Macroscopic cross-section database construction method, dose determination method, apparatus and device

Abstract

The application relates to the technical field of medical data processing, and discloses a macroscopic cross-section database construction method, a dose determination method, a device and equipment, which comprises the following steps: setting multiple boron concentration values for the tissue types corresponding to a human body model; selecting corresponding neutron source parameters for the tissue types and the multiple boron concentration values, wherein the neutron source parameters comprise energy range values; dividing the energy range values into multiple discrete neutron energy groups; performing particle transport simulation based on the selected neutron source parameters, jumping between different neutron energy groups in the simulation process, and obtaining neutron reaction rates and neutron fluxes of the tissue types under the multiple boron concentration values and the multiple neutron energy groups; and obtaining neutron macroscopic cross-section values of the tissue types under the multiple boron concentration values and the multiple neutron energy groups based on the neutron reaction rates and the neutron fluxes, and constructing a multi-group macroscopic cross-section database. The application can improve the calculation efficiency of the radiation dose and improve the adaptability to individual metabolic differences during dose calculation.

Description

Technical Field

[0001] This invention relates to the field of medical data processing technology, specifically to a method for constructing a macroscopic cross-section database, a method for determining dosage, an apparatus, and equipment. Background Technology

[0002] Boron neutron capture therapy (BNCT) is a binary targeted radiotherapy method based on nuclear reactions. Currently, the effectiveness of BNCT largely depends on the accuracy of dose calculation, and the treatment planning system (TPS) is a crucial component of the entire BNCT process. Accurate dose calculation ensures that tumor tissue receives a sufficiently high radiation dose, effectively killing tumor cells, improving local tumor control, and reducing the risk of tumor recurrence. Current BNCT treatment planning systems all employ the Monte Carlo method as the core algorithm for dose calculation. Through extensive random sampling and statistical analysis, they simulate the transport process of neutrons in human tissue and their interaction with boron atoms, thus obtaining relatively accurate dose distribution results. During the simulation, detailed physical parameters need to be input, such as the neutron energy spectrum, the composition and density of human tissue, and the concentration of boron drugs. Through multiple simulations, the energy deposition at different locations is statistically analyzed, thereby deriving the dose distribution.

[0003] Monte Carlo programs based on continuous energy spectra (such as the Monte Carlo N-Particle Transport Code (MCNP) and the Open Monte Carlo Neutron Transport Code (OpenMC)) are highly accurate, but they usually require simulating particle reactions at every energy level, resulting in low computational efficiency. Consequently, when dealing with complex structures (such as complete human models) and multiple sets of boron concentration parameters, the computation time is extremely long, making it difficult to meet the need for "rapid dose feedback". Summary of the Invention

[0004] In view of this, the present invention provides a method for constructing a macroscopic cross-section database, a method for determining dose, an apparatus and equipment to solve the problem of low efficiency in determining radiation dose.

[0005] In a first aspect, the present invention provides a method for constructing a macroscopic cross-section database, comprising:

[0006] Multiple boron concentration values ​​were set for the tissue types corresponding to the human body model;

[0007] For each tissue type and the corresponding multiple boron concentration values, a corresponding neutron source parameter is selected, wherein the neutron source parameter includes an energy range value;

[0008] The energy range is divided into multiple discrete neutron energy groups;

[0009] Based on the selected neutron source parameters, particle transport simulations are performed, with the simulation process jumping between different neutron energy groups to obtain the neutron reaction rate and neutron flux of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups.

[0010] Based on the neutron reaction rate and the neutron flux, the neutron macroscopic cross-section values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups are obtained, and a multi-group macroscopic cross-section database is constructed. Each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-section value.

[0011] In one optional implementation, obtaining the neutron macroscopic cross-sectional values ​​of the tissue type at multiple boron concentration values ​​and multiple neutron energy groups based on the neutron reaction rate and the neutron flux includes:

[0012] Based on the neutron reaction rate and neutron flux distribution of each neutron energy group in each voxel grid corresponding to the tissue type, the neutron macroscopic cross-sectional value of each voxel grid is obtained;

[0013] Based on the neutron macroscopic cross-sectional values ​​of each voxel grid, the neutron macroscopic cross-sectional values ​​corresponding to the tissue type are obtained.

[0014] In one optional implementation, the neutron energy group is divided in a non-uniform manner;

[0015] The neutron energy group partitioning interval for thermal neutrons and / or hyperthermal neutrons is smaller than that for fast neutrons.

[0016] In a second aspect, the present invention provides a dosage determination method, the method comprising:

[0017] A multi-group macroscopic cross-section database is obtained, wherein each data in the multi-group macroscopic cross-section database consists of tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-section value;

[0018] Obtain the target parameters of the target object; the target parameters include the target tissue type and the target boron concentration value corresponding to the target tissue.

[0019] Based on the multi-group macroscopic cross-section database, the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameters is obtained;

[0020] Based on the target neutron macroscopic cross-sectional value, the dose for the target object is obtained.

[0021] In one optional implementation, prior to acquiring the multi-group macroscopic cross-section database, the method further includes:

[0022] Multiple boron concentration values ​​were set for the tissue types corresponding to the human body model;

[0023] For each tissue type and the corresponding multiple boron concentration values, a corresponding neutron source parameter is selected, wherein the neutron source parameter includes an energy range value;

[0024] The energy range is divided into multiple discrete neutron energy groups;

[0025] Based on the selected neutron source parameters, particle transport simulations are performed, with the simulation process jumping between different neutron energy groups to obtain the neutron reaction rate and neutron flux of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups.

[0026] Based on the neutron reaction rate and the neutron flux, the neutron macroscopic cross-section values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups are obtained, and the multi-group macroscopic cross-section database is constructed.

[0027] In one optional implementation, obtaining the neutron macroscopic cross-sectional values ​​of the tissue type at multiple boron concentration values ​​and multiple neutron energy groups based on the neutron reaction rate and the neutron flux includes:

[0028] Based on the neutron reaction rate and neutron flux distribution of each neutron energy group in each voxel grid corresponding to the tissue type, the neutron macroscopic cross-sectional value of each voxel grid is obtained;

[0029] Based on the neutron macroscopic cross-sectional values ​​of each voxel grid, the neutron macroscopic cross-sectional values ​​corresponding to the tissue type are obtained.

