JP2017009393A - Neutron measurement device, neutron measurement method, and treatment apparatus for boron neutron capture therapy - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide neutron measurement technology for measuring dose of neutrons by energy in real time, and a treatment apparatus for a boron neutron capture therapy using the measured dose of the neutrons.SOLUTION: A neutron measurement device 10 is provided with: a recoil proton type detector 11 which has a substance generating inelastic scattering with neutrons to be irradiated, and generates a first signal when the inelastic scattering due to the neutrons is generated; a nuclear reaction type detector 12 which has a substance generating a nuclear reaction with the neutrons, and generates a second signal when the nuclear reaction due to the neutrons is generated; a mixed type detector 13 which has a substance generating the inelastic scattering or nuclear reaction with the neutrons, and generates a third signal when the inelastic scattering or nuclear reaction is generated; an event quantity counting part 16 which counts event quantities generated in the respective detectors, respectively based on the first signal, the second signal, and the third signal; and a dose by energy calculation part 17 which calculates dose of the neutrons by energy based on the counted event quantities in the respective detectors.SELECTED DRAWING: Figure 1

Description

  Embodiments of the present invention relate to a neutron measurement technique for measuring a neutron dose, and a treatment apparatus for boron neutron capture therapy that performs treatment by irradiating a patient to which a medicine containing boric acid is administered with neutrons.

  In boron neutron capture therapy (BNCT) and material inspection that use neutron radiation for treatment, it is necessary to accurately grasp the state of the neutron beam to be irradiated. A measuring device for measuring the profile is required.

  Note that BNCT is one of radiotherapy methods, and a drug containing boron is preliminarily used by utilizing the property that a drug containing boron is accumulated at a higher concentration in tumor cells (for example, cancer cells) than normal cells. Drugs are accumulated in the tumor area when injected into the patient, and the neutron generated by the nuclear reactor or accelerator is irradiated to the tumor area to cause a nuclear reaction in the tumor cell, and the tumor is selectively selected by the energy given to the tumor cell. It is a cure that destroys it. In particular, it is expected as an effective treatment method for cancers with high malignancy such as brain tumors in which cancer cells and normal cells are mixed.

  By the way, since neutrons do not have a charge, when they enter a material, reactions such as elastic scattering, inelastic scattering, and neutron capture occur with nuclei in the material. These reaction cross sections are determined by the energy of the neutron and the type of the nucleus of the material. The larger the reaction cross section, the easier the reaction.

  As a method of measuring the dose of neutrons, a method of measuring protons, which are charged particles blown away by scattering, protons, deuterium (Triton), alpha rays (helium nuclei) generated by reaction with neutrons, etc. There are two main methods for measuring charged particles. Regardless of the method, charged particles generated by the reaction can be measured with a scintillator detector or a semiconductor detector.

  Development of devices that measure neutron dose has been underway, and as detectors for measuring neutron beams, gas detectors / scintillator detectors containing helium and boron with a large capture cross-section for neutrons, A recoil-proton detector using a liquid scintillator containing a large amount of hydrogen having a large inelastic scattering cross section with respect to is disclosed. Since the sensitivity changes according to the neutron energy, a detector suitable for the neutron energy to be measured can be selected and used.

  In particular, liquid scintillators and plastic scintillators with high sensitivity to neutrons with high energy such as epithermal neutrons can discriminate neutrons and gamma rays, and neutron energy can be measured by using an unfolding method.

  In addition, as a neutron dosimetry device applied to BNCT, a method of measuring the neutron dose at the center of the beam from the surrounding neutron data by providing an opening on the neutron beam to be measured, and the amount of activation of a gold thin film, etc. A technique for measuring a neutron dose by evaluating the above is disclosed.

JP 2014-190754 A JP 2004-233168 A

Radiation Measurement Handbook 4th Edition

  The above-mentioned neutron dosimetry technology can individually evaluate the dose of high-energy or low-energy neutrons because the sensitivity of the detector depends on the energy of neutrons. The dose of neutron radiation could not be evaluated.

  In addition, the technology that evaluates the amount of activation of gold thin films, etc. can evaluate the energy characteristics of neutrons by changing the type of thin film while reducing the effect on neutron radiation, but it can accurately measure the dose of neutrons in real time. It is difficult to measure well. For this reason, there is a problem that application is difficult when neutron dose data such as BCNT is required in real time.

  The present invention has been made in view of such circumstances, and provides a neutron measurement technique for measuring a dose of neutrons according to energy in real time, and a treatment apparatus for boron neutron capture therapy using the measured dose of neutrons. For the purpose.

