Method for generating boron neutron capture therapy plan, computer equipment and storage medium

Abstract

The invention relates to the technical field of radiotherapy, and discloses a method for generating a boron neutron capture therapy plan, computer equipment and a storage medium. Comprising the following steps: based on current state data of neutrons, determining first section data, second section data and a current transport distance of the neutrons at a current position, and judging whether the neutrons collide or not; wherein the first cross section data is used for describing the macroscopic cross section of nuclides except boron, and the second cross section data is used for describing macroscopic cross section neutrons obtained through calculation according to the boron concentration of the current grid and the microcosmic cross section of the nuclide boron; and under the condition that neutrons do not collide and the neutrons are transmitted from the current grid to the next grid through the surface, if the two grids are made of the same material, updating the second cross section data based on the boron concentration of the next grid to calculate the dose, and finally generating a boron neutron capture therapy plan by using a dose distribution calculation result. Therefore, only the cross section data related to the boron concentration is updated in the penetrating surface, and the calculation efficiency is optimized.

Description

Technical Field

[0001] The present application relates to the technical field of radiotherapy, and in particular to a method, computer device, and storage medium for generating a boron neutron capture therapy plan. Background Art

[0002] Boron neutron capture therapy (BNCT) is widely studied and planned for use in tumor treatment due to its ability to target cancer cells. The principle is to irradiate the tumor with a neutron beam. The neutrons react with boron (B-10) accumulated within the tumor cells, generating secondary neutrons with a high linear energy density. The energy deposited by these neutrons can kill cancer cells in a targeted manner.

[0003] Before BNCT treatment, it is very important for the treatment planning system to calculate the dose. However, there are still limitations in the calculation of dose distribution in related technologies. A method is needed to improve the accuracy and efficiency of dose distribution calculation. Summary of the Invention

[0004] This application aims to solve, at least to some extent, one of the technical problems in the related art. To this end, this application proposes a method, computer device, and storage medium for generating a boron neutron capture therapy plan. The main technical solutions adopted in this application include: In a first aspect, an embodiment of the present application provides a method for generating a boron neutron capture therapy plan, which is used in a boron neutron capture therapy process, the method comprising: determining first cross-sectional data, second cross-sectional data and current transport distance of the neutron at the current position based on current state data of the neutron; wherein the first cross-sectional data is used to describe the macroscopic cross-sectional data of nuclides other than boron, and the second cross-sectional data is used to describe the macroscopic cross-sectional data calculated using the boron concentration of the current grid at the current position and the microscopic cross-sectional data of the nuclide boron; using the first cross-sectional data, the second cross-sectional data and the current transport distance to determine whether the neutron collides; in the case where the neutron does not collide and the neutron is transported through the surface from the current grid to the next grid, if the materials of the current grid and the next grid are the same, updating the second cross-sectional data based on the boron concentration of the next grid, performing dose distribution calculation based on the updated second cross-sectional data, and generating a boron neutron capture therapy plan using the dose distribution calculation result.

[0005] Optionally, when the boron concentration of the next grid is not equal to the boron concentration of the current grid, the second cross-sectional data is updated based on the boron concentration of the next grid.

[0006] Optionally, the first cross-sectional data, the second cross-sectional data and the current transport distance are used to determine whether the neutron has collided, including: accumulating the current equivalent transport distance of the neutron using the first cross-sectional data, the second cross-sectional data and the current transport distance to obtain an updated equivalent transport distance; and determining whether the neutron has collided by comparing the updated equivalent transport distance with a collision reference value; wherein the collision reference value is a dimensionless equivalent distance used to determine whether the neutron has collided.

[0007] Optionally, the collision reference value is determined by: extracting a current random number; determining a probability that the neutrons do not collide based on the random number; and taking a logarithm of the probability to obtain the collision reference value.

[0008] Optionally, the current equivalent transport distance of the neutron is accumulated using the first cross-section data, the second cross-section data and the current transport distance to obtain an updated equivalent transport distance, including: summing the first cross-section data and the second cross-section data to obtain the current total cross-section of the neutron at the current position; multiplying the current total cross-section and the current transport distance to obtain an equivalent transport distance increment; and adding the equivalent transport distance increment to the current equivalent transport distance to obtain an updated equivalent transport distance.

[0009] Optionally, the first cross-sectional data is determined by: determining the nuclide information corresponding to the current grid based on a basic material matrix that does not include boron concentration information; and determining the first cross-sectional data based on the nuclide information and neutron energy information in the current state data.

[0010] Optionally, the second cross-sectional data is determined by: determining the boron concentration of the current grid using boron concentration distribution information under the voxel grid; and determining the second cross-sectional data based on the boron concentration and neutron energy information of the current grid.

[0011] Optionally, the method further includes: in the event of a neutron collision, determining the post-collision situation; if the post-collision situation indicates that the neutron is alive, updating the current state data of the neutron, and repeating the above-mentioned dose distribution calculation steps; if the post-collision situation indicates that the neutron is dead and releases secondary neutrons, storing the secondary neutrons in a secondary neutron library; if the post-collision situation indicates that the neutron is dead and no secondary neutrons are released, determining whether there are residual neutrons in the secondary neutron library, and if so, performing the above-mentioned dose distribution calculation steps for the remaining neutrons.

[0012] In a second aspect, an embodiment of the present application provides a device for generating a boron neutron capture therapy plan, the device comprising: an information acquisition module for determining first cross-sectional data, second cross-sectional data and current transport distance of the neutron at a current position based on current state data of the neutron; wherein the first cross-sectional data is used to describe the macroscopic cross-sectional data of nuclides other than boron, and the second cross-sectional data is used to describe the macroscopic cross-sectional data obtained by calculating the boron concentration of the current grid at the current position and the microscopic cross-sectional data of the nuclide boron; a collision judgment module for judging whether a neutron collision occurs using the first cross-sectional data, the second cross-sectional data and the current transport distance; a dose calculation module for updating the second cross-sectional data based on the boron concentration of the next grid if the materials of the current grid and the next grid are the same, when no neutron collision occurs and the neutron is transported through the surface from the current grid to the next grid, so as to perform dose distribution calculation based on the updated second cross-sectional data.

[0013] In a third aspect, the present application further provides a computer device comprising a memory and a processor, wherein the memory stores a computer program, and the processor implements the steps of any of the above methods when executing the computer program.

[0014] In a fourth aspect, the present application also provides a computer-readable storage medium having a computer program stored thereon, which implements the steps of any of the above methods when the computer program is executed by a processor.

[0015] In a fifth aspect, the present invention provides a computer program product, comprising a computer program, which implements the steps of any of the above methods when executed by a processor.

