JP2018200235A - Neutron detector and neutron measurement device - Google Patents

Abstract

To provide a neutron detector and neutron measurement device capable of obtaining the neutron spectrum in an epithermal region with high accuracy.SOLUTION: A neutron detector 1 includes: a Li containing scintillator 5 receiving flying neutrons to generate light emission; a first shell 10 surrounding the Li containing scintillator 5; and a second shell 20 covering the outside of the first shell 10. The first shell 10 is constituted of a moderator for moderating the neutrons, and the second shell 20 is constituted of a material having a main peak of the cross-sectional area of resonance absorption reaction in a range of 1 eV or more and 10 keV or less. A neutron measurement device includes the neutron detector 1 and a photo-multiplier connected to the Li containing scintillator 5 via a light guide body.SELECTED DRAWING: Figure 1

Description

  The present invention relates to a Bonner sphere type neutron detector having a peculiar sensitivity to neutrons in the outside region, and a neutron measuring apparatus including the same.

One type of radiation therapy for treating cancer is boron neutron capture therapy (BNCT). Boron neutron capture therapy irradiates boron compounds selectively accumulated in cancer cells with neutrons and destroys cancer cells with alpha particles and lithium nuclei generated by the nuclear reaction of ¹⁰ B (n, α) ⁷ Li Is a cure. Since the range of alpha particles and lithium nuclei is about the same as that of cells, boron neutron capture therapy can selectively destroy only cancer cells without damaging normal cells. .

In boron neutron capture therapy, a boron compound containing boron 10 ( ¹⁰ B) is administered to a patient, and cancer cells in which the boron compound is accumulated are irradiated with neutrons for treatment. The lower the neutron energy, the greater the cross-sectional area of the reaction and the damage of normal cells can be avoided, while high energy is required to reach the depth of the patient's tissue. Therefore, the neutron beam used for treatment is required to have high epithermal neutron intensity, a low rate of fast neutron contamination, and a low ratio of thermal neutron flux to epithermal neutron flux.

  The neutron source of boron neutron capture therapy is shifting from a research reactor to an accelerator, and the high-speed neutron beam emitted from the neutron source is decelerated to a region outside the thermal region and used for treatment. In order to properly prepare a treatment plan for boron neutron capture therapy, it is important to accurately grasp the effects of neutron treatment and the patient's exposure dose. It is also desirable to know an accurate dose in real time during treatment. Under such circumstances, the neutron spectrum is measured to evaluate the intensity and dose of the neutron field, including the boron neutron capture therapy treatment system.

As a neutron detector for detecting neutrons, scintillation detection using helium 3 ( ³ He), a proportional counter using BF ₃ containing boron 10 ( ¹⁰ B), or a glass scintillator containing lithium 6 ( ⁶ Li) Typical examples are vessels. When thermal neutrons are the object of measurement, moderators are combined, and when fast neutrons are the object of measurement, proportional counters that detect recoil protons are also used. In addition, activation methods using metal foils and metal wires are also used for neutrons in general. The foil activation method and the Bonner sphere type neutron detector are used to measure the treatment system of boron neutron capture therapy.

  The foil activation method is performed by activating a foil-like metal with less self-absorption with neutrons and then measuring the emitted gamma rays. After activation, gamma rays are measured using a semiconductor detector, scintillator, etc., and the neutron flux at the time of activation is absolutely measured by calculating using decay constants. In a treatment system for boron neutron capture therapy that requires evaluation of the epithermal region, gold foil is often used as the metal foil. In the measurement of epithermal neutron flux, cadmium or the like is used in combination.

On the other hand, the Bonner sphere type neutron detector generally includes a proportional counter and a scintillator and a moderator covering the proportional counter. Neutron spectra in the range including the outer region are estimated by counting neutrons for each energy using Bonner spheres with different moderator thicknesses and different response functions, and analyzing the results by the unfolding method. . Non-Patent Document 1 discloses the result of analyzing the response function of a Bonner sphere equipped with a ³ He proportional counter by changing the thickness of the moderator.

Vladimir Mares et al. “Calculated neutron response of a Bonner sphere spectrometer with 3He counter”, Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment, 1911, 307 (2-3), p. 398-412

  Conventionally, the foil activation method and the Bonner sphere type neutron detector which are generally used can obtain the neutron spectrum of the neutron field including the treatment system of boron neutron capture therapy. Moreover, the neutron spectrum in the thermal region can be obtained with a certain degree of accuracy by using the threshold reaction in the foil activation method or by performing appropriate measurement by changing the thickness of the moderator of the Bonner sphere. .