  In the neutron measurement apparatus according to the embodiment of the present invention, a recoil proton that has a substance that generates inelastic scattering with an irradiated neutron and generates a first signal when the inelastic scattering by the neutron occurs. A nuclear reaction detector having a substance that causes a nuclear reaction with the neutron, and generating a second signal when the nuclear reaction by the neutron occurs, the neutron, and the inelasticity A mixed detector that has a substance that causes scattering or the nuclear reaction, and generates a third signal when the inelastic scattering or the nuclear reaction occurs, and the first signal and the second signal And an event amount counter for counting the amount of events generated in each detector based on the third signal, and calculating the dose of the neutrons for each energy based on the event amount in each detector counted. By energy Characterized in that it comprises a quantity calculating section.

  In the neutron measurement method according to the embodiment of the present invention, a first recurrence occurs when the inelastic scattering due to the neutron occurs using a recoil proton type detector having a material that causes inelastic scattering and the neutron to be irradiated. A step of receiving a signal of 1, a step of receiving a second signal generated when the nuclear reaction by the neutron occurs using a nuclear reaction detector having a substance that causes a nuclear reaction with the neutron, Receiving a third signal generated when the inelastic scattering or the nuclear reaction is generated using a neutron and a mixed detector having a substance that causes the inelastic scattering or the nuclear reaction; 1, counting the amount of events generated by each detector based on the second signal, and the third signal, respectively, and counting the event at each counted detector. Characterized in that it comprises a step of calculating the dose of the neutrons by energy based on the cement amount.

  According to an embodiment of the present invention, a neutron measurement technique for measuring a neutron dose according to energy in real time and a treatment apparatus for boron neutron capture therapy using the measured neutron dose are provided.

The block diagram of the neutron measuring apparatus which concerns on 1st Embodiment. (A) Schematic sectional drawing which shows an example of joining of the recoil-proton type detector and optical transmission part (optical fiber) applied to this embodiment, (B) II sectional drawing of FIG. 2 (A). (A) The schematic sectional drawing which shows an example of joining of the recoil-proton type detector applied to this embodiment, and an optical transmission part (wavelength shift bar), (B) II-II sectional drawing of FIG. 3 (A). (A) The schematic sectional drawing which shows an example of joining of the recoil-proton type detector applied to this embodiment, and an optical transmission part (wavelength shift bar), (B) III-III sectional drawing of FIG. 4 (A). The figure which shows the structure of the event amount counting part which discriminate | determines the event which generate | occur | produces by gamma ray origin. (A) Explanatory drawing which showed an example of the energy spectrum of neutron in BNCT, (B) Explanatory drawing which showed the detection sensitivity of each detector with respect to the energy of neutron. The flowchart which shows the measurement procedure of the neutron measuring apparatus which concerns on 1st Embodiment. The block diagram of the neutron measuring apparatus which concerns on 2nd Embodiment. The figure which shows the structure of the treatment apparatus for boron neutron capture therapy to which the neutron measuring apparatus which concerns on 3rd Embodiment was applied. (A) The figure which shows the structure of the treatment apparatus for boron neutron capture therapy to which the modification of the neutron measuring apparatus which concerns on 3rd Embodiment was applied, (B) The figure which shows the state which attached the several detection unit to the affected part of a patient .

(First embodiment)
Hereinafter, embodiments of the present invention will be described with reference to the accompanying drawings.
As shown in FIG. 1, a neutron measurement apparatus 10 according to the first embodiment includes a recoil-proton detector 11, a nuclear reaction detector 12, a mixed detector 13, and an optical transmission unit 14 (14 ₁ , 14 ₁ , 14 ₂ , 14 ₃ ), a light detection unit 15 (15 ₁ , 15 ₂ , 15 ₃ ), an event amount counting unit 16, and a dose-by-energy calculation unit 17. FIG. 1 shows a configuration example in the case where a scintillator type is used as the three neutron detectors of the recoil proton type detector 11, the nuclear reaction type detector 12, and the mixed type detector 13.

  The recoil-proton detector 11 contains a neutron to be irradiated and a substance that generates inelastic scattering inside, and neutron detection in which a first signal is generated by the recoil proton generated when the neutron is inelastically scattered. It is a vessel. A light element such as hydrogen is used as a substance that causes inelastic scattering with neutrons.

  Examples of the recoil proton detector 11 include a plastic scintillator, a liquid scintillator, and a still benzene scintillator. In addition to a scintillator, a detector (particularly a proportional counter) in which a gas containing a light element such as hydrogen is enclosed may be used. The first signal generated when inelastic scattering occurs varies depending on the type of detector. When the scintillator type is used, the first signal becomes scintillator light, and when a gas detector such as a proportional counter is used, the first signal Become. The shape of the detector is not limited to a rectangular shape such as a cylindrical shape.

The first signal generated by the recoil proton in the recoil proton detector 11 depends on the energy of the irradiated neutron, and the recoil proton detector 11 is sensitive to neutrons in a high energy region. The scintillator-type recoil-proton detector 11 can detect neutrons of about 100 keV or higher, while for neutrons having energy lower than this, a nuclear reaction detector containing lithium ( ⁶ Li) or the like. Compared to 12, the reaction cross section is several orders of magnitude smaller.