[0016] In the above embodiment, by distinguishing between first cross-sectional data that does not contain boron and second cross-sectional data that does contain boron, repeated calculations of fixed nuclide cross-sections are avoided, significantly reducing computational effort and data storage requirements. Furthermore, when traversing adjacent grids made of the same material, only the second cross-sectional data related to the boron concentration is updated, thereby reducing cross-sectional update operations and significantly improving computational efficiency. This overcomes the limitations of Monte Carlo calculations in BNCT caused by uneven boron concentration distribution, achieving a dose distribution calculation method that balances accuracy and speed, and improving the reliability and efficiency of BNCT plan generation. BRIEF DESCRIPTION OF THE DRAWINGS

[0017] In order to more clearly illustrate the specific implementation methods of the present application or the technical solutions in the prior art, the following is a brief introduction to the drawings required for use in the specific implementation methods or the description of the prior art. Obviously, the drawings described below are some implementation methods of the present application. For ordinary technicians in this field, other drawings can be obtained based on these drawings without any creative work.

[0018] Figure 1aA flowchart of a method for generating a boron neutron capture therapy plan according to one embodiment of the present application; Figure 1b A schematic diagram of a neutron transport process provided according to one embodiment of the present application; Figure 2 This is a flowchart of a collision determination method provided according to one embodiment of the present application; Figure 3 This is a structural block diagram of an apparatus for generating a boron neutron capture therapy plan according to one embodiment of the present application; Figure 4 The figure is a diagram of the internal structure of a computer device according to one embodiment of the present application. DETAILED DESCRIPTION

[0019] To make the purpose, technical solutions, and advantages of the embodiments of the present application more clear, the technical solutions in the embodiments of the present application will be clearly and completely described below in conjunction with the drawings in the embodiments of the present application. Obviously, the described embodiments are part of the embodiments of the present application, not all of the embodiments. Based on the embodiments in the present application, all other embodiments obtained by those skilled in the art without creative work are within the scope of protection of this application.

[0020] Boron neutron capture therapy (BNCT) has been widely studied and planned for use in the treatment of tumors such as gliomas and melanomas due to its targeted cancer cell killing properties. The fundamental principle is to irradiate the tumor site with a beam of thermal or epithermal neutrons. Neutrons undergo capture reactions with boron (B-10) enriched within tumor cells, generating secondary neutrons with a high linear energy density. The energy deposited by these neutrons can target and kill cancer cells. Before BNCT treatment, dose calculation within the treatment planning system is crucial. Technological advances have shifted dose calculations from the early, cruder methods to more refined ones, enabling precise treatment planning and achieving more effective treatment outcomes. Specifically, Monte Carlo methods have been used, shifting from multi-cluster to fine-grained continuous energy calculations, compared to approximate discrete methods. Furthermore, the geometric model used in the calculations has been refined, with voxel grid resolution reaching up to pixel-level human phantoms. Furthermore, consideration of the impact of detailed boron concentration on dose calculations is crucial.

[0021] In related methods, when estimating tissue concentration from blood boron concentration, it is assumed that boron is uniformly distributed in the tissue, thereby ignoring the actual spatial concentration differences and resulting in inaccurate dose calculations. At the same time, this method of replacing the boron concentration in different regions with a single value not only fails to take into account the real-time metabolic changes and microscopic distribution differences of boron in tissues, but also fails to reflect the subtle effects of concentration gradients on dose distribution. In addition, this homogenization treatment method may overestimate the dose in low-concentration areas (such as tumor edge areas). On the one hand, it may mask the risk of insufficient treatment and make the treatment plan conservative. On the other hand, it leads to incomplete treatment of low-boron areas due to overestimation of the dose, increasing the possibility of recurrence.

[0022] With the advancement of positron emission tomography-computed tomography (PET-CT) technology, it has become possible to obtain boron concentration distributions based on a pixel-level grid. However, related methods process boron concentration distributions based on a voxel grid by adding detected boron to body tissue materials according to a gradient, thereby increasing the number of materials to achieve boron-containing neutron transport. If there are m types of human tissue materials and n concentration gradients, m times n materials are required to fully represent the varying boron concentrations in different tissues. While this method can account for the effect of boron concentration on neutron transport to some extent, determining the appropriate concentration gradient is extremely difficult. If the intervals defined for the concentration gradients are too large, the influence of boron concentration cannot be properly accounted for during the transport process. If the intervals defined are too small, more material types are required, significantly increasing the number of material changes during the transport process. This increases the number of macroscopic cross-sections and the number of collision distances sampled, affecting computational efficiency.

[0023] In summary, conventional general-purpose Monte Carlo methods often rely on boron concentration gradients for multiple material settings. This gradient boron concentration data, on the one hand, introduces an approximation of the boron concentration distribution and fails to accurately reflect the fine-scale boron concentration within the tissue. Furthermore, it significantly reduces computational efficiency in models with non-uniform boron concentrations.

[0024] Based on this, according to an embodiment of the present application, an embodiment of a dose distribution calculation method is provided. It should be noted that the steps shown in the flowchart of the accompanying drawings can be executed in a computer system such as a set of computer executable instructions, and although a logical order is shown in the flowchart, in some cases, the steps shown or described can be executed in an order different from that shown here.

[0025] In this embodiment, a method for generating a boron neutron capture therapy plan is provided. Figure 1a FIG. 1 is a flow chart of a method for generating a boron neutron capture therapy plan according to an embodiment of the present application. Figure 1a As shown, the process includes the following steps: S110 , based on the current state data of the neutron, determining the first cross-sectional data, the second cross-sectional data and the current transport distance of the neutron at the current position.

[0026] Here, the neutron may refer to a thermal neutron or epithermal neutron beam during boron neutron capture therapy; the current state data of the neutron may refer to a particle in a Monte Carlo simulation, i.e., a set of real-time physical properties of the neutron at a specific moment. Specifically, the current state data of the neutron may include the neutron's current position information, current energy information, and current direction of motion. Exemplarily, the current position information of the neutron may be the neutron's precise position in a voxel-level geometric model, including the specific grid within a three-dimensional voxel grid where the neutron resides, i.e., the neutron's current grid information and the neutron's precise spatial coordinates within the current grid; the current energy information of the neutron may refer to the neutron's current kinetic energy; and the current direction of motion of the neutron may be vector data containing a direction, used to reflect the trend of the neutron's motion. Exemplarily, the neutron may refer to a neutron from a radioactive source, and its current state data may be obtained by sampling the neutrons from the radioactive source. Sampling based on predefined neutron source distribution parameters can yield the neutron's position, energy, and direction of motion.

[0027] The current transport distance refers to the distance that the neutron travels within the voxel grid, that is, the track length of the neutron within the voxel grid; the current transport distance can also be determined based on the current state data of the neutron.

