A fast calculation method for dose in boron neutron capture therapy

Abstract

The present invention discloses a method for rapidly calculating the dose in boron neutron capture therapy. First, a nuclear database suitable for the human body and based on a deterministic method is prepared. Then, a corresponding model is established according to human body model parameters and treatment plan design parameters. Finally, a highly efficient deterministic program is used to perform neutron-photon transport calculations on the established model to obtain the dose distribution results, thereby realizing rapid calculation of the dose in boron neutron capture therapy.

Description

Technical Field

[0001] The present invention relates to the field of medical and engineering integration technology, and in particular to a method for rapid dose calculation in boron neutron capture therapy (BNCT). Background Art

[0002] Boron neutron capture therapy (BNCT) is a rapidly developing new cancer treatment approach internationally. It is a dual therapy that combines boron-targeted drugs with neutron irradiation, effectively killing cancer cells selectively at the cellular level. It represents a new generation of cutting-edge precision cancer treatment technology.

[0003] In the actual application of boron neutron capture therapy, corresponding treatment plans need to be formulated for different cases, so the treatment plans need to be evaluated. This requires particle transport calculations to obtain the irradiation dose of tumors and other tissues under the treatment plan.

[0004] Particle transport calculations in current well-known treatment planning software systems are all based on the Monte Carlo method, which has the problems of long calculation time and large statistical variance for deep penetration problems. Summary of the Invention

[0005] In order to overcome the problems existing in the above-mentioned prior art, the purpose of the present invention is to provide a method for rapid dose calculation in boron neutron capture therapy. First, a nuclear database suitable for the human body and based on a deterministic method is prepared. Then, a corresponding model is established according to the human body model parameters and the treatment plan design parameters. Finally, a high-efficiency deterministic program is used to perform neutron-photon transport calculations on the established model to obtain the dose distribution results, thereby realizing rapid dose calculation in boron neutron capture therapy.

[0006] In order to achieve the above objectives, the present invention adopts the following technical solutions to be implemented:

[0007] A method for rapidly calculating the dose in boron neutron capture therapy comprises the following steps:

[0008] Step 1: Create a nuclear database suitable for the human body based on deterministic methods;

[0009] To produce deterministic cluster cross-section data applicable to the human body, it is first necessary to create a basic fine cluster cross-section database. Then, based on a human body model, resonance self-screening calculations are performed to obtain an effective self-screening cross-section suitable for boron neutron capture therapy. The basic fine cluster cross-section database is combined with the basic fine cluster cross-section database to obtain a fine cluster cross-section database suitable for boron neutron capture therapy. Finally, the database is merged to obtain a broad cluster cross-section database suitable for boron neutron capture therapy.

[0010] The following is an introduction to the creation of a basic fine group database, the calculation of resonance self-screening based on a human body model, and the merging and creation of a wide group database:

[0011] Radiative transport requires consideration of both neutron and photon transport, so the basic fine-group cross-section database must include fine-group neutron and photon cross-section data. The fine-group neutron-photon coupling database uses the MATXS format. Fine-group neutron cross sections and neutron-photon scattering matrix data can be obtained from the evaluation nuclear data through resonance reconstruction and linearization, Doppler broadening, effective self-screening cross sections in the unresolved resonance region, thermal neutron scattering processing, and energy group merging. For the calculation of fine-group photoatom cross sections, since photoatom reactions do not require consideration of temperature effects and resonant self-screening effects, only the cross-section data given in the evaluation nuclear database need to be resonantly reconstructed and linearized, and energy group merged, resulting in the basic fine-group database, or MATXS format.

