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.

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.

Methods implemented by at least one electronic processor for generating pointwise fast neutron spectra may include receiving composition data; receiving source data or calculating the source data; receiving nuclear data; and calculating the pointwise fast neutron spectrum based on numerical integration using the composition, source, and nuclear data. Systems for generating pointwise fast neutron spectra may include a bus; at least one electronic processor connected to the bus; an input device connected to the bus; and a communication link connected to the bus. The at least one electronic processor may be configured to receive composition data from the input device via the bus, to receive source data from the input device via the bus or to calculate the source data, to receive nuclear data from the communication link via the bus, and to calculate the pointwise fast neutron spectrum based on numerical integration using the composition, source, and nuclear data.

A radiation detector (16) having a first detector layer (24) and a second detector layer (26) encircles an examination region (14). Detectors of the first layer include scintillators (72) and light detectors (74), such as avalanche photodiodes. The detectors of the second detector layer (26) include scintillators (62) and optical detectors (64). The scintillators (72) of the first layer have a smaller cross-section than the scintillators (62) of the second layers. A group, e.g., nine, of the first layer scintillators (72) overlay each second group scintillator (62). In a CT mode, detectors of the first layer detect transmission radiation to generate a CT image with a relatively high resolution and the detectors of the second layer detect PET or SPECT radiation to generate nuclear data for reconstruction into a lower resolution emission image. Because the detectors of the first and second layers are aligned, the transmission and emission images are inherently aligned.

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.

The invention discloses a nuclear cross section data processing optimization method in MC (Monte Carlo) particle transport simulation. On the traditional MC particle transport simulation continuity point cross section processing and application method, energy grid points of all nuclides involved in the merging problem form a uniform energy grid, one time of searching is only needed to be conducted to the energy grid for each step of transport, the specific position of the current energy in each nuclide energy grid is directly found through a nuclide pointer array obtained through preprocessing, thus the operation of repetitively searching different energy grids of each nuclide in a material in the traditional method is avoided, the times of searching of the energy grid in MC particle transport calculation are greatly reduced and the calculation speed is improved on the premise that the calculation precision is not lost; and according to the distribution characteristics of nuclear data energy points, an energy segmentation concept is introduced to form a dual-layer searching mode of energy grids, thus the times of ineffective searching in areas with higher distribution density are reduced and the grid searching efficiency per time is improved.

The invention discloses a nuclear power station fuel module damage online detection device. The nuclear power station fuel module damage online detection device comprises a gas sampling chamber, a gamma radiation detector, a beta radiation detector, a gas extracting device, a gas sucking device and a nuclear data processing device, wherein the gamma radiation detector and the beta radiation detector are arranged on the gas sampling chamber, the gamma radiation detector detects gamma radiation particles inside the gas sampling chamber, and the beta radiation detector detects beta radiation particles inside the gas sampling chamber. The nuclear power station fuel module damage online detection device is based on beta-gamma compounded radioactivity detection technology, reduces the influenceof device background on environmental factors through association of beta-gamma rays to increase the detection lower limit and further to ensure determination of whether a fuel module is damaged on alow radioactive level, thereby improving the detecting accuracy.

The invention discloses an explicit indistinguishable resonance region continuous energy section construction method, which comprises: sampling average resonance energy level spacing to obtain formant positions of an indistinguishable resonance region, and sampling resonance widths to obtain resonance widths of different reaction channels at each formant; during the period, performing interpolation on resonance parameters in the evaluation kernel database to obtain resonance parameters at all formant positions not in energy points given by the evaluation kernel database; and then, through a multi-energy-level Blade-Wigner formula and a Psi-Chi Doppler broadening formula, obtaining continuous energy point sections at different temperatures. According to the method, the continuous energy cross section of the indistinguishable resonance region can be explicitly constructed, so that the high-precision self-shielding cross section calculation requirement of the indistinguishable resonance region of a modern reactor can be met.

The invention discloses a method for solving the physical response sensitivity of a reactor. The method comprises the steps: firstly obtaining a sampling sample of multi-group nuclear data through covariance matrix decomposition and a multi-dimensional normal distribution sampling technology; and secondly, realizing a forward sensitivity analysis process based on a reduced-order model by using the sampling technology so as to obtain the sensitivity of the physical generalized response of the reactor to the nuclear data. According to the method for solving the physical response sensitivity of the reactor by using covariance matrix decomposition sampling provided by the invention, by combining the reduced-order model and the covariance matrix decomposition sampling technology, the complexity caused by using a generalized perturbation theory is completely avoided, meanwhile, the calculation amount of forward sensitivity analysis is effectively reduced, sensitivity with precision equivalent to that of a direct disturbance method is provided.