[0028] Specifically, the current transport distance can include the distance the neutron travels from the current position to the current grid boundary, which is used for collision determination and final dose calculation during subsequent simulated transport processes. For example, using the neutron's current precise x-, y-, and z-axis coordinates and motion direction vector data, the distances from the neutron's current position to the x-, y-, and z-planes that intersect with the next voxel grid can be calculated. These three distances can serve as the distance the neutron travels from the current position to the current grid boundary. Furthermore, the current transport distance can also include the distance the neutron travels across the current grid to the next grid boundary. For example, also using the neutron's current coordinates and motion vector, the distance the neutron travels from the x-, y-, or z-plane of the grid across an entire voxel grid to the x-, y-, or z-plane of the next voxel grid can be calculated. This distance across two voxel grids can serve as the distance the neutron travels across the current grid to the next grid boundary. This distance can be used to directly simulate crossing two grids based on the distance when the adjacent grid materials are the same, to avoid redundant calculations.

[0029] For example, the two-dimensional coordinate level is used for explanation. Please refer to Figure 1b In the figure, the neutron is at point A=(x a ,y a ) position, and its movement direction is Vector. The distance a neutron travels from point A to point B is the distance from its current position to the current grid boundary; the distance from point B to point C is the distance the neutron travels across the entire voxel grid to the next grid boundary. Note that this explanation is based on a two-dimensional scale, but the same concept applies to three-dimensional scales.

[0030] Furthermore, based on the current position information of the neutron, first cross-sectional data and second cross-sectional data of the neutron at the current position can be obtained.

[0031] The first cross-sectional data may be used to describe the macroscopic cross-sectional data of nuclides other than boron. For example, the first cross-sectional data may refer to the probability of radioactive neutrons interacting in a base material (excluding boron), i.e., an indicator for measuring the likelihood of collision.

[0032] Optionally, the first cross-sectional data can be determined in the following manner: first, the nuclide information corresponding to the current grid can be determined based on the basic material matrix that does not contain boron concentration information, and then the first cross-sectional data can be determined based on the nuclide information and the neutron energy information in the current state data.

[0033] The base material matrix can be an integer matrix with the same spatial dimensions as the voxel grid, with each element in the matrix storing material information corresponding to a particular voxel grid. For example, the material information can refer to the types and concentrations of nuclides within a particular voxel grid. It is understood that the nuclide concentration can be represented by nuclide density.

[0034] It should be noted that before constructing the basic material matrix, it is first necessary to predefine a voxel-level basic database, where voxel is the abbreviation of volume pixel, which is the smallest unit for discretizing objects in three-dimensional space, and voxel level refers to the degree of detail or accuracy of voxelizing objects in three-dimensional space so that each part of the object is represented by a voxel category. Specifically, the basic database contains basic material information (nuclide type and density) of different human tissues. That is, the basic material information includes the category of human tissue to which the voxel grid belongs, and also includes the category of different nuclides in the voxel grid and their corresponding content, so the basic database can also be called a material database. Specifically, using the material database, not only can we accurately determine whether a voxel grid is muscle, bone, fat or other tissues, but we can also obtain the content of different nuclides in the grid, such as carbon, nitrogen, oxygen or hydrogen.

[0035] Specifically, the basic material matrix can be constructed based on the material database. Using the grayscale values ​​of medical images (e.g., CT scans), different tissues or organs in the medical images are divided into different regions at the same voxel level as the material database. These regions are then matched one-to-one with the material database, thereby establishing a material number matrix based on the voxel level. This material number matrix can then serve as the basic material matrix. It should be noted that this basic material matrix does not include boron concentration information.

[0036] Furthermore, after obtaining the basic material matrix that does not contain boron concentration information, it can be used to determine the nuclide information corresponding to the current grid.

[0037] It is understood that in transport simulations, the current position of a neutron can be determined by its three-dimensional spatial coordinates. By mapping these coordinates to the spatial relationships between the grid cells in the voxel grid, the current grid unit where the neutron is located can be precisely located, that is, a specific voxel grid unit can be determined. A matching search can then be performed in the base material matrix. For example, using the grid number (e.g., row and column) of the current grid where the neutron is located, a matching search can be performed in the base material matrix, directly reading the nuclide information corresponding to the current grid.

[0038] Then, the neutron energy information in the current state data can be used in combination with the nuclide information to retrieve the macroscopic cross-section of the target nuclide (such as carbon, hydrogen, oxygen, and nitrogen, etc.) in the cross-section database established by the nuclide. Among them, the cross-section database can be a table describing the correspondence between neutron energy values ​​and cross-section values, and the corresponding macroscopic cross-section can be directly found using the neutron energy. If there is no direct matching data for a certain energy value, it can also be generated by interpolation of adjacent energy values. For example, the neutron energy information is used to match and search in the cross-section database of the nuclide carbon to determine the cross-section value of the carbon. Similarly, the neutron energy information is used to match and search in the cross-section database of the nuclide oxygen to determine the cross-section value of the oxygen. Subsequently, the density of all nuclides except boron is used to perform a nuclide density weighted calculation on all the cross-section values ​​obtained to finally obtain the first cross-section data.

[0039] Optionally, when calculating the first cross-sectional data, the same method can be used to search the conversion coefficient database to obtain the flux-dose conversion coefficient for subsequent dose calculations. Thus, using the base material matrix that does not contain boron concentration information, macroscopic cross-sectional data for nuclides other than boron can be obtained, providing reliable data input for subsequent precise dose calculations.

[0040] Furthermore, the second cross-sectional data can be used to describe the macroscopic cross-sectional data calculated using the boron concentration of the current grid at the current location and the microscopic cross-sectional data of the nuclide boron. Specifically, the second cross-sectional data describes the contribution of boron to the macroscopic cross-sectional data. It should be noted that the second cross-sectional data is not a microscopic cross-sectional data, but rather a component of the macroscopic cross-sectional data after dynamic adjustment of the boron concentration, directly reflecting the impact of the local boron concentration on the neutron capture reaction.

[0041] Optionally, the second cross-sectional data may be determined by first determining the boron concentration of the current grid using boron concentration distribution information under the voxel grid, and then determining the second cross-sectional data based on the boron concentration and neutron energy information of the current grid.

[0042] Specifically, pixel values ​​or radioactivity values ​​obtained from positron emission tomography (PET) images are analyzed and converted into boron concentration values ​​to establish a voxel-level boron concentration distribution matrix. The converted boron concentration values ​​are then precisely matched to the spatial locations of the voxel grid, ensuring that each voxel grid has a unique boron concentration value, thereby obtaining a boron concentration distribution within the voxel-by-voxel grid.

[0043] Furthermore, the neutron energy information can be used to retrieve the corresponding microscopic cross-sectional data in the cross-sectional database of the nuclide boron alone. Subsequently, the boron concentration of the current grid is multiplied by the microscopic cross-sectional data, and the microscopic cross-sectional data of boron is multiplied by the boron concentration to obtain the second cross-sectional data that describes the contribution of boron to the macroscopic cross-sectional data. Optionally, when calculating the microscopic cross-sectional data, the microscopic flux dose conversion coefficient of neutrons to nuclide boron can also be found in the conversion coefficient database for subsequent dose calculations. At this point, by integrating the boron concentration distribution with the microscopic cross-sectional data, dynamic and flexible cross-sectional calculations are achieved. The decoupling design from the first cross-sectional data allows the boron distribution to be dynamically and independently updated, significantly improving the calculation efficiency.

