1418 results about "Nuclear fuel" patented technology

One variation of a system includes a pressure vessel: including a wall; defining a primary working fluid circuit extending vertically within the wall; defining a secondary working fluid circuit extending vertically within the wall and fluidly isolated from the primary working fluid circuit; and configured to store a nuclear fuel, a primary working fluid, and a secondary working fluid. The wall of the pressure vessel: defines a heat exchanger configured to transfer thermal energy from the primary working fluid flowing through the primary working fluid circuit into the secondary working fluid flowing through the secondary working fluid circuit; and defines a radiation shield configured to attenuate radiation emitted by the nuclear fuel.

The present application relates to nuclear fuel claddings, and more particularly to zirconium alloy tubes with a ceramic covering. A first coating layer is applied to the ceramic covering. A second coating layer is over the first coating layer. The first coating layer includes Nb. The second coating layer includes Cr.

The invention relates to the technical field of deformation monitoring, in particular to a nuclear fuel cladding tube biaxial creep deformation monitoring system which comprises an acquisition monitoring module, a deformation recognition module, a sample aggregation module, a deformation training module and a risk early warning module. According to the method, the deformation abrupt change section is screened out on the basis of multiple parameters, the same-direction change trend of each physical parameter in the same section is further extracted, the collaborative change fragment is constructed, and the physical inducement of the cladding tube structure abrupt change can be accurately locked. On the basis, quantitative analysis is carried out on the time difference between response inflection point time and deformation rate peak time of different parameters, a difference value mode reflecting system response lag is formed, and the time dynamic sensing capacity of deformation behaviors is enhanced. According to the method, samples with different lag characteristics are extracted, a linkage relation point set among parameters is reconstructed, and a multi-dimensional response interval is constructed in combination with Euclidean distance, so that clustering classification of deformation density and change trend is realized.

The invention relates to a plate type new fuel assembly transportation container which comprises a hollow cylindrical barrel, a first shock absorber, a second shock absorber, a first cover and a second cover, and the barrel is used for containing a new fuel assembly to be transported; the first shock absorber and the second shock absorber are arranged at the two ends of the cylinder in a sleeving mode respectively, and a multi-layer heat insulation structure and a steel pipe and steel plate wrapping design are adopted. The first cover body and the second cover body are located on the inner side of the top of the barrel to form a double-layer containing structure. The internal buffer assembly comprises a buffer gasket, a fireproof cloth sleeve and an accelerometer mounting part, provides dual protection for the fuel assembly, and meanwhile, dynamic changes in the transportation process are detected through an accelerometer to ensure that the fuel assembly is intact. The transportation container provided by the invention has excellent shock absorption, heat insulation and inclusiveness, is suitable for safe transportation of the plate-type new fuel assembly, and has a wide application prospect in the field of nuclear fuel transportation.

The invention discloses a method for preparing high-purity uranium carbide microspheres by a microfluidic-assisted internal gelation process, which comprises the following steps of: designing a uranium glue solution component containing carbon black to obtain a uranium glue solution which is stable for a long time at room temperature and quickly gelates at high temperature; the preparation method comprises the following steps: preparing gel microspheres with uniform size by using a uranium glue solution as a micro-fluidic dispersion phase through a micro-fluidic assisted inner gel process; the pore structure in the gel microspheres is maintained through washing and soaking with propylene glycol monomethyl ether, so that the crack-free gel microspheres are obtained; the UC nuclear fuel microspheres which are uniform in size, good in sphericity degree and high in phase purity are prepared through the carbon thermal reduction reaction of the non-cracking gel microspheres under the optimized sintering schedule, and development of ultrahigh-temperature gas cooled reactor nuclear fuel is expected to be promoted.