In the nuclear reaction type detector 12, a substance that causes a nuclear reaction with the neutron to be irradiated is sealed or coated inside the detector, and the first reaction is caused by radiation generated when the substance and the neutron cause a nuclear reaction. This is a neutron detector that generates two signals. As a substance that causes a nuclear reaction, at least one of boron ( ¹⁰ B, lithium ( ⁶ Li), and helium ( ³ He) is used.

  The nuclear reaction type detector 12 generates charged particles by a reaction as shown in Table 1 when a nuclear reaction is caused by irradiated neutrons.

The nuclear reaction type detector 12 may be any detector that can measure charged particles generated by this reaction, and either a scintillator type or a gas detection type is applicable. For example, a surface of a scintillator such as a LiCAF scintillator or a plastic scintillator coated with ¹⁰ B, ⁶ Li, ³ He or the like can be used. The second signal generated when a nuclear reaction due to neutrons occurs differs depending on the type of detector. When the scintillator type is used, it becomes scintillator light, and when a gas detector such as a proportional counter is used, an electric signal It becomes. The shape of the detector is not limited to a rectangular shape such as a cylindrical shape.

When the nuclear reaction detector 12 is irradiated with neutrons, neutron capture depending on the cross-sectional area is generated. In particular, in the case of a LiCaF scintillator containing ⁶ Li, when a neutron having an energy of 1 keV is incident, a (n, α) reaction occurs.

The nuclear reaction is determined by the atomic density of ⁶ Li and the energy dependence of the reaction cross section is roughly proportional to the reciprocal of the neutron velocity. For this reason, the nuclear reaction type detector 12 has high sensitivity to neutrons in a low energy region.

  The mixed type detector 13 encloses or is coated with a substance that causes inelastic scattering or nuclear reaction together with neutrons, and generates a third signal when inelastic scattering or nuclear reaction occurs. .

The mixed detector 13 is made of a material containing a light element such as hydrogen that is easily inelastically scattered with neutrons and an element (at least one of ¹⁰ B, ⁶ Li, and ³ He) that easily reacts with neutrons. Enclosed or coated inside.

Examples of the mixed detector 13 include a plastic scintillator containing ¹⁰ B and ⁶ Li, and a liquid scintillator. It should be noted that both elements contained in the detector need not be mixed uniformly, and either element may be applied to the surface.

  The third signal generated when an inelastic or nuclear reaction by neutrons occurs depends on the type of detector. If a scintillator type is used, the third signal becomes scintillator light, and a gas detector such as a proportional counter is used. In the case, it becomes an electric signal. The shape of the detector is not limited to a rectangular shape such as a cylindrical shape.

  Since the mixed detector 13 is a detector having both a recoil proton type and a nuclear reaction type configuration, it has sensitivity from a high energy region neutron of 100 keV to a low energy region neutron.

  The three detectors of the recoil proton detector 11, the nuclear reaction detector 12, and the mixed detector 13 need only be installed in a range where the energy distribution can be regarded as substantially the same. For example, a detector having a total of 6 mm square may be formed by arranging detectors of 5 mm square side by side or by arranging detectors of 2 mm × 6 mm × 6 mm horizontally. Moreover, you may arrange | position the reflective material etc. which the scintillator light which generate | occur | produces in each detector does not interfere between detectors.

The optical transmission unit 14 (14 ₁ , 14 ₂ , 14 ₃ ) transmits a signal (scintillator light) generated by each detector to the light detection unit 15 (15 ₁ , 15 ₂ , 15 ₃ ).

  The light transmission unit 14 is a transmission unit for transmitting scintillator light generated by the detector to the light detection unit 15 at the subsequent stage, and any device that can transmit light such as an optical fiber or a wavelength shift bar can be applied. In addition, since it becomes possible to maintain detection efficiency high by reducing the transmission error of the light in the coupling | bond part of each detector and the optical transmission part 14, it is desirable to couple and use it with optical grease.

2 to 4 shows the bonding structure of the anti跳陽Ko detector 11 and the optical transmission unit 14 _1. Incidentally, Figures 2-4, there shall as a representative example of the joining structure shown joining example anti跳陽Ko detector 11 and the optical transmission unit 14 _1, the nuclear reaction detector 12 and the optical transmission unit 14 _2, junction structure of a mixed type detector 13 and the optical transmission section 14 ₃ also becomes similar.

FIG. 2A shows an example of a recoil proton detector 11 junction when an optical fiber is used as the optical transmission unit 14, and FIG. 2B shows a cross-sectional view taken along the line II. Scribed portion of the anti跳陽Ko detector 11, optical fibers have been fitted, the scintillator light reaching the distal end portion of the optical fiber is transmitted to the light detector 15 _1.