[0044] S120 , using the first cross-sectional data, the second cross-sectional data, and the current transport distance to determine whether a neutron collision occurs.

[0045] It should be noted that the collision event here refers to the collision between a neutron and all nuclides in the medium, including but not limited to boron. In other words, the collision can refer to the collision between a neutron and a target nucleus, not just boron, but any nucleus in the medium.

[0046] Specifically, a cumulative collision probability threshold can be predefined. Since the first cross-sectional data represents the probability of a neutron colliding with all nuclides except boron per unit length, and the second cross-sectional data represents the probability of a neutron colliding with the nuclide boron per unit length, and the current transport distance represents the actual physical path length of the neutron's flight since the last state update, the first and second cross-sectional data can be used to first calculate the collision probability of the neutron with all nuclides (including boron) per unit length, and then combined with the physical path length of the neutron's flight through the voxel grid in the simulation to determine a collision. Specifically, when the calculated distance the neutron actually flew exceeds the cumulative collision probability threshold accumulated for that path segment, a collision can be determined to have occurred.

[0047] S130. In the case where no neutron collision occurs and the neutron is transported through the surface from the current grid to the next grid, if the materials of the current grid and the next grid are the same, update the second cross-sectional data based on the boron concentration of the next grid, perform dose distribution calculation based on the updated second cross-sectional data, and generate a boron neutron capture therapy plan using the dose distribution calculation result.

[0048] Among them, the next grid can refer to the next pixel grid that the neutron is about to enter based on the current direction of movement and position during the neutron transport process. It can be adjacent to the edge of the current grid or adjacent to the corner of the current grid. Specifically, when the neutron passes through the current grid boundary along the motion trajectory, it can be determined which coordinate plane it intersects with based on its motion direction and motion trend, thereby determining which surrounding grid unit the neutron may enter. It should be noted that passing through the surface can be understood as the neutron crossing the boundaries of two adjacent grids, or transporting from the current grid to the next grid diagonally opposite. For example, please continue to refer to Figure 1b Based on the direction and trend of the neutron's movement, it can be determined that the neutron will pass through the dividing line of X=X1 from position A, and thus it can be determined that the neutron will enter the grid unit to the right of point B in the figure.

[0049] Furthermore, if the neutron does not collide, after determining the next grid to which the neutron will be transported, the boron concentration of the next grid can be determined using the grid number. That is, the boron concentration of the next grid can be obtained from the boron concentration distribution matrix. This boron concentration can then be used to update the second cross-sectional data describing the boron contribution to the macroscopic cross-sectional data.

[0050] It's important to note that if the current grid and the next grid share the same material—that is, if the current and next grids have the same nuclide type and content, except for boron—then the first cross-sectional data for the current grid can be reused directly, eliminating the need to recalculate the first cross-sectional data for the next grid. In other words, the second cross-sectional data can be updated based solely on the boron concentration of the next grid, and then superimposed with the first cross-sectional data to obtain total cross-sectional data reflecting the total collision probability for the next grid, which can then be combined with the updated current transport data for further collision determination. Ultimately, this reuse and local update method is repeated multiple times to perform global dose calculations.

[0051] Optionally, the boron concentration of the next grid may be equal to or different from the boron concentration of the current grid. In the case that the boron concentration of the next grid is not equal to the boron concentration of the current grid, the second cross-sectional data can be updated based on the boron concentration of the next grid. That is, if the boron concentration of the next grid is not equal to the boron concentration of the current grid, the above method can be used to directly reuse the first cross-sectional data when the material of the current grid and the next grid is the same, and the second cross-sectional data can be updated using the boron concentration distribution information of the next grid to perform dose distribution calculation based on the updated second cross-sectional data. Through this method, when passing through the surface to the grid with the same material, the repeated calculation of the cross-sectional data is reduced, the calculation efficiency is improved, and at the same time, only the boron concentration effect is dynamically superimposed, and there is no need to generate new material types for different boron concentrations, which significantly reduces memory usage. In addition, this continuous processing of voxel-level boron concentration avoids gradient discretization errors to the greatest extent, ensuring the accuracy of dose calculation.

[0052] It is understandable that after the basic material matrix is ​​established, a statistical counting grid can be further created that is completely consistent with the voxel grid size and spatial position of the basic material matrix. The statistical counting grid is used to record the dose accumulation information of a specific area of ​​interest, and its initial value can be set to zero. Subsequently, during the Monte Carlo simulation process, when neutrons are transported through the statistical counting grid, the corresponding counting results can be dynamically accumulated based on the actual transport length and energy deposition information of the neutrons in the grid. Finally, after the simulation is completed, the cumulative counting results in the statistical counting grid are normalized to output the global dose distribution. For example, the normalized statistical counting results can be multiplied by the flux dose conversion coefficient to obtain the final dose distribution result.

[0053] Furthermore, after obtaining the dose distribution results, the neutron beam parameters or irradiation time in the irradiation plan can be selectively adjusted according to the prescription dose requirements and dose limitation requirements of different areas (such as tumors or normal tissues), and a boron neutron capture treatment plan can be accurately generated to ensure that the tumor area can obtain sufficient treatment dose while protecting normal tissue from excessive irradiation.

[0054] It should be noted that the content of generating a boron neutron capture therapy plan based on the dose distribution result can be understood by those skilled in the art and will not be elaborated here.

[0055] Optionally, the method further comprises, in the event of a neutron collision, determining a post-collision condition.

[0056] Determining the post-collision state can be used to characterize the change in the physical state of the neutron after it interacts with the nuclides in the medium. That is, based on the different collision types during the neutron collision, the subsequent state of the neutron can be determined. For example, the post-collision state may include the following three possible scenarios: if the neutron undergoes elastic scattering, the neutron will continue to be transported with a new emission direction and energy; if the neutron is absorbed after the collision (such as a boron-10 capture reaction), it means that the neutron has died. At this time, the neutron may disappear directly or release secondary neutrons.

[0057] If the post-collision results indicate that the neutron survived, the current state data is updated, and the cross-sectional data update calculation steps are repeated. Specifically, if scattering occurs and the neutron is deemed to be alive, its energy and direction data are updated, and the transport process is re-entered. The updated state data is used to calculate the boron concentration of the next grid, update the cross-sectional data, and continue the iteration until the next collision or trans-surface transport occurs.

[0058] If the post-collision conditions indicate that the neutron died and released a secondary neutron, the secondary neutron is stored in the secondary neutron bank. Specifically, if the post-collision conditions indicate that the neutron was absorbed and a secondary neutron was produced, the secondary neutron and its state data, including its position, energy, and direction, can be stored in the secondary neutron bank to facilitate subsequent simulation of its transport process and ensure that the dose contribution of the secondary neutron is accounted for.

