36 results about "Nuclear data" patented technology

A method is for establishing a nuclear reactor core loading pattern (LP) for fuel assemblies and burnable absorbers (BAs). The method establishes an optimum LP through the steps of: a) providing nuclear data representing fuel assemblies and BAs in a nuclear reactor core; b) depleting the nuclear data to form a reference core depletion; c) incorporating the nuclear data into a system of linear equations of a nuclear design quality flux solution method; d) defining the system of linear equations to include constraints which accurately represent the neutron physics of the reactor; employing the equations as a constraint matrix for a MIP solver to find an optimum core pattern solution; f) repeating steps b) through e) updating the constraints and objective functions to satisfy specified engineering requirements and establish an optimum core loading pattern. An algorithm for deriving the system of equations is also disclosed.

An explicit method for constructing continuous energy cross-sections in the indistinguishable resonance region. The position of the resonance peak in the indistinguishable resonance region is obtained by sampling the average resonance energy level spacing, and the resonance widths of different reaction channels at each resonance peak are obtained by sampling the resonance width. Resonance parameters at each resonance peak position not given in the energy point in the evaluation nuclear database are obtained by interpolating the resonance parameters in the evaluation nuclear database; The continuous energy point section at different temperatures can be obtained by using the Le broadening formula; the invention can explicitly construct the continuous energy section of the indistinguishable resonance region, which is used to meet the high-precision calculation requirements of the self-screen section of the indistinguishable resonance region of the modern reactor.

The invention discloses a method for calculating the thermal neutron scattering effect of a silicon carbide coating molten salt reactor, which comprises the following steps: carrying out nuclear dataprocessing on the basis of the thermal neutron scattering rates of liquid molten salt, solid graphite and solid silicon carbide to obtain a thermal neutron scattering cross section library; calling athermal neutron scattering cross section library of liquid molten salt, solid graphite and solid silicon carbide, and carrying out assembly homogenization of the molten salt reactor by adopting a Monte Carlo neutron transport method to obtain homogenization group parameters of typical assemblies in the reactor core and near a reflecting layer; and calling few-group uniform parameters of typical components of the molten salt reactor, and calculating the influence of the thermal neutron scattering effect of the molten salt reactor on the key neutronics parameters of the molten salt reactor for the liquid fuel silicon carbide coating molten salt reactor on the basis of a neutron transport model or a neutron diffusion space-time kinetic model by considering the flow influence of the liquid molten salt. According to the method, the defect that a traditional deterministic theory method is difficult to completely adapt to the complex geometry of the molten salt reactor is overcome, and the obtained homogenization parameters of the molten salt reactor assembly are more accurate.

The present invention discloses a whole energy spectrum neutron radiation damage precise simulation system, which mainly comprises a radiation damage nuclear data module, a high energy physical nuclear reaction module and a radiation damage calculation module. The algorithm of the system comprises: 1, inputting a neutron flux and material component or selecting a fixed radiation environment; 2, inputting the neutron flux and material component into a radiation damage nuclear data module and a high energy physical nuclear reaction module, carrying out radiation damage cross section calling and calculation, and generating the neutron radiation damage DPA and the gas generation cross section of a specified nuclide whole energy spectrum; and 3, according to the called or calculated neutron radiation damage cross section in the step 2, inputting into a radiation damage calculation module, combing with the radiation time of the material and the neutron flux and material component, and performing the whole energy spectrum neutron radiation damage precise calculation in any radiation environments. The system of the present invention can completely meet the radiation damage calculation and simulation of a variety of nuclear energy systems.

A method to minimize the sampling sample size of nuclear reactor physical uncertainty analysis, firstly determine the multi-group cross-sectional population covariance matrix of the nuclear reactor inner element to be analyzed, and then use the Latin hypercube sampling to obtain a group from the same dimension standard normal distribution population The sample, so that the rank of its covariance matrix is ​​not less than the rank of the overall covariance matrix of the multi-group section, linearly transforms the sample, and solves a transformation matrix for linear transformation to ensure that the transformed sample mean and covariance are respectively equal to The mean and covariance of the multi-group cross-section population, and then obtain the calculation samples of the multi-group cross-section input parameters. The invention reconstructs the uncertainty of nuclear data with the minimum sample size, ensures the convergence of uncertainty analysis results, and solves the problems of loss of nuclear data uncertainty information and huge sample size required by traditional sampling methods; the inventive method is easy to implement and easy to use. The calculation efficiency is significantly improved, and the convergence result is accurate and reliable, which is of great significance to the analysis of nuclear reactor physical uncertainty.

