350 results about "Nuclear facilities" patented technology

Decontamination of nuclear facilities is necessary to reduce the radiation field during normal operations and decommissioning of complex equipment. In this invention, we discuss gel and foam based diphosphonic acid (HEDPA) chemical solutions that are unique in that these solutions can be applied at room temperature; provide protection to the base metal for continued applications of the equipment; and reduce the final waste form production to one step. The HEDPA gels and foams are formulated with benign chemicals, including various solvents, such as ionic liquids and reducing and complexing agents such as hydroxamic acids, and formaldehyde sulfoxylate. Gel and foam based HEDPA processes allow for decontamination of difficult to reach surfaces that are unmanageable with traditional aqueous process methods. Also, the gel and foam components are optimized to maximize the dissolution rate and assist in the chemical transformation of the gel and foam to a stable waste form.

The invention discloses a method and system for real-time online evaluation of off-site results of nuclear facility accidents, and belongs to the technical field of nuclear facility accident result analysis. The method comprises a wind field prediction method of a simulation area, a wind field diagnosis method, a Lagrange puff atmospheric diffusion simulation method, an estimation method of gaseous radionuclide-caused radiation dosages and prevention measures, and an airborne passing food chain radioactivity estimation method. Through the method and system provided by the invention, the wind field prediction and diagnosis, and the contaminant atmospheric diffusion simulation and radiation dosage estimation of a plurality of accident points from the happening of a nuclear facility accident to the end moment in a simulation area range which takes a nuclear facility as a center can be completed, so that result evaluation data basis is provided for the accident emergency command processing after the nuclear facility accident.

A system for providing a simulated total dose exposure measurement during a nuclear facility training exercise by locating participants using a real time location system, modeling incremental exposure as a function of location and summing incremental exposure to produce a total dose for each of the participants. Total dose may be displayed via a wireless link to a simulated dosimiter worn by each participant. Radiation sources may also have location tags and the exposure model may be modified in real time according to the tracked location of the radiation source. In one embodiment, the locating technology comprises near field locating technology based on comparing near field signal characteristics. Alternative locating technologies may be used.

The invention provides a plasma volume reducing system and method for processing nuclear facility solid waste. The plasma volume reducing system comprises a fusion burning system and a tail gas processing system. The fusion burning system further comprises a fusion furnace and a secondary burning chamber. Low-and-intermediate level radioactive solid waste is processed through a combination mode of plasma pyrolysis fusion and secondary combustion. Tail gas generated by secondary combustion is processed in a combination mode of a half-drying-quick-cooling method, a cloth bag, a wet method and denitration. Various kinds of waste can be processed, the size of the waste is lowered to 1 / 6, radionuclide can be stably solidified in a solidifying body, and exhaust gas emission can reach the national standard. Comprehensively, the high volume reduction ratio and the high safety in the volume reduction technology can be achieved.

By using a scintillation surveymeter with good calibration performance evaluation for a secondary standard radiation field, and a working standard part obtaining an ambient dose equivalent rate, in cooperation with a portable irradiator, and an irradiator lifter, a laser range finder and a laser locator of a relevant radiation source, in-situ calibration is capable of being performed on fixed, or large-scale, or continuous monitoring type radiation monitors to be calibrated stationed in nuclear power plants, nuclear medical departments, and other nuclear facility operating institutions. Moreover, a time-efficient and effective in-situ calibration method is further provided, which can be performed based upon a standard calibration field that is achieved using a portable 137Cs radiation source. The in-situ calibration method is capable of saving the trouble of delivering large-scale monitors, or monitors difficult to move, or monitors requiring continuous monitoring to calibration laboratories for scheduled calibration.