[0030] In one optional implementation, obtaining the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameter based on the multi-group macroscopic cross-section database includes:

[0031] From the multi-group macroscopic cross-section database, obtain the first neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the first boron concentration value, and the second neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the second boron concentration value; the first boron concentration value and the second boron concentration value are two boron concentration values ​​adjacent to the target boron concentration value in the multi-group macroscopic cross-section database;

[0032] The target neutron macroscopic cross-section value is determined based on the first boron concentration value, the second boron concentration value, the target boron concentration value, the first neutron macroscopic cross-section value, and the second neutron macroscopic cross-section value.

[0033] In one optional implementation, obtaining the dose for the target object based on the target neutron macroscopic cross-sectional value includes:

[0034] The neutron flux corresponding to each neutron energy group, the energy released when a macroscopic neutron reaction occurs, and the density of the target tissue are obtained.

[0035] The dose is obtained based on the neutron flux corresponding to each neutron energy group, the energy released when a macroscopic neutron reaction occurs, the density of the target tissue, and the macroscopic cross-section value of the target neutron.

[0036] Thirdly, the present invention provides a dosage determination device, the device comprising:

[0037] The boron concentration setting module is used to set multiple boron concentration values ​​for the tissue types corresponding to the human body model.

[0038] The neutron source parameter selection module is used to select the corresponding neutron source parameters for the tissue type and the corresponding multiple boron concentration values, wherein the neutron source parameters include energy range values;

[0039] The energy group partitioning module is used to divide the energy range value into multiple discrete neutron energy groups;

[0040] The particle transport module is used to perform particle transport simulation based on the selected neutron source parameters. The simulation process jumps between different neutron energy groups to obtain the neutron reaction rate and neutron flux of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups.

[0041] The database construction module is used to obtain the neutron macroscopic cross-sectional values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups based on the neutron reaction rate and the neutron flux, and to construct a multi-group macroscopic cross-sectional database. Each data in the multi-group macroscopic cross-sectional database consists of tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-sectional value.

[0042] Fourthly, the present invention provides a dosage determination device, the device comprising:

[0043] The database acquisition module is used to acquire a multi-group macroscopic cross-section database. Each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-section value.

[0044] The target parameter acquisition module is used to acquire the target parameters of the target object; the target parameters include the target tissue type and the target boron concentration value corresponding to the target tissue.

[0045] The cross-section data acquisition module is used to acquire the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameters based on the multi-group macroscopic cross-section database.

[0046] The dose simulation module is used to obtain the dose for the target object based on the macroscopic cross-sectional value of the target neutron.

[0047] Fifthly, the present invention provides a computer device, comprising: a memory and a processor, the memory and the processor being communicatively connected to each other, the memory storing computer instructions, and the processor executing the computer instructions to perform the macroscopic cross-section database construction method of the first aspect or any corresponding embodiment thereof, or to perform the dose determination method of the second aspect or any corresponding embodiment thereof.

[0048] In a sixth aspect, the present invention provides a computer-readable storage medium storing computer instructions, which are used to cause a computer to perform the macroscopic cross-section database construction method of the first aspect or any corresponding embodiment thereof, or to perform the dose determination method of the second aspect or any corresponding embodiment thereof.

[0049] In a seventh aspect, the present invention provides a computer program product, including computer instructions, which are used to cause a computer to execute the macroscopic cross-section database construction method of the first aspect or any corresponding embodiment thereof, or to execute the dose determination method of the second aspect or any corresponding embodiment thereof.

[0050] The macroscopic cross-section database construction method, dose determination method, apparatus, and equipment provided in this embodiment construct a three-dimensional multi-group macroscopic cross-section database with practical physical meaning by using tissue type, boron concentration, and neutron energy groups as joint variables. This three-dimensional multi-group macroscopic cross-section database is constructed based on discrete neutron energy groups, that is, based on the multi-group Monte Carlo (MND) method. The multi-group Monte Carlo method divides the continuous energy range of neutrons into several discrete neutron energy groups, pre-calculates and stores the neutron macroscopic cross-section values ​​of each neutron energy group, thereby greatly reducing the computational complexity, significantly improving the computational speed, and thus improving the efficiency of radiation dose determination.

[0051] The dosage determination method provided in this embodiment is applicable to the efficient simulation of dosage distribution and individualized treatment assessment in boron neutron capture therapy (BNCT) systems.

[0052] Furthermore, this invention also achieves a precise response of neutron macroscopic cross-sectional values ​​to physiological tissue variability and boron drug concentration changes. Specifically, individual metabolic differences lead to differences in boron distribution. Therefore, this invention establishes a homogenized multi-group macroscopic cross-sectional database related to boron concentration. This database includes neutron macroscopic cross-sectional data corresponding to multiple boron concentrations in multiple different tissue types, thereby enabling a dynamic response of neutron macroscopic cross-sectional values ​​to changes in boron concentration. This meets the clinical need for modeling boron distribution differences caused by individual metabolic variations and enhances the adaptability of the treatment planning system to individual differences. Attached Figure Description

[0053] To more clearly illustrate the technical solutions in the specific embodiments or related technologies of the present invention, the drawings used in the description of the specific embodiments or related technologies will be briefly introduced below. Obviously, the drawings described below are some embodiments of the present invention. For those skilled in the art, other drawings can be obtained from these drawings without creative effort.

[0054] Figure 1 This is a flowchart illustrating a method for constructing a macroscopic cross-section database according to an embodiment of the present invention;

[0055] Figure 2 This is a schematic flowchart of a dosage determination method according to an embodiment of the present invention;

[0056] Figure 3 This is a structural block diagram of a macroscopic cross-section database construction apparatus according to an embodiment of the present invention;

[0057] Figure 4 This is a structural block diagram of a dose determination device according to an embodiment of the present invention;

[0058] Figure 5 This is a schematic diagram of the hardware structure of a computer device according to an embodiment of the present invention. Detailed Implementation

[0059] To make the objectives, technical solutions, and advantages of the embodiments of the present invention clearer, the technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort are within the scope of protection of the present invention.

[0060] According to an embodiment of the present invention, a method for constructing a macroscopic cross-section database is provided. It should be noted that the steps shown in the flowchart in the accompanying drawings can be executed in a computer system such as a set of executable computer instructions. Furthermore, although a logical order is shown in the flowchart, in some cases, the steps shown or described may be executed in a different order than that shown here.

[0061] This embodiment provides a method for constructing a macroscopic cross-section database, which can be used on various computer devices. Figure 1 This is a flowchart of a macroscopic cross-section database construction method according to an embodiment of the present invention, such as... Figure 1 As shown, the process includes the following steps:

[0062] Step S101: Set multiple boron concentration values ​​for the tissue type corresponding to the human body model.