  However, the measurement by the foil activation method has a problem that the neutron spectrum at the time of irradiation cannot be known in real time because the gamma ray is measured after the metal is activated. Moreover, in order to measure a neutron spectrum, the measurement using a different kind of metal foil is needed, and the gamma ray detector according to it is needed. From the standpoint of appropriately proceeding with the treatment plan for boron neutron capture therapy, a method capable of obtaining the neutron spectrum in the epithermal region with high accuracy and accuracy is desired. It is not in line with application to treatment systems.

  Further, as described in Non-Patent Document 1, a general Bonner sphere type neutron detector has a problem that the resolution of energy in the outside region is low. When obtaining a neutron spectrum with a general Bonner sphere type neutron detector, it is necessary to measure a large number of energy groups by changing the thickness of the moderator in various energy ranges with lower resolution. For this reason, there is a possibility that the reproducibility of the measurement is lowered in time or space, and there is a high possibility that the exposure will occur due to the installation or replacement of the Bonner sphere.

In addition, since a general Bonner sphere type neutron detector is often provided with a ³ He proportional counter, the outer diameter of the moderator increases according to the size of the proportional counter, and the overall diameter is several cm. To a few tens of centimeters. For this reason, there are problems that the Bonner sphere itself disturbs the neutron field and that it is difficult to secure a space for arranging the Bonner sphere, thereby impairing the accuracy of neutron spectrum measurement.

  Therefore, an object of the present invention is to provide a neutron detector and a neutron measuring apparatus that can obtain a neutron spectrum in the outside region with high accuracy.

  In order to solve the above problems, a neutron detector according to the present invention covers a Li-containing scintillator that generates light upon receiving flying neutrons, a first shell that surrounds the Li-containing scintillator, and an outer side of the first shell. A second shell, wherein the first shell is made of a moderator that decelerates neutrons, and the second shell is a material having a main peak of the cross-sectional area of the resonance absorption reaction in the range of 1 eV to 10 keV. It is structured.

  The neutron measurement apparatus according to the present invention includes a Li-containing scintillator that generates light upon receiving flying neutrons, a first shell that surrounds the Li-containing scintillator, and a second shell that covers the outside of the first shell, The first shell is composed of a moderator that decelerates neutrons, and the second shell is a neutron detector composed of a material having a main peak of the cross-sectional area of the resonance absorption reaction in the range of 1 eV to 10 keV; And a photomultiplier tube connected to the Li-containing scintillator via a light guide.

  According to the present invention, it is possible to provide a neutron detector and a neutron measuring apparatus that can obtain a neutron spectrum in the outside region with high accuracy. The neutron detector and the neutron measurement device can be provided in a small size, and can reduce the disturbance given to the neutron field and enable quick measurement.

It is sectional drawing which shows typically an example of the neutron detector which concerns on embodiment of this invention. It is sectional drawing which shows typically the other example of the neutron detector which concerns on embodiment of this invention. It is a figure which shows the response function of the neutron detector whose outer shell material is gold. It is a figure which shows the difference of the sensitivity by the outer shell of gold | metal | money. It is a figure which shows the response function of the neutron detector whose outer shell material is indium. It is a figure which shows the difference of the sensitivity by the outer shell of indium. It is a figure which shows the response function of the neutron detector whose outer shell material is silver. It is a figure which shows the difference of the sensitivity by the outer shell of silver. It is a figure which shows the response function of the neutron detector whose outer shell material is manganese. It is a figure which shows the difference of the sensitivity by the outer shell of manganese. It is a figure which shows the response function of the neutron detector whose outer shell material is tungsten. It is a figure which shows the difference of the sensitivity by the outer shell of tungsten. It is a figure which shows the response function of the neutron detector whose outer shell material is nickel. It is a figure which shows the difference of the sensitivity by the outer shell of nickel. It is a figure which shows the response function of the neutron detector whose material is cadmium. It is a figure which shows the response function of a neutron detector. It is a figure which shows the response function of a neutron detector. It is a figure which shows the neutron spectrum calculated | required using the unfolding method. It is a figure which shows the principal part of the neutron measuring apparatus which concerns on embodiment of this invention.