The invention provides a 3dsMax-based nuclear facility model radiation field dosage simulation method. The method comprises the following steps of: constructing a model by 3dsMax software according to determined nuclear facility parameters, and storing a file in a 3DS format; importing a 3DS nuclear facility model file and obtaining model parameters; decomposing the 3dsMax nuclear facility model into voxels by using an octree method; writing determined voxel parameters and material information into an input card; importing the input card into a point nuclear integration program; calculating an accumulation factor; calculating a mean free path, in a radiation field, of a gamma ray; establishing a flux rate-dosage rate conversion factor, quality decay factors of chemical elements and materials and a single-layer accumulation factor database by utilizing an SQLite database engine; carrying out combined operation on a box by using a Boolean connective operator, and constructing a complicated radiation field geometric structure; and calculating a three-dimensional radiation field dosage by using a point nuclear integration method. The method provided by the invention is capable of realizing radiation field dosage calculation of complicated 3dsMax nuclear facility models with sizes, materials and energy parameters.

The invention relates to a digital nuclear facility environmental radiation monitoring system. The digital nuclear facility environmental radiation monitoring system comprises a control device and field measuring devices. The control device comprises a central station. The field measuring devices include an environmental medium sampler, a meteorological sensor and a radiation detector. The radiation detector is a digital NaI (TI) spectrometer internally provided with a high-range dosage rate detector. The environmental medium sampler, the meteorological sensor and the digital NaI (TI) spectrometer are connected with a local industrial computer, measuring signals and sampling signals are transmitted to the local industrial computer which is connected with the central station in a wired communication mode and in a wireless communication mode, and the central station remotely monitors the working state of all the field measuring devices. According to the system, nuclide recognition can be carried out while a radiation dose rate gamma is given, communication is reliable, the bandwidth is high, interference is not likely to happen, and the data acquiring rate of the system can be guaranteed.

The invention discloses a Fe-Cr-B system corrosion-resistant block amorphous alloy and a preparation method thereof and belongs to the field of amorphous alloys. The chemical components of the Fe-Cr-B system corrosion-resistant block amorphous alloy are designed according to the atomic ratio as follows: FeaCrbMocBdMeRfXg and are characterized in that M is one or more of Mn, Co and Ni, R is one or more of Ti, Zr, Nb, Hf, Ta and W, and X is one or more of Si and P, wherein a is larger than 20 and no larger than 78, b is no less than 5 and no larger than 45, c is no less than 0 and no larger than 20, d is no less than 6 and no larger than 30, e is no less than 0 and no larger than 40, f is no less than 2 and no larger than 15, g is no less than 0 and no larger than 10, and a+b+c+d+e+f+g=100. The Fe-Cr-B system corrosion-resistant block amorphous alloy is characterized in that the Fe-Cr-B system corrosion-resistant block amorphous alloy is low in cost and can be smelted by using an industrial raw material, is high in boron content, is good in amorphous formation capability and is particular suitable for storage and transportation of nuclear wastes, is high in Cr content without containing C so as to ensure excellent corrosion resistance, and is also very high in compressive strength and microhardness. Therefore, the iron-based amorphous alloy disclosed by the invention has a good application prospect in the aspects of metal material corrosion resistance, nuclear facilities and wear-resistant parts.

The invention belongs to the technical field of numerical value simulation, and particularly relates to a mumerical value simulating method for nuclear facility waste gas diffusion under the support of a GIS. The method includes the following steps of firstly, data preprocessing; secondly, simulation calculating; thirdly, GIS simulation output. The first step includes the substeps of firstly, digitally forming a layer of a data table, importing the data table in a database, and discretizing the data table into control point data and grid layer data; secondly, establishing an (x, y, z) coordinate system according to the dosage of waste gas; thirdly, establishing a coordinate system according to meteorological data, and determining the range of a pollution area. The second step includes thesubsteps of firstly, imputing the data preprocessed in the first step in a nuclide diffusion model, and establishing an optimal model; secondly, obtaining absolute coordinates and concentration data of grid central points; thirdly, calculating the time-varying result of nuclide concentration data under the same coordinates. The method can simulate the pollution risk space, quantify the evaluatingresult and provide the technical support for related decision making.

The invention relates to a chemical decontaminating agent and a decontamination method. The chemical decontaminating agent contains acids, bromides, molybdates, surfactants, corrosion inhibitors, anti-settling agents and deionized water. The chemical decontaminating agent is mainly applied to eliminate radioactive contamination during decommissioning and operation of nuclear facilities, reduces corrosion to facilities of different materials, and has a high decontamination factor (DF) of common nuclides. Besides, the chemical decontaminating agent has a good stability.