3A shows an example of joining with the recoil-type proton detector 11 when a wavelength shift bar is used as the optical transmission unit 14, and FIG. 3B shows a cross-sectional view taken along the line II-II. . Wavelength shift bar so as to penetrate the inside of the anti跳陽Ko detector 11 and is inserted, scintillator light reaching the surface of the wavelength shift bars are transmitted to the light detector 15 ₁ is shifted wavelength.

Similarly, FIG. 4A shows an example of joining with the recoil-type proton detector 11 when a wavelength shift bar is used as the optical transmission unit 14, and FIG. 4B is a cross-sectional view taken along the line III-III. Show. Wavelength shift bar to the side of the anti跳陽Ko detector 11 has been arranged, scintillator light reaching the surface of the wavelength shift bars are transmitted to the light detector 15 ₁ is shifted wavelength.

  Compared with the case where an optical fiber is used as the optical transmission unit 14, when the wavelength shift bar is used, the area for capturing the scintillator light increases, so that high detection efficiency can be realized.

The light detection unit 15 (15 ₁ , 15 ₂ , 15 ₃ ) inputs the scintillator light generated by each detector via the optical transmission unit 14, and receives each scintillator light as an electric signal (s ₁ , s _2). , S ₃ ) and output to the event amount counter 16.

As the photodetection unit 15, optical signals such as photomultiplier tubes (PMTs), silicon PMT (Si-PN), photodiodes (PD), avalanche photodiodes (APD), CCD / CMOS, etc. are used as electrical signals. Any one that can be converted may be used. The converted electrical signals (s ₁ , s ₂ , s ₃ ) also include information such as the amount of scintillator light emitted. In addition, when the light detection part 15 can be installed in the vicinity of each detector without the light detection part 15 being damaged by the influence of radiation, the transmission distance between each detector and the light transmission part 14 can be shortened.

The event amount counting unit 16 receives electric signals (s ₁ , s ₂ , s ₃ ) from the light detection unit 15, and recoils proton type detector 11, nuclear reaction type detector 12, and mixed type detector 13. Each event amount generated by neutron detection is counted. Then, the counted event amount due to the neutron in each detector is output to the energy-specific dose calculation unit 17.

  Thereby, the event amount by the neutron in the high energy region, the event amount by the neutron in the low energy region, and the event amount by the neutron from the low region to the high region are counted. Further, the event amount counter 16 subtracts the event amount corresponding to the recoil proton detector 11 and the event amount corresponding to the nuclear reaction detector 12 from the neutron event amount corresponding to the mixed detector 13. Thus, the event amount in the intermediate energy region may be calculated.

  In addition, each of the recoil proton detector 11, the nuclear reaction detector 12, and the mixed detector 13 is sensitive to gamma rays. For this reason, when measuring in an environment where gamma rays are present together with neutrons, it is necessary to exclude events due to gamma rays.

  FIG. 5 shows an example of the configuration of the event amount counting unit 16 that discriminates events generated due to gamma rays.

  The event amount counting unit 16 includes a preamplifier 18, an A / D converter 19, a wave height discriminating unit 20, and a counting unit 21.

The electric signals (s ₁ , s ₂ , s ₃ ) input from the light detection unit 15 (FIG. 1) are amplified by the preamplifier 18, converted to a digital signal by the A / D converter 19, and the pulse height discrimination unit. 20 is output.

The wave height discriminating unit 20 discriminates events generated by gamma rays based on the wave height value of the electrical signal s ₂ corresponding to the nuclear reaction type detector 12. Specifically, the crest value of the signal with respect to the energy of the gamma ray is evaluated in advance, and the crest value of the obtained electric signal s ₂ is measured to discriminate the event due to the gamma ray.

For example, in the case of BNCT, the energy of gamma rays by neutron capture of dominant hydrogen is 2.2 MeV, whereas nuclear reaction (for example, nuclear reaction between ⁶ Li and neutrons) is performed by the nuclear reaction detector 12. The generated energy is 5.4 MeV, which is twice as large. That is, an event caused by gamma rays can be discriminated by evaluating in advance the peak value of the signal with respect to the energy of gamma rays.

  The counting unit 21 counts the event amount detected by each detector, and counts the event amount due to the discriminated gamma ray. Further, the counting unit 21 may subtract the event amount due to gamma rays from the event amount of each detector, and extract only the event due to neutrons in each detector.

  The energy-specific dose calculation unit 17 calculates the neutron dose for each energy based on the counted event amount in each detector. Further, when considering the event amount due to gamma rays, the dose of gamma rays is obtained based on the counted events due to gamma rays, and the neutron dose is calculated for each energy based on the event amount excluding the gamma ray dose.