  Hereinafter, a neutron detector and a neutron detection apparatus including the neutron detector according to an embodiment of the present invention will be described in detail with reference to the drawings. In addition, about the structure which is common in each figure, the same code | symbol is attached | subjected and the overlapping description is abbreviate | omitted.

FIG. 1 is a cross-sectional view schematically showing an example of a neutron detector according to an embodiment of the present invention.
As shown in FIG. 1, the neutron detector 1 according to the present embodiment includes a Li-containing scintillator 5 that detects neutrons, a light guide 6 connected to the Li-containing scintillator 5, and an inner shell ( A first shell) 10 and an outer shell (second shell) 20 covering the outside of the inner shell (first shell) 10.

  The neutron detector 1 is a device that detects neutrons in a neutron field, and is a Bonner sphere type neutron detector that detects neutrons flying to the device after decelerating with a moderator. The neutron detector 1 has a peculiar sensitivity with respect to neutrons in the outer region, and can be used for measuring a neutron spectrum in the outer region due to a difference in response function. In this specification, neutrons whose kinetic energy is less than 0.5 eV are defined as thermal neutrons, neutrons between 0.5 eV and 10 keV are defined as epithermal neutrons, and neutrons exceeding 10 keV are defined as fast neutrons. Are referred to as a thermal region, a non-thermal region, and a high-speed region.

  As shown in FIG. 1, the neutron detector 1 has a spherical outer shape, and includes a solid-state Li-containing scintillator 5 that emits light upon receiving neutrons at the center of the sphere. The Li-containing scintillator 5 is supported in a state where it is adhered to the end surface of the core of the light guide 6 that propagates an optical signal generated by light emission. As the light guide 6, for example, an appropriate member capable of propagating an optical signal such as an optical fiber or a light guide is used. The material of the light guide 6 may be any of fluorine-based polymer, polycarbonate, polystyrene, quartz glass, and the like, for example.

The Li-containing scintillator 5 contains lithium 6 ( ⁶ Li) and causes a reaction of ⁶ Li (n, α) ³ H when neutrons are incident. The Q value of this nuclear reaction is about 4.78 MeV. The α particles and the tritium particles generated by this nuclear reaction excite electrons in the Li-containing scintillator 5, and light emission occurs with subsequent deexcitation.

As a material of the Li-containing scintillator 5, for example, lithium calcium aluminum fluoride (LiCaAlF ₆ ) can be used. Lithium calcium aluminum fluoride is suitable as a material for the Li-containing scintillator 5 for detecting neutrons because it exhibits a high light emission amount, a short light emission lifetime, and chemical stability that is difficult to deliquesce. In addition, compared with a ³ He proportional counter equipped with a general Bonner sphere type neutron detector, sufficient detection efficiency can be obtained even if it is very small, so that the outer diameter of the inner shell 10, outer shell 20, etc. Can be reduced.

The range of α particles and tritium particles generated by the nuclear reaction of lithium 6 ( ⁶ Li) is several tens of μm. On the other hand, the range of fast electrons generated when gamma rays mixed in the neutron field are incident on lithium calcium aluminum fluoride is about 1 to 2 mm. Therefore, by using lithium calcium aluminum fluoride that is sufficiently smaller than the range of high-speed electrons as the Li-containing scintillator 5, neutron rays can be distinguished from gamma rays and detected.

  The lithium calcium aluminum fluoride can be doped with one or more of europium, cerium, sodium and praseodymium. As a material of the Li-containing scintillator 5, lithium calcium aluminum fluoride (Eu: LiCaAlF) doped with europium is particularly preferable from the viewpoint of the amount of light emission and the like. In lithium calcium calcium aluminum, calcium may be substituted with one or more of magnesium, strontium, and barium, and aluminum may be substituted with one or more of gallium and scandium.

  The lithium calcium calcium fluoride is preferably a single crystal. Single crystal lithium calcium aluminum fluoride can be obtained by appropriate crystal growth methods such as Bridgman method, Czochralski method, Bernoulli method, hydrothermal method, flux method, chemical vapor deposition method and physical vapor deposition method. It can be manufactured.

  Lithium calcium aluminum fluoride preferably has a short-side length of 100 μm or more and 700 μm or less, and more preferably 200 μm or more and 500 μm or less. If the length in the short direction is in such a range, the detection rate can be increased while discriminating gamma rays mixed in the neutron field.

