KR102143063B1 - Apparatus and method for dose calculation of neutron beam and recording medium storing dose calculation program - Google Patents

Abstract

A method of calculating the radiation dose of a neutron beam applied to a patient′s treatment planning system includes the steps of: calculating each response parameter for neutrons and photons generated by a linear accelerator for neutron treatment, and storing the same as a data file in advance; and calculating the radiation dose of the patient in a deterministic method by performing convolution and superposition based on the direction, energy and size of the beam to be irradiated to the patient using the stored response parameter.

Description

A recording medium in which the method, device, and radiation dose calculation program are stored in the neutron beam radiation dose calculation method. {APPARATUS AND METHOD FOR DOSE CALCULATION OF NEUTRON BEAM AND RECORDING MEDIUM STORING DOSE CALCULATION PROGRAM}

A method for calculating the radiation dose of a neutron beam, a device and a recording medium in which a radiation dose calculation program is stored, and more specifically, the radiation dose of a neutron beam for calculating the amount of radiation entering the human body through radiation therapy or radiation exposure using neutrons. It relates to a recording medium in which a calculation method, an apparatus and a radiation dose calculation program are stored.

With the development of science and technology today, technologies using radioactive materials are being used in a variety of ways, and interest in their usefulness and the impact on the environment or life caused by radiation exposure is increasing.

Typically, radiation therapy has been used as a technology for treating patients in the medical field. In particular, neutron therapy has a high rate of mortality of cancer cells and is known to be effective in treating cancers such as elderly cancer patients and intractable cancers, but it is a reality that it has been difficult to use due to issues such as technology development.

With recent advances in technology, the development of linear accelerator technology for generating neutron beams in Korea as well as overseas has been made, and accordingly, radiation therapy using neutron beams is expected to be actively progressed.

Since there are many cases and studies that even a small amount of radiation exposure can be harmful to the human body, it is most important to properly plan the radiation exposure dose to the patient due to radiation treatment. Be able to calculate quickly.

In the calculation of radiation dose, two largely deterministic and probabilistic methods can be used. In conventional X-ray therapy, a deterministic algorithm using the Convolution/Superposition method is mainly applied. However, in the case of calculating the neutron dose, the response pattern is different compared to that of X-rays, and the dose component is complex, so a new dose calculation algorithm different from the conventional X-ray and electron beam treatments is required.

In the conventional neutron therapy, the radiation dose was calculated using a simulation method based on the Monte Carlo method, which is a probabilistic algorithm. However, this method has a problem that the accuracy is similar to that of the calculation using the deterministic algorithm, but it takes more time to calculate, so it is not suitable for application to the treatment of patients who need to immediately calculate the radiation dose and establish a treatment plan.

International Atomic Energy Agency, "Current status of neutron capture therapy", IAEA-TECDOC-1223, 2001

An object of the present invention is to provide an algorithm for calculating a neutron radiation dose, and to provide a method for calculating a radiation dose of a neutron beam for calculating a radiation dose entering a human body in the case of radiation therapy using neutrons using the same.

In order to achieve the above object, according to an aspect of the present invention, as a method for calculating the radiation dose of a neutron beam applied to a treatment planning system of a patient, each response parameter for neutrons and photon rays generated from a linear accelerator for neutron therapy Calculating and storing in advance as a data file, and using the stored response parameter, the radiation dose of the patient by convolution and superposition based on the beam direction, energy and size of the beam to be irradiated to the patient There is provided a radiation dose calculation method comprising the step of calculating.

According to an embodiment of the present invention, the step of calculating the reaction parameter and storing it as a data file in advance may include storing a calculation result using a Monte Carlo (MC) simulation in a library.

According to an embodiment of the present invention, when applying Boron neutron capture therapy (BNCT) radiation therapy to the patient, the response parameter includes a neutron weighting factor and TEGMA (Total Energy Generated per Unit Mass). I can.

According to an embodiment of the present invention, when applying BNCT radiation therapy to the patient, the radiation dose includes a gamma dose and a neutron dose, and the neutron dose may be calculated as shown in Equation 1.

[Equation 1]

는 지점(i, j, k)에서 흡수된 중성자 선량을 의미하고, k는 커널,

는 뇌 밀도,

는 중성자 소스 에너지 가중 인자,

는 에너지 그룹,

는 TEGMA, 즉 지점(i, j, k)에서 각 에너지 그룹의 단위 질량당 생성된 총 에너지를 나타낸다.here,

Is the absorbed neutron dose at point (i, j, k), k is the kernel,

Is the brain density,

Is the neutron source energy weighting factor,

Is the energy group,

Denotes TEGMA, i.e., the total energy generated per unit mass of each energy group at points (i, j, k).

는 식 2와 같이 산출될 수 있다.According to an embodiment of the present invention, the

Can be calculated as Equation 2.