[0012] After obtaining the basic fine group cross section database, it is necessary to perform resonant self-screening calculations based on the human body model to obtain the effective self-screening cross section suitable for boron neutron capture therapy. The macroscopic total cross section and macroscopic scattering cross section of the effective self-screening are calculated by the coupled iterative method of formula (1):

[0013]

[0014] Where:

[0015] Σ t,i (E)——the total macroscopic cross section of neutrons with energy E in region i;

[0016] φ i (E)——the neutron flux rate of energy E in region i;

[0017] φ j (E′) — the neutron flux rate of energy E′ at region j;

[0018] V i ——the volume of region i;

[0019] P j→i (E) — the probability of a neutron with energy E colliding from region j to region i;

[0020] Σ s,j (E′) — macroscopic scattering cross section of neutrons with energy E′ at region j;

[0021] f(E′→E)dE′—the probability that a neutron with energy E′ collides with an atom and the neutron’s energy is within dE′ of E′.

[0022] For the non-resonance energy range, the calculation of the effective self-screening cross section requires interpolating the resonance integral table of the background cross section to obtain the effective self-screening cross section of the nuclide;

[0023] The human body model in boron neutron capture therapy is a non-uniform model. For the non-uniform model, the background cross section can be calculated using formula (2):

[0024]

[0025] Where:

[0026] ——background cross section of nuclide r in region I;

[0027] N I,r —nucleon density of nuclide r in region I;

[0028] ——the total microscopic cross section of nuclide j in region I;

[0029] — macroscopic cross section of region I;

[0030] P I→J ——Collision probability from area I to area J;

[0031] After performing resonance self-screening calculations based on a human body model, a fine group cross section database suitable for boron neutron capture therapy can be obtained. To consider the impact of complex problems on computational efficiency, it is necessary to merge the fine group database into energy groups to obtain a wide group cross section database with a smaller number of energy groups to improve computational efficiency. The human body model is selected for calculation using a refined transport program to obtain typical weight spectra of important organs or parts of the human body. Based on the conservation of reaction rate, the fine group cross section database after resonance self-screening is merged into energy groups to obtain a wide group cross section database.

[0032] Step 2: Construct a boron neutron capture therapy model based on the human body model parameter data measured by the instrument and the treatment technology design parameters (dose parameters, collimator parameters, irradiation field, irradiation position, material parameters, etc.);

[0033] Step 3: Based on the wide group cross-section database obtained in step 1, a highly efficient deterministic program is used to perform neutron-photon transport calculations on the established boron neutron capture therapy model to obtain the dose distribution results.

[0034] Compared with the prior art, the present invention has the following advantages:

[0035] In the method of the present invention, an efficient deterministic method is proposed as the core of neutron-photon transport calculation in boron neutron capture therapy, which provides an effective method for rapid dose calculation in boron neutron capture therapy, thereby realizing rapid calculation of boron neutron capture therapy dose without statistical variance, saving a lot of computing time and computing resources. BRIEF DESCRIPTION OF THE DRAWINGS

[0036] Figure 1 Create a flow chart for the neutron-photon coupled MATXS database;

[0037] Figure 2 Flowchart of the rapid dose calculation method for boron neutron capture therapy. DETAILED DESCRIPTION

[0038] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments:

[0039] The present invention is a method for rapidly calculating the dose in boron neutron capture therapy. Figure 2 As shown, the present invention first creates a nuclear database suitable for the human body and based on a deterministic method, then builds a corresponding boron neutron capture therapy model according to the human body model parameters and the treatment plan design parameters, and finally uses a highly efficient deterministic program to perform neutron-photon transport calculations on the established boron neutron capture therapy model to obtain the dose distribution results, thereby realizing rapid calculation of the dose in boron neutron capture therapy.

[0040] The method specifically comprises the following steps:

[0041] Step 1: Create a nuclear database suitable for the human body based on deterministic methods;

[0042] To produce deterministic cluster cross-section data applicable to the human body, it is first necessary to create a basic fine cluster cross-section database. Then, based on a human body model, resonance self-screening calculations are performed to obtain an effective self-screening cross-section suitable for boron neutron capture therapy. The basic fine cluster cross-section database is combined with the basic fine cluster cross-section database to obtain a fine cluster cross-section database suitable for boron neutron capture therapy. Finally, the database is merged to obtain a broad cluster cross-section database suitable for boron neutron capture therapy.