The invention discloses a prismatic TRISO particle dispersion additive fuel element as well as a preparation method and application thereof, and relates to the technical field of nuclear fuel. The method comprises the following steps: preparing the SiC shell by adopting a binder injection molding process or a photocuring molding process; the SiC shell is subjected to degreasing treatment; carrying out chemical vapor infiltration pretreatment on the SiC shell subjected to degreasing treatment; the pretreated SiC shell is filled with simulated TRISO particles and SiC matrix powder, high-solid-content SiC slurry is adopted for packaging, then an enhanced chemical vapor infiltration method is adopted for densifying the dried additive SiC element, and the prismatic TRISO particle dispersion additive fuel element is obtained. The TRISO fuel particles with high inherent safety and the nuclear-grade SiC ceramic material with excellent high-temperature resistance and irradiation resistance are combined, the additive manufacturing process is matched, the design flexibility and expansibility are improved, and the TRISO fuel particles have large application potential in the high-temperature gas cooled reactor.

The invention belongs to the field of protective coatings, and relates to a Cr-based protective coating with a multi-layer structure and a preparation method of the Cr-based protective coating. The Cr-based protective coating comprises a plurality of periodic layers, and each periodic layer is composed of a Cr layer and a Cr-Al-Si-N layer. The element composition of the Cr-Al-Si-N layer is CrxAlySizN1-x-y-z, x, y and z are atomic ratios, x is larger than or equal to 0.33 and smaller than or equal to 0.43, y is larger than or equal to 0.20 and smaller than or equal to 0.27, and z is larger than or equal to 0.04 and smaller than or equal to 0.08; and the Cr-Al-Si-N layer is of a two-phase composite structure in which nanocrystals are dispersed and distributed in compact noncrystals. The multi-layer structure coating provided by the invention combines the double advantages of radiation resistance and high-temperature oxidation resistance, shows good comprehensive protection performance, and can provide effective protection for the nuclear fuel cladding tube in a complex and severe reactor environment.

To provide a comparison method and a search method of strings to be optimized mainly focusing on Chinese search, based on a string search process of data information of a nuclear fuel management process.SOLUTION: These methods include steps of: splitting and preprocessing a target string and integrating a corresponding language library; calculating an editing distance and performing preliminary filtering using a character distance algorithm; and searching a Chinese string and verifying its collation degree using the Aho-Corasick algorithm, and caching frequently searched pattern strings after searching.SELECTED DRAWING: Figure 1

The invention relates to a preparation method of an irradiation-resistant and accident-resistant SiC / SiC nuclear fuel cladding. The preparation method comprises the following steps: (1) alternately carrying out the following step (A) and step (B) from step (A) to step (A) to obtain a tubular SiC fiber preform with at least two PyC interface layers and at least one SiC nanowire interface layer: (A) depositing the PyC interface layers on the surface of the tubular SiC fiber preform through a chemical vapor deposition process; (B) loading a catalyst on the PyC interface layer through vacuum impregnation, and generating a SiC nanowire interface layer on the surface of the PyC interface layer in situ through a chemical vapor deposition process; (2) carrying out densification treatment on the tubular SiC fiber preform by adopting a chemical vapor infiltration process; and (3) depositing a SiC coating on the surface of the tubular SiC fiber preform by adopting a chemical vapor deposition process to obtain the SiC / SiC nuclear fuel cladding tube.

One variation of a system includes: a nuclear reactor; a shield arranged about the nuclear reactor; a metallic coolant; and a set of melt seals. The nuclear reactor includes a pressure vessel and nuclear fuel arranged within a lower region of the pressure vessel. The metallic coolant includes a mixture of metals and is configured to: occupy a liquid state within an operating temperature range; occupy an interstitial volume between the nuclear reactor and the shield; and occupy the lower region of the pressure vessel encompassing the nuclear fuel. The set of melt seals are arranged on the pressure vessel and configured to open to enable transfer of a volume of the metallic coolant from the interstitial volume into the lower region of the pressure vessel in response to temperatures within the pressure vessel exceeding the operating temperature range.