[Equation 2]

는 Q-value를 의미하고,

는 각 에너지 그룹의 매크로 단면적의 합,

는 지점(i, j, k)에서 각 에너지 그룹의 중성자 플럭스(flux)를 나타낸다.here,

Means Q-value,

Is the sum of the macro cross-sectional areas of each energy group,

Represents the neutron flux of each energy group at points (i, j, k).

According to another aspect of the present invention, as a device for calculating the radiation dose of a neutron beam for a treatment plan of a patient, the response parameters for neutrons and photons generated from a linear accelerator for neutron therapy are calculated and stored in advance as a data file. One or more processors that control and calculate the radiation dose of the patient by performing convolution and superposition calculations (convolution and superposition) based on the direction of the beam to be irradiated to the patient, energy and size of the beam using the stored response parameter, and the There is provided a radiation dose calculation apparatus including a display that outputs a distribution of radiation dose as a 2D or 3D image.

According to an embodiment of the present invention, the at least one processor may control to store a calculation result using a Monte Carlo (MC) simulation in a library.

According to an embodiment of the present invention, when BNCT radiation therapy is applied to the patient, the response parameter may include a neutron weighting factor and TEGMA.

According to another aspect of the present invention, a storage medium storing instructions, wherein when the instructions are executed by at least one processor, the at least one processor causes the radiation dose calculation method of a neutron beam to be applied to a patient's treatment planning system. The at least one step is set to perform at least one step of, wherein the at least one step includes calculating reaction parameters for each of the neutrons and photons generated in the linear accelerator for neutron therapy and storing them in advance as a data file, and the stored A program for executing an operation comprising calculating the radiation dose of the patient by performing convolution and superposition calculations based on the direction of the beam to be irradiated to the patient, the energy and size of the beam using a response parameter. A recorded computer-readable storage medium is provided.

The method of calculating the radiation dose of a dialogue neutron beam according to various embodiments of the present invention has established a new concept, Total Energy Generated per Unit Mass (TEGMA), so that the convolution/superposition deterministic method can be applied to the neutron dose calculation. This method can improve the accuracy of the calculation of the radiation dose by neutrons and photons in various fields (medical fields, etc.) using a neutron beam, and at the same time, can significantly reduce the calculation time required. In addition, since the dose can be calculated within a short time with high reliability, it can be applied to an actual treatment plan from a practical point of view. Through this, it can be utilized as a radiation dose calculation engine for an essential treatment planning system along with the development of treatment technology using a neutron beam.

The effects obtainable in the present invention are not limited to the above-mentioned effects, and other effects not mentioned can be clearly understood by those of ordinary skill in the technical field to which the present invention belongs from the following description. will be.

1 is a block diagram of a treatment planning system (TPS) using a method of calculating a radiation dose of a neutron beam according to an embodiment of the present invention.
2 is a block diagram of a dose calculation algorithm module for BNCT according to an embodiment of the present invention.
3 is a graph showing a radiation dose graph of relative BNCT dose elements according to depths of each of the MC and CS methods for a BNCT beam field size of 1x1 (2.5x2.5 mm 2) according to an embodiment of the present invention.
4 is a graph showing a radiation dose graph of relative BNCT dose elements according to depths of each of the MC and CS methods for a BNCT beam field size of 3x3 (7.5x7.5 mm 2) according to an embodiment of the present invention.

Advantages and features of the present invention, and methods for achieving them will be clarified with reference to embodiments described below in detail together with the accompanying drawings. However, the present invention is not limited to the embodiments disclosed below, but may be implemented in various different forms, and only the embodiments allow the disclosure of the present invention to be complete, and common knowledge in the technical field to which the present invention pertains. It is provided to fully inform the holder of the scope of the invention, and the invention is only defined by the scope of the claims. The same reference numerals refer to the same components throughout the specification.

In the description of the embodiments of the present invention, when it is determined that a detailed description of known functions or configurations may unnecessarily obscure the subject matter of the present invention, the detailed description will be omitted, and terms to be described below are used in the embodiments of the present invention. The terms are defined in consideration of the function of. It can be changed according to the intention or custom of the user or operator. Therefore, the definition should be made based on the contents throughout this specification.

Combinations of each block of the attached block diagram and each step of the flowchart may be executed by computer program instructions (execution engine), and these computer program instructions are used on a processor of a general purpose computer, special purpose computer or other programmable data processing equipment. As it can be mounted, its instructions executed by the processor of a computer or other programmable data processing equipment generate means for performing the functions described in each block of the block diagram or each step of the flowchart.

These computer program instructions can also be stored in computer readable or computer readable memory that can be oriented to a computer or other programmable data processing equipment to implement a function in a particular way, so that computer readable or computer readable memory The instructions stored in it are also possible to produce an article of manufacture containing instructions means for performing the functions described in each block of the block diagram or in each step of the flowchart.