  In this way, the amount of neutron events generated by the three types of neutron detectors, the recoil-proton detector 11, the nuclear reaction detector 12, and the mixed detector 13, which have different detectable neutron energy ranges, is calculated. By counting and calculating the neutron dose for each energy based on the event amount in each detector, the irradiation dose for each neutron energy can be measured in real time.

  In addition, the energy-specific dose calculation unit 17 sets the energy range of a desired neutron in advance, the neutron dose in this energy range, the neutron dose having energy higher than the maximum value of the energy range, and the minimum value of the energy range or less The neutron dose in three regions of the neutron dose having the energy of may be calculated.

  In the case of BNCT, the energy of neutrons used for treatment depends on the apparatus, but neutrons of several keV to several tens keV are generally used. FIG. 6A is an explanatory diagram showing an example of the energy spectrum of neutrons in BNCT, and the vertical axis represents the dose of neutrons. Thus, BNCT has a certain range of energy rather than monochromatic energy neutrons.

  In the case of neutrons with low energy, they are absorbed on the surface of the target without reaching the affected area, so that the contribution to the treatment is small and the influence on the increase in exposure dose is large. On the other hand, if the energy is high, neutrons are scattered by the hydrogen contained in the body, and the probability that it will be released outside the body without being absorbed by the affected area increases, so the contribution to treatment is small and the effect on the increase in exposure dose Becomes larger.

  When the energy of the irradiated neutron changes, the sensitivity of the detector to neutrons and the calculated calculation of the exposure dose change greatly, so that the evaluation of the irradiation dose to the patient greatly changes as shown in Table 2 below (see: Isotope notebook (11th edition)).

For this reason, from the viewpoint of reducing the exposure dose of the patient due to neutron irradiation, out of the neutron energy range included in the neutron source to the patient, the neutron energy range E reaching the affected depth and the maximum value of the energy range E It is necessary to accurately grasp each neutron dose equal to or higher than E _High and equal to or lower than the minimum value E _Low of the energy range E.

FIG. 6B is an explanatory diagram showing the detection sensitivity of each detector with respect to neutron energy. Neutrons higher than E _High are mainly detected by the recoil proton detector 11, while neutrons lower than E _Low are mainly detected by the nuclear reaction detector 12. Then, neutrons in the energy range E of neutrons reaching a certain depth of the affected part are mainly detected by the mixed detector 13.

Therefore, the energy-specific dose calculation unit 17 calculates the dose of neutrons equal to or higher than E _High using the event amount corresponding to the first signal, and corresponds the dose of neutrons equal to or lower than E _Low to the second signal. Calculate from the event amount.

  Then, the neutron dose in the energy range E is obtained by subtracting the event amount corresponding to the first signal and the second signal from the event amount corresponding to the third signal.

  A specific neutron dose calculation method will be described using the measurement flow of the neutron measurement apparatus 10 shown in FIG. 7 (see FIG. 1 as appropriate). Here, the effect of gamma rays is taken into consideration.

  The irradiated neutron beam is simultaneously measured by the recoil proton detector 11, the nuclear reaction detector 12, and the mixed detector 13 (S10, S11, S12).

  The light detection unit 15 inputs each of the scintillator lights generated by the three detectors, and converts each scintillator light into an electric signal (S13). The converted electrical signal is output to the event amount counter 16.

Event amount counter 16 evaluates the peak value of an electrical signal at a nuclear reaction detector 12, the event of gamma: counting the N γ _(S14). Then, the event amount counting unit 16 counts event amounts (N ₁ , N ₂ , N ₃ ) in each detector (S15).

Energy per dose calculating section 17, the event of counted gamma event amount counter 16: calculates the gamma dose X _gamma based on the N _{γ (S16).} The following formula (1) shows a calculation formula for calculating the gamma dose _Xγ . Α ₁ is a conversion coefficient for obtaining the dose from the event amount.

X _γ = α ₁ × N _γ Expression (1)

Energy per dose calculating section 17, to the exclusion of gamma dose X _gamma, it calculates a high-energy neutron dose _{X high} on the basis of the event amount _{N 1} in the anti跳陽Ko detector 11 (S17). The following formula (2) shows a calculation formula for calculating the dose X _high of high energy neutrons. Α ₂ means a conversion coefficient for obtaining the dose from the event amount.

X _high = α ₂ × N ₁ −X _γ (2)

Energy per dose calculating section 17, to the exclusion of gamma dose X _gamma, it calculates the low-energy neutron dose _{X low} based on the event the amount _{N 2} in a nuclear reaction detector 12 (S18). The following formula (3) shows a calculation formula for calculating the low energy neutron dose X _low . Β ₁ and β ₂ are conversion coefficients for obtaining a dose from the event amount.

