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.

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 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 radiation source intensity reconstruction method based on reverse particle transport. The method comprises the following steps: establishing a particle reverse transport nuclear database; obtaining particle energy and intensity distribution information at a measurement device, and establishing a reverse Monte Carlo radiation transport calculation model in combination with initial source space distribution information, medium image information and measurement device information; sampling the source particles according to particle energy and intensity distribution information at the measurement equipment, and simulating the particles by using a Monte Carlo reverse transport method until the particles are transported to a radiation source; finally, performing reverse transport simulation on all particles, and carrying out normalization processing on particle intensity distribution information at the radiation source obtained through statistics to obtain particle source intensity distribution at the radiation source. According to the invention, particle source intensity distribution can be effectively obtained, and accurate radiation source inversion can be carried out.

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 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 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.