The invention discloses a welding seam detection system and method for a lower end plug of a nuclear fuel rod, and relates to the technical field of nuclear power equipment manufacturing automation, the welding seam detection system comprises a data acquisition module and a data analysis module, and the data acquisition module comprises a camera, a movement mechanism and an endoscopic detection probe installed on the movement mechanism. A data analysis module is used for positioning a welding seam detection position on an environment image of a target to-be-detected lower end plug shot by a camera after a nuclear fuel assembly hovers, sending a motion control instruction to control a motion mechanism to bear an endoscopic detection probe to reach the welding seam detection position, and controlling the endoscopic detection probe to reach the welding seam detection position; the endoscope video is collected, and the small-size endoscopic detection probe goes deep into the weld joint detection position, so that the limitation of a tiny space is overcome, and the full-coverage collection of the weld joint image of the lower end plug to be detected is realized. Furthermore, the data analysis module is used for achieving welding seam anomaly detection based on endoscope videos, and the accuracy of anomaly detection results is improved.

A system is described that includes a sputter target and a magnetic element array including multiple sets of magnets arranged to have a Hall-Effect region that extends along a length of the sputter target. The elongated sputtering electrode material tube is interposed between the magnetic array and an object to be deposited with a sputtered material from the sputter target. During a direct current high-power impulse magnetron sputtering operation, the system performs a depositing on a surface of the object by generating and controlling an ion and neutral particle flux by: providing a vacuum apparatus containing a sputter target holder electrode; first generating a high-power pulsed plasma magnetron discharge with a high-current negative direct current (DC) pulse to the sputter a target holder electrode; and second generating a configurable positive voltage kick pulse to the sputter target holder electrode after terminating the negative DC pulse.

System and procedure for the inspection of the surface of a nuclear fuel rod for the automatic detection, localization and characterization of defects, comprising a plurality of point cloud acquisition systems and translation means for the nuclear fuel rod; where the system also comprises processing means capable of transforming the point cloud into a height map, locally thresholding for the segmentation of the height map and obtaining the discontinuities in said map, classifying the discontinuities in different geometric shapes to obtain the area, length, width, position and orientation, projecting the defect on the y-axis to obtain its depth and calculating the equivalent function and its second derivative to obtain the points where the depth of the defect is maximum.

The invention provides a nuclear fuel grillwork clamping device and a vibration acceleration control method, and relates to the technical field of nuclear fuel grillwork clamping devices.The nuclear fuel grillwork clamping device comprises the steps that a plurality of clamping pieces are slidably installed in a nuclear reaction box body in the first direction and connected with the multiple clamping pieces at the same time through a transmission mechanism, and the clamping pieces are connected with the nuclear reaction box body through the structural characteristics of the transmission mechanism; and the plurality of clamping pieces are enabled to vibrate in two opposite directions of the first direction. Based on the structural design, the mode that the clamping pieces are installed in the nuclear reaction box body is changed into sliding connection from fixed connection, and the multiple clamping pieces are connected at the same time through the transmission mechanism. In the nuclear reaction process, the transmission effect of the transmission mechanism can be utilized to enable the plurality of nuclear fuel grillages to vibrate along opposite directions; therefore, the vibration amplitude of the nuclear fuel grillwork is reduced, and finally the purpose of reducing the failure rate of the nuclear fuel grillwork is achieved.

Disclosed herein are methods and systems for the recovery of uranium from used fuel salts using nonaqueous chemistry without electrometallurgy techniques. The used fuel salts are maintained in a molten state in a chlorination chamber, and are then chlorinated for a period of time sufficient to convert at least some UCl3 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.

The invention provides natural graphite powder for a high-temperature gas cooled reactor nuclear fuel element and a preparation method of the natural graphite powder. The preparation method comprises the following steps: performing screening, crushing, shaping and grading treatment on graphite fine powder to obtain powder with the particle size of 25-35 microns; carrying out primary purification and secondary purification on the powder by using a mixed acid solution, heating the obtained chemically purified graphite powder in vacuum, carrying out high-temperature purification, introducing a mixed gas of argon and carbon tetrafluoride when the temperature reaches 1800 DEG C, and starting a circular reaction; then continuously carrying out gradient heating to 2000 DEG C, 2200 DEG C and 2400 DEG C, and respectively carrying out circular reaction at each temperature; then cooling to 300-400 DEG C, introducing mixed gas of oxygen and argon, and carrying out oxidation treatment; and finally, scattering and homogenizing to obtain a natural graphite powder finished product. The natural graphite powder provided by the invention realizes lower total impurity element content and total boron equivalent and higher specific surface area.