X _low = β ₂ N ₂ −β ₁ X _γ Formula (3)

Then, the energy-specific dose calculation unit 17 subtracts the high energy neutron dose X _high , the low energy neutron dose X _low , and the gamma dose X _γ from the event amount N ₃ in the mixed detector 13. to calculate the neutron dose _{X E} in the range E (S19). Formula (4) shows a calculation formula for calculating the neutron dose X _E energy range E. C ₁ and C ₂ are conversion factors for obtaining a dose from the event amount.

X _E = C ₂ N ₃ -C ₁ (X _high + X _low + X _γ ) (4)

  In this way, the amount of neutron events generated by the three types of neutron detectors, the recoil-proton detector 11, the nuclear reaction detector 12, and the mixed detector 13, which have different detectable neutron energy ranges, is calculated. By counting and setting the desired energy range, the neutron dose in the energy range according to the purpose can be measured in real time by calculating the neutron dose for each energy based on the event amount in each detector. .

  In order to accurately obtain the event amount N in the energy range E, the calculated value on the right side of the equation (4) considering the error needs to be positive in consideration of the calculation error.

  Therefore, a statistic / measurement time necessary for evaluating N is determined using the energy spectrum of the neutron source evaluated in advance and a function of detection sensitivity of each detector.

Specifically, by setting the energy range E and integrating the product of the function of the energy spectrum of the neutron source and the detection sensitivity of each detector in each energy range, the amount of event can be estimated, and statistical error Can be calculated. Then, a sum of errors of each event amount is calculated, and a time necessary for accumulating event amounts necessary to minimize the sum of the calculation errors is set as a measurement time. Thus, it is possible to improve the accuracy of the neutron dose X _E energy range E.

(Second Embodiment)
FIG. 8 shows a configuration diagram of the neutron measurement apparatus 10 according to the second embodiment. In FIG. 8, parts having the same configuration or function as those in FIG. 1 are denoted by the same reference numerals, and redundant description is omitted.

  The neutron measurement apparatus 10 according to the second embodiment is different from the first embodiment in that the sensitivity of gamma rays is higher than the sensitivity to neutrons together with the recoil-type proton detector 11, the nuclear reaction detector 12, and the mixed detector 13. The gamma ray detector 22 is provided.

  The gamma ray detector 22 is disposed in the irradiation field of neutrons together with each detector, and measures the dose of gamma rays. As the gamma ray detector 22, a semiconductor detector such as Ge, CdTe, CdTe, CdZnTe, or TlBr can be used.

  The energy-specific dose calculation unit 17 inputs the gamma ray dose measured by the gamma ray detector 22 and calculates the neutron dose for each energy based on the event amount excluding the gamma dose.

By additionally arranging the gamma ray detector 22 having high detection sensitivity for gamma rays, the above-described gamma ray event amount N _γ can be measured with high accuracy, so that the error in the above equation (4) can be reduced. can, it is possible to calculate the neutron dose X _E energy range E with high accuracy.

(Third embodiment)
FIG. 9 shows a configuration example of a treatment apparatus 50 for boron neutron capture therapy to which the neutron measurement apparatus 10 according to the third embodiment is applied. 9, parts having the same configuration or function as those of the neutron measurement apparatus 10 (FIG. 1) according to the first embodiment are denoted by the same reference numerals, and redundant description is omitted. The optical transmission unit 14 and the light detection unit 15 are required for each detector, but are omitted in FIG.

  First, the configuration of a treatment apparatus 50 for boron neutron capture therapy that treats a patient who has received a drug containing boron by irradiating neutrons will be briefly described.

  The treatment device 50 for boron neutron capture therapy includes a proton beam accelerator 51, an irradiation unit 52, a neutron irradiation control device 53, and a treatment table 57.

  The proton beam accelerator 51 accelerates protons using a cyclotron or the like (not shown) to generate a proton beam (proton beam). The generated proton beam is guided to the irradiation unit 52.

  The irradiation unit 52 is a unit that irradiates the affected area with the generated neutrons, and includes a target 54, a moderator 55, and a collimator 56. The target 54 is irradiated with a proton beam guided from the proton beam accelerator 51 to generate neutrons. The moderator 55 decelerates the generated neutron energy to the energy required for treatment. The collimator 56 forms a flow path for guiding neutrons generated by the target 54 to the irradiation field.

  The neutron irradiation control device 53 controls the start / stop of neutron irradiation output from the irradiation unit 52 and the irradiation time and irradiation intensity via the proton beam accelerator 51. The neutron irradiation control device 53 stores three-dimensional information including position data of the patient region and boron concentration distribution data. This three-dimensional information is acquired by a medical imaging apparatus such as MRI, X-ray CT, and PET before the start of treatment.

  In the detection unit 23, the recoil proton detector 11, the nuclear reaction detector 12, and the mixed detector 13 are bound (or bonded), and the patient is in a position that does not affect the treatment. It is attached around the affected area. The detection unit 23 has a marker that serves as a mark when measuring the three-dimensional position from the outside. The detection unit 23 may contain three detectors in a container shape as long as the material does not affect neutron measurement.