[0059] If the post-collision situation indicates that the neutron is dead and no secondary neutrons are released, the secondary neutron reservoir is determined to contain residual neutrons. If so, the dose distribution calculation steps described above can be performed for the residual neutrons. Specifically, if the post-collision situation indicates that the neutron was absorbed and disappeared without generating any secondary neutrons, the secondary neutron reservoir can be checked to see if there are any previously stored and unprocessed residual neutrons. If so, the residual neutrons in the secondary neutron reservoir are removed and the transport simulation continues. If not, the current neutron transport simulation process is terminated, and the cumulative count results within the counting grid are normalized to output the dose distribution data for that area.

[0060] Optionally, if there are no remaining neutrons in the secondary neutron reservoir, it is possible to further check whether the number of source neutrons in the voxel grid has reached the preset number for the transport simulation. If not, it is possible to return to the step of sampling the neutron source distribution parameters and re-acquire a new round of current state data of the neutrons to repeat the iterative simulation process.

[0061] At this point, by dynamically processing the different states after neutron collision, accurate simulation and efficient calculation of the neutron transport process have been achieved. By taking different measures for different collision situations, while ensuring calculation efficiency, detailed consideration of the voxel-level boron concentration distribution has been achieved, thus providing a more accurate dose assessment scheme for boron neutron capture therapy.

[0062] In the above implementation, by distinguishing between first cross-sectional data free of boron and second cross-sectional data with boron contribution, repeated calculations of fixed nuclide cross-sections are avoided, significantly reducing computational effort and data storage requirements. Furthermore, when traversing adjacent grids of the same material, only the second cross-sectional data related to boron concentration is updated, thereby reducing cross-sectional update operations and significantly improving computational efficiency. This overcomes the limitations of Monte Carlo calculations in BNCT caused by uneven boron concentration distribution, achieving a dose distribution calculation method that balances accuracy and speed, and improving the reliability and efficiency of BNCT plan generation.

[0063] In some embodiments, please refer to the attached Figure 2 , using the first cross-section data, the second cross-section data and the current transport distance to determine whether the neutrons collide, including: S210 , accumulating the current equivalent transport distance of neutrons using the first cross-sectional data, the second cross-sectional data, and the current transport distance to obtain an updated equivalent transport distance.

[0064] Among them, the current equivalent transport distance can be used to quantify the total collision probability experienced by neutrons on the transport path. It is not a physical transport distance value, but a dimensionless collision probability value.

[0065] Specifically, in the first round of simulation, the current equivalent transport distance of neutrons can be initialized to zero.

[0066] It can be understood that the updated equivalent transport distance can be used to represent the accumulated collision probability of the neutron when it arrives at a new position after the transport state changes, and is also a dimensionless collision probability value.

[0067] Specifically, the total collision probability of a neutron with all nuclides, including boron, per unit length can be calculated using the first and second cross-sectional data. This is then combined with the physical path length of the neutron in the simulation, i.e., the current transport distance, to obtain the cumulative collision probability increment during the neutron's transport flight. It should be noted that this collision probability increment and the current equivalent transport distance have the same dimension, so it can be used to accumulate the current equivalent transport distance to obtain the updated equivalent transport distance.

[0068] Optionally, accumulating the current equivalent transport distance of neutrons using the first cross-sectional data, the second cross-sectional data, and the current transport distance to obtain an updated equivalent transport distance may include: First, the first and second cross-section data are summed to obtain the current total cross-section of the neutron at the current position. Then, the current total cross-section is multiplied by the current transport distance to obtain the equivalent transport distance increment. Finally, the equivalent transport distance increment is added to the current equivalent transport distance to obtain the updated equivalent transport distance.

[0069] Among them, the current total cross section can refer to the probability density of a neutron colliding with all nuclides including boron at the current position, which represents the total probability of a neutron colliding with any atomic nuclide when transported in the current grid. Specifically, after determining the current grid where the neutron is located and the boron concentration distribution information of the current grid, the first cross-sectional data unrelated to boron and the second cross-sectional data related to boron can be obtained respectively. The two are then summed and superimposed to obtain the current total cross section describing the total collision probability of the neutron with all nuclides in the current grid. Exemplarily, the current total cross section is equal to the first cross-sectional data (macroscopic cross section excluding boron) plus the second cross-sectional data (microscopic cross-sectional data of boron multiplied by the boron concentration).

[0070] Furthermore, by multiplying the current total cross section by the current transport distance, we can obtain the equivalent transport distance increment. This equivalent transport distance increment is the additional collision probability contribution generated by neutron transport within the current grid. It can be understood as the additional cumulative collision probability increment when neutrons fly within the current grid. Like the current equivalent transport distance, it is dimensionless data and can be numerically superimposed.

[0071] It's important to reiterate that the current transport distance refers to the physical distance a neutron actually travels within the current voxel grid. Its units are length dimensions, such as mm, and can be directly derived from geometric calculations. The current equivalent transport distance, equivalent transport distance increment, and updated equivalent transport distance are dimensionless values ​​obtained by multiplying the physical distance by the cross-sectional data. Essentially, they all quantify the contribution to the collision probability. By multiplying the first and second cross-sectional data at the neutron's current position by the actual current transport distance, the physical distance can be converted into a dimensionless probability value. This requires only multiplication and addition operations, avoiding the repeated logarithmic operations in traditional Monte Carlo methods and significantly improving the speed of the neutron transport simulation process.

[0072] S220 , determining whether the neutrons collide by comparing the updated equivalent transport distance with the collision reference value.

[0073] The collision reference value can be a dimensionless equivalent distance used to determine whether a neutron collision occurs, that is, a dimensionless probability threshold for triggering a collision event. It should be emphasized that the collision reference value and the updated equivalent transport distance are essentially contribution values ​​used to quantify the probability of collision. Therefore, the collision reference value and the updated equivalent transport distance can be compared in the same dimension to determine whether a neutron collision occurs. For example, if the collision reference value is less than the updated equivalent transport distance, it means that the neutron will collide during the current transport process; if the collision reference value is greater than the updated equivalent transport distance, it means that the neutron will not collide during the current transport process.

[0074] Optionally, the collision reference value may be determined by first extracting a current random number, then determining the probability of no neutron collision based on the random number, and finally taking the logarithm of the probability to obtain the collision reference value.

[0075] Specifically, at the beginning of the neutron transport simulation, a random number can be drawn. For example, this random number can be any positive integer or any random number from a uniform distribution between [0, 1]. This random number can then be normalized to map it to the standard [0, 1] interval, serving as the probability that the neutron will not collide. As will be appreciated, because the free path of neutrons in the medium follows an exponential distribution, the calculated uniform random number must be converted to an exponentially distributed value. Specifically, a negative logarithmic transformation is applied to the probability of a neutron not collide, thereby obtaining a collision reference value that represents the actual attenuation distance.