[0043] The following is an introduction to creating a basic fine group database, conducting resonance self-screening calculations based on typical problems, and merging to create a wide group database:

[0044] Radiation transport needs to consider both neutron and photon transport, so the basic fine group cross section database needs to include fine group neutron and photon cross section data. The production process of the fine group neutron-photon coupling database is as follows: Figure 1The database uses the MATXS format. Fine-group neutron cross sections and neutron-photon scattering matrix data can be obtained from the evaluation nuclear data through resonance reconstruction and linearization, Doppler broadening, effective self-screening cross sections in the unresolved resonance region, thermal neutron scattering processing, and energy group merging. For the calculation of fine-group photoatom cross sections, since photoatom reactions do not require consideration of temperature effects and resonant self-screening effects, only resonance reconstruction and linearization, as well as energy group merging, are required on the cross section data given in the evaluation nuclear database to obtain the basic fine-group cross section database.

[0045] After obtaining the basic fine group cross section database, it is necessary to perform resonant self-screening calculations based on the human body model to obtain the effective self-screening cross section suitable for boron neutron capture therapy. The macroscopic total cross section and macroscopic scattering cross section of the effective self-screening are calculated by the coupled iterative method of formula (1):

[0046]

[0047] Where:

[0048] Σ t,i (E)——the total macroscopic cross section of neutrons with energy E in region i;

[0049] φ i (E)——the neutron flux rate of energy E in region i;

[0050] φ j (E′) — the neutron flux rate of energy E′ at region j;

[0051] V i ——the volume of region i;

[0052] P j→i (E) — the probability of a neutron with energy E colliding from region j to region i;

[0053] Σ s,j (E′) — macroscopic scattering cross section of neutrons with energy E′ at region j;

[0054] f(E′→E)dE′—the probability that a neutron with energy E′ collides with an atom and the neutron’s energy is within dE′ of E′.

[0055] For the non-resonance energy range, the calculation of the effective self-screening cross section requires interpolating the resonance integral table of the background cross section to obtain the effective self-screening cross section of the nuclide;

[0056] The human body model in boron neutron capture therapy is a non-uniform model. For the non-uniform model, the background cross section can be calculated using formula (2):

[0057]

[0058] Where:

[0059] ——background cross section of nuclide r in region I;

[0060] N I,r —nucleon density of nuclide r in region I;

[0061] ——the total microscopic cross section of nuclide j in region I;

[0062] — macroscopic cross section of region I;

[0063] P I→J ——Collision probability from area I to area J;

[0064] After performing resonance self-screening calculations based on a typical human body model, a fine group cross section database suitable for boron neutron capture therapy can be obtained. In order to consider the impact of complex problems on computational efficiency, it is necessary to merge the fine group database into energy groups to obtain a wide group cross section database with a smaller number of energy groups to improve computational efficiency. A typical human body model is selected for calculation using a refined transport program to obtain a typical weight spectrum of important organs or parts of the human body. Based on the conservation of reaction rate, the fine group cross section database after resonance self-screening is merged into energy groups to obtain a wide group cross section database.

[0065] Step 2: Construct a boron neutron capture therapy model based on the human body model parameter data measured by the instrument and the treatment technology design parameters (dose parameters, collimator parameters, irradiation field, irradiation position, material parameters, etc.);

[0066] Step 3: Based on the wide group cross-section database obtained in step 1, a highly efficient deterministic program is used to perform neutron-photon transport calculations on the established boron neutron capture therapy model to obtain the dose distribution results.

[0067] The ingenuity of the method of the present invention lies in: proposing the use of an efficient deterministic method as the core of neutron-photon transport calculation in boron neutron capture therapy, providing an effective method for rapid dose calculation in boron neutron capture therapy, thereby realizing rapid dose calculation in boron neutron capture therapy without statistical variance, saving a large amount of computing time and computing resources.