  The position measurement unit 24 measures the three-dimensional position information of the detection unit 23 using the marker of the detection unit 23 as a mark. Any method for measuring the position information may be used as long as it can measure the position of the detection unit 23, such as a method using a laser distance meter or a stereo vision method.

  The positional information associating unit 25 associates and outputs the neutron energy-specific dose (dose information) calculated by the energy-specific dose calculating unit 17 and the measured positional information. Thereby, even if a patient's position fluctuates during treatment, the position shift and the accompanying neutron irradiation dose can be monitored online.

  The neutron irradiation control device 53 inputs neutron dose information associated with position information. Then, a simulation analysis is performed using the stored three-dimensional information in the affected area of the patient and neutron dose information to calculate the exposure dose of the patient. And the irradiation intensity | strength of neutron is controlled according to the calculated exposure amount.

  Thereby, even if the patient's position fluctuates during the treatment and changes from the irradiation dose of neutron evaluated before the start of the treatment, the fluctuation can be fed back to the neutron irradiation control device 53. Treatment can be performed with appropriate neutron irradiation intensity.

  FIG. 10A shows a schematic configuration diagram of a treatment apparatus 50 for boron neutron capture therapy to which a modification of the neutron measurement apparatus 10 according to the third embodiment is applied. FIG. 10B shows a state where a plurality of detection units 23 are attached to the affected area of the patient.

  In this modification, as shown in FIG. 10B, a plurality of detection units 23 are attached around the affected area of the patient.

  The position measurement unit 24 measures the three-dimensional position information of each detection unit 23 using the marker of each detection unit 23 as a mark. Then, the position information association unit 25 associates and outputs the acquired position information of the detection unit 23 and neutron dose information corresponding to the detection unit 23.

  The dose distribution evaluation unit 26 obtains a neutron dose distribution around the affected area based on the output doses of the respective detection units 23 for each energy. Thereby, even if a patient's position fluctuates during treatment, the dose distribution of neutrons around the affected area of the patient can be accurately obtained online. Further, the neutron dose distribution can be evaluated before treatment, and the neutron dose at an arbitrary position such as the center of the neutron irradiation position can be evaluated by comparing with the prior information.

  The neutron irradiation control device 53 inputs neutron dose information associated with each measured position information. Then, a simulation is performed using the stored three-dimensional information and neutron dose information to calculate the exposure dose of the patient. The neutron irradiation intensity is controlled according to the calculated exposure dose.

  As a result, even when the position of the patient fluctuates during the treatment and fluctuates from the neutron irradiation dose evaluated before the start of the treatment, the fluctuation can be more accurately fed back to the neutron irradiation control device 53. It becomes possible.

  According to the neutron measurement apparatus of each embodiment described above, three types of neutrons, that is, a recoil-type proton detector 11, a nuclear reaction detector 12, and a mixed detector 13, which have different detectable neutron energy ranges. By counting the neutron event amount generated by the detector and calculating the neutron dose for each energy based on the event amount in each detector, the irradiation dose for each neutron energy can be measured in real time.

  Although several embodiments of the present invention have been described, these embodiments are presented by way of example and are not intended to limit the scope of the invention. These novel embodiments can be implemented in various other forms, and various omissions, replacements, and changes can be made without departing from the scope of the invention. These embodiments and modifications thereof are included in the scope and gist of the invention, and are included in the invention described in the claims and the equivalents thereof.

  In addition, each component of the event amount counting unit, the energy-specific dose calculating unit, the position information associating unit, and the dose distribution evaluating unit can be realized by a processor of a computer that processes the received input signal. The code may be realized by executing it in an electronic circuit such as a processor, and is not limited to such software processing. For example, the code may be configured as a unit or computer realized by hardware processing using an electronic circuit such as an ASIC. Alternatively, it may be configured as a unit or a computer realized by combining software processing and hardware processing.

DESCRIPTION OF SYMBOLS 10 ... Neutron measuring apparatus, 11 ... Recoil proton type detector, 12 ... Nuclear reaction type detector, 13 ... Mixed type detector, 14 (14 ₁ , 14 ₂ , 14 ₃ ) ... Optical transmission part, 15 (15 ₁ , 15 ₂ , 15 ₃ ) ... light detection unit, 16 ... event amount counting unit, 17 ... dose calculation unit by energy, 18 ... preamplifier, 19 ... A / D converter, 20 ... wave height discrimination unit, 21 ... counting , 22 ... gamma ray detector, 23 ... detection unit, 24 ... position measurement unit, 25 ... position information association unit, 26 ... dose distribution evaluation unit, 50 ... treatment apparatus for boron neutron capture therapy, 51 ... proton beam accelerator, 52 ... Irradiation unit, 53 ... Neutron irradiation control device, 54 ... Target, 55 ... Moderator, 56 ... Collimator.