  In FIG. 1, the Li-containing scintillator 5 has a cubic shape and is bonded to the end face of the light guide 6. However, the shape of the Li-containing scintillator 5 and the Li-containing scintillator 5 and the light guide 6 are in contact with each other. The shape of the contact surface is not particularly limited. The shape of the Li-containing scintillator 5 can be an appropriate shape such as a flat plate shape, a prismatic shape, a cylindrical shape, or a spherical shape. Moreover, the shape of the contact surface where the Li-containing scintillator 5 and the light guide 6 are in contact with each other can be an appropriate shape such as a flat surface, a spherical surface, and a light condensing shape. The Li-containing scintillator 5 may be coated with a reflective material such as a tape made of polytetrafluoroethylene on the surface other than the emission surface from which light emitted by light emission is extracted.

In FIG. 1, the inner shell (first shell) 10 forms a spherical shell outside the Li-containing scintillator 5 and surrounds the entire Li-containing scintillator 5. The inner shell 10 is made of a moderator that decelerates neutrons. Specifically, the material of the inner shell 10 is preferably a material in which polyethylene, paraffin, or boron is mixed. A particularly preferable material for the inner shell 10 is high density polyethylene having a density of about 0.94 g / cm ³ or more.

  For example, the inner shell 10 may have a split structure, or may have a multilayer structure in which a plurality of moderators are arranged in the radial direction or a divided structure in which the moderators are arranged in the circumferential direction. The thickness of the inner shell 10 is determined according to the energy of the neutron to be detected. For example, the radial thickness of the inner shell 10 can be several hundred μm or more and 1 cm or less.

  In FIG. 1, the outer shell (second shell) 20 forms a spherical shell outside the inner shell (first shell) 10, and covers the outer surface of the inner shell (first shell) 10. The outer shell 20 is made of a material having a main peak of the microscopic cross section of the resonance absorption reaction in the range of 1 eV or more and 10 keV or less. The main peak of the microscopic cross section of the resonance absorption reaction means a peak having the maximum integrated area on the neutron energy-microscopic cross section diagram.

  Since the outer shell 20 undergoes a resonance absorption reaction with neutrons applied from the hot outer region to the high-speed region, neutrons flying to the apparatus are shielded from reaching the inner shell 10 and the Li-containing scintillator 5. Therefore, when a material having a maximum absorption cross-sectional area in different energy regions is used, a specific sensitivity can be obtained for each energy region corresponding to the material type. That is, a different response function can be obtained for each material type of the outer shell 20 between the neutron detectors.

  Specifically, the material of the outer shell 20 is preferably gold, indium, silver, manganese, tungsten, nickel, cadmium, or an alloy thereof. A particularly preferable material for the outer shell 20 is gold, indium, silver, manganese, tungsten, or nickel.

  For example, the outer shell 20 may be formed of a foil, may be divided into two parts, or may be formed of a divided structure in which a plurality of members are arranged in the circumferential direction. The thickness of the outer shell 20 is set to such a thickness that the epithermal neutrons in a predetermined energy range to be detected are sufficiently absorbed. For example, the thickness of the outer shell 20 in the radial direction can be several tens of μm or more and several thousand μm or less.

FIG. 2 is a cross-sectional view schematically showing another example of the neutron detector according to the embodiment of the present invention.
As shown in FIG. 2, the neutron detector 1 according to the present embodiment includes a Li-containing scintillator 5 that detects neutrons, a light guide 6 connected to the Li-containing scintillator 5, and an interior that surrounds the Li-containing scintillator 5. In addition to the shell (first shell) 10 and the outer shell (second shell) 20 that covers the outer side of the inner shell (first shell) 10, the thermal neutron shielding shell that covers the outer side of the outer shell (second shell) 20 The neutron detector 2 may further include a (third shell) 30.

  In FIG. 2, the thermal neutron shielding shell (third shell) 30 forms a spherical shell on the outer side of the outer shell (second shell) 20 and covers the outer surface of the outer shell (second shell) 20. ing. The thermal neutron shielding shell 30 is made of a material that absorbs thermal neutrons.

  Since the thermal neutron shielding shell 30 absorbs thermal neutrons mixed in the neutron field, thermal neutrons flying to the apparatus are shielded from reaching the inner shell 10, the outer shell 20, and the Li-containing scintillator 5. Therefore, by using the neutron detector 2 having the thermal neutron shielding shell 30 or using the neutron detector 2 having the thermal neutron shielding shell 30 and the neutron detector 1 not having the thermal neutron shielding shell 30 in combination. It becomes possible to measure the energy in the heat region and to reduce the influence of the background.