[0063] In this embodiment, a human body model can be constructed based on tissue type. A single human body model can correspond to multiple tissue types or only one. If a human body model corresponds to multiple tissue types, multiple model regions can be constructed to form the human body model, with each model region corresponding to one tissue type, and all having consistent structure and uniform material density. In this embodiment of the invention, through homogenization processing, a single region of a complex human body model is considered as a volumetric unit with consistent structure and uniform material density, thereby obtaining representative macroscopic neutron cross-sectional values ​​and improving the representativeness and physical consistency of the data. Tissue types include, for example, bones, blood, and various organs. Specific tissue type classifications can be based on reports published by the International Commission on Radiological Protection (ICRP).

[0064] The multiple boron concentration values ​​can be set from 0 to 100 ppm with an interval of 1 ppm.

[0065] Step S102: For the tissue type and the corresponding multiple boron concentration values, select corresponding neutron source parameters, whereby the neutron source parameters include an energy range value. This energy range value can be, for example, 0-10 MeV. Neutron source parameters may also include direction and / or angle.

[0066] Step S103: Divide the energy range value into multiple discrete neutron energy groups.

[0067] Step S104: Perform particle transport simulation based on the selected neutron source parameters. The simulation process jumps between different neutron energy groups to obtain the neutron reaction rate and neutron flux of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups.

[0068] Specifically, the open-source Monte Carlo particle transport simulation program (OpenMC) can be used to simulate particle transport. In the boron neutron capture therapy scenario, the neutron flux and neutron reaction rate of each neutron energy group in different tissue regions (corresponding to different tissue types) and under different boron concentrations in a human model can be calculated. Other continuous-energy neutron transport simulation programs can also be used. Furthermore, before conducting particle transport simulations, it is necessary to construct a continuous-energy neutron source and establish a moderator.

[0069] In this embodiment of the invention, neutron energy can be divided from a continuous interval into discrete neutron energy groups according to the statistics of the open-source Monte Carlo particle transport simulation program (OpenMC). Specifically, based on the nuclear reaction cross-section, neutron energy can be divided into several neutron energy groups, using a non-uniform method to divide the neutron energy groups, with dense division for thermal neutrons and hyperthermal neutrons, and sparse division for fast neutrons. That is, the neutron energy group division interval for thermal neutrons and / or hyperthermal neutrons is smaller than that for fast neutrons.

[0070] Step S105: Based on the neutron reaction rate and the neutron flux, obtain the neutron macroscopic cross-section values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups, and construct a multi-group macroscopic cross-section database. That is, for a given tissue type and a given boron concentration value, the neutron macroscopic cross-section value corresponding to a given neutron energy group is constructed as a single data entry in the multi-group macroscopic cross-section database: the neutron macroscopic cross-section value corresponding to a given tissue type and a given boron concentration value for a given neutron energy group.

[0071] In this embodiment of the invention, particle transport simulations are performed for each tissue type at multiple different boron concentration values ​​to construct a macroscopic cross-section database under multiple tissue types, multiple boron concentration values, and multiple neutron energy groups. Subsequently, neutron macroscopic cross-section values ​​for any tissue type, at any boron concentration, and for any neutron energy group can be obtained through interpolation methods.

[0072] The macroscopic cross-section database construction method provided in this embodiment uses tissue type, boron concentration, and neutron energy group as joint variables to construct a three-dimensional multi-group macroscopic cross-section database with practical physical meaning. This three-dimensional multi-group macroscopic cross-section database is constructed based on discrete neutron energy groups, that is, based on the multi-group Monte Carlo (multi-group Monte Carlo) method. The multi-group Monte Carlo method divides the continuous energy range of neutrons into several discrete neutron energy groups, pre-calculates and stores the neutron macroscopic cross-section values ​​of each neutron energy group, thereby greatly reducing the computational complexity, significantly improving the computational speed, and thus improving the efficiency of radiation dose determination.

[0073] In other words, the multi-group Monte Carlo (MRC) method does not track the precise continuous energy of particles. Instead, it divides the neutron energy into several discrete neutron energy groups and performs probability jumps between groups without tracking the continuous energy. It only counts the reaction behavior within each neutron energy group. The particle propagation, reaction, and energy deposition behavior of each neutron energy group are described by the corresponding macroscopic cross-sectional value of the group neutrons, which greatly reduces the computational complexity and improves the computational speed.

[0074] The dosage determination method provided in this embodiment is applicable to the efficient simulation of dosage distribution and individualized treatment assessment in boron neutron capture therapy (BNCT) systems.

[0075] Furthermore, the macroscopic cross-section database provided in this embodiment of the invention can express the variation law of neutron macroscopic cross-section values ​​of the same tissue under different boron concentrations, enabling the treatment planning system to accurately model individual pharmacokinetics differences and achieve precise response of neutron macroscopic cross-section values ​​to physiological tissue differences and boron drug concentration changes. Specifically, individual metabolic differences lead to differences in boron distribution. Therefore, this embodiment of the invention establishes a homogenized multi-group macroscopic cross-section database related to boron concentration. The multi-group macroscopic cross-section database includes neutron macroscopic cross-section value data corresponding to multiple different boron concentrations for multiple different tissue types, thereby enabling dynamic response of neutron macroscopic cross-section values ​​to changes in boron concentration. This meets the clinical need for modeling boron distribution differences caused by individual metabolic differences and enhances the adaptability of the treatment planning system to individual differences.

[0076] In some optional embodiments, step S105, namely, obtaining the neutron macroscopic cross-sectional values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups based on the neutron reaction rate and the neutron flux, includes:

[0077] Step S1051: Based on the neutron reaction rate and neutron flux distribution of each neutron energy group in each voxel grid corresponding to the tissue type, obtain the neutron macroscopic cross-sectional value of each voxel grid.

[0078] In this embodiment, when obtaining the neutron reaction rate and neutron flux distribution of each voxel grid based on particle transport simulation, the statistics are based on the neutron energy group. Therefore, the data obtained are the neutron reaction rate and neutron flux distribution data of each neutron energy group in each voxel network. That is, the neutron reaction rate and neutron flux distribution of each neutron energy group in all voxel grids are statistically analyzed for each tissue type. The neutron reaction rate may include, for example, the total macroscopic reaction rate, absorption reaction rate, scattering reaction rate, and / or trapping reaction rate. Accordingly, the neutron macroscopic cross-section value in this embodiment may include the total reaction cross-section value, scattering reaction cross-section value, absorption reaction cross-section value, and / or trapping reaction cross-section value. The formula for calculating the neutron macroscopic cross-section value of a voxel network for a certain tissue type may be, for example:

[0079]

[0080] in, For a voxel network of a tissue type at a boron concentration value, the first The macroscopic cross-section value of the neutron energy group. For this tissue type, the voxel grid is the first Neutron reaction rate of each neutron energy group For this tissue type, the voxel grid is the first Neutron flux of a neutron energy group.