The invention discloses a radioactive waste resin metering system, comprising a waste resin storage tank, a waste resin metering tank, a dehydrating pump, a water metering tank and a waste barrel, wherein the waste resin metering tank is connected with and receives water-containing waste resin from the waste resin storage tank, the dehydrating pump is connected between the waste resin meter tank and the water metering tank, dehydrates stored objects in the waste resin metering tank, and discharges the dehydrated water into the water metering tank, and the waste resin metering tank and the water metering tank are respectively connected to the waste barrel to add the dehydrated waste resin and water into the waste barrel according to the preset ratio of a solidification formula. Compared with the prior art, the radioactive waste resin metering system has the advantages that the problems of difficulty in metering and large metering error of radioactive waste resins of nuclear power plants or other nuclear facilities are solved by the mode of separating the waste resin from the water and then accurately metering the waste resin and the water, and the aims of realizing the precise preparation of the solidification formula and solving the reuse problem of radioactive waste water are achieved simultaneously.

The invention belongs to the field of Gamma ray radiation measurement, particularly relates to a Gamma ray imaging device which comprises a Gamma camera and a data acquisition processing system, wherein the Gamma camera comprises a pinhole collimator, a scintillation crystal, a position sensitive photo-multiplier tube and resistor grids; and electrical signals output by the resistor grids are input to the data acquisition processing system arranged outside the Gamma camera. The Gamma ray imaging device produced according to the technical scheme of the invention can measure the large-scaled decommissioning site of nuclear facilities in real time to obtain a region distribution map of hot points, which includes nuclide positioning, nuclide identification and nuclide dosage.

The invention provides a tritium water collecting device. According to the tritium water collecting device, a vacuum booster pump, an air cooling coil, a cold trap cooler, a molecular sieve absorber, a liquid nitrogen temperature molecular sieve absorber, a cleaning bed and a control panel are connected through a pipeline. According to the working procedure, the vacuum booster pump presses tritium gas into the air cooling coil, the tritium gas is cooled into the temperature about 40 DEG C and then enters the cold trap cooler to be cooled, tritium water is absorbed by a molecular sieve, the tritium water is absorbed by the molecular sieve under the liquid nitrogen temperature, and the left tritium gas is exhausted to a tritium purifying system through a blowoff valve. By means of the tritium water collecting device, the service life of an alloy hydrogen absorption bed and a catalytic oxidation bed of the tritium purifying system can be prolonged by about 50 percent, the using efficiency can be improved by more than 50 percent, and the radioactive waste of the produced molecular sieve dry adsorbent can be reduced by more than 80 percent. The tritium water collecting device can be applied to nuclear emergency work of high-radiation tritium gas of nuclear facilities.

The invention provides a 3dsMax nuclear facility model radiation field dosage distribution simulation method. The method comprises the steps that 1, 3dsMax software constructs a model according to determined nuclear facility parameters, and a file is saved as a 3DS format; 2, a 3DS nuclear facility model file is imported, and model parameters are obtained; 3, a bounding box and a tetrahedron are constructed; 4, determined voxel parameters and texture information are written as an input card; 5, a plane triangle is dispersed into pixels; 6, according to attribute values of voxels, entity voxels are constructed; 7, the determined voxel parameters and physical property information are written into the input card; 8, a point kernel integration method is used for calculating three-dimensional radiation field dosage distribution. Rapid assignment is directly performed on the facility size, texture and energy parameters inside the 3dsMax software; radiation field dosage distribution calculation of the complex nuclear facility model internally with a hole is achieved.

The invention aims to disclose an ambient neutron dose equivalent meter. The ambient neutron dose equivalent meter is characterized by comprising a neutron probe, a neutron energy response adjustor and a control system, wherein a neutron detection element is arranged in the neutron probe, the neutron detection element is europium-doped lithium iodide scintillation crystal, the signal output end of the neutron probe is connected with the signal input end of the control system, and the neutron probe is placed at the position of the geometric center of the neutron energy response adjustor. Because the photomultiplier with high sensitivity is used as a photoelectric converter, the meter has light weight, portability, sensitive neutron response and good angle response, can accurately measure the ambient neutron dose equivalent level, has sound and light alarm function, is used for ambient neutron radiation monitoring of large nuclear facilities such as reactors, high-energy particle accelerators and the like, ensures neutron radiation safety of the ambient and the public, and fulfills the purpose of the invention.