  Specifically, the material of the thermal neutron shielding shell 30 is cadmium, gadolinium, hafnium, tantalum, boron, an alloy or a compound thereof. A particularly preferred material for the thermal neutron shielding shell 30 is cadmium.

  For example, the thermal neutron shielding shell 30 may be formed of a foil, may be divided into two parts, or may be formed of a divided structure in which a plurality of members are arranged in the circumferential direction. The thickness of the thermal neutron shielding shell 30 is set such that thermal neutrons are sufficiently absorbed. For example, the radial thickness of the thermal neutron shielding shell 30 can be 200 μm or more and 2000 μm or less.

  Next, the response function of the neutron detector will be described by showing the results calculated using the Monte Carlo method.

FIGS. 3-14 has shown the calculation result of the response function supposing the neutron detector 2 shown in FIG. As a calculation system, a Li-containing scintillator 5, an inner shell 10, an outer shell 20, a thermal neutron shielding shell 30, and a virtual neutron detector 2 without a light guide 6, in a vacuum, We set up a system that irradiates neutrons from a disk-shaped neutron surface source that exceeds the diameter of the detector. As the Li-containing scintillator 5, Eu: LiCaAlF ₆ having a density of 2.988 g / cm ³ was set as a sphere having a radius of 200 μm. Further, as the inner shell 10, polyethylene having a density of 0.95 g / cm ³ was set with a radial thickness of 0.8 cm or 2.0 cm. Moreover, as the outer shell 20, various metals were set with a radial thickness of 0.1 cm or 0.5 cm. Further, as the thermal neutron shielding shell 30, cadmium was set with a radial thickness of 0.05 cm. PHITS was used as a calculation code, and the sensitivity [cm ² ] was obtained by dividing the neutron count rate by the unit neutron flux, and the neutron energy was divided into 30 on the logarithmic axis.

FIG. 3 is a diagram showing a response function of a neutron detector whose outer shell material is gold. FIG. 4 is a diagram showing a difference in sensitivity due to a gold outer shell.
FIG. 3 shows the response function of a neutron detector in which a 0.8 cm thick polyethylene inner shell is covered with a 0.1 cm thick gold outer shell and a 0.05 cm thick cadmium thermal neutron shielding shell. is there. In FIG. 4, the response function of the neutron detector whose outer shell material shown in FIG. 3 is gold, and the outer shell obtained by directly covering the 0.8 cm thick polyethylene inner shell with 0.05 cm thick cadmium. The difference from the response function of the neutron detector that is not provided is shown.

  Gold has a maximum peak of absorption cross section at about 4.89 eV. Therefore, as shown in FIG. 3, the sensitivity for detecting neutrons in the vicinity of about 4.89 eV is low. As shown in FIG. 4, when the difference between the response function of the neutron detector whose outer shell material is gold and the response function of the neutron detector without the outer shell are taken, the difference is remarkable in the vicinity of about 4.89 eV. There is a sensitivity difference. On the other hand, in the energy range higher than the vicinity of about 4.89 eV, a large sensitivity difference does not occur. Therefore, by using gold as the outer shell material and using a thermal neutron shielding shell as needed, only a response in the energy region of about 4.89 eV in the outer region can obtain a unique response function. I understand.

FIG. 5 is a diagram showing a response function of a neutron detector whose outer shell material is indium. FIG. 6 is a diagram showing the difference in sensitivity due to the outer shell of indium.
FIG. 5 shows a response function of a neutron detector in which an inner shell of polyethylene having a thickness of 0.8 cm is covered with an outer shell of indium having a thickness of 0.1 cm and a thermal neutron shielding shell having a thickness of 0.05 cm. is there. In FIG. 6, the response function of the neutron detector in which the outer shell material shown in FIG. 5 is indium, and the outer shell obtained by directly covering the 0.8 cm thick polyethylene inner shell with 0.05 cm thick cadmium. The difference from the response function of the neutron detector that is not provided is shown.

  Indium has a maximum absorption cross section peak at about 1.46 eV. Therefore, as shown in FIG. 5, the sensitivity for detecting neutrons in the vicinity of about 1.46 eV is low. As shown in FIG. 6, when the difference between the response function of the neutron detector whose outer shell material is indium and the response function of the neutron detector without the outer shell is taken, it is noticeable in the vicinity of about 1.46 eV. There is a sensitivity difference. On the other hand, in the energy range higher than the vicinity of about 1.46 eV, a large sensitivity difference does not occur. Therefore, by using indium as the outer shell material and using a thermal neutron shielding shell as required, a unique response function can be obtained only for the response in the energy region near about 1.46 eV in the outer region. I understand.