In addition, since computer program instructions can be mounted on a computer or other programmable data processing equipment, a series of operation steps are performed on a computer or other programmable data processing equipment to create a computer-executable process. It is also possible for the instructions to perform the data processing equipment to provide steps for executing the functions described in each block in the block diagram and in each step in the flowchart.

In addition, each block or each step can represent a module, segment, or portion of code that includes one or more executable instructions for executing specified logical functions, and in some alternative embodiments referred to in blocks or steps It should be noted that it is also possible for functions to occur out of sequence. For example, two blocks or steps shown in succession may in fact be performed substantially simultaneously, and the blocks or steps may be performed in the reverse order of a corresponding function as necessary.

The present invention relates to a method for calculating a radiation dose of a neutron beam and a recording medium storing a radiation dose calculation program. Particularly, in the present specification, the calculation of the radiation dose of a neutron beam for neutron treatment such as boron neutron capture therapy (BNCT) is described as an example, but is not limited thereto. The algorithm for calculating the radiation dose of the neutron beam proposed in the present invention can be applied to any field requiring rapid radiation dose calculation, such as in the case of radiation exposure.

Hereinafter, preferred embodiments of the present invention will be described with reference to the accompanying drawings. The embodiments of the present invention disclosed in the text are exemplified for the purpose of explanation only, and the embodiments of the present invention may be implemented in various forms and should not be construed as being limited to the embodiments described in the text. .

The present invention is not intended to limit the present invention to a specific disclosed form, as various changes may be added and various forms may be applied, and all changes, equivalents or substitutes included in the spirit and scope of the present invention It should be understood to include.

1 is a block diagram of a treatment planning system (TPS) 100 using a method of calculating a radiation dose of a neutron beam according to an embodiment of the present invention.

Referring to FIG. 1, a treatment planning system 100 according to an embodiment of the present invention includes an RT-structure module, an RT-plan module, an RT-dosage module and a dose calculation algorithm module 110.

For example, the treatment planning system 100 may establish an irradiation plan of a neutron beam using a DICOM CT image, calculate a dose distribution through a dose calculation algorithm, and visually output it through a viewer to assist in establishing a treatment plan. have.

Specifically, the treatment planning system 100 contours the Target/OAR structure using the DICOM CT image in the RT-structure module, and based on this, the beam direction of the neutron beam of the radiation beam system in the RT-plan module, Energy, size, etc. can be determined. The dose calculation algorithm module 110 calculates the radiation dose of the neutron beam using information such as the beam direction, energy, and size of the neutron beam determined as described above. Using the calculated radiation dose, the RT-dose module calculates a 2D or 3D dose distribution, and makes it easy to visually check the dose distribution through the RT PACS viewer.

The treatment planning system 100 according to an embodiment of the present invention may be implemented in the form of an application on an electronic device to output a dose distribution during neutron beam treatment with respect to a patient's DICOM CT image input through a viewer. I can. In this case, the RT-structure module, the RT-plan module, the RT-dose module, and the dose calculation algorithm module 110 may all be executed in the electronic device, and at least one of them may be executed in an external device of the electronic device. Data transmission/reception between the modules may be performed through a direct physical connection or a wired/wireless communication network. The operations of the RT-structure module, the RT-plan module, the RT-dose module and the dose calculation algorithm module 110 may be controlled by one or more processors. That is, the operations of the RT-structure module, the RT-plan module, the RT-dose module and the dose calculation algorithm module 110 are stored in one or more storage media in the form of a computer program, and can be controlled to be executed by one or more processors. have. For example, the electronic device may be a smartphone, a tablet personal computer, a mobile phone, a video phone, an e-book reader, a desktop PC, a laptop PC, a netbook computer, and a work. It may include at least one of a workstation, a server, a personal digital assistant (PDA), a media box, an electronic dictionary, or a wearable device. Wearable devices include accessory types (e.g. watches, rings, bracelets, anklets, necklaces, glasses, contact lenses, or head-mounted-devices (HMDs), fabric or clothing integrals (e.g. electronic clothing), body It may include at least one of an attachment type (for example, a skin pad or a tattoo), or an implantable circuit In various embodiments, the electronic device is flexible, or It may be a combination of two or more of the aforementioned various devices, but the electronic device is not limited to the aforementioned devices.

The treatment planning system 100 according to an embodiment of the present invention may further include an input/output device (not shown) for receiving and outputting a command. The input/output device may receive commands or data such as conditions for calculating radiation dose from a user or an external device, or output 2D or 3D image data of a dose distribution. For example, the input device may include a touch panel, a keyboard, a mouse, a pen sensor, a microphone, and the like, and the output device may include a display, audio, and the like.

In order to be able to easily calculate the radiation dose of a patient to be treated with a neutron beam through such a treatment planning system 100, the present invention calculates the radiation dose of the dose calculation algorithm module 110, which provides a certain level of accuracy and has a high calculation speed. Suggest a method.