[0081] Specifically, for a voxel network of a given tissue type, the total macroscopic reaction rate, absorption reaction rate, scattering reaction rate, and capture reaction rate of a neutron energy group can be divided by the neutron flux of that neutron energy group to obtain the total reaction cross section, scattering reaction cross section, absorption reaction cross section, and capture reaction cross section of that neutron energy group.

[0082] Step S1052: Based on the neutron macroscopic cross-sectional values ​​of each voxel grid, obtain the neutron macroscopic cross-sectional value corresponding to the tissue type. For example, the neutron macroscopic cross-sectional value corresponding to the tissue type can be obtained by calculating the average of the neutron macroscopic cross-sectional values ​​of each voxel network of the same tissue type.

[0083] According to an embodiment of the present invention, a dosage determination method embodiment is provided. It should be noted that the steps shown in the flowchart in the accompanying drawings can be executed in a computer system such as a set of executable computer instructions. Furthermore, although a logical order is shown in the flowchart, in some cases, the steps shown or described may be performed in a different order than that shown here.

[0084] This embodiment provides a dosage determination method that can be used with various computer devices. Figure 2 This is a flowchart of a dosage determination method according to an embodiment of the present invention, such as... Figure 2As shown, the process includes the following steps:

[0085] Step S201: Obtain a multi-group macroscopic section database (also known as a multi-group macroscopic section parameter library). Each data in the multi-group macroscopic section database consists of tissue type, boron concentration value, neutron energy group, and neutron macroscopic section value.

[0086] Specifically, a multi-group macroscopic cross-section database can include multiple data entries, each of which must include at least: a specific tissue type, a specific boron concentration value, and a neutron macroscopic cross-section value corresponding to a specific neutron energy group. Furthermore, among these multiple data entries, at least one of the three parameters—tissue type, boron concentration value, and neutron energy group—must differ in value.

[0087] The specific storage format of the data in the multi-group macroscopic cross-section database can be tabular or other formats, which is not limited here.

[0088] Specifically, the aforementioned multi-group macroscopic cross-section database can be a homogenized multi-group macroscopic cross-section library for boron neutron capture therapy systems, built based on the open-source Monte Carlo particle transport simulation program (OpenMC) platform. The specific construction process of the multi-group macroscopic cross-section database can be found in the above-mentioned implementation examples of macroscopic cross-section database construction methods, and will not be repeated here.

[0089] Tissue types include bones, blood, and various organs, and specific tissue type classifications can be made according to reports published by the International Commission on Radiological Protection (ICRP).

[0090] Step S202: Obtain the target parameters of the target object; the target parameters include the target tissue type and the target boron concentration value corresponding to the target tissue.

[0091] Specifically, the target object can be a patient, and the target tissue can be the tissue where the lesion is located. The boron concentration in the target tissue can be estimated by examining the target object's positron emission tomography-computed tomography (PET-CT) images and combining them with a pre-established pharmacokinetic model. This model is based on extensive preclinical studies (animal experiments) and early clinical trial data.

[0092] Step S203: Based on the multi-group macroscopic cross-section database, obtain the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameters.

[0093] Specifically, after obtaining the target tissue type and target boron concentration value, these two determined parameters are used to search (or match) in a multi-group macroscopic cross-section database. If the target tissue type and target boron concentration value corresponding to the target tissue can be found directly, then the target neutron macroscopic cross-section value corresponding to each neutron energy group is directly obtained. If the same boron concentration value cannot be found directly in the multi-group macroscopic cross-section database, then the target neutron macroscopic cross-section value corresponding to each neutron energy group is obtained using the following method.

[0094] Specifically, step S203, namely, obtaining the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameters based on the multi-group macroscopic cross-section database, includes:

[0095] Step S2031: Obtain from the multi-group macroscopic cross-section database the first neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the first boron concentration value, and the second neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the second boron concentration value; the first boron concentration value and the second boron concentration value are two boron concentration values ​​adjacent to the target boron concentration value in the multi-group macroscopic cross-section database.

[0096] That is, the target boron concentration value is greater than the first boron concentration value and less than the second boron concentration value. If the boron concentration values ​​in the multi-group macroscopic cross-section database are 100 integer values ​​from 1 to 100, then the target boron concentration value can be rounded down to obtain the first boron concentration value, and then 1 can be added to the first boron concentration value to obtain the second boron concentration value.

[0097] Step S2032: Based on the first boron concentration value, the second boron concentration value, the target boron concentration value, the first neutron macroscopic cross section value, and the second neutron macroscopic cross section value, determine the target neutron macroscopic cross section value corresponding to each neutron energy group.

[0098] For example, the macroscopic cross-sectional value of the target neutron can be obtained using a linear interpolation algorithm. For a given tissue type and a neutron energy group at a given target concentration, the macroscopic cross-sectional value of the target neutron can be calculated using the following formula:

[0099]

[0100] in, For the target neutron macroscopic cross-sectional value, This represents the first macroscopic cross-sectional value of neutrons at the first boron concentration. This represents the macroscopic cross-sectional value of the second neutron at the second boron concentration. The target boron concentration value, This is the first boron concentration value. This is the second boron concentration value.

[0101] The fast neutron macroscopic cross-sectional value acquisition method based on linear interpolation algorithm provided in this invention supports real-time calculation of multiple Monte Carlo groups at arbitrary boron concentrations, reducing calculation time while ensuring computational efficiency.

[0102] This invention addresses the critical neutron macroscopic cross-section in boron neutron capture therapy by establishing a multi-group macroscopic cross-section database with boron concentration as the independent variable. Through an interpolation algorithm, the neutron macroscopic cross-section values ​​corresponding to continuous concentrations are obtained in real time during multi-group Monte Carlo simulations. This enables flexible adaptation to individualized treatment needs and improves simulation efficiency while maintaining reasonable accuracy, making it more suitable for rapid preclinical dose planning and optimization.

[0103] Furthermore, the neutron energy group range in the multi-group macroscopic cross-section database can be set to be relatively wide, while the actual neutron energy group range required for dose simulation of the target object can be narrower than the neutron energy group range in the multi-group macroscopic cross-section database. In this case, when obtaining the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameters, it is only necessary to obtain the neutron macroscopic cross-section value corresponding to the required neutron energy group.