The invention discloses a nuclear fuel pellet out-of-pile cracking in-situ monitoring system and method. The system comprises an experiment box, a sample table, a first thermometer, a second thermometer, a blue light source, a spectrometer, an infrared camera, an optical camera and a graphic server. The spectrometer divides the light of the light guide measurement channel into visible light and infrared light which are respectively irradiated to the optical camera and the infrared camera. Since the first temperature collected by the first thermometer and the second temperature collected by the second thermometer are more accurate temperatures, the infrared photo can be calibrated, and more accurate temperature distribution can be obtained. The cracking morphology of the pellet sample can be captured in real time by a picture shot by the optical camera, so that the temperature field of the infrared picture and the cracking morphology of the optical picture are synchronously matched, non-contact temperature-optical coupling in-situ measurement in a pellet sample out-of-pile heating experiment is realized, the influence of a measuring instrument on the cracking process of the pellet sample is reduced, and the measurement accuracy is improved. The research on the evolution mechanism of the cracking of the pellet sample under different thermal stress conditions is facilitated.

The present invention relates to a nuclear fuel assembly suitable for a small modular reactor (SMR), the nuclear fuel assembly comprising: a spacer grid (110) having 17x17 grid cells; a plurality of fuel rods (120) arranged in the grid cells of the spacer grid (110); guide tubes (130) arranged and fixed in the grid cells of the spacer grid (110); and an upper end fitting (140) and a lower end fitting (150) fixed to the upper end and the lower end of the guide tubes (130), respectively, wherein the guide tubes (130) are characterized in that eight guide tubes are arranged in the main axis directions (Ixx)(Iyy) with respect to the central cell of the spacer grid (110), four guide tubes are arranged in the diagonal direction (Ixy), and sixteen guide tubes are arranged in the non-diagonal directions (Ixxy)(Ixyy).

The invention mainly provides a nuclear fuel cladding composite material and a preparation method thereof. The nuclear fuel cladding composite material comprises a zirconium alloy matrix, the niobium-rich diffusion barrier layer is formed on the surface of the zirconium alloy matrix; the Cr coating covers the niobium-rich diffusion barrier layer; and an Al-rich layer formed on the surface of the Cr coating. The preparation method comprises the following steps: firstly, carrying out surface pretreatment on the zirconium alloy matrix; a niobium-rich diffusion barrier layer is formed on the surface of the zirconium alloy matrix through ion implantation; then depositing a Cr coating on the niobium-rich layer; and finally, an Al element is injected into the surface of the Cr coating, and an Al-rich layer is formed so that an Al2O3 protective layer can be formed under the high-temperature working condition. According to the technical scheme, interface mutual diffusion can be effectively inhibited, the stability of the coating is improved, meanwhile, the Cr coating structure is optimized, the oxidation process is delayed, a gradient protection system is constructed, and the oxidation resistance of a high-temperature material is enhanced.

The invention belongs to the technical field of nuclear fuel assembly design, and particularly relates to a high-safety fuel assembly capable of measuring temperature and conducting strong flow guiding. Comprising a temperature-measurable instrument tube, a strong flow-guiding grillwork, a fuel rod, a guiding tube, a foreign-matter-catching-preventing lower tube seat, a low-pressure-drop upper tube seat and a connecting structure thereof. Through the arrangement of the structure and the connection of all the components, the high thermotechnical safety margin of the fuel assembly is realized. By arranging the temperature-measurable instrument tube, a temperature sensor can be placed in the temperature-measurable instrument tube, the temperature of a coolant entering the temperature-measurable instrument tube from the radially uniformly distributed drain holes can be continuously measured on line in real time, the actual thermotechnical safety margin of the fuel assembly can be known, operation can be fed back, operation strategy optimization can be guided, and transient state and low accident incidence of the fuel assembly are realized; by measuring and feeding back the coolant temperature of the hot spot part of the fuel assembly, the fuel assembly and the power plant can operate at safe and as high as possible parameters for a long time based on the mastered safety margin, and the economical efficiency of the nuclear power plant is improved.