In general, two types of deterministic and probabilistic methods can be used to calculate radiation dose. In conventional X-ray treatment, a deterministic algorithm using the convolution/superposition method has been mainly applied. However, in the case of neutron dose calculation in neutron therapy, the response pattern is different from that of X-rays, and the dose component is complex, so a new dose calculation algorithm is required.

The Monte Carlo simulation method, which is a probabilistic algorithm, was used for the conventional neutron dose calculation. However, in the case of the Monte Carlo method, the accuracy is similar compared to the calculation using the deterministic algorithm, but the calculation time is much longer, and thus it is difficult to use it in a real-time treatment plan for neutron treatment.

In order to solve this problem, the present invention introduces the Convolution/Superposition algorithm to the neutron radiation dose calculation. The Convolution/Superposition algorithm according to an embodiment of the present invention stores calculation results using Monte Carlo, for example, weighting factors, TEGMA, kernel, and TERMA, in a data library, and loads the stored data during calculation, convolutions and overlaps Calculate the dose in a way. This makes it possible to calculate the dose with similar accuracy at a much faster rate compared to conventional Monte Carlo simulation methods.

The method of calculating the radiation dose of a neutron beam according to an embodiment of the present invention uses Monte Carlo simulation to pre-calculate reaction parameters for neutrons and photons generated in a linear accelerator for neutron therapy, and call them when calculating Convolution/Superposition. Process to save the data file.

In addition, the present invention proposes a radiation dose calculation program for commercialization in actual neutron therapy. The corresponding program code can be implemented using a variety of programming languages (eg, C language). The code can be loaded into the treatment planning system, and the code can be constructed by introducing parallel operations to speed up the calculation. Such a program can be executed by one or more processors.

2 is a block diagram of a dose calculation algorithm module 110 for Boron neutron capture therapy (BNCT) according to an embodiment of the present invention.

BNCT is one of the targeted radiation therapy. After injection of boron drugs, the B-10 (boron-10) compound accumulates in tumor cells, while the boron concentration in healthy tissues decreases. Neutrons are irradiated at the site of the tumor, which is thermally neutronized through tissue, and interacts with B-10 nuclei. In particular, B-10 with a large absorption cross-sectional area for thermal neutrons can significantly reduce thermal neutron flux in tumor cells. As a result, alpha and lithium particles with high linear energy transfer (LET) are produced, which selectively destroy tumor cells.

In contrast to photon radiation therapy, the transport of high-temperature neutrons contains radioactive components with different relative biological effectiveness (RBE) and therefore must be considered separately. BNCT has four dose components: boron, nitrogen, fast neutron, and gamma radiation dose. ¹⁰ B(n, α) ⁷ Boron dose produced by Li reaction, ¹⁴ N(n, p) ¹⁴ C reaction solution nitrogen dose, ¹ H(n, n') ¹ H reaction fast neutron dose and ¹ There are internal gamma doses from H(n, γ) ² H, ¹⁰ B(n, α) ⁷ Li reactions and external gamma doses from some contaminated gamma rays from beam shaping assembly (BSA). For computerized treatment planning, techniques for calculating the effect of each of these dose components in BNCT are required.

There is no uniform international dosimetry guideline for BNCT radiation therapy. Only MC (Monte Carlo)-based treatment planning system was applied to clinical BNCT. However, as described above, the dose calculation using the MC energetic simulation method takes a very long time and requires the development of a faster deterministic method.

The CS (Convolution/Superposition) method is a deterministic dose calculation method using MC simulation results. It is usually applied to the dose calculation method for photon beam radiation therapy. The calculation speed of the CS method is faster than that of the MC method. The calculation result of the MC method is more accurate than the CS method for the photon beam, but the difference is not large.

The present invention proposes a method of applying the CS method to the BNCT dose calculation algorithm in both gamma rays and neutron dose components. To this end, hereinafter'TEGMA' is defined as an abbreviation of'Total Energy Generated per unit Mass'. This is the same concept as TERMA (Total Energy Released per unit Mass) used in the CS method, and the difference is that it represents the energy generated by the nuclear reaction. TEGMA is convolutional to the neutron energy deposition kernel, and the neutron dose is calculated. After calculating the dose with the CS algorithm, the results can be compared with the MC simulation to compare the accuracy and time required.

CS

Convolution is a convolution operation, which refers to an integral operation in which one function is superimposed on another function. Superposition is a superposition operation, and the superposition principle represents a linear system. The net response generated by two or more stimuli is the sum of the responses generated individually by each stimulus. The superposition method is a variant of the convolution method.