[0076] It's important to reiterate that, in voxel grids with non-uniform boron concentrations, the boron concentration may vary across each voxel. Therefore, in order to accurately calculate the dose distribution, the related art requires resampling a random geometric distance based on the total cross-section (including the macroscopic cross-section of the base material and the cross-section related to boron concentration) for each neutron transport. This involves a large number of logarithmic operations and cross-sectional data calculations, resulting in low computational efficiency. However, this method converts the dimensional geometric distance sampling into dimensionless probability accumulation. This allows for a single collision reference value to be sampled and calculated at the beginning of the simulation, which serves as the global collision threshold for the neutron. This avoids repeated sampling and redundant logarithmic calculations, significantly improving computational efficiency.

[0077] In the above embodiment, by dynamically accumulating equivalent transport distances and comparing them with the initially generated dimensionless collision reference values, efficient optimization of neutron collision judgment under non-uniform boron concentration is achieved, and the repeated geometric distance sampling and logarithmic operations in related technologies are simplified to simple dimensionless value comparisons. The physical distance sampling is replaced by the accumulation of dimensionless probability values. On the one hand, the computational complexity is significantly reduced, and on the other hand, the influence of voxel-level boron concentration on the collision probability can be accurately reflected, thereby greatly improving the computational efficiency while ensuring the accuracy of dose calculation.

[0078] The embodiment of this specification further provides a method for generating a boron neutron capture therapy plan, the method comprising the following steps: S302, establishing a basic material matrix and a statistical counting grid based on a material database.

[0079] S304. Sampling source neutrons according to neutron source distribution parameters to obtain current state data of the neutrons.

[0080] S306 , extracting a current random number, determining the probability of neutrons not colliding based on the random number, and taking the logarithm of the probability to obtain a collision reference value.

[0081] S308: Determine the current transport distance of the neutron at the current position based on the current state data of the neutron. The current transport distance includes the distance the neutron needs to travel from the current position to the current grid boundary and the distance the neutron needs to travel across the current grid to the next grid boundary.

[0082] S310 , determining nuclide information corresponding to the current grid based on a basic material matrix that does not include boron concentration information.

[0083] S312. Determine first cross-sectional data based on the nuclide information and the neutron energy information in the current state data, wherein the first cross-sectional data is used to describe the macroscopic cross-sectional view of nuclides other than boron.

[0084] S314 , using the boron concentration distribution information under the voxel grid to determine the boron concentration of the current grid.

[0085] S316. Determine second cross-sectional data based on the boron concentration and neutron energy information of the current grid, wherein the second cross-sectional data is used to describe the macroscopic cross-sectional data obtained by calculating the boron concentration of the current grid at the current position and the microscopic cross-sectional data of the nuclide boron.

[0086] S318. Sum the first cross-section data and the second cross-section data to obtain the current total cross-section of the neutron at the current position.

[0087] S320. Multiply the current total cross-section by the current transport distance to obtain an equivalent transport distance increment.

[0088] S322. Add the equivalent transport distance increment to the current equivalent transport distance to obtain an updated equivalent transport distance.

[0089] S324. Determine whether the neutrons collide by comparing the updated equivalent transport distance with a collision reference value, wherein the collision reference value is a dimensionless equivalent distance used to determine whether the neutrons collide.

[0090] S326. When no neutron collision occurs and the neutron is transported through the surface from the current grid to the next grid, if the material of the current grid and the next grid is the same, and the boron concentration of the next grid is not equal to the boron concentration of the current grid, then the second cross-sectional data is updated based on the boron concentration of the next grid, so as to update the updated total cross-sectional area and updated transport distance of the neutron at the new position based on the updated second cross-sectional data, and jump to S320 to continue the updated calculation of the cross-sectional data.

[0091] S328. In the event of a neutron collision, determine the post-collision situation.

[0092] S330: If the post-collision situation indicates that the neutron survives, the current state data of the neutron is updated, and the process returns to S306 and repeats the above-mentioned cross-sectional data updating calculation steps.

[0093] S332. If the post-collision situation indicates that the neutron is dead and releases secondary neutrons, the secondary neutrons are stored in a secondary neutron bank.

[0094] S334. If the post-collision situation indicates that the neutron is dead and no secondary neutrons are released, determine whether there are residual neutrons in the secondary neutron reservoir. If so, return to S306 and execute the above-mentioned dose distribution calculation step for the residual neutrons.

[0095] S336: If there are no remaining neutrons in the secondary neutron reservoir, return to S304 to resample and execute the above-mentioned dose distribution calculation steps.

[0096] S338. After the simulation is completed, the dose distribution calculation is performed, and the dose distribution calculation results are used to generate a boron neutron capture therapy plan.

[0097] It should be understood that, although the various steps in the above flowchart are shown in sequence as indicated by the arrows, these steps are not necessarily performed in the order indicated by the arrows. Unless otherwise specified herein, there is no strict order restriction on the execution of these steps, and these steps can be performed in other orders. Moreover, at least a portion of the steps in the above flowchart may include multiple steps or multiple stages, and these steps or stages are not necessarily performed at the same time, but can be performed at different times. The execution order of these steps or stages is not necessarily to be performed in sequence, but can be performed in turn or alternately with other steps or at least a portion of steps or stages in other steps.

[0098] The embodiment of this specification also provides a device 300 for generating a boron neutron capture therapy plan, such as Figure 3 As shown, the apparatus 300 for generating a boron neutron capture therapy plan includes: an information acquisition module 310, a collision judgment module 320, and a therapy plan generation module 330, wherein: The information acquisition module 310 is used to determine the first cross-sectional data, the second cross-sectional data and the current transport distance of the neutron at the current position based on the current state data of the neutron; wherein the first cross-sectional data is used to describe the macroscopic cross-sectional data of the nuclides other than boron, and the second cross-sectional data is used to describe the macroscopic cross-sectional data calculated using the boron concentration of the current grid at the current position and the microscopic cross-sectional data of the nuclide boron.

[0099] The collision judgment module 320 is used to judge whether a neutron collision occurs by using the first cross-section data, the second cross-section data and the current transport distance.

[0100] The treatment plan generation module 330 is used to update the second cross-sectional data based on the boron concentration of the next grid if the materials of the current grid and the next grid are the same, when neutrons do not collide and neutrons are transported through the surface from the current grid to the next grid, perform dose distribution calculation based on the updated second cross-sectional data, and generate a boron neutron capture treatment plan using the dose distribution calculation result.

[0101] In some embodiments, the apparatus 300 for generating a boron neutron capture therapy plan further includes a dose calculation module for updating the second cross-sectional data based on the boron concentration of the next grid if the boron concentration of the next grid is not equal to the boron concentration of the current grid.