The method and the database for obtaining the thermal neutron scattering database of solid and liquid fluorine salts, the method steps are: 1. For solid fluoride salt crystals and liquid fluoride salt solutions, use the lattice dynamics program based on the principle of first quantum mechanics to calculate Its phonon spectrum distribution; In addition, for liquid fluoride salt solution, use the statistical physics method based on molecular dynamics to calculate the diffusion coefficient of each fluoride salt ion; 2, the phonon spectrum distribution and diffusion coefficient obtained in step 1 are provided to The nuclear data processing program NJOY can obtain the thermal neutron scattering data of solid and liquid fluorine salts; 3. According to different application purposes, the corresponding nuclear database index file is provided, and the index file and the thermal neutron scattering data obtained in step 2 form a complete set The database of thermal neutron scattering; the present invention also discloses the database obtained by providing the method; it fills the gap in the field of thermal neutron scattering database production for multi-group transport and subgroup resonance calculation of solid fluoride salt crystals and liquid fluoride salt solutions in the world blank.

The invention belongs to the technical field of nuclear data measurement, and particularly relates to a method and a system for measuring a reaction cross section of a certain element generating a certain nuclide, and the method comprises the following steps: preparing a sample, and carrying out neutron irradiation on the sample; then measuring the total count of all-energy peaks of characteristic gamma rays for generating the same nuclide in the sample after neutron irradiation; and finally, obtaining a generation section of a certain nuclide by using a calculation formula for generating a reaction section of a certain nuclide by a certain element. According to the method, the scientific definition of the reaction section of the certain element for generating the certain nuclide is given, the general calculation formula of the reaction section of the certain element for generating the certain nuclide is given, the formula principle is clear, the calculation is simple and rapid, and the calculation efficiency is high. The method has important significance in production of important isotopes, selection of reactor structure materials, relevant nuclear physics basic research and application research and the like.

The invention belongs to the field of nuclear data measurement. In order to solve the problem of complicated operation of an existing separation method, the invention provides a group separation method. The group separation method comprises the following steps: 1, preparing a mixed carrier solution; 2, preparing a sample to be separated; 3, filling into an HZ201 anion resin column, and leaching with 7 to 10mol / L hydrochloric acid, so as to obtain an Ag solution; 4, leaching with 1 to 3mol / L hydrochloric acid, so as to obtain a Pu-containing solution; 5, leaching with concentrated hydrochloricacid to obtain a solution containing Pd and Cd; 6, leaching with 1 to 2mol / L hydrochloric acid and then leaching with 8 to 10mol / L nitric acid; desorbing to obtain an Sn-containing solution; 7, leaching with de-ionized water and then leaching with 0.2 to 1mol / L ammonia water; desorbing to obtain an Sb-containing solution. According to the group separation method, group separation of plutonium, palladium, silver, cadmium, tin and antimony is realized by adopting one HZ201 anion resin column; separation operation is simplified, and the separation efficiency and recycling efficiency are remarkably improved.

The invention discloses an online detection device for nuclear power plant fuel assembly damage, which comprises a gas sampling chamber, a gamma radiation detector, a beta radiation detector, an air extraction device, an air suction device and a nuclear data processing device; the gamma radiation detector and the beta radiation detector The radiation detector is arranged on the gas sampling chamber, and the gamma radiation detector detects the gamma ray particles in the gas sampling chamber, and the beta radiation detector detects the beta ray particles in the gas sampling chamber. The present invention is an online detection device for nuclear power plant fuel assembly damage, and proposes an online sipping detection device for nuclear power plant fuel assembly damage based on radioactivity measurement β-γ coincidence technology, and uses the correlation of β-γ rays to reduce the background of the device to environmental factors Therefore, the lower detection limit of the device can be improved, thereby ensuring that the device can accurately determine whether the fuel assembly is damaged even at a low radioactivity level, and improve the accuracy of detection.

The invention discloses a method and device for monitoring the running state of a load-carrying vehicle, wherein the method includes: obtaining vehicle model information, driving speed, vibration response measurement data and road surface structure parameters of the load-carrying vehicle, and the vehicle model information includes: nuclear load data and vehicle parameters ; According to the model information of the loaded vehicle, the driving speed and the road surface structure parameters, determine the theoretical data of the vibration response of the loaded vehicle; compare the measured data of the vibration response of the loaded vehicle with the theoretical data of the vibration response, and compare the results according to the comparison Monitor the running status of the truck. The invention can realize the monitoring of the running state of the load-carrying vehicle without arranging the weighing equipment on the built road facilities, reduces the difficulty of monitoring, improves the monitoring efficiency, and ensures the normal passage of the lane.