The invention discloses a method for calculating a multi-group slow luminescence photon yield by using a fission yield and decay data. The method comprises the following steps: firstly, calculating a multi-group decay photon yield of each nuclide based on a decay photon energy spectrum given in a decay sub-library of an evaluation nuclear database; constructing decay information of each nuclide according to the decay mode and the branching ratio data; secondly, fission yield data under different neutron incident energies are selected according to different requirements; then constructing decay chain information of all fission products until each fission product decays to a stable nuclide; according to fission yield data, adding decay photons generated in the process of decaying fission products to stable nuclides in different energy ranges, so as to obtain the multi-group slow luminescence photon yield of the fission nuclides; and finally, in order to ensure the accuracy of heat release calculation of the slow luminophores, correcting the yield of the multi-group slow luminophores obtained by calculation by using the average fission slow luminophore energy given in the evaluation kernel database. The method is high in calculation precision and calculation efficiency.

The invention relates to the technical field of nuclear reactor cores, and particularly discloses a method for accelerating solving of a generalized conjugate neutron transport equation. The method comprises the following steps: respectively solving a neutron transport equation and a conjugate neutron transport equation by utilizing a characteristic line method, and respectively obtaining corresponding neutron flux distribution; calculating to obtain a source item of the generalized conjugate equation according to a response which specifically needs to be solved, and constructing a generalizedconjugate neutron transport equation; constructing a fixed source solver to solve a generalized conjugate neutron transport equation; and solving to obtain a generalized conjugate neutron transport equation. According to the method, the solving accuracy of the generalized conjugate neutron transport equation is ensured through characteristic line scanning, and the high efficiency of the generalized conjugate neutron transport equation is ensured through coarse net finite difference acceleration; when the method is used for solving the sensitivity coefficient of physical parameters such as thereaction rate ratio and the average fission power of the reactor to nuclear data, the calculation time can be remarkably shortened, and the efficiency is improved.

The invention relates to a method for evaluating keff uncertainty caused by nuclear cross section deviation, and belongs to the technical field of nuclear data evaluation. The method comprises the following steps: S1, selecting nuclides needing to be evaluated in a nuclear system to be analyzed; S2, selecting a nuclear cross section, which needs to be evaluated, of each nuclide; S3, respectively calculating the sensitivity of each nuclear cross section selected in the step S2; S4, selecting covariance data corresponding to each kernel cross section, and adjusting or estimating the covariance data; S5, respectively calculating the uncertainty Unck of each covariance matrix to the keff; and S6, calculating the uncertainty Unc of the keff caused by the nuclear cross section deviation of the nuclear system. According to the method provided by the invention, the influence of each covariance matrix on the keff uncertainty is calculated through the nuclear cross section sensitivity data and the covariance data, so that the uncertain value of the keff caused by the nuclear cross section deviation can be quantitatively obtained, and design, analysis and evaluation can be guided.

The invention discloses a method for manufacturing a fusion reactor multi-group shielding database based on a Monte Carlo method. The method comprises the following steps: 1, processing an evaluation nuclear database into a continuous energy ACE format nuclear database; 2, adding information in all scattering reaction channels of each nuclide in the ACE format database to obtain a continuous energy scattering response coefficient database; 3, designing a simplified one-dimensional model according to the real three-dimensional structure of the fusion reactor; 4, calculating a simplified model based on the ACE format database and the continuous energy scattering response coefficient database by adopting a Monte Carlo method; 5, obtaining a multi-group high-order scattering cross section according to the obtained low-order flux, high-order flux and high-order scattering reaction rate; and 6, combining the multi-group high-order scattering cross section with the total cross section, the neutron generation cross section, the absorption cross section, the low-order scattering cross section and the response function library in the multi-group shielding database to generate a new fusion reactor multi-group shielding database. According to the database, fusion reactor radiation shielding calculation is more accurate, and the safety margin is reduced.

The invention discloses a method and equipment for solving fuel consumption calculation response sensitivity, and the method can solve the sensitivity of calculation response of a transportation fuel consumption coupling program to nuclear data on the premise of not using a generalized perturbation theory. According to the method, the reduced-order model and forward sensitivity analysis are combined, and the sensitivity of all burn-up responses to all nuclear data can be obtained. Compared with a first-order perturbation theory, the method is simpler and more convenient to implement, has universality and is convenient to transplant to other reactor physical design programs.