The invention relates to the technical field of nuclear reactor design, in particular to an air-cooled nuclear reactor system for series thermal ionization magnetohydrodynamic power generation. Comprising a reactor core which comprises a fuel grid which is formed by a plurality of grid unit plates which are crossed transversely and longitudinally; each grid unit plate is made of sheet-shaped fuel; the thermal ionization magnetohydrodynamic power generation element is arranged outside the reactor core and is used for providing a magnetic field for the fuel grid; the cooling system is used for providing rare gas for the reactor core and is configured to enable the rare gas to exchange heat with the fuel grid and to be converted into conductive fluid in a generated magnetic field, and the conductive fluid flows in the generated magnetic field to generate induced electromotive force so as to realize thermoelectric conversion. While high thermoelectric conversion efficiency is ensured, the engineering feasibility of an air-cooled nuclear reactor system is improved, and the problems that gaseous nuclear fuel is prone to corrode structural materials, prone to leakage and large in aftertreatment difficulty are solved.

The invention provides a remote communication method, device and equipment for a mixing and clarifying tank stirring device, belongs to the technical field of nuclear fuel post-processing, and solves the problem of control signal fluctuation caused by electromagnetic interference, cable aging and environment abrupt change in a complex industrial scene. The method comprises the following steps: acquiring signal quality data; preprocessing the signal quality data to obtain preprocessed signal quality data; determining predicted signal attenuation data according to the preprocessed signal quality data; comparing and analyzing the predicted signal attenuation data with a preset threshold value to obtain a control instruction; and controlling the mixing and clarifying tank to communicate according to the control instruction. According to the scheme, the anti-interference performance and the prediction precision can be improved, and the communication failure rate and the energy consumption are reduced.

A reactor vessel that includes a reactor core mounted within a volume of the reactor vessel, the reactor core comprising one or more nuclear fuel assemblies configured to generate a nuclear fission reaction, a riser positioned above the reactor core, the riser forming a primary coolant flow path, a steam generator thermally coupled to the riser, the steam generator communicatively coupled to a steam turbine through a steam inlet that includes a steam inlet valve, a secondary coolant flow path that extends through the steam generator, the secondary coolant flow path coupled to a coolant pump, and a control system coupled to both the steam inlet valve and the coolant pump, the control system configured to control a power output of the nuclear fission reaction by adjusting one or more parameters of the steam inlet valve or the coolant pump.

The invention provides a device and method for continuously measuring the diameter of a fuel rod, and belongs to the technical field of post-irradiation inspection of nuclear fuel circulation, and the method comprises the steps: enabling a support to move relative to a sliding rail assembly so as to adjust the detection point position of a diameter measuring instrument in the axial direction of the fuel rod, and enabling the diameter measuring instrument to move relative to the support so as to adjust the detection height of the diameter measuring instrument, the measurement of different axial positions of the fuel rod is realized; the chuck can clamp the fuel rod placed on the supporting piece, and the chuck drives the fuel rod to rotate so as to realize continuous measurement of different radial positions of the fuel rod, so that the diameter value and the ovality of a detection point are obtained, the detection continuity of the diameter and the ovality of the fuel rod is improved, and the detection efficiency of the diameter and the ovality of the fuel rod is improved; meanwhile, the driving device drives the chuck to rotate so as to drive the fuel rod to rotate, it can be ensured that the fuel rod rotates in place, the rotating accuracy of the fuel rod is improved, and the accuracy of data collection and the corresponding accuracy can be improved.

