161 results about "Nuclear technology" patented technology

The invention discloses welding and heat treatment processes of large-diameter and thick-wall nuclear process pipes, which are suitable for welding and heat treatment of large-diameter and thick-wall pipes. The pipes welded by the welding and heat treatment processes are steady and reliable in the quality of welding lines, and can completely meet the quality requirement of nuclear technology.

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

The invention discloses a method for preparing a Cu-Nb alloy with high strength, high conductivity and high-temperature softening resistance. The method comprises the following steps of: (1) preparing Cu-Nb nanocrystal solid solution powder; (2) performing vacuum hot pressed sintering to prepare a Cu-Nb alloy billet; and (3) performing copper clad, manufacturing a sheath, sealing, and heating the copper sheathed billet to 850DEG C for hot extrusion. The Cu-Nb alloy prepared by the method has the sigma b between 600 and 800MPa, the relative electric-conductivity between 84 and 89 IACS, and the soft annealing resistant temperature between 900 and 1,100 DEG C, and can be applied to the fields such as electro vacuum, resistance welding electrodes, high voltage switches, electronic components, nuclear technology and the like.

The invention discloses a method of removing sulfonamide antibiotics in water by utilizing an ionization radiation technology, belonging to the technical field of nuclear technology application and environmental protection. The method can be used for effectively removing sulfonamide antibiotic drugs (for example sulfamethazine) in water without adjusting pH of wastewater and adding any chemical agent by applying the ionization radiation (high-power electron beams generated by electron accelerator or gamma rays generated by a radioactive source) technology. The method disclosed by the invention is simple, effective, easy to operate and can be used for treating antibiotic wastewater.

The invention relates top a car radiation imaging and detecting system, which belongs to the nuclear technology application field. The invention is characterized by comprising a ray source, a front and a back collimators, an array detector, other detecting components and a translational pulling mechanism, wherein the translational pulling mechanism is arranged between the ray source and the array detector, which forms a detecting channel for a measured foreign body to get in and get out and finish scanning in uniform speed with the ray source, the array detector and other detecting components. The car radiation imaging and detecting system is used to get vertical perspective image of cars or microbus or double perspective image in vertical or horizontal direction. The car radiation imaging and detecting system is used to do nondestructive detection for large-scale objects such as cars, microbus and the like.

The invention relates to a thickness and convexity detection device for plates and strips and belongs to the field of nuclear technology application. The device comprises a C-shaped frame, two ray sources which are arranged in an upper arm of the C-shaped frame along a width direction of a steel plate to be detected at an interval, two rows of high-pressure aeration ionization chamber detector arrays which are arranged in a lower arm of the C-shaped frame along a motion direction of a plate or a strip at intervals, a collimator which is arranged below the two ray sources and used for making each ray source emit a ray to the corresponding one row of detector array, a preamplifier connected with the detector arrays, a data acquirer connected with the preamplifier, a data processing and display computer connected with the data acquirer, a steam service system for guaranteeing running of a system and monitoring, and a control system. The thickness and convexity detection device is simple in mechanical structure and high in dynamic measurement precision; furthermore, the detector has the advantages of low temperature drift, high radiation resistance, high space resolution, high cost performance and the like.

The invention relates to the field of nuclear technology and radiochemistry, more specifically to the production and isolation of radionuclides for medical purposes. The method for producing actinium-225 and isotopes of radium comprises irradiating a solid block of metallic thorium of a thickness of 2 to 30 mm, which is contained within a hermetically sealed casing made of a material which does not react with thorium, with a flow of accelerated charged particles with high intensity. The irradiated metallic thorium is removed from the casing and is either heated with the addition of lanthanum and the distillation of radium or is dissolved in nitric acid with the recovery of actinium-225 by extraction. A target for implementing this method consists of blocks of metallic thorium of a thickness of 2 to 30 mm, which are contained within a hermetically scaled casing made of different materials which do not react with thorium.

The invention discloses a displacement damage dose detecting method for a spacecraft device, which belongs to the technical field of radiation detection in nuclear technology. The method comprises the following steps: scaling a photoelectric coupler integrated with an LED and a PD by using a radiation source to acquire a response curve of the increment of current transmission rate reciprocal of the photoelectric coupler during normalization and the displacement damage dose; then radiating the photoelectric coupler by using a ground displacement damage effect testing and evaluating source to acquire the increment of the current transmission rate reciprocal of the photoelectric coupler; and according to the increment of the current transmission rate reciprocal of the photoelectric coupler during specific radiation, searching the displacement damage dose corresponding to the increment in the response curve so as to detect the displacement damage dose at the moment. The method overcomes the influence of non-ideal single energy and secondary radiation of a general testing simulation source; and compared with the conventional method which only uses the LED to perform displacement damage detection, the testing system is simple and practical, and the error of optical coupling introduction is eliminated.