According to an embodiment of the present invention, the CS algorithm is applied to both gamma and neutron dose calculations for BNCT. For example, the unit sizes of the kernel, TERMA, internal gamma effect, and neutron reaction rate used in the CS method are 61x61x120, 1x1x60, 61x61x60, and 41x41x60 voxels, respectively. The voxel size is 2.5mm x 2.5mm x 2.5mm. Convolution is performed along the z-axis parallel to the beam direction, and superposition is performed along the x and y axes of the beam field. After performing convolution and superposition of the product and sum of voxels, the dose result is derived.

Dose component

According to an embodiment of the present invention, in order to calculate the BNCT dose using the CS method, the BNCT dose components are classified into five types. First, it is divided into two categories: gamma dose and neutron dose, which are particles generated from nuclear reaction deposition dose. Then, the gamma dose is divided into external and internal gamma doses 210 and 220. The external gamma dose 210 is a gamma ray emitted from the BSA produced during the generation of a neutron source. The internal gamma dose 220 is generated by nuclear reactions with neutrons and substances in the human body during treatment. The last one is the neutron dose 230, which includes boron, nitrogen and fast neutron doses.

External gamma dose

The external gamma dose 210 reaches the patient due to incident photons from the BSA. The external gamma dose 210 is calculated by the conventional CS method. The absorbed dose, D(i, j, k), is expressed as the convolution of the dose variance kernel and the total energy released per unit mass known as TERMA.

Here, TERMA is defined as the total energy released during a photon's interaction with a substance per unit mass, that is, the kinetic energy of the resulting secondary electron, including the energy released from other interactions, such as the excited state. This radiant energy lost by photon interactions can be absorbed locally or away. TERMA relies on the primary photon energy effect.

The dose distribution kernel forms the probability per unit volume of energy deposited in the vicinity of the first-order photon interaction. The kernel depends on its location. The dose distribution kernel results from the sum of the volumetric components at points (i, j, k) and the energy deposition of single energy photons at the interaction points (i', j', k').

는 <수학식 1>과 같이 나타낼 수 있다.Dose absorbed at points (i, j, k) of external gamma dose 210 according to an embodiment of the present invention

Can be expressed as <Equation 1>.

는 가중 인자(weighting factor)이고,

는 에너지 그룹으로

를 나타낸다.

는 커널을 나타내고,

는 TERMA, 즉 상호작용 지점(i', j', k')에서 각 에너지 그룹의 단위 질량당 방출된 총 에너지를 의미하고, <수학식 2>와 같이 나타낼 수 있다. here,

Is the weighting factor,

As an energy group

Indicates.

Represents the kernel,

Denotes TERMA, that is, the total energy released per unit mass of each energy group at the interaction points (i', j', k'), and can be expressed as <Equation 2>.

는 상호작용 지점(i', j', k')에서 각 에너지 그룹의 광자 에너지 영향을 의미하고,

는 각 에너지 그룹의 질량 감쇠 계수를 의미한다. here,

Means the photon energy effect of each energy group at the point of interaction (i', j', k'),

Denotes the mass decay coefficient of each energy group.

As shown in Equation 1 above, three parameters of a weighting factor, TERMA, and a kernel are required for the convolution & superposition 207 for calculating the external gamma dose 210.

According to an embodiment of the present invention, the weighting factor can be obtained from the gamma energy spectrum 203 originating from the BSA of an accelerator-based Boron Neutron Capture Therapy (A-BNCT) facility. For example, the source of the gamma energy spectrum 203 may be created by MC simulation using the MCNP (Monte Carlo N-Particle) 6.1 code. For example, the MC simulation classifies gamma particles from BSA according to their energy and divides them into 71 energy groups, and a value obtained by counting the number of particles by 71 energy groups can be calculated as a weighting factor of each energy group.

According to an embodiment of the present invention, TERMA may be calculated manually along the beam direction by multiplying the photon energy effect and the mass attenuation coefficient in brain matter (ICRU-44) of each energy group. This can be calculated to a depth of 15 cm at 2.5 mm intervals.

According to an embodiment of the present invention, the kernel may be pre-calculated by MC simulation using edknrc, which is an EGSnrc code for calculating the energy deposition kernel of photon interaction. It was created from a homogeneous brain component with a 15 cm sphere radius. The electron cut-off energy was set to 512 keV, and the photon cut-off energy was set to 10 keV. The energy deposition kernel is represented in a 2D polar grid. It consists of a cone and a sphere, and the spacing is chosen equally at 1˚ and 2.5mm respectively. After creating the kernel, its coordinates are converted from spherical to Cartesian. Then, the kernel value of the MC simulation included in the new Cartesian kernel voxel is matched to each of these Cartesian voxels. The matched kernel values are averaged and displayed as the new kernel values in all voxels of the Cartesian kernel.