A telescoping rod control cluster assembly change tool for moving control rod assemblies among fuel assemblies in a nuclear facility. The operation of the tool is completely mechanical and the telescoping feature enables the tool to have a relatively low profile when it is being moved and stored without housing a control rod assembly. Rigidly supported alignment cards guide a gripper that attaches to the control rod assembly as the control rod assembly is withdrawn into the tool with the alignment cards preventing any lateral or rotational movement of the gripper.

The invention discloses a decorated board for shielding an X ray, which is characterized by comprising the following raw materials by the weight portion: 1-5 of quartz powder, 2-8 of pulverized fuel ash, 3-10 of reinforced fiber, 6-15 of sulphate aluminium cement, 65-80 of permanent white, 5-10 of zinc powder and 3-5 of sodium silicate. The decorated board for shielding the X ray has better comprehensive properties and favorable function of X ray radiation prevention, can be widely applied to construction, military, medicine, power plant or nuclear facility and many other fields and has light weight and low production cost.

There is provided high power laser systems, high power laser tools, and methods of using these tools and systems for cutting, sectioning and removing structures objects, and materials, and in particular, for doing so in difficult to access locations and environments, such as offshore, underwater, or in hazardous environments, such as nuclear and chemical facilities. Thus, there is also provided high power laser systems, high power laser tools, and methods of using these systems and tools for removing structures, objects, and materials located offshore, under bodies of water and under the seafloor.

The invention discloses an estimation method of gaseous radionuclide-induced radiation doses and protective measures and suggestions and belongs to the nuclear facility accident off-site assessment technical field. The method is a calculation method for calculating potential doses and expected doses to which people in each grid in each period of the early stage of an accident of a region are exposed, as well as avertable doses and residual doses of intervention actions such as taking iodine, concealment and withdraw, under a condition that gaseous radioactive substances are released unevenly at any time span in nuclear accident consequence assessment. With the method of the invention adopted, the calculation of doses to which people are exposed when passing through different paths in withdraw and relocation actions and the estimation of the concentration of key nuclides in reservoirs and milk are realized; and potential doses and expected doses to which people are exposed during different time periods, the avertable doses and residual doses of intervention actions, suggested intervention action regions, directions, administrative divisions, demographics and other information can be provided; and necessary information support can be provided for emergency decision makers.

The invention belongs to the technical field of nuclear facility decommissioning and radioactive waste treatment, and particularly relates to an AHP (Analytic Hierarchy Process)-based high-level radioactive waste geological repository site suitability quantitative evaluation method. The method comprises the following steps: 1, indexes for high-level radioactive waste geological repository site suitability quantitative evaluation are determined; 2, a high-level radioactive waste geological repository site suitability evaluation hierarchical structure model is built; 3, a high-level radioactive waste geological repository site suitability evaluation hierarchical structure model judgment matrix is built; and 4, different site selection schemes are ranked comprehensively. Three difficult problems of how to build the hierarchical structure model, how to build the judgment matrix and how to comprehensively and quantitatively judge advantages and disadvantages of different site selection schemes when the AHP theory is applied to the high-level radioactive waste geological repository site suitability quantitative evaluation can be solved, and quantitative evaluation on the suitability of a large, complicated and extra long-safety evaluation period site for the high-level radioactive waste geological repository can be carried out.

The invention discloses a radiation-resistant power mechanical hand. The radiation-resistant power mechanical hand comprises a control chamber, a barrier shield and a hot chamber. A control system is arranged in the control chamber and at least comprises a central control cabinet, a travelling crane control cabinet, a mechanical arm control cabinet, an operation platform and a hand operation box, a penetration piece is arranged on the barrier shield, and an operating execution system is arranged in the hot chamber and at least comprises a travelling crane, a telescopic cylinder, monitoring equipment and a tool shelf. Control lines of the central control cabinet, the travelling crane control cabinet and the mechanical arm control cabinet all penetrate through the penetration piece on the barrier shield to be connected to the travelling crane, the monitoring equipment and the tool shelf. By the adoption of radiation resistance design, the mechanical hand can clamp, carry and cut radioactive waste through teleoperation in low-and-intermediate radiation environments like the hot chamber, and the situation that working staff make direct or indirect contact with radioactive waste is reduced; the mechanical hand can replace workers to dispose radioactive waste, and the nuclear waste disposal efficiency is improved; meanwhile, the mechanical hand can perform maintenance work like replacement on nuclear facility parts.