The invention relates to a method for removing cyanides in wastewater by using the ionizing radiation which is combined with ozone oxidation, in particular to the method for removing the simple and complex cyanides in the wastewater by utilizing the ionizing radiation technology (such as Gamma-ray or electron beam) which is combined with the ozone oxidation, and the method belongs to the technology field of nuclear technology application and environmental protection. Although the currently adopted alkaline chlorine oxidation method, the ozone oxidation method, the H2O2 oxidation method and the like for the cyanide-containing wastewater with the low concentration can oxidize the simple cyanides, the methods can not effectively damage the complex cyanides. The method adopts the combined use of the ionizing radiation and ozone to generate hydroxyl free radical. OH with stronger oxidation capacity and lower selectivity, the oxidation-reduction potential (E<0> is equal to 2.8V) is higher than the ozone (E<0> is equal to 2.07V) by 35 percent, and the reaction rate is 100 to 1000 times of the ozone. The method can not only reduce the radiation dose which is needed by the individual ionizing radiation treatment, but can oxidize and decompose the complex cyanides which are difficult to degrade by the alkaline chlorine oxidation method, the ozone oxidation method, the H2O2 oxidation, and the like.

The invention discloses a method for preparing a Cu-Nb alloy with high strength, high conductivity and high-temperature softening resistance. The method comprises the following steps of: (1) preparing Cu-Nb nanocrystal solid solution powder; (2) performing vacuum hot pressed sintering to prepare a Cu-Nb alloy billet; and (3) performing copper clad, manufacturing a sheath, sealing, and heating thecopper sheathed billet to 850DEG C for hot extrusion. The Cu-Nb alloy prepared by the method has the sigma b between 600 and 800MPa, the relative electric-conductivity between 84 and 89 IACS, and thesoft annealing resistant temperature between 900 and 1,100 DEG C, and can be applied to the fields such as electro vacuum, resistance welding electrodes, high voltage switches, electronic components,nuclear technology and the like.

The invention relates to a method for preparing lithium deuteride by using a high-temperature direct manner and belongs to the crossing field of material sciences and nuclear technology. According to the method for preparing lithium deuteride by using the high-temperature direct manner, lithium metal reacts with deuterium gas at high temperature to generate lithium deuteride. By virtue of the method, the problems of industrialization scale preparation technology and the preparation purity of lithium deuterid are solved; a complex production device does not need to be arranged; the required preparation device is simple and easy to purchase; a lithium deuteride production line can be safely constructed under the current factory conditions; the method for preparing lithium deuteride by using the high-temperature direct manner is simple and easy to operate, is free of high-temperature pressurizing and depressurizing device in the preparation process, adopts inert gas protection, and is safe and reliable; the product is relatively high in yield and purity and good in stability, and can be stored for a long time; three wastes are not produced in the production process; the method is free of harm to the human body and the environment, environmental-friendly and green, and has important significance for safely preparing high-purity lithium deuteride in large industrial scales and applying lithium deuteride in scales.

The invention provides a microchannel plate type fast neutron position gas detector and a detection method thereof, and relates to the field of nuclear technology applications. The detector comprises a gas chamber with an intake hole and an exhaust hole, a drift cathode, a microchannel structure plate, a gas electron multiplier, a high voltage dividing circuit, a read out PCB and a working gas. The drift cathode is arranged in the gas chamber with the intake hole and the exhaust hole. The microchannel structure plate is arranged below the drift cathode. The gas electron multiplier is arranged below the microchannel structure plate. The read out PCB is arranged below the gas electron multiplier. The working gas fills the whole chamber and a microchannel in the microchannel structure plate. The microchannel plate type fast neutron position gas detector provided by the invention can normally work at a high counting rate, and can be used for large part non-destructive detection and container scanning security check.

A three-dimensional dosage inversion method is disclosed. In the method, two-dimensional dosage acquired by measuring out of a die body is inversed so as to obtain the three-dimensional dosage in the die body; through a dosage guiding plan, the irradiation dosage can be guaranteed to be consistent with the plan dosage. The method can be used in radiation dosage calculating, three dimensional dosage measurement, a dosimetry verification and quality guarantee direction which belong to an interdisciplinary field, such as nuclear physics, a nuclear technology and application and the like. The three dimensional dosage in the irradiated die body can be accurately and rapidly acquired. By using a traditional point measurement or two-dimensional measurement, the three dimensional dosage in the die body can be hardly reflected. By using the method of the invention, the above problem can be solved. Besides, through error feedback of real time dosage reconstruction, controlling flux of a radioactive source and guiding the radioactive source to carry out accurate irradiation with quality assurance according to a plan, implement of the accurate guiding irradiation plan can be realized.