FIG. 7 is a diagram showing a response function of a neutron detector whose outer shell material is silver. FIG. 8 is a diagram showing the difference in sensitivity due to the silver outer shell.
FIG. 7 shows a response function of a neutron detector in which a 0.8 cm thick polyethylene inner shell is covered with a 0.1 cm thick silver outer shell and a 0.05 cm thick cadmium thermal neutron shielding shell. is there. In FIG. 8, the response function of the neutron detector in which the outer shell material shown in FIG. 7 is silver and the outer shell obtained by directly covering the 0.8 cm thick polyethylene inner shell with 0.05 cm thick cadmium. The difference from the response function of the neutron detector that is not provided is shown.

  Silver has a maximum peak of absorption cross section at about 5.2 eV. Moreover, a secondary peak exists in the vicinity of about 10 eV. Therefore, as shown in FIG. 7, the sensitivity for detecting neutrons in the vicinity of about 5.2 eV or about 10 eV is low. As shown in FIG. 8, when the difference between the response function of the neutron detector whose outer shell material is silver and the response function of the neutron detector without the outer shell are taken, the difference is noticeable in the vicinity of about 5.2 eV. There is a sensitivity difference. On the other hand, in the energy region higher than the vicinity of about 10 eV, a large sensitivity difference does not occur. Therefore, by using silver as the outer shell material and using a thermal neutron shielding shell as necessary, a response function in which the response in the energy region in the vicinity of about 5.2 eV or in the vicinity of about 10 eV is unique. It can be seen that can be obtained.

FIG. 9 is a diagram showing a response function of a neutron detector whose outer shell material is manganese. FIG. 10 is a diagram showing the difference in sensitivity due to the outer shell of manganese.
FIG. 9 shows a response function of a neutron detector in which a 2.0 cm thick polyethylene inner shell is covered with a 0.5 cm thick manganese outer shell and a 0.05 cm thick cadmium thermal neutron shielding shell. is there. In FIG. 10, the response function of the neutron detector whose outer shell material shown in FIG. 9 is manganese and the outer shell obtained by directly covering the inner shell of polyethylene having a thickness of 2.0 cm with cadmium having a thickness of 0.05 cm. The difference from the response function of the neutron detector that is not provided is shown.

  Manganese has a maximum absorption cross section peak at about 341 eV. A secondary peak is in the vicinity of about 1 keV to about 10 keV. Therefore, as shown in FIG. 9, the sensitivity of detecting neutrons in the vicinity of about 341 eV or in the vicinity of about 1 keV to about 10 keV is low. As shown in FIG. 10, when a difference is taken between the response function of the neutron detector whose outer shell material is manganese and the response function of the neutron detector without the outer shell, there is a significant sensitivity difference in the vicinity of about 341 eV. Has occurred. Therefore, by using manganese as the outer shell material and using a thermal neutron shielding shell as needed, the response in the energy region in the vicinity of about 341 eV or in the vicinity of about 1 keV to about 10 keV in the outer region is unique. You can see that you can get the function.

FIG. 11 is a diagram showing a response function of a neutron detector whose outer shell material is tungsten. FIG. 12 is a diagram showing the difference in sensitivity due to the outer shell of tungsten.
FIG. 11 shows the response function of a neutron detector in which a 2.0 cm thick polyethylene inner shell is covered with a 0.5 cm thick tungsten outer shell and a 0.05 cm thick cadmium thermal neutron shielding shell. is there. In FIG. 12, the response function of the neutron detector whose outer shell material shown in FIG. 11 is tungsten, and the outer shell obtained by directly covering the 2.0 cm thick polyethylene inner shell with 0.05 cm thick cadmium. The difference from the response function of the neutron detector that is not provided is shown.

  Tungsten has a maximum absorption cross section peak at about 18.8 eV. A secondary peak is in the vicinity of about 1 keV to about 10 keV. Therefore, as shown in FIG. 11, the sensitivity of detecting neutrons in the vicinity of about 18.8 eV or in the vicinity of about 1 keV to about 10 keV is low. As shown in FIG. 12, when the difference between the response function of the neutron detector whose outer shell material is tungsten and the response function of the neutron detector without the outer shell is taken, it is noticeable in the vicinity of about 18.8 eV. There is a sensitivity difference. Therefore, by using tungsten as the outer shell material and using a thermal neutron shielding shell as necessary, the response in the energy region of about 18.8 eV or about 1 keV to about 10 keV in the outer region is unique. It can be seen that a simple response function can be obtained.