Decontamination of nuclear facilities is necessary to reduce the radiation field during normal operations and decommissioning of complex equipment. In this invention, we discuss gel and foam based diphosphonic acid (HEDPA) chemical solutions that are unique in that these solutions can be applied at room temperature; provide protection to the base metal for continued applications of the equipment; and reduce the final waste form production to one step. The HEDPA gels and foams are formulated with benign chemicals, including various solvents, such as ionic liquids and reducing and complexing agents such as hydroxamic acids, and formaldehyde sulfoxylate. Gel and foam based HEDPA processes allow for decontamination of difficult to reach surfaces that are unmanageable with traditional aqueous process methods. Also, the gel and foam components are optimized to maximize the dissolution rate and assist in the chemical transformation of the gel and foam to a stable waste form.

The invention discloses an on-site calibration device of an airborne tritium monitoring instrument. The on-site calibration device of the airborne tritium monitoring instrument comprises an air processor, a bubbling unit, a transmission instrument, an aspirator pump and a recovery unit, all of which are sequentially connected, wherein the air processor is used for filtering and drying inlet air, the bubbling unit is used for bubbling tritiated water to form saturated tritium flow, the transmission instrument is used for being connected with a calibrated detection instrument and measuring the activity and concentration of the tritium flow generated on site, the aspirator pump is used for pumping the tritium flow into the recovery unit for recovery after the activity and concentration of the tritium flow generated on site are detected, and the recovery unit is used for cooling and recovering the recovered tritium flow. The ranges of the activity and concentration of the tritium flow generated by the on-site calibration device of the airborne tritium monitoring instrument in a bubbling mode depend on the total activity of the tritium flow for bubbling and can meet the calibration of most existing nuclear facilities using the measuring span of the tritium monitoring instrument. The uncertainty of the synthesis standard of the activity and concentration of the bubbled tritium flow is less than or equal to 10 percent, and the recovery rate of the tritiated water vapor by the recovery unit is more than 99 percent.

The invention provides a representative data selection method based on sensitivity analysis. According to the method, through the recipe survey around nuclear facilities and the introduction of radiation protection dosimetry, statistics and parameter sensitivity analysis method in the statistic process, the recommended representative personnel recipe can reflect the actual radiation dosage receiving condition of residents around the plant site. The method provided by the invention has the advantages that the representative problems of representative personnel recipe selection in the existing recipe survey can be effectively solved, an effective and optimized selection method of representative personnel recipe data can be provided, and more reasonable and more precise evaluation results can be obtained in the personnel radiation dosage evaluation of nuclear facilities.

The invention relates to leakage testing equipment in the nuclear field, and in particular relates to a halogen generator for inspecting leakage rate of iodine adsorber machinery in a ventilation and purification system of a nuclear facility. The halogen generator structurally comprises a pressure-bearing storage tank used for storing a freon liquid reagent, wherein the pressure-bearing storage tank is respectively connected with a compressed air pipeline, a freon charging pipeline and a freon gasification outlet pipeline, the compressed air pipeline comprises two gas flow paths, one of the gas flow paths is connected with the pressure-bearing storage tank through a three-way switch valve, the other gas flow path is connected with the freon gasification outlet pipeline, and a heater is arranged on the freon gasification outlet pipeline. The halogen generator provided by the invention can continuously and uniformly generate halogen gas with stable concentration, can easily generation conditions, is wide in gas generation range, stable in generation rate, convenient and rapid to operate, and good in adaptability to the scene.