The invention belongs to the technical field of monitoring of nuclear technology, and discloses a method for measuring the isotope ratio of uranium in a single particle. The method comprises the following steps: firstly seeking a uranium-containing particle by using a scanning electron microscope, using a micro-operation system to transfer the uranium-containing particle to a conductive adhesive with a marked metal mesh on the surface, positioning and fixing the uranium-containing particle, and then analyzing the isotope ratio by using a laser ablation-inductively coupled plasma mass spectrometry. The method has the advantages of simple and quick pre-treatment method, simplified analysis flow, and high analysis precision which can reach to 1%.

The invention discloses a method of preparing a ceramic grade uranium dioxide ball. The method sequentially comprises the following steps: carrying out ball milling on uranium oxide, wherein the particle size of the uranium oxide after ball milling is 0.1-0.3 micron; adding the uranium oxide into a polymer monomer, and uniformly stirring to obtain polymer slurry; producing slurry drops by virtue of the polymer slurry; adding the slurry drops into a dispersion column which contains heated carbon tetrachloride or liquid paraffin, and descending the slurry drops to obtain a green body ball in the dispersion column under the surface tension, wherein the temperature in the dispersion column is 80-90 DEG C; washing the green body ball by adopting an ethanol solution; drying the green body ball, and removing an organic matter on the surface of the green body ball to obtain a dry calcined ball; and carrying out reduction sintering on the dry calcined ball in a hydrogen atmosphere to obtain the ceramic grade UO2 ball. The method is simple in whole process, convenient in realization, small in introduced impurities, high in purity of the prepared UO2 ball and beneficial to popularizing and applying the UO2 ball in the field of nuclear technology.

The invention discloses an irradiation reactor for continuously treating industrial wastewater under cooperation of an electron accelerator, and a method for the irradiation reactor, and belongs to the field of application of nuclear technologies and environmental protection. The irradiation reactor comprises nozzles 2 and an irradiation container 4; and the irradiation container 4 is established 200 to 400mm below a scanning window 6 of the electron accelerator, each of the side walls of the irradiation container 4 is provided with one flat nozzle 2, and wastewater 1 with certain pressure and flow velocity is sprayed by using the nozzles 2 to form a parabolic water film with certain thickness and width. A certain dose of electron beams are irradiated on the wastewater, and then the irradiated wastewater is discharged by a material discharge port 5, and then enters the next process. The irradiation reactor can be matched with electron accelerators with various types and energy for use, and has the advantages of simple structure, low energy consumption, uniform dose, and the like and is easy to control.

The invention relates to a hierarchically reconstructed non-uniform rational B-spline (NURBS) curve interpolator and a configuration method thereof. A field programmable gate array (FPGA) chip and an internet protocol (IP) nuclear technology are adopted in the interpolator, and the interpolator comprises a physical layer, a data layer and an application layer. The physical layer consists of a microprocessor, an interrupt manager, a timer, a pipeline burst cache (PLB) bus, a serial port IP nucleus, an IP interface, an interface control module, a manual mode module, a fine interpolation module and a feedback signal measurement module; the data layer contains a host access buffer area and an interpolation feedback register set; the data layer stores NURBS curve parameters written in from a numerical controller, processing commands and axis displacement error calculated by the feedback signal measurement module; and the application layer contains a human-machine interface module and a rough interpolation module, and the application layer receives field configuration information and realizes NURBS curve rough interpolation. Field configuration is easily realized by a three-layer control structure, time and technical requirements required by reconfiguration of the interpolator are reduced, stability and reliability are improved, and the processing capacity of a numerical control system is expanded.

The invention provides a liquid fuel heat pipe reactor, belongs to the field of nuclear engineering and nuclear technologies, and particularly relates to a liquid fuel heat pipe reactor. The problemsthat a fuel layer of a solid reactor core is deformed, and local heating is not uniform at high temperature are solved. The liquid fuel heat pipe reactor comprises a heat pipe, a fuel medium and a reactor core. The novel heat pipe reactor is mainly provided and used for generating heat energy, and then the heat energy is converted into electric energy or kinetic energy.