[0102] In some embodiments, the collision judgment module 320 is further used to accumulate the current equivalent transport distance of the neutron using the first cross-sectional data, the second cross-sectional data and the current transport distance to obtain an updated equivalent transport distance; by comparing the updated equivalent transport distance and the collision reference value, it is determined whether the neutron has collided; wherein the collision reference value is a dimensionless equivalent distance used to determine whether the neutron has collided.

[0103] In some embodiments, the collision judgment module 320 is further configured to determine a collision reference value by: extracting a current random number; determining a probability that the neutrons do not collide based on the random number; and taking a logarithm of the probability to obtain a collision reference value.

[0104] In some embodiments, the device 300 for generating a boron neutron capture therapy plan further includes a transport distance judgment module, which is used to accumulate the current equivalent transport distance of the neutron using the first cross-sectional data, the second cross-sectional data, and the current transport distance to obtain an updated equivalent transport distance, including: summing the first cross-sectional data and the second cross-sectional data to obtain the current total cross-sectional area of ​​the neutron at the current position; multiplying the current total cross-sectional area by the current transport distance to obtain an equivalent transport distance increment; and accumulating the equivalent transport distance increment on the current equivalent transport distance to obtain an updated equivalent transport distance.

[0105] In some embodiments, the information acquisition module 310 is further used to determine the first cross-sectional data in the following manner: determining the nuclide information corresponding to the current grid based on the basic material matrix that does not contain boron concentration information; and determining the first cross-sectional data based on the nuclide information and the neutron energy information in the current state data.

[0106] In some embodiments, the information acquisition module 310 is further used to determine the second cross-sectional data by: determining the boron concentration of the current grid using the boron concentration distribution information under the voxel grid; and determining the second cross-sectional data based on the boron concentration and neutron energy information of the current grid.

[0107] In some embodiments, the apparatus 300 for generating a boron neutron capture therapy plan is further configured to determine a post-collision situation in the event of a neutron collision; if the post-collision situation indicates that the neutron is alive, update the current state data of the neutron, and repeat the above steps; if the post-collision situation indicates that the neutron is dead and a secondary neutron is released, store the secondary neutron in a secondary neutron reservoir; if the post-collision situation indicates that the neutron is dead and no secondary neutron is released, determine whether there are any remaining neutrons in the secondary neutron reservoir, and if so, perform the above-mentioned dose distribution calculation steps for the remaining neutrons.

[0108] The specific definition of a device for generating a boron neutron capture therapy plan can be found in the definition of a method for generating a boron neutron capture therapy plan described above and will not be repeated here. The various modules in the above-mentioned device for generating a boron neutron capture therapy plan can be implemented in whole or in part through software, hardware, or a combination thereof. The above-mentioned modules can be embedded in or independent of a processor in a computer device in the form of hardware, or can be stored in the memory of the computer device in the form of software, so that the processor can call and execute the operations corresponding to the above-mentioned modules.

[0109] In this embodiment, a device for generating a boron neutron capture therapy plan is presented in the form of a functional unit, where the unit refers to an ASIC (Application Specific Integrated Circuit) circuit, a processor and memory that executes one or more software or fixed programs, and / or other devices that can provide the above functions.

[0110] The embodiment of the present application further provides a computer device, which may be a terminal, and its internal structure diagram may be as shown in FIG. Figure 4 As shown. The computer device includes a processor, a memory, a communication interface, a display screen and an input device connected via a system bus. The processor of the computer device is used to provide computing and control capabilities. The memory of the computer device includes a non-volatile storage medium and an internal memory. The non-volatile storage medium stores an operating system and a computer program. The internal memory provides an environment for the operation of the operating system and the computer program in the non-volatile storage medium. The communication interface of the computer device is used to communicate with an external terminal in a wired or wireless manner, and the wireless manner can be achieved through WIFI, a mobile cellular network, NFC (near field communication) or other technologies. When the computer program is executed by the processor, a method for generating a boron neutron capture therapy plan is implemented. The display screen of the computer device can be a liquid crystal display screen or an electronic ink display screen, and the input device of the computer device can be a touch layer covering the display screen, or a button, trackball or touchpad provided on the computer device housing, or an external keyboard, touchpad or mouse.

[0111] Those skilled in the art will understand that Figure 4 The structure shown in the figure is only a block diagram of a part of the structure related to the solution of the present application, and does not constitute a limitation on the computer device to which the solution of the present application is applied. The specific computer device may include more or fewer components than shown in the figure, or combine certain components, or have a different component arrangement.

[0112] The embodiments of the present application also provide a computer-readable storage medium. The above-mentioned method according to the embodiment of the present application can be implemented in hardware, firmware, or implemented as a computer code that can be recorded in a storage medium, or implemented as a computer code that is originally stored in a remote storage medium or a non-temporary machine-readable storage medium and downloaded through a network and will be stored in a local storage medium, so that the method described herein can be stored in such software processing on a storage medium using a general-purpose computer, a dedicated processor, or programmable or dedicated hardware. Among them, the storage medium can be a magnetic disk, an optical disk, a read-only storage memory, a random access memory, a flash memory, a hard disk or a solid-state drive, etc.; further, the storage medium can also include a combination of the above-mentioned types of memory. It can be understood that a computer, a processor, a microprocessor controller or programmable hardware includes a storage component that can store or receive software or computer code. When the software or computer code is accessed and executed by a computer, a processor or hardware, the method shown in the above embodiment is implemented.

[0113] An embodiment of the present application provides a computer program product, which includes computer instructions stored in a computer-readable storage medium. A processor of a computer device reads the computer instructions from the computer-readable storage medium and executes the computer instructions, causing the computer device to perform a method according to any embodiment of the present application.

[0114] The dose distribution calculation method, computer device, and storage medium described in the above embodiments can be implemented by a computer chip or entity, or by a product with certain functions. A typical implementation device is a computer. Specifically, the computer can be, for example, a personal computer, laptop computer, cellular phone, camera phone, smartphone, personal digital assistant, media player, navigation device, email device, game console, tablet computer, wearable device, or any combination of these devices.

[0115] For the convenience of description, the above devices are described as being divided into various units according to their functions. Of course, when implementing this application, the functions of each unit can be implemented in the same or multiple software and / or hardware.

[0116] Those skilled in the art will appreciate that the embodiments of the present application can be provided as methods, systems, or computer program products. Therefore, the present application can adopt the form of a complete hardware embodiment, a complete software embodiment, or an embodiment in combination with software and hardware. Moreover, the present application can adopt the form of a computer program product implemented on one or more computer-usable storage media (including but not limited to magnetic disk storage, CD-ROM, optical storage, etc.) that contain computer-usable program code.