Internal gamma dose

The internal gamma dose 220 is generated by photons derived from neutron capture reactions in human tissue. There can be two types of inner gamma, one of which is produced by the capture of ¹ H(n, γ) ² H hydrogen. Hydrogen in tissue absorbs thermal neutrons and releases 2.2 MeV photons. The other is the gamma rays from the ¹⁰ B(n, α) ⁷ Li reaction, ie the n-alpha reaction. From this reaction, about 94% of the recoiling Li-7 ions are produced in the excited state and emit 0.48 MeV gamma rays during de-excitation.

The phase-space file 201 of the neutron source from the BSA of A-BNCT is obtained by MC simulation using the MCNP 6.1 code.

는 <수학식 3>과 같이 나타낼 수 있다.Dose absorbed at points (i, j, k) of the internal gamma dose 220 according to an embodiment of the present invention

Can be expressed as <Equation 3>.

는 중성자 소스 에너지 가중 인자이고,

는 에너지 그룹으로

를 나타낸다.

는 지점(i, j, k)에서 각 에너지 그룹의 광자 에너지 플루엔스(fluence) 나타내고,

는 각 에너지 그룹의 질량 에너지 흡수 계수를 의미한다.here,

Is the neutron source energy weighting factor,

As an energy group

Indicates.

Denotes the photon energy fluence of each energy group at point (i, j, k),

Denotes the mass energy absorption coefficient of each energy group.

According to the ICRU-44 report, the formula for calculating the internal gamma dose (220) by convolution & superposition (209) in Equation 3 requires three parameters: weighting factor, photon energy fluence, and mass absorption coefficient in brain matter. To The product of the photon energy fluence and the mass absorption coefficient represents the deposited energy per unit mass.

According to an embodiment of the present invention, first, the internal gamma energy of 0.1 MeV to 10 MeV is divided into 100 groups at intervals of 100 keV, and then the weighting factor of each energy group is obtained from the neutron energy spectrum.

The photon energy fluence is the product of the photon fluence and the energy of the fluence. The photon fluence for each energy group is obtained from MC simulation using MCNP MC code and F4 tally. The median energy of the energy group is used as the energy multiplied by the fluence of the photon.

Neutron dose

The neutron dose 230 is generated by a nuclear reaction between a neutron and an atomic nucleus. There are three representative nuclear reactions in BNCT, a neutron trapping reaction of ¹⁴ N(n, p) ¹⁴ C, ¹⁰ B(n, α) ⁷ Li, which contributes to the dose of boron and nitrogen, respectively, and ¹ H ( n, n') is an elastic scattering reaction of ¹ H.

According to an embodiment of the present invention, the neutron source 201 of the phase-space file from the BSA of the A-BNCT facility is also obtained by MC simulation using the MCNP 6.1 code.

는 <수학식 4>와 같이 나타낼 수 있다.Dose absorbed at the point (i, j, k) of the neutron dose 230 according to an embodiment of the present invention

Can be expressed as <Equation 4>.

는 커널을 나타내고,

은 국부 축적을 의미한다.

는 뇌 밀도이다.

는 중성자 소스 에너지 가중 인자이고,

는 에너지 그룹으로, 여기서는

에너지를 복수의 에너지 그룹으로 나눈다.

는 TEGMA, 즉 지점(i, j, k)에서 각 에너지 그룹의 단위 질량당 생성된 총 에너지를 의미하고, <수학식 5>와 같이 나타낼 수 있다. here,

Represents the kernel,

Means local accumulation.

Is the brain density.

Is the neutron source energy weighting factor,

Is the energy group, where

Divide energy into multiple energy groups.

Is TEGMA, that is, the total energy generated per unit mass of each energy group at points (i, j, k), and can be expressed as <Equation 5>.

는 Q-value를 의미하고,

는 각 에너지 그룹의 평균 매크로 단면적을 의미한다.

는 지점(i, j, k)에서 각 에너지 그룹의 중성자 플럭스를 의미한다.here,

Means Q-value,

Means the average macro cross-sectional area of each energy group.

Denotes the neutron flux of each energy group at points (i, j, k).

In order to calculate the neutron dose 230 by convolution & superposition (209), three parameters are needed: kernel, brain density, and TEGMA. The above <Equation 4> and <Equation 5> are derived from the CS equation used for calculating the photon dose.

)를 로그 스케일로 199 개의 에너지 그룹으로 나누었다.According to an embodiment of the present invention, energy (

) Was divided into 199 energy groups on a logarithmic scale.

The kernel, which is the same concept as the photon energy deposition kernel, refers to the diffusion of energy generated from nuclear reactions. Since energy is accumulated locally, the kernel value is estimated to be 1. This can be confirmed by the stopping force of the Li particles and alpha particles produced from the n-alpha reaction. The alpha and Li particles have large linear energy transfers of 150 keV/µm and 175 keV/µm, respectively, and have a short irregularity within about 4.5 to 10 µm in the tissue. This distance is the cell diameter, that is, the size of the local accumulation.