The invention relates to a collimation mechanism for a convexity meter, and belongs to the field of nuclear technology application. The collimation mechanism includes two sets of collimation devices with the same structure, and each set of collimation devices includes 3-stage collimators, and the collimators from the first stage to the third stage are arranged sequentially along the radiation direction of a radiation source. The first-level collimator includes a first tungsten plate; the second-level collimator includes second to fifth tungsten plates, two mounting plates, a bottom plate, a flat plate, and a fixing block, and the two mounting plates are placed on the flat plate, The two mounting plates are respectively connected with the second to the fifth tungsten plates, the two mounting plates are connected with the flat plate, the two mounting plates are connected, the flat plate is connected with the bottom plate, and a spring is pressed between the flat plate and the fixed block; the third level collimation The device includes the sixth tungsten plate, and the sixth tungsten plate is fixed by the top wire; after this three-level collimation, the ray beam emitted by the ray source becomes a collimated ray beam, which solves the problem that the convexity instrument lacks the collimation function for the ray beam emitted by the ray source Problems, improve the accuracy of product detection.

Belonging to the field of nuclear technologies, the invention specifically relates to an iodine adsorber performance test method and device. The method includes: forming circulating air between the inlet and outlet of a test iodine adsorber, adopting a frequency adjustment fan to accurately adjust the flow of circulating air, controlling the humidity of the circulating air by a heat pump system to make the air humidity on the inlet side and outlet side of the iodine adsorber consistent, injecting a test source into the inlet side of the iodine adsorber, conducting sampling at the inlet and outlet of the iodine adsorber respectively to test the purification performance of the iodine adsorber. The method and the device provided by the invention combine air circulation, frequency adjustment technology, heat pump technology and other energy saving technologies, and reach the comprehensive benefits of environmental protection, energy saving, safety and high efficiency.

The invention (2) relates to a device (2) that is simply constructed and displays a high fail safeness for using heated thermoelements (HT) and unheated thermoelements (UHT) as transmitters for measuring a liquid in a liquid container, in particular in the reactor container (4) of a nuclear installation, makes an especially precise and reliable measurement of the level height (H) possible. In addition, the invention proposes a plurality of elongated measuring tubes (6), at a distance from one another, wherein each measuring tube (6) exhibits a number of thermoelements (HT, UHT) arranged such that they are distributed in a longitudinal direction, and wherein an unheated thermoelement (UHT), acting as a reference transmitter, is allocated to a thermoelement (HT), acting as a primary transmitter, arranged at a first measuring tube (6) and heated by a heating element (HE) and which is arranged at a measuring tube (6) that is different from the first measuring tube (6).

The invention relates to a radiation processing and nuclear technology accelerator, disclosing a multi-filament wide-beam electronic curtain. The multi-filament wide-beam electronic curtain comprises a control system, a power supply system, a water cooling system, a nitrogen supply system and a vacuum system, wherein the power supply system comprises a high-voltage power supply, a head power supply and an isolation transformer; the head power supply is connected with high-voltage ceramics by a high-voltage inlead; the vacuum system comprises an electronic gun, an external barrel and filaments, wherein a vacuum cavity is formed between the electronic gun and the external barrel; the lower side of the electronic gun is provided with a second grid electrode of the flat plate shape; the second grid electrode is provided with a second grid aperture; a titanium foil and an acceleration electrode are arranged just below the second grid electrode; and the titanium foil is provided with a cooled copper grid. In the radiation processing and nuclear technology accelerator, the density of the filament is reasonably set to overcome the distribution unevenness of end beam current because of titanium foil dispersion; the second grid electrode is provided with the second grid aperture to reduce cut beams on the titanium foil copper grid; and an outer layer of shielding cover is also provided with a distribution target so as to monitor beam current distribution in real time to greatly improve the performance of the whole accelerator.

The invention relates to nuclear technology, and to irradiation targets and their preparation. One embodiment of the present invention includes a method for preparation of a target containing intermetallic composition of antimony Ti—Sb, Al—Sb, Cu—Sb, or Ni—Sb in order to produce radionuclides (e.g., tin-117m) with a beam of accelerated particles. The intermetallic compounds of antimony can be welded by means of diffusion welding to a copper backing cooled during irradiation on the beam of accelerated particles. Another target can be encapsulated into a shell made of metallic niobium, stainless steel, nickel or titanium cooled outside by water during irradiation. Titanium shell can be plated outside by nickel to avoid interaction with the cooling water.

The invention discloses an irradiation processing method for acrylonitrile waste water, which belongs to the field of nuclear technology application and environmental protection. The method mainly makes use of the principle that ionizing radiation can induce acrylonitrile to have a polyreaction, enables the acrylonitrile in the waste water to have the polyreaction through the ionizing radiation; and then acrylonitrile polymer deposit is separated from the water phase through a filtration method so as to realize the aim of removing the acrylonitrile from the waste water. The method can effectively remove the biologically toxic acrylonitrile pollutant in the high-concentration acrylonitrile waste water. The method is a waste water processing technology with high efficiency, economy and environmental friendliness.