Claims (10)

A recoil-proton detector that has a substance that causes inelastic scattering and neutrons to be irradiated, and generates a first signal when the inelastic scattering by the neutron occurs;
A nuclear reaction detector having a substance that causes a nuclear reaction with the neutron and generating a second signal when the nuclear reaction by the neutron occurs;
A mixed detector that includes the neutron and a substance that causes the inelastic scattering or the nuclear reaction, and generates a third signal when the inelastic scattering or the nuclear reaction occurs;
An event amount counter for counting the amount of events generated by each detector based on the first signal, the second signal, and the third signal,
A neutron measurement apparatus comprising: a dose-by-energy calculation unit that calculates the dose of neutrons by energy based on the event amount in each detector counted. The recoil proton detector, the nuclear reaction detector, and the mixed detector are scintillators,
An optical transmission unit for transmitting scintillator light generated by each detector;
The neutron measurement apparatus according to claim 1, further comprising: a light detection unit that converts the inputted scintillator light into an electric signal and outputs the electric signal to the event amount counting unit. The event amount counting unit is a wave height discriminating unit that discriminates events generated by gamma rays based on the crest value of the electrical signal corresponding to the nuclear reaction type detector;
A counting unit that counts the amount of the event detected by each detector and counts the event due to the gamma ray that has been discriminated,
The dose calculation unit by energy calculates the dose of the gamma ray based on the counted event by the gamma ray, and calculates the dose of the neutron by energy based on the event amount excluding the dose of the gamma ray. The neutron measurement apparatus according to claim 2, wherein   The dose calculation unit for each energy sets a desired energy range of the neutron, and uses the dose of the neutron having energy larger than the maximum value of the energy range using an event amount corresponding to the first signal. Calculating a dose of the neutron having an energy smaller than a minimum value of the energy range using an event amount corresponding to the second signal, and calculating the dose of the neutron in the energy range by the third amount. 4. The neutron measurement apparatus according to claim 1, wherein the neutron measurement apparatus is obtained by subtracting an event amount corresponding to the first signal and the second signal from an event amount corresponding to the signal. 5. A gamma ray detector having a gamma ray sensitivity higher than the sensitivity to the neutron with each detector is further provided,
The dose calculation unit according to energy calculates the dose of the neutron according to energy based on the event amount excluding the dose of gamma rays detected by the gamma ray detector. The neutron measuring apparatus according to claim 1. A detection unit having a marker capable of measuring the position from the outside by binding the recoil-type detector, the nuclear reaction detector, and the mixed detector;
Using the marker as a mark, a position measurement unit that acquires position information on the detection unit in three dimensions;
The position information association part which correlates and outputs the acquired position information and the dose according to the energy of the neutron calculated by the energy dose calculation part. The neutron measurement apparatus according to any one of 5. A plurality of the detection units are arranged,
The position information associating unit associates and outputs each position information of the acquired detection unit and the dose by energy of the neutron calculated by the dose calculation unit by energy,
The neutron measurement apparatus according to claim 6, further comprising a dose distribution evaluation unit that evaluates a three-dimensional dose distribution of the neutron based on the output dose of each detection unit for each energy.   Calculate the sum of the error of the event amount by each detector using the energy spectrum of the neutron source evaluated in advance and the detection sensitivity of each detector, and the events necessary to minimize this error sum The neutron measurement apparatus according to any one of claims 1 to 7, wherein a time necessary for accumulating quantities is set as a measurement time. Receiving a first signal generated when the inelastic scattering due to the neutron occurs using a recoil proton type detector having a material that causes irradiation and inelastic scattering; and
Receiving a second signal generated when the nuclear reaction by the neutron occurs using a nuclear reaction detector having a substance that causes a nuclear reaction with the neutron; and
Receiving a third signal generated when the inelastic scattering or the nuclear reaction is generated using a mixed detector having the neutron and a substance that causes the inelastic scattering or the nuclear reaction;
Counting each event amount generated in each detector based on the first signal, the second signal, and the third signal;
Calculating a dose of the neutron for each energy based on the event amount counted in each detector. A treatment device for boron neutron capture therapy that treats a patient who has been administered a drug containing boron by irradiating neutrons,
The neutron measurement apparatus according to claim 6 or 7,
An irradiation unit for irradiating the affected area of the patient with the generated neutron;
A neutron irradiation control device for controlling the irradiation intensity of the neutron irradiated to the affected area of the patient,
The detection unit is attached around the affected area of the patient,
The neutron irradiation control device calculates an exposure dose of the patient based on the dose of the neutron calculated by the dose calculation unit according to energy, and controls the irradiation intensity of the neutron according to the exposure dose. A therapeutic device for boron neutron capture therapy.

Images