[0104] Step S204: Based on the target neutron macroscopic cross-sectional value, obtain the dose for the target object. Specifically, it is the dose for the target tissue of the target object.

[0105] Specifically, the aforementioned doses, also known as absorbed doses, include: boron dose, photon dose, nitrogen dose, and hydrogen dose. The boron dose is determined by the high-energy particles (α and β) produced by the nuclear reaction between neutrons and boron-10. 7 The energy of Li deposited in tissues, photon dose is the energy of secondary electrons (such as recoil electrons) generated by γ photons deposited in tissues, and nitrogen dose is mainly composed of... 14 N(n,p) 14 The C reaction produces a recoil gas. 14 The C nucleus and 583 keV protons, the hydrogen dose is the energy of recoil protons deposited in the tissue by elastic scattering of neutrons with hydrogen atoms.

[0106] Monte Carlo (MC) simulations are considered the "gold standard" in radiation dosimetry. In this embodiment of the invention, MC simulations are used to calculate the boron neutron capture therapy dose. MC simulations can accurately simulate neutron slowing, transport, and the spatial distribution of the four dose components (boron dose, nitrogen dose, hydrogen dose, and photon dose). Monte Carlo simulations simulate the macroscopic behavior of the entire radiation field by tracking the random walk history (from creation to destruction) of a large number of individual particles (such as photons, electrons, and neutrons). Finally, the dose is calculated by statistically analyzing the energy deposited by these particles in the region of interest. Monte Carlo simulations rely on the neutron macroscopic cross section (Σ) for particle transport calculations. The neutron macroscopic cross section is the core of the MC particle transport algorithm, determining how far a particle can travel before its next interaction and what type of interaction will occur next.

[0107] Furthermore, as mentioned above, Monte Carlo simulations are computationally extremely expensive, requiring the tracking of a vast number of particles (millions to hundreds of millions) to obtain results with low statistical noise, which is very time-consuming (from hours to days). During Monte Carlo simulations, it is necessary to frequently query relevant macroscopic cross-section databases to obtain neutron macroscopic cross-section values ​​for particles at specific energies, a computationally intensive operation. In this embodiment of the invention, the continuous energy range is divided into discrete neutron energy groups, and a multi-group macroscopic cross-section database based on these neutron energy groups is pre-constructed. This database stores the neutron macroscopic cross-section values ​​for each neutron energy group. Therefore, during the simulation, particles only need to find the corresponding neutron energy group based on their energy and then directly call the pre-stored neutron macroscopic cross-section values, avoiding a large amount of real-time interpolation calculations and database queries, thus greatly accelerating the simulation speed.

[0108] It should be noted that the dosage determination method provided in this embodiment of the invention provides dosages for doctors' reference only, and the specific dosage shall be ultimately determined by the doctor based on the actual situation.

[0109] The dose determination method provided in this embodiment uses tissue type, boron concentration, and neutron energy groups as joint variables to construct a three-dimensional multi-group macroscopic cross-section database with practical physical meaning. This database is constructed based on discrete neutron energy groups, specifically the multi-group Monte Carlo (MCM) method. MCM divides the continuous energy range of neutrons into several discrete neutron energy groups, pre-calculates and stores the macroscopic cross-section values ​​of each neutron energy group, thereby greatly reducing computational complexity, significantly improving computational speed, and ultimately enhancing the efficiency of radiation dose determination. The dose determination method provided in this embodiment is suitable for efficient simulation of dose distribution and individualized treatment assessment in boron neutron capture therapy (BNCT) systems.

[0110] Furthermore, this invention also achieves a precise response of neutron macroscopic cross-sectional values ​​to physiological tissue variability and boron drug concentration changes. Specifically, individual metabolic differences lead to differences in boron distribution. Therefore, this invention establishes a homogenized multi-group macroscopic cross-sectional database related to boron concentration. This database includes neutron macroscopic cross-sectional data corresponding to multiple boron concentrations in multiple different tissue types, thereby enabling a dynamic response of neutron macroscopic cross-sectional values ​​to changes in boron concentration. This meets the clinical need for modeling boron distribution differences caused by individual metabolic variations and enhances the adaptability of the treatment planning system to individual differences.

[0111] In one optional implementation, step S204, namely obtaining the dose for the target object based on the target neutron macroscopic cross-sectional value, includes:

[0112] Step S2041: Obtain the neutron flux corresponding to each neutron energy group, the energy released when a neutron-related reaction occurs, and the density of the target tissue.

[0113] The neutron-related reactions here include the overall reaction, scattering reaction, absorption reaction and / or capture reaction.

[0114] Step S2042: Based on the neutron flux corresponding to each neutron energy group, the energy released when a neutron-related reaction occurs, the density of the target tissue, and the macroscopic cross-sectional value of the target neutron, the dose is obtained.

[0115] Specifically, as described above, the target neutron macroscopic cross-section value is a plurality of neutron macroscopic cross-section values ​​corresponding to each neutron energy group under the target parameters (including the type of target tissue of the target object and the target boron concentration value of the target tissue). The dose within a single voxel grid, including boron dose, photon dose, nitrogen dose, and hydrogen dose, can be calculated using the following formula:

[0116]

[0117] in, The dose per unit time (boron dose, photon dose, nitrogen dose, or hydrogen dose, in Gy / s). For the first Neutron flux of each neutron energy group (in n / m) 2 ·s), For the first The macroscopic cross-section of the target neutron energy group (in units of) ), Energy released during a given reaction (overall reaction, scattering reaction, absorption reaction, or capture reaction) is expressed in kJ. Density of the target tissue (unit: kg / m³) 3The density of the target tissue can be obtained from data acquired by medical devices (such as CT image data).

[0118] In this embodiment of the invention, the process of determining the dose for the target object can be implemented by a MultiGroup-MC (MultiGroup Monte Carlo) dosing calculation program. In other words, the dose determination method provided in this embodiment of the invention can be implemented by the MultiGroup-MC program. The MultiGroup-MC program skips the complex microscopic data processing and reaction sampling process in traditional continuous energy Monte Carlo simulations, significantly improving the speed and scalability of dose calculation in boron neutron capture therapy plans, and is suitable for large-scale patient simulations and protocol evaluations. The neutron flux corresponding to each neutron energy group and the energy released during each reaction are obtained by the MultiGroup-MC program during the dose simulation process.

[0119] That is, the dose determination method provided in this embodiment of the invention, after using OpenMC to construct a macroscopic cross-section library of neutrons in each neutron energy group under different tissue types and different boron concentrations, obtains the tissue type and boron concentration for each voxel grid, and then uses a linear interpolation strategy in the multi-group macroscopic cross-section database according to the tissue type and boron concentration corresponding to the current grid to obtain the neutron macroscopic cross-section value of each neutron energy group under a specific boron concentration.