The invention provides a radial modular nuclear reactor, belongs to the technical field of nuclear engineering and nuclear technology, in particular relates to the radial modular nuclear reactor, andsolves the problems of large volume, inconvenient replacement of fuel components, easy occurrence of accidents and poor heat exchange effect of the existing nuclear reactor. The radial modular nuclearreactor comprises a reactor core and control rods, wherein the reactor core comprises a plurality of fuel assembly modules, the plurality of fuel assembly modules are arranged along the circumferential direction to form an annular structure, the center of the annular structure is a central channel, the control rods pass through the fuel assembly modules along the axial direction, the fuel assembly modules comprise a plurality of fuel rods and heat transfer substances, the plurality of fuel rods are distributed along the radial direction of the reactor core, the plurality of fuel rods are distributed in an array and are provided with gaps, and the outer wall of the fuel rods is in contact with the heat transfer substances. The radial modular nuclear reactor is mainly used for radially arranged modular nuclear reactors.

The invention discloses a method for measuring the energy of an accelerator, comprising the following steps: a measuring step for measuring a measurement curve that an attenuation multiple of an X ray from an accelerator is changed along with the thickness of a shielding material; and a comparing step for comparing the measurement curve with a reference curve to determine the energy of the X ray from the accelerator. The method can accurately judge the actual energy of the X ray beam from the electronic accelerator. The method is not limited from the restriction condition of the accelerator, and has the characteristics of low cost, simple operation, vivid data, high precision and small error caused by artificial factor. The method is not only suitable for the quality control of the production, debugging and check of the accelerator in the nuclear technology but also capable of guiding the operator to debug the accelerator precisely.

The invention relates to the field of nuclear technology application, in particular to a method and the device completing centering and positioning under the condition that a radiation source is connected with a detector non-rigidly when large component nondestructive detection is carried out. The centering method of the invention used for the radiation source and the detector during the nondestructive detection of large components comprises the steps as follows: (a) the radiation source is roughly centered with the detector; (b) a centered shielding pattern is rotated to a shielding position; (c) the position of the shielding pattern is observed after the image forming; (d) the position from the radiation source to the shielding pattern is adjusted to a certain imaging error, thus completing the centering. The corresponding device of the invention comprises a centering positioning plate, the shielding pattern and a shielding collimator; wherein, the shielding pattern is arranged at the front end of the shielding collimator by the centering positioning plate. The method and the device of the invention realize the exact centering of the radiation source and the detector during the nondestructive detection application of large and super-large components, and can calculate the distance from the radiation source point to the detector, thus achieving the basic condition of the radiation image forming and completing the detection.

The invention discloses a method for improving the biodegradability of papermaking wastewater by using a combined process of coagulation and irradiation in the field of nuclear technology application and environment protection. The method comprises the following steps: the papermaking wastewater is treated by adopting a method combining coagulation and ionizing radiation; and firstly, most organic matters with cutoff molecular weights above 3000 kDa in the papermaking wastewater are removed by utilizing a coagulation and sedimentation process, and then toxic and difficultly biodegraded matters in the waste water, such as residual lignin, cellulose, aromatic compounds, and the like, are transformed to non-toxic / low-toxic and easily biodegraded matters through ionizing radiation, thereby improving the biodegradability of the papermaking wastewater. By adopting the combined process of coagulation and irradiation, difficultly biodegraded organic matters can be removed, and the radiation dose can be reduced to the maximum extent, thereby improving the economical efficiency of ionizing radiation treatment of the papermaking wastewater.

The invention provides a preparation method for uranium molybdenum alloy powder, and the method is used for preparing the uranium molybdenum alloy powder meeting the demand on a certain particle size and belongs to the field of nuclear technique application. The preparation method provided by the invention mainly comprises the following steps of: cleaning and activating before activating the uranium molybdenum alloy. By cleaning the uranium molybdenum alloy with aqua regia, distilled water, absolute ether, absolute ethyl alcohol, and the like, an oxide layer on the surface of the uranium molybdenum alloy can be efficiently removed, thereby achieving the clean surface of the uranium molybdenum alloy before absorbing hydrogen. Parameters of an activating process are efficiently controlled, thereby accurately controlling the particle size of the uranium molybdenum alloy powder. The preparation method for the uranium molybdenum alloy powder provided by the invention has the characteristics of less working procedures, simple process, small investment, and the like, and is suitable for being applied to the field of nuclear fuel cycle.