shunt pipe which are coaxially connected from outside to inside, the heat insulation pipe is communicated with a gas inlet pipe and a gas outlet pipe, the pressure pipe is communicated with a liquid inlet pipe, and the shunt pipe comprises an outer pipe and an inner pipe. The top of the inner pipe is communicated with a liquid outlet pipe, the lower portion of the inner pipe is communicated with a fuel element, a neutron absorption screen surrounds the fuel element, two end ports of the neutron absorption screen penetrate out of the side wall of the pressure pipe, a thermocouple is arranged between the outer pipe and the inner pipe, the measuring end of the thermocouple is arranged around the fuel element through a lead pipe, and the lead end of the thermocouple penetrates out of the shunt pipe. Through the multi-layer sleeve structure composed of the heat insulation pipe, the pressure pipe and the shunt pipe and the reasonable layout of the gas inlet pipe, the gas outlet pipe, the liquid inlet pipe, the liquid outlet pipe and the neutron absorption screen, the transient irradiation test requirement of the fuel element is met, the complexity of the structure is reduced, and the integration level of the structure is improved.

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.

A method of segmenting a calandria within vault of a nuclear reactor is provided. The method comprises: providing at least one cutting tool; cutting a portion of the vault to form an opening; forming a shielded volume defined by a shielding wall, the vault, and end shields of the calandria, the shielding wall configured to absorb radiation from within the shielded volume; segmenting a shell and internal components of the calandria within the shielded volume; and removing the shell and internal components of the calandria from the shielded volume. Optionally, a filler, such as concrete, may be used to immobilize the calandria. Once the calandria shell and internal components are removed the end shields may be segmented or detached from the vault and transferred intact by positioning the tube sheet potion of each end shield against each other.

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.