The invention relates to a method for electrokinetic decontamination of a porous solid medium, which comprises:a) extracting the contaminating species present in the solid medium in an electrolyte appearing as an essentially inorganic gel, this extraction being obtained by applying an electric current between two electrodes positioned at the surface and / or in the interior of the solid medium, the contact of at least one of these electrodes with said solid medium being ensured by a layer of said gel,b) drying the gel containing the thereby extracted contaminating species until a dry residue is obtained which fractures, andc) removing the thereby obtained dry residue from said solid residue.Applications: decontamination of materials with a cement matrix, notably within the scope of dismantlement of military or industrial installations, in particular nuclear installations, or the rehabilitation of installations likely to have been contaminated by ecotoxic or radiological chemical species; clearing pollution of geological soils, sediments and sludges.

The invention provides an accident sorting method based on environmental safety accident response. The method comprises the following steps: (1) accident situation analysis; (2) collecting, arrangingand analyzing data in a preset evaluation range of the nuclear facility; (3) estimating accident consequences; and (4) evaluating accident consequences and forming an accident sequence based on environmental safety. According to the method provided by the invention, specific accident estimation can be carried out on an accident scene in INES classification, the environmental influence possibly caused by an accident and the dosage value caused by the public can be mastered in advance, and corresponding technical support is provided for accurately providing accident response measures.

A method for evaluating the public toxicity impact of consequences of nuclear facilities accidents includes the following steps: (1) accident state analysis; (2) radionuclide toxicity analysis and index determination; (3) identification of key resident groups and representative individuals; (4) radiation effects caused by radioactive gaseous effluents and estimation of compound concentration; and(5) accident consequence evaluation. The method provided by the present invention includes not only radiation environmental impact assessment, but also toxicity assessment of toxic compounds (nuclidesand toxic compounds) to the public. In the nuclear accident consequence evaluation, the result can be more accurate.

The invention relates to a rapid analysis method for multinuclides in a radioactive contaminated soil. The method comprises the steps that 1) drying, reducing, grinding and sieving are carried out onthe radioactive contaminated soil to acquire a radioactive contaminated soil sample; 2) 1 .0g radioactive contaminated soil is weighed for digestion; 3) after digestion is completed, liquid in a teflon tank is slowly evaporated at low temperature to nearly dry; 4) the sample is dissolved with a 10mL nitric acid solution and transferred to a centrifuge tube; 5) 0.5mL sample acquired in the step 4)is moved to a small stainless steel plate to prepare an alpha measurement source plate; 6) the total alpha and total beta of the sample are measured on the liquid scintillation counter, and the ratioof 241Am+328Pu and 239+240Pu in the total alpha is acquired; 7) the Cherenkov count of the sample is measured prior to the liquid scintillation counter, and after the end of the measurement, 137Cs and60Co in the sample are measured again on the gamma spectrum; and 8) the acidity is adjusted to 5.0 mol / L, and the U content in the sample is measured through a 30% TBP extraction spectrophotometric method. According to the invention, more than 3,000 radioactive contaminated soils in the decommissioning process of nuclear facilities are analyzed and determined.

A spent fuel storage rack 1 according to the present invention is installed in a fuel storage pool of a nuclear facility, and has a rectangular parallelepiped shape forming a plurality of lattice-like cells 2a that are configured to separately accommodate a plurality of fuel assemblies in a matrix of rows and columns. As shown in FIG. 1, the spent fuel storage rack 1 includes: a base 17 configured to support lower parts of fuel assemblies, the base 17 forming a bottom surface of the spent fuel storage rack 1; an outer frame 3 located above the base 17, the outer frame 3 forming an outermost periphery of the spent fuel storage rack 1; and a lattice body 2 disposed inside the outer frame 3, the lattice body 2 forming the lattice-like cells 2a. The lattice body 2 includes: a main lattice 4 having a height equivalent to an active length of the fuel assembly; an upper lattice 5 disposed above the main lattice 4 so as to be fitted to an upper end of the main lattice 4; and a lower lattice 6 disposed between the base 17 and the main lattice 4 so as to be fitted to a lower end of the main lattice 4. Thus, the main lattice 4 can be formed without welding, whereby the main lattice 4 can be made of a boron-added stainless steel to which a sufficient amount of boron is added to absorb neutrons.