A covering for reinforcing a base layer of a nuclear fuel cladding is provided. The covering comprises a first layer configured to cover a first portion of the outer surface of the base layer of the nuclear fuel cladding, a second layer surrounding the first layer and the base layer and a third layer surrounding the second layer. The first layer comprises a fiber based material, the second layer comprises an interfacing material and the third layer comprises Chromium. A reinforced cladding for nuclear fuel and a method for producing a reinforced nuclear fuel cladding are also provided.

A system is described that includes a sputter target and a magnetic element array including multiple sets of magnets arranged to have a Hall-Effect region that extends along a length of the sputter target. The elongated sputtering electrode material tube is interposed between the magnetic array and an object to be deposited with a sputtered material from the sputter target. During a direct current high-power impulse magnetron sputtering operation, the system performs a depositing on a surface of the object by generating and controlling an ion and neutral particle flux by: providing a vacuum apparatus containing a sputter target holder electrode; first generating a high-power pulsed plasma magnetron discharge with a high-current negative direct current (DC) pulse to the sputter a target holder electrode; and second generating a configurable positive voltage kick pulse to the sputter target holder electrode after terminating the negative DC pulse.

The invention relates to a nuclear fuel cladding produced with a substrate (14) which is made of pure zirconium or of a zirconium based alloy and a multilayer protective coating (16) which covers a surface (14B) of the substrate (14), the protective coating (16) comprising a main layer (18) made of pure chromium and one or more additional layers (20), each additional layer (20) being made of pure chromium or from a material made of chromium and, additionally, oxygen and / or nitrogen, with the possible presence of unavoidable impurities.

A fast neutron nuclear reactor contains a nuclear reactor core having an array of device locations. Some device locations in the nuclear reactor core contain fissile and fertile nuclear fuel assembly devices. One or more other device locations in the nuclear reactor core contain Doppler reactivity augmentation devices that amplify the negativity of the Doppler reactivity coefficient within the nuclear reactor core. In some implementations, a Doppler reactivity augmentation device can also reduce the coolant temperature coefficient within the nuclear reactor core. Accordingly, a Doppler reactivity augmentation device contributes to a more stable nuclear reactor core.

A nuclear fuel cell cladding that includes a zirconium-carbon covetic material. The zirconium-carbon covetic material has a carbon component associated with the surface of zirconium particles. The amount of carbon present in the zirconium-carbon covetic material is in a range of greater than 0.1 wt % to about 25 wt % of the zirconium-carbon covetic material. The carbon component may include carbon nanotubes, carbon nanomaterials, graphene, or graphene nanoplatelets. The carbon component may be uniformly distributed within the zirconium matrix. The zirconium-carbon covetic material may be formed from a zirconium alloy. The zirconium-carbon covetic material may be configured for use in various types of nuclear reactors. The synthesis of the cladding involves a process of plasma-enhanced chemical vapor deposition. The resulting nuclear fuel cell cladding offers improved performance and reliability for nuclear reactor applications.

The invention provides a nuclear fuel rod pile internal damage characteristic recovery experiment device, and relates to the technical field of nuclear fuel grillwork experiment devices, and the nuclear fuel rod pile internal damage characteristic recovery experiment device comprises the following steps: fixing a grillwork in an experiment box by using a first clamping assembly; the second clamping assembly is installed on the vibration assembly, and the vibration assembly is installed in the experiment box. And clamping the fuel rod by using a second clamping assembly. The vibration assembly drives the second clamping assembly and the fuel rod to vibrate, so that the fuel rod and the elastic reed rub with each other, and an abrasion experiment of the fuel rod and the grillwork elastic reed can be completed; and the fuel rod can also apply an acting force to the elastic reed, so that the fatigue test of the elastic reed can be completed. The experimental device can drive the fuel rods to repeatedly rub with the elastic reeds in the through holes of the grillwork or enable the fuel rods to repeatedly extrude the elastic reeds, so that the technical problem that an existing experimental device is difficult to meet two experiments at the same time is solved.