[0120] In summary, the dosage determination method provided in this invention is applicable to boron neutron capture therapy (BNCT) systems. It employs a homogenized multi-group neutron macroscopic cross-section calculation method and a multi-group Monte Carlo dosage calculation method. Because the macroscopic cross-section database is pre-calculated and the dosage calculation is efficient and controllable, it meets the real-time requirements of boron neutron capture therapy planning systems. This method has the capability to be embedded in a TPS (Total Photonics System) and can serve as a technical support module for real-time applications such as pre-treatment assessment and online optimization in boron neutron capture therapy. Specifically, this invention not only considers tissue spatial heterogeneity and energy spectrum characteristics but also supports differences in physical parameters caused by variations in boron concentration, ultimately achieving rapid and precise personalized treatment dosage assessment.

[0121] This embodiment also provides a macroscopic cross-section database construction apparatus, which is used to implement the above embodiments and preferred embodiments; details already described will not be repeated. As used below, the term "module" can be a combination of software and / or hardware that implements a predetermined function. Although the apparatus described in the following embodiments is preferably implemented in software, hardware implementation, or a combination of software and hardware, is also possible and contemplated.

[0122] This embodiment provides a macroscopic cross-section database construction device, such as... Figure 3 As shown, it includes:

[0123] Boron concentration setting module 301 is used to set multiple boron concentration values ​​for the tissue type corresponding to the human body model;

[0124] The neutron source parameter selection module 302 is used to select corresponding neutron source parameters for the tissue type and the corresponding multiple boron concentration values, wherein the neutron source parameters include energy range values;

[0125] The energy group partitioning module 303 is used to partition the energy range value into multiple discrete neutron energy groups;

[0126] The particle transport module 304 is used to perform particle transport simulation based on the selected neutron source parameters. The simulation process jumps between different neutron energy groups to obtain the neutron reaction rate and neutron flux of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups.

[0127] The database construction module 305 is used to obtain the neutron macroscopic cross-section values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups based on the neutron reaction rate and the neutron flux, and to construct a multi-group macroscopic cross-section database. Each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-section value.

[0128] In some optional embodiments, the database construction module 305 includes:

[0129] The first cross-sectional value acquisition unit is used to acquire the neutron macroscopic cross-sectional value of each voxel grid based on the neutron reaction rate and neutron flux distribution of each voxel grid corresponding to the tissue type.

[0130] The second cross-sectional value acquisition unit is used to acquire the neutron macroscopic cross-sectional value corresponding to the tissue type based on the neutron macroscopic cross-sectional value of each voxel grid.

[0131] In some optional embodiments, the neutron energy group is divided in a non-uniform manner;

[0132] The neutron energy group partitioning interval for thermal neutrons and / or hyperthermal neutrons is smaller than that for fast neutrons.

[0133] Further functional descriptions of the above modules and units are the same as those in the corresponding embodiments described above, and will not be repeated here.

[0134] In this embodiment, the macroscopic cross-section database construction device is presented in the form of functional units. Here, a unit refers to an ASIC (Application Specific Integrated Circuit) circuit, a processor and memory that execute one or more software or fixed programs, and / or other devices that can provide the above functions.

[0135] This embodiment also provides a dosage determination device for implementing the above embodiments and preferred embodiments; details already described will not be repeated. As used below, the term "module" can refer to a combination of software and / or hardware that performs a predetermined function. Although the device described in the following embodiments is preferably implemented in software, hardware implementation, or a combination of software and hardware, is also possible and contemplated.

[0136] This embodiment provides a dosage determination device, such as... Figure 4 As shown, it includes:

[0137] The database acquisition module 401 is used to acquire a multi-group macroscopic section database, wherein each data in the multi-group macroscopic section database includes tissue type, boron concentration value, neutron energy group and neutron macroscopic section value.

[0138] The target parameter acquisition module 402 is used to acquire the target parameters of the target object; the target parameters include the target tissue type and the target boron concentration value corresponding to the target tissue.

[0139] The cross-section data acquisition module 403 is used to acquire the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameters based on the multi-group macroscopic cross-section database.

[0140] The dose simulation module 404 is used to obtain the dose for the target object based on the target neutron macroscopic cross-sectional value.

[0141] In some optional embodiments, the dose determination device further includes:

[0142] The boron concentration setting module is used to set multiple boron concentration values ​​for the tissue types corresponding to the human body model.

[0143] The neutron source parameter selection module is used to select the corresponding neutron source parameters for the tissue type and the corresponding multiple boron concentration values, wherein the neutron source parameters include energy range values;

[0144] The energy group partitioning module is used to divide the energy range value into multiple discrete neutron energy groups;

[0145] The particle transport module is used to perform particle transport simulation based on the selected neutron source parameters. The simulation process jumps between different neutron energy groups to obtain the neutron reaction rate and neutron flux of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups.

[0146] The database construction module is used to obtain the neutron macroscopic cross-sectional values ​​of the tissue type under multiple boron concentration values ​​and multiple neutron energy groups based on the neutron reaction rate and the neutron flux, and to construct a multi-group macroscopic cross-sectional database. Each data in the multi-group macroscopic cross-sectional database consists of tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-sectional value.

[0147] In some optional implementations, the database construction module includes:

[0148] The first cross-sectional value acquisition unit is used to acquire the neutron macroscopic cross-sectional value of each voxel grid based on the neutron reaction rate and neutron flux distribution of each voxel grid corresponding to the tissue type.

[0149] The second cross-sectional value acquisition unit is used to acquire the neutron macroscopic cross-sectional value corresponding to the tissue type based on the neutron macroscopic cross-sectional value of each voxel grid.

[0150] In some optional embodiments, the cross-sectional data acquisition module 403 includes:

[0151] The matching and filtering unit is used to obtain, from the multi-group macroscopic cross-section database, the first neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the first boron concentration value, and the second neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the second boron concentration value; the first boron concentration value and the second boron concentration value are two boron concentration values ​​adjacent to the target boron concentration value in the multi-group macroscopic cross-section database;

[0152] The target macroscopic cross-section value determination unit is used to determine the target neutron macroscopic cross-section value based on the first boron concentration value, the second boron concentration value, the target boron concentration value, the first neutron macroscopic cross-section value, and the second neutron macroscopic cross-section value.