Claims

1. A method for rapidly calculating the dose in boron neutron capture therapy, characterized by: The following steps are involved: Step 1: Create a nuclear database suitable for the human body based on deterministic methods; To produce deterministic cluster cross-section data applicable to the human body, it is first necessary to create a basic fine cluster cross-section database. Then, based on a human body model, resonance self-screening calculations are performed to obtain an effective self-screening cross-section suitable for boron neutron capture therapy. The basic fine cluster cross-section database is combined with the basic fine cluster cross-section database to obtain a fine cluster cross-section database suitable for boron neutron capture therapy. Finally, the database is merged to obtain a broad cluster cross-section database suitable for boron neutron capture therapy. The specific contents of step 1 are as follows: Radiation transport needs to consider both neutron and photon transport, so the basic fine group cross-section database needs to include fine group neutron and photon cross-section data; the fine group neutron-photon coupling database uses the MATXS format; the fine group neutron cross section and neutron-photon scattering matrix data are obtained from the evaluation nuclear data through resonance reconstruction and linearization, Doppler broadening, indistinguishable resonance region effective self-screening cross section processing, thermal neutron scattering processing, and energy group merging; for the calculation of fine group photoatom cross sections, since photoatom reactions do not need to consider temperature effects and resonance self-screening effects, it is only necessary to perform resonance reconstruction and linearization and energy group merging on the cross-section data given in the evaluation nuclear database to finally obtain the basic fine group cross-section database; After obtaining the basic fine group cross section database, it is necessary to perform resonant self-screening calculations based on the human body model to obtain the effective self-screening cross section suitable for boron neutron capture therapy. The macroscopic total cross section and macroscopic scattering cross section of the effective self-screening are calculated by the coupled iterative method of formula (1): Where: Σ t,i (E)——the total macroscopic cross section of neutrons with energy E in region i; φ i (E)——the neutron flux rate of energy E in region i; φ j (E′) — the neutron flux rate of energy E′ at region j; V i ——the volume of region i; P j→i (E) — the probability of a neutron with energy E colliding from region j to region i; Σ s,j (E′) — macroscopic scattering cross section of neutrons with energy E′ at region j; f(E′→E)dE′—the probability that a neutron with energy E′ collides with an atom and the neutron’s energy is within dE′ of E′. For the non-resonance energy range, the calculation of the effective self-screening cross section requires interpolating the resonance integral table of the background cross section to obtain the effective self-screening cross section of the nuclide; The human body model in boron neutron capture therapy is a non-uniform model. For the non-uniform model, the background cross section is calculated by formula (2): Where: ——background cross section of nuclide r in region I; N I,r —nucleon density of nuclide r in region I; ——the total microscopic cross section of nuclide j in region I; — macroscopic cross section of region I; P I→J ——Collision probability from area I to area J; After performing resonance self-screening calculations based on a human body model, a fine group cross-section database suitable for boron neutron capture therapy is obtained. It is necessary to merge the fine group cross-section database into energy groups to obtain a wide group cross-section database with a small number of energy groups to improve computational efficiency. A human body model is selected for calculations using a refined transport program to obtain typical weight spectra of important organs or parts of the human body. Based on the conservation of reaction rates, the fine group cross-section database after resonance self-screening is merged into energy groups to obtain a wide group cross-section database. Step 2: Construct a boron neutron capture therapy model based on the human body model parameter data measured by the instrument and the treatment technology design parameters; Step 3: Based on the wide group cross-section database obtained in step 1, a deterministic program is used to perform neutron-photon transport calculations on the established boron neutron capture therapy model to obtain the dose distribution results.

2. The method for rapid calculation of dose in boron neutron capture therapy according to claim 1, characterized in that: The treatment technology design parameters described in step 2 include dose parameters, collimator parameters, irradiation field, irradiation position and material parameters.

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