[0117] The present application is described with reference to the flow chart and / or block diagram of the method, device (system), and computer program product according to the embodiment of the present application. It should be understood that each flow process and / or box in the flow chart and / or block diagram and the combination of the flow process and / or box in the flow chart and / or block diagram can be realized by computer program instructions. These computer program instructions can be provided to the processor of a general-purpose computer, a special-purpose computer, an embedded processing machine or other programmable data processing device to produce a machine, so that the instructions executed by the processor of the computer or other programmable data processing device produce a device for realizing the function specified in one flow chart flow or multiple flows and / or one box or multiple boxes of the block diagram.

[0118] These computer program instructions may also be stored in a computer-readable memory that can direct a computer or other programmable data processing device to operate in a specific manner, so that the instructions stored in the computer-readable memory produce a product including an instruction device that implements the functions specified in one or more processes in the flowchart and / or one or more boxes in the block diagram.

[0119] These computer program instructions can also be loaded onto a computer or other programmable data processing device so that a series of operating steps are executed on the computer or other programmable device to produce a computer-implemented process, so that the instructions executed on the computer or other programmable device provide steps for implementing the functions specified in one or more processes in the flowchart and / or one or more boxes in the block diagram.

[0120] Throughout this specification, reference to terms such as "one embodiment," "some embodiments," "examples," "specific examples," or "some examples" means that a specific feature, structure, material, or characteristic described in conjunction with that embodiment or example is included in at least one embodiment or example of the present application. In this specification, schematic representations of the above terms do not necessarily refer to the same embodiment or example. Furthermore, the specific features, structures, materials, or characteristics described may be combined in any suitable manner in any one or more embodiments or examples.

[0121] Furthermore, the terms "first" and "second" are used for descriptive purposes only and should not be construed as indicating or implying relative importance or implicitly specifying the number of the technical features being referred to. Thus, a feature defined as "first" or "second" may explicitly or implicitly include at least one of such features. Throughout the description of this application, "plurality" means at least two, for example, two, three, etc., unless otherwise specifically defined.

[0122] It should also be noted that the terms "comprises," "includes," or any other variations thereof are intended to encompass non-exclusive inclusion, such that a process, method, commodity, or apparatus that includes a series of elements includes not only those elements but also other elements not explicitly listed, or includes elements inherent to such process, method, commodity, or apparatus. In the absence of further limitations, an element defined by the phrase "comprises a ..." does not exclude the presence of other identical elements in the process, method, commodity, or apparatus that includes the element.

[0123] Each embodiment in this specification is described in a progressive manner. Similar parts between the embodiments can be referred to in detail. Each embodiment focuses on the differences from other embodiments. Since it is basically similar to the method embodiment, the description is relatively simple. For relevant parts, refer to the partial description of the method embodiment.

[0124] The foregoing is merely an embodiment of the present application and is not intended to limit the present application. For those skilled in the art, the present application may have various changes and variations. Any modifications, equivalent replacements, improvements, etc. made within the spirit and principles of the present application should all be included within the scope of the claims of the present application.

[0125] Although the embodiments of the present application have been described with reference to the accompanying drawings, those skilled in the art may make various modifications and variations without departing from the spirit and scope of the present application, and such modifications and variations shall fall within the scope defined by the appended claims.

Claims

1. A method for generating a boron neutron capture therapy plan, characterized in that: The method comprises: Determining first cross-sectional data, second cross-sectional data, and current transport distance of the neutron at a current position based on current state data of the neutron; wherein the first cross-sectional data is used to describe a macroscopic cross-sectional view of nuclides other than boron, and the second cross-sectional data is used to describe a macroscopic cross-sectional view calculated using a boron concentration of a current grid at the current position and a microscopic cross-sectional view of the nuclide boron; The first cross-sectional data, the second cross-sectional data, and the current transport distance are used to determine whether the neutrons collide. When the neutrons do not collide and are transported from the current grid through the surface to the next grid, if the current grid and the next grid are made of the same material, the second cross-sectional data are updated based on the boron concentration of the next grid, a dose distribution calculation is performed based on the updated second cross-sectional data, and a boron neutron capture therapy plan is generated using the dose distribution calculation result.

2. The method according to claim 1, characterized in that When the boron concentration of the next grid is not equal to the boron concentration of the current grid, the second cross-sectional data is updated based on the boron concentration of the next grid.

3. The method according to claim 1 or 2, characterized in that The determining whether the neutrons collide using the first cross-sectional data, the second cross-sectional data, and the current transport distance includes: Accumulating the current equivalent transport distance of the neutrons using the first cross-sectional data, the second cross-sectional data, and the current transport distance to obtain an updated equivalent transport distance; By comparing the updated equivalent transport distance and a collision reference value, it is determined whether the neutrons collide. The collision reference value is a dimensionless equivalent distance used to determine whether the neutrons collide.

4. The method according to claim 3, characterized in that The collision reference value is determined by: Extract the current random number; determining a probability that the neutrons do not collide based on the random number; Taking the logarithm of the probability, the collision reference value is obtained.

5. The method according to claim 3, characterized in that The accumulating the current equivalent transport distance of the neutrons by using the first cross-sectional data, the second cross-sectional data, and the current transport distance to obtain an updated equivalent transport distance includes: Summing the first cross-sectional data and the second cross-sectional data to obtain a current total cross-sectional area of ​​the neutron at the current position; Multiplying the current total cross-section by the current transport distance to obtain an equivalent transport distance increment; The equivalent transport distance increment is accumulated on the current equivalent transport distance to obtain the updated equivalent transport distance.

6. The method according to any one of claims 1 or 2, characterized in that The first cross-sectional data is determined by: Determining the nuclide information corresponding to the current grid based on a basic material matrix that does not include boron concentration information; The first cross-sectional data is determined based on the nuclide information and neutron energy information in the current state data.

7. The method according to any one of claims 1 or 2, characterized in that The second cross-sectional data is determined by: Determining the boron concentration of the current grid using boron concentration distribution information under the voxel grid; The second cross-sectional data is determined based on the boron concentration of the current grid and the neutron energy information.

8. The method according to any one of claims 1 or 2, characterized in that The method further comprises: in the event of a collision of said neutrons, determining a post-collision condition; If the post-collision condition indicates that the neutron survives, updating the current state data of the neutron and repeating the above-mentioned dose distribution calculation step; If the post-collision condition indicates that the neutron is dead and releases secondary neutrons, storing the secondary neutrons in a secondary neutron bank; If the post-collision situation indicates that the neutron is dead and no secondary neutrons are released, it is determined whether there are residual neutrons in the secondary neutron reservoir. If so, the above-mentioned dose distribution calculation step is performed for the residual neutrons.

9. A computer device, characterized in that: include: A memory and a processor, wherein the memory and the processor are communicatively connected to each other, the memory stores computer instructions, and the processor executes the method according to any one of claims 1 to 8 by executing the computer instructions.

10. A computer-readable storage medium, characterized in that The computer-readable storage medium stores computer instructions, and the computer instructions are used to enable a computer to execute the method according to any one of claims 1 to 8.

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