[0153] In some optional embodiments, the dose simulation module 404 is specifically used to obtain the neutron flux corresponding to each neutron energy group, the energy released when a neutron-related reaction occurs, and the density of the target tissue; and to obtain the dose based on the neutron flux corresponding to each neutron energy group, the energy released when a neutron-related reaction occurs, the density of the target tissue, and the target neutron macroscopic cross-sectional value.

[0154] Further functional descriptions of the above modules and units are the same as those in the corresponding embodiments described above, and will not be repeated here.

[0155] In this embodiment, the dose determination device is presented in the form of a functional unit. Here, a unit refers to an ASIC (Application Specific Integrated Circuit) circuit, a processor and memory that execute one or more software or fixed programs, and / or other devices that can provide the above functions.

[0156] This invention also provides a computer device having the above-described features. Figure 3 The macroscopic cross-section database construction device shown or the above Figure 4 The dose determination device shown.

[0157] Please see Figure 5 , Figure 5 This is a schematic diagram of the structure of a computer device provided in an optional embodiment of the present invention, such as... Figure 5 As shown, the computer device includes one or more processors 10, memory 20, and interfaces for connecting the components, including high-speed interfaces and low-speed interfaces. The components communicate with each other via different buses and can be mounted on a common motherboard or otherwise installed as needed. The processors can process instructions executed within the computer device, including instructions stored in or on memory to display graphical information of a GUI on external input / output devices (such as display devices coupled to the interfaces). In some alternative implementations, multiple processors and / or multiple buses can be used with multiple memories and multiple memory modules, if desired. Similarly, multiple computer devices can be connected, each providing some of the necessary operations (e.g., as a server array, a group of blade servers, or a multiprocessor system). Figure 5 Take a processor 10 as an example.

[0158] Processor 10 may be a central processing unit, a network processor, or a combination thereof. Processor 10 may further include a hardware chip. The hardware chip may be an application-specific integrated circuit (ASIC), a programmable logic device (PLD), or a combination thereof. The programmable logic device may be a complex programmable logic device (CAMP), a field-programmable gate array (FPGA), a general-purpose array logic (GPA), or any combination thereof.

[0159] The memory 20 stores instructions executable by at least one processor 10 to cause the at least one processor 10 to perform the method shown in the above embodiments.

[0160] The memory 20 may include a program storage area and a data storage area. The program storage area may store the operating system and applications required for at least one function; the data storage area may store data created based on the use of the computer device. Furthermore, the memory 20 may include high-speed random access memory and may also include non-transitory memory, such as at least one disk storage device, flash memory device, or other non-transitory solid-state storage device. In some alternative embodiments, the memory 20 may optionally include memory remotely located relative to the processor 10, and these remote memories may be connected to the computer device via a network. Examples of such networks include, but are not limited to, the Internet, intranets, local area networks, mobile communication networks, and combinations thereof.

[0161] The memory 20 may include volatile memory, such as random access memory; the memory may also include non-volatile memory, such as flash memory, hard disk or solid-state drive; the memory 20 may also include a combination of the above types of memory.

[0162] The computer device also includes an input device 30 and an output device 40. The processor 10, memory 20, input device 30, and output device 40 can be connected via a bus or other means. Figure 5 Taking the example of a connection between China and Israel via a bus.

[0163] Input device 30 can receive input numerical or character information, and generate key signal inputs related to user settings and function control of the computer device, such as a touchscreen, keypad, mouse, trackpad, touchpad, joystick, one or more mouse buttons, trackball, joystick, etc. Output device 40 may include display devices, auxiliary lighting devices (e.g., LEDs), and haptic feedback devices (e.g., vibration motors). The aforementioned display devices include, but are not limited to, liquid crystal displays, light-emitting diodes, displays, and plasma displays. In some alternative embodiments, the display device may be a touchscreen.

[0164] The computer device also includes a communication interface for communicating with other devices or communication networks.

[0165] This invention also provides a computer-readable storage medium. The methods described above according to embodiments of the invention can be implemented in hardware or firmware, or implemented as computer code that can be recorded on a storage medium, or implemented as computer code downloaded via a network and originally stored on a remote storage medium or a non-transitory machine-readable storage medium and then stored on a local storage medium. Thus, the methods described herein can be processed by software stored on a storage medium using a general-purpose computer, a dedicated processor, or programmable or dedicated hardware. The storage medium can be a magnetic disk, optical disk, read-only memory, random access memory, flash memory, hard disk, or solid-state drive, etc.; further, the storage medium can also include combinations of the above types of memory. It is understood that computers, processors, microprocessor controllers, or programmable hardware include storage components capable of storing or receiving software or computer code, which, when accessed and executed by the computer, processor, or hardware, implements the methods shown in the above embodiments.

[0166] A portion of this invention can be applied as a computer program product (specifically, a MultiGroup Monte Carlo dosing calculation program (MultiGroup-MC)), such as computer program instructions, which, when executed by a computer, can invoke or provide the methods and / or technical solutions according to the invention through the operation of the computer. Those skilled in the art will understand that the computer program instructions exist in computer-readable media in forms including, but not limited to, source files, executable files, installation package files, etc. Correspondingly, the ways in which the computer executes the computer program instructions include, but are not limited to: the computer directly executing the instructions, or the computer compiling the instructions and then executing the corresponding compiled program, or the computer reading and executing the instructions, or the computer reading and installing the instructions and then executing the corresponding installed program. Here, the computer-readable medium can be any available computer-readable storage medium or communication medium accessible to a computer.

[0167] Although embodiments of the invention have been described in conjunction with the accompanying drawings, those skilled in the art can make various modifications and variations without departing from the spirit and scope of the invention, and such modifications and variations all fall within the scope defined by the appended claims.

Claims

1. A macroscopic cross-section database construction method characterized by comprising: The method comprises: setting a plurality of boron concentration values for a tissue type corresponding to a human body model; selecting corresponding neutron source parameters for the tissue type and the plurality of corresponding boron concentration values, the neutron source parameters including energy range values; dividing the energy range values into a plurality of discrete neutron energy groups; performing particle transport simulation based on the selected neutron source parameters, with the simulation process jumping between different neutron energy groups, to obtain neutron reaction rates and neutron fluxes of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups; based on the neutron reaction rates and the neutron fluxes, obtaining neutron macroscopic cross-section values of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups, and constructing a multi-group macroscopic cross-section database, wherein the composition of each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group, and neutron macroscopic cross-section value.