The invention provides a nuclear reactor physical simulation method and device based on singular value decomposition transformation, computer equipment and a storage medium, which are used for realizing correlation control on a random sample space and reducing the calculation amount of nuclear data sampling, so that an obtained simulation calculation result reaches expectation based on an obtained final sample. The method mainly comprises the following steps: determining the number of physical corresponding parameters of a nuclear reactor; the number of samples needing to be extracted; obtaining a random sample space corresponding to the number of the parameters; performing singular value decomposition on the covariance matrix of the random sample space to obtain a singular value matrix and a left singular matrix, and establishing a corresponding diagonal matrix; constructing a new random sample space according to the relational expression; checking whether the number of parameters in the random sample space reaches the number of parameters; if yes, calculating a final sample through the relational expression, and outputting the final sample; and inputting the final sample into a simulation system to obtain a nuclear reactor physical simulation result.

The invention relates to a statistical and physical combined (n, alpha) reaction cross section experimental data evaluation method, which belongs to the technical field of nuclear data evaluation, and comprises the following steps: investigating and collecting the related information of the existing (n, alpha) reaction cross section experimental data, determining the experimental method of the experimental data, and carrying out mathematical and physical analysis on the experimental data to obtain the experimental data of the (n, alpha) reaction cross section. The distribution rule of experimental data is known, and hidden variables are found; performing classified analysis and evaluation on the experimental data according to the implicit variables to obtain an experimental data evaluation result; and on the basis, a recommended excitation function is obtained by means of mathematical fitting and the like. According to the method, hidden variables influencing data accuracy in Simpson paradox are found from the two aspects of mathematics and physics, the PPP problem can be effectively avoided from the perspective of experimental data evaluation, and the accuracy of evaluation data and the efficiency of evaluation work are effectively improved.

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.

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 discloses a low-voltage power supply for an out-of-reactor nuclear instrument system simulation component and an application method of the low-voltage power supply, relates to the fieldof on-line nuclear instrument and on-line reactor nuclear data measurement of reactors, and solves the problem that the influence of a high-frequency component of a switching power supply on the performance of a conditioning part cannot be eliminated. The low-voltage power supply comprises a low-voltage power supply body supplying power to three channels of an out-of-reactor nuclear instrument system, wherein a rectifier transformer is used for converting alternating-current voltage into four paths of independent direct-current voltage, and a voltage stabilizing circuit independently performslow ripple output processing on the four paths of independent direct-current voltage respectively, so that low ripple output of the direct-current voltage is that VP-P<=10 mV and VR-MS<=2 mV. The low-voltage power supply can eliminate the influence of the high-frequency component of an existing low-voltage power supply on the operation of the nuclear instrument system simulation component, improves the efficiency of the low-voltage power supply, reduces the heating of the low-voltage power supply, and improves the performance of the out-of-reactor nuclear instrument system.

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 relates to a nuclear data measuring device, in particular to a gamma ray angle correlation measuring device and a measuring method based on the same, and solves the problems that when anexisting gamma ray angle correlation measuring device is adopted for measurement, time and labor are consumed, early-stage and later-stage measurement of nuclide with a short half-life period cannotbe considered simultaneously, and the measurement accuracy is difficult to guarantee. The device is characterized in that the device comprises at least six detectors which are circumferentially distributed along a detector laying ring; the central axes of the detectors are positioned on the same plane and penetrate through the circle center of the detector laying ring; the detector laying ring isa virtual circle; at least one included angle of the included angles of the central axes of every two detectors of the at least six detectors is 90 degrees, one included angle is 180 degrees, eight included angles are greater than 90 degrees and less than 180 degrees, or the number of included angles than 90 degrees and less than 180 degrees plus the number of included angles after the included angles greater than 0 degree and less than 90 degrees are converted according to the angle correlation symmetry principle is greater than or equal to 8, and the eight angles are not equal to one another.

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 a method, a device, equipment and a medium for reducing the uncertainty of nuclear data correlation calculation. The method comprises the following steps: carrying out nuclear data correlation sensitivity analysis on a target industrial reactor core and a plurality of critical physical experiments, and obtaining sensitivity coefficients of physical response parameters of all reactor cores to all nuclear data; forming a sensitivity vector of the physical response parameter of the target industrial reactor core to the nuclear data and a sensitivity vector of each critical physical experiment; constructing a relative covariance matrix according to all the nuclear data, and combining the sensitivity vector to obtain the nuclear data related uncertainty of the target industrial reactor core and the critical physical experiment; calculating a similarity coefficient between the target industrial reactor core and the critical physical experiment and a correlation factor between the experiments; and according to the similarity coefficient and the correlation factor, calculating the nuclear data correlation posteriori of the target reactor core and the nuclear data correlation uncertainty of the posteriori. According to the method, critical physical experiment data can be fully utilized, and a nuclear data adjustment process is avoided.

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.