는 핵반응으로부터 생성된 에너지, 광자에 대한 CS 방법에서 TERMA의 동의어일 수 있다. 이러한 개념은 본 발명에서 처음으로 제안되었다. 가중 인자,

는 중성자 소스 에너지 스펙트럼으로부터 계산된다. Q-value는 핵반응 동안 흡수되거나 방출되는 에너지의 양을 의미한다. 예를 들어, ¹⁰B(n, α)⁷Li 반응에서, Q-value는 2.33MeV (94%) 및 2.81MeV (6%)이다. ¹⁴N(n, p)¹⁴C 반응에서, Q-value는 0.63MeV이다.According to an embodiment of the present invention, TEGMA,

May be synonymous with TERMA in CS method for photons, energy generated from nuclear reaction. This concept was first proposed in the present invention. Weighting factor,

Is calculated from the neutron source energy spectrum. Q-value refers to the amount of energy absorbed or released during a nuclear reaction. For example, in a ¹⁰ B(n, α) ⁷ Li reaction, the Q-values are 2.33 MeV (94%) and 2.81 MeV (6%). ^{In the 14} N(n, p) ¹⁴ C reaction, the Q-value is 0.63 MeV.

는 TEGMA, 즉 지점(i, j, k)에서 각 에너지 그룹의 단위 질량당 생성된 총 에너지를 의미하고, <수학식 6> 및 <수학식 7>과 같이 정의될 수 있다. ^{In 1} H(n, n') ¹ H reaction,

Is TEGMA, that is, the total energy generated per unit mass of each energy group at points (i, j, k), and may be defined as <Equation 6> and <Equation 7>.

는 지점(i, j, k)에서 각 에너지 그룹의 중성자 플럭스를 의미한다. ¹H(n, n')¹H 반응의 경우, Q-value 대신에 입사 중성자 에너지의 반, 즉 평균 에너지 양성자가 중성자로 탄성 산란을 겪을,

가 이용된다. 즉, E_g는 입사 중성자가 속한 에너지 그룹의 평균 값으로, 중성자의 탄성 산란으로 생성된 양성자의 에너지를 입사 중성자 에너지의 절반으로 가정할 수 있음을 의미한다. here,

Denotes the neutron flux of each energy group at points (i, j, k). ^{For 1} H(n, n') ¹ H reaction, instead of Q-value, half of the incident neutron energy, that is, the average energy proton undergoes elastic scattering as a neutron,

Is used. That is, E _g is the average value of the energy group to which the incident neutron belongs, and it means that the energy of the proton generated by the elastic scattering of the neutron can be assumed to be half the energy of the incident neutron.

According to an embodiment of the present invention, the macro cross section is the effective target area of all nuclei contained in the volume of material for B-10, N-14 and H-1. Multiplying the macro cross section by the neutron flux gives the reaction rate. For example, obtained by MC simulation with MCNP 6.1 for each energy group and reaction type.

The results of comparing the dose calculated from the above-described MC method and CS method are shown in FIGS. 3 and 4.

According to an embodiment of the present invention, FIG. 3 shows a graph of the radiation dose of relative BNCT dose elements according to the depth of each of the MC and CS methods for a BNCT beam field size of 1x1 (2.5x2.5 mm 2 ). In addition, FIG. 4 shows a graph of the radiation dose of relative BNCT dose elements according to the depth of each of the MC and CS methods for a BNCT beam field size of 3x3 (7.5x7.5 mm 2 ). At this time, to calculate the dose by the MC method, the radiation dose of the relative BNCT dose elements according to the depth may be calculated and compared using the MCNP and the PHITS code.

Referring to FIGS. 3 and 4, in the MC simulation method for calculating the neutron dose, the MCNP code is conventionally mainly used, and in the case of MCNP, it can be confirmed that the error rate is less than 1% when compared to the CS method. In an embodiment of the present invention, since the CS method pre-calculates the parameters for calculating the radiation dose as described above by MC simulation and stores it as data, very similar dose results can be obtained. In addition, when comparing the dose by the CS method and PHITS, a difference of 10% or more was shown in the external and internal gamma doses, but similar results were observed in the neutron dose.

In addition, a comparison between the CS method and the dose calculation and calculation speed by MCNP and PHITS is shown in Table 1.

CPU (Intel) GPU Calculation time Uncertainty
@ 2cm depth core Processor
Fundamental frequency core Convolution/
superposition One 2.80 GHz 3,584
CUDA cores 1.5GHz 34.9 sec - MCNP 48 2.67 GHz - 70 min Less than 1% PHITS 4 2.80 GHz - 24 h 3~5%
48 2.67 GHz - 4 h

That is, as shown in <Table 1>, in the case of the CS method, there is no significant difference between the calculation result and the accuracy of the result by the MC method, and an improved effect of reducing the calculation speed to within tens of seconds can be expected.