2. The macroscopic cross-section database construction method according to claim 1, characterized by, The method comprises: setting a plurality of boron concentration values for a tissue type corresponding to a human body model; selecting corresponding neutron source parameters for the tissue type and the plurality of corresponding boron concentration values, the neutron source parameters including energy range values; 3. The macroscopic cross-section database construction method according to claim 1 or 2, characterized by, dividing the energy range values into a plurality of discrete neutron energy groups; performing particle transport simulation based on the selected neutron source parameters, with the simulation process jumping between different neutron energy groups, to obtain neutron reaction rates and neutron fluxes of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups; 4. A method of dose determination, characterized in that based on the neutron reaction rates and the neutron fluxes, obtaining neutron macroscopic cross-section values of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups, and constructing a multi-group macroscopic cross-section database, wherein the composition of each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group, and neutron macroscopic cross-section value. The method comprises: setting a plurality of boron concentration values for a tissue type corresponding to a human body model; selecting corresponding neutron source parameters for the tissue type and the plurality of corresponding boron concentration values, the neutron source parameters including energy range values; dividing the energy range values into a plurality of discrete neutron energy groups; performing particle transport simulation based on the selected neutron source parameters, with the simulation process jumping between different neutron energy groups, to obtain neutron reaction rates and neutron fluxes of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups; based on the neutron reaction rates and the neutron fluxes, obtaining neutron macroscopic cross-section values of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups, and constructing a multi-group macroscopic cross-section database, wherein the composition of each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group, and neutron macroscopic cross-section value. The method comprises: setting a plurality of boron concentration values for a tissue type corresponding to a human body model; selecting corresponding neutron source parameters for the tissue type and the plurality of corresponding boron concentration values, the neutron source parameters including energy range values; dividing the energy range values into a plurality of discrete neutron energy groups; 5. The dose determination method of claim 4, characterized in that, performing particle transport simulation based on the selected neutron source parameters, with the simulation process jumping between different neutron energy groups, to obtain neutron reaction rates and neutron fluxes of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups; based on the neutron reaction rates and the neutron fluxes, obtaining neutron macroscopic cross-section values of the tissue type under a plurality of boron concentration values and a plurality of neutron energy groups, and constructing a multi-group macroscopic cross-section database, wherein the composition of each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group, and neutron macroscopic cross-section value. The neutron macroscopic cross-section value of each voxel grid is obtained based on the neutron reaction rate and the neutron flux distribution of each voxel grid corresponding to the tissue type of each neutron energy group; The neutron macroscopic cross-section value corresponding to the tissue type is obtained based on the neutron macroscopic cross-section value of each voxel grid.

6. The dose determination method of claim 4, wherein, The target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameter is obtained based on the multi-group macroscopic cross-section database, including: From the multi-group macroscopic cross-section database, the first neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the first boron concentration value, and the second neutron macroscopic cross-section value of each neutron energy group under the target tissue type and the second boron concentration value are obtained; the first boron concentration value and the second boron concentration value are two boron concentration values adjacent to the target boron concentration value in the multi-group macroscopic cross-section database; The target neutron macroscopic cross-section value is determined based on the first boron concentration value, the second boron concentration value, the target boron concentration value, the first neutron macroscopic cross-section value and the second neutron macroscopic cross-section value.

7. The dose determination method of claim 4, wherein, The dose for the target object is obtained based on the target neutron macroscopic cross-section value, including: The neutron flux corresponding to each neutron energy group, the energy released when the neutron-related reaction occurs, and the density of the target tissue are obtained; The dose is obtained based on the neutron flux corresponding to each neutron energy group, the energy released when the neutron-related reaction occurs, and the density of the target tissue and the target neutron macroscopic cross-section value.

8. A macro cross section database construction apparatus characterized by comprising: It includes: The boron concentration setting module is used to set multiple boron concentration values for the tissue type corresponding to the human body model; The neutron source parameter selection module is used to select corresponding neutron source parameters for the tissue type and the corresponding multiple boron concentration values, respectively, and the neutron source parameters include energy range values; The energy group division module is used to divide the energy range values into multiple discrete neutron energy groups; The particle transport module is used to perform particle transport simulation based on the selected neutron source parameters, and the simulation process jumps between different neutron energy groups to obtain the neutron reaction rate and the neutron flux of multiple neutron energy groups under multiple boron concentration values of the tissue type; The database construction module is used to obtain the neutron macroscopic cross-section value of multiple neutron energy groups under multiple boron concentration values of the tissue type based on the neutron reaction rate and the neutron flux, and construct a multi-group macroscopic cross-section database, and the composition of each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-section value.

9. A dose determination device, characterized in that The device includes: The database acquisition module is used to obtain a multi-group macroscopic cross-section database, and the composition of each data in the multi-group macroscopic cross-section database includes tissue type, boron concentration value, neutron energy group and neutron macroscopic cross-section value; The target parameter acquisition module is used to obtain the target parameter of the target object; the target parameter includes the target tissue type corresponding to the target tissue and the target boron concentration value; The cross-section data acquisition module is used to obtain the target neutron macroscopic cross-section value corresponding to each neutron energy group under the target parameter based on the multi-group macroscopic cross-section database; a dose simulation module configured to obtain a dose for the target object based on the target macroscopic cross section value of neutrons; the dose determination apparatus further comprises: a boron concentration setting module configured to set a plurality of boron concentration values for a tissue type corresponding to a human model; a neutron source parameter selection module configured to select a corresponding neutron source parameter for the tissue type and the corresponding plurality of boron concentration values, the neutron source parameter comprising an energy range value; a group division module configured to divide the energy range value into a plurality of discrete neutron energy groups; a particle transport module configured to perform particle transport simulation based on the selected neutron source parameter, the simulation process jumping between different neutron energy groups, to obtain a neutron reaction rate and a neutron flux of the tissue type at a plurality of boron concentration values and a plurality of neutron energy groups; a database construction module configured to obtain a macroscopic cross section value of neutrons of the tissue type at a plurality of boron concentration values and a plurality of neutron energy groups based on the neutron reaction rate and the neutron flux, and to construct the multi-group macroscopic cross section database.

10. A computer device, comprising: comprising: a memory and a processor, which are in communication connection with each other, and the memory stores computer instructions, and the processor executes the computer instructions to perform the macroscopic cross section database construction method of any one of claims 1 to 3 or the dose determination method of any one of claims 4 to 7.

11. A computer readable storage medium, characterized in that, the computer readable storage medium stores computer instructions, and the computer instructions are used to make the computer execute the macroscopic cross section database construction method of any one of claims 1 to 3 or the dose determination method of any one of claims 4 to 7.

12. A computer program product, characterised in that, comprising computer instructions, and the computer instructions are used to make the computer execute the macroscopic cross section database construction method of any one of claims 1 to 3 or the dose determination method of any one of claims 4 to 7.

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