As shown in FIGS. 3 and 4, when the CS method proposed in the present invention is applied to calculate the radiation dose of the neutron beam, it can be seen that the calculated result is similar to the MC method. Since the CS method can significantly reduce the computational time compared to the MC method, it can be effectively applied to the treatment planning system of patients applying neutron beam therapy.

As described above, the present invention newly established the concept of TEGMA and introduced it into the neutron dose calculation, thereby enabling a deterministic method of CS in neutron dose calculation unlike the prior art. Through this, since the calculation speed of neutron dose is rapidly accelerated, it is possible to calculate the dose very quickly, and it may be used in various fields requiring neutron dose calculation, such as in the case of exposure as well as neutron therapy.

In the above-described specific embodiments, constituent elements included in the invention are expressed in the singular or plural according to the presented specific embodiments. However, the singular or plural expression is selected appropriately for the situation presented for convenience of description, and the above-described embodiments are not limited to the singular or plural constituent elements, and even constituent elements expressed in plural are composed of the singular or However, even a component expressed in a singular number may be composed of a plurality.

Meanwhile, although specific embodiments have been described in the description of the present invention, various modifications may be made without departing from the scope of the technical idea implied by various embodiments. Therefore, the scope of the present invention is limited to the described embodiments and should not be defined, but should be defined by the claims and equivalents as well as the claims to be described later.

Claims (11)

As a method of calculating the radiation dose of a neutron beam applied to a patient's treatment planning system,
Calculating reaction parameters for neutrons and photons generated in a linear accelerator for neutron therapy, and storing them as a data file in advance; And
Comprising the step of calculating the radiation dose of the patient by performing convolution and superposition calculations based on the beam direction to be irradiated to the patient, energy and size of the beam using the stored response parameter, . The method of claim 1,
The step of calculating the reaction parameter and storing it in advance as a data file,
A radiation dose calculation method comprising the step of storing the calculation result using the MC (Monte Carlo) simulation in a library. The method of claim 1,
When applying Boron neutron capture therapy (BNCT) radiation therapy to the patient, the response parameter includes a neutron weighting factor and TEGMA (Total Energy Generated per Unit Mass). The method of claim 1,
When applying BNCT radiation therapy to the patient, the radiation dose includes a gamma dose and a neutron dose,
The neutron dose is calculated as in Equation 1, the radiation dose calculation method.
[Equation 1]

here,

Is the absorbed neutron dose at point (i, j, k), k is the kernel,

Is the brain density,

Is the neutron source energy weighting factor,

Is the energy group,

Represents TEGMA, i.e. the total energy produced per unit mass of each energy group at points (i, j, k). The method of claim 4,
remind

Is calculated as Equation 2, the radiation dose calculation method.
[Equation 2]

here,

Means Q-value,

Is the sum of the average macro cross-sectional areas of each energy group,

Represents the neutron flux of each energy group at points (i, j, k). As a device for calculating the radiation dose of a neutron beam for a patient's treatment plan,
Controls to calculate each response parameter for neutrons and photon rays generated in the linear accelerator for neutron therapy and store it in advance as a data file, and the beam direction, energy and size of the beam to be irradiated to the patient using the stored response parameter At least one processor for calculating the radiation dose of the patient by performing convolution and superposition based on; And
A display for outputting the distribution of the radiation dose as a 2D or 3D image. The method of claim 6,
The one or more processors,
A radiation dose calculation device that controls to store a calculation result using MC (Monte Carlo) simulation in a library. The method of claim 6,
When applying BNCT radiation therapy to the patient, the response parameter comprises a neutron weighting factor and TEGMA. The method of claim 6,
When applying BNCT radiation therapy to the patient, the radiation dose includes a gamma dose and a neutron dose,
The radiation dose calculation device, wherein the neutron dose is calculated as shown in Equation 1.
[Equation 1]

here,

Is the absorbed neutron dose at point (i, j, k), k is the kernel,

Is the brain density,

Is the neutron source energy weighting factor,

Is the energy group,

Represents TEGMA, i.e. the total energy generated per unit mass of each energy group at points (i, j, k). The method of claim 9,
remind

A radiation dose calculation device that is calculated as in Equation 2.
[Equation 2]

here,

Means Q-value,

Is the sum of the average macro cross-sectional areas of each energy group,

Represents the neutron flux of each energy group at points (i, j, k). As a storage medium that stores instructions,
The instructions are set to cause the at least one processor to perform at least one step of a method for calculating a radiation dose of a neutron beam applied to a patient's treatment planning system when executed by at least one processor, the at least one step ,
Calculating reaction parameters for neutrons and photons generated in a linear accelerator for neutron therapy, and storing them as a data file in advance; And
Using the stored response parameter, a beam direction to be irradiated to the patient, a convolution and superposition based on the energy and size of the beam, and calculating the radiation dose of the patient. Computer-readable storage medium on which the program is recorded.

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