FIG. 13 is a diagram showing a response function of a neutron detector whose outer shell material is nickel. FIG. 14 is a diagram showing the difference in sensitivity due to the nickel outer shell.
FIG. 13 shows a response function of a neutron detector in which a 2.0 cm thick polyethylene inner shell is covered with a 0.5 cm thick nickel outer shell and a 0.05 cm thick cadmium thermal neutron shielding shell. is there. In FIG. 14, the response function of the neutron detector in which the outer shell material shown in FIG. 13 is nickel, and the outer shell obtained by directly covering the inner shell of polyethylene having a thickness of 2.0 cm with cadmium having a thickness of 0.05 cm. The difference from the response function of the neutron detector that is not provided is shown.

  Nickel has a maximum absorption cross section peak at about 15 keV. Therefore, as shown in FIG. 13, the sensitivity for detecting neutrons in the vicinity of about 15 keV is low. As shown in FIG. 14, when the difference between the response function of the neutron detector whose outer shell material is nickel and the response function of the neutron detector without the outer shell, a significant sensitivity difference is about 15 keV. Has occurred. Therefore, it can be seen that by using nickel as the material of the outer shell and using a thermal neutron shielding shell as necessary, a response function with a unique response in the energy region of about 15 keV in the outer region can be obtained.

FIG. 15 is a diagram showing a response function of a neutron detector whose material is cadmium.
FIG. 15 shows the result of obtaining the response function of a neutron detector in which the Li-containing scintillator 5 is directly covered with cadmium having a thickness of 0.05 cm in the same manner as the calculation system shown in FIGS. It can be seen that cadmium is effective in reducing background effects because it has the ability to shield neutrons in the thermal region.

  Next, the result of having obtained the neutron spectrum from a predetermined neutron source using the unfolding method using a neutron detector is shown.

  As a calculation system, a total of 8 types of neutron detectors are arranged in an annular shape with a diameter of 15 cm, and a total of 8 types of neutron detectors are irradiated with a neutron beam from a disk-shaped neutron surface source with a diameter of 16 cm. It was set. The neutron beam irradiated from the neutron plane source was a spectrum simulating a neutron slow-down irradiation apparatus that emits neutron beams in the epithermal region used for boron neutron capture therapy. Since the calculation of the neutron spectrum by unfolding is an inverse problem of an unfavorable condition, a spectrum simulating this neutron moderator was used as the initial estimated spectrum (initial value). Further, a SANDII code was used as the unfolding code.

16 and 17 are diagrams showing a response function of the neutron detector.
In the calculation by the unfolding method, a total of eight types of neutron detectors having response functions shown in FIGS. 16 and 17 were set. That is, a single neutron detector of lithium calcium aluminum fluoride of a Li-containing scintillator, a neutron detector in which an inner shell of polyethylene having a thickness of 0.8 cm is directly covered with cadmium having a thickness of 0.05 cm, and having a thickness of 0.8 cm A neutron detector (see Fig. 3) with a polyethylene outer shell covered with a 0.1 cm thick gold outer shell and a 0.05 cm thick cadmium thermal neutron shielding shell. Neutron detector (see FIG. 5) with inner shell covered with 0.1 cm thick indium outer shell and 0.05 cm thick cadmium thermal neutron shielding shell, 2.0 cm thick polyethylene inner shell Neutron detector directly covered with 0.05 cm thick cadmium, 2.0 cm thick polyethylene inner shell, 0.5 cm thick manganese outer shell, 0.05 cm thick cadmium Neutron detector covered with thermal neutron shielding shell (see Fig. 9), 2.0cm thick polyethylene inner shell, 0.5cm thick tungsten outer shell and 0.05cm thick cadmium thermal neutron Neutron detector covered with shielding shell (see FIG. 11), 2.0 cm thick polyethylene inner shell, 0.5 cm thick nickel outer shell and 0.05 cm thick cadmium thermal neutron shielding shell The neutron detector covered with (see FIG. 13) was set.

FIG. 18 is a diagram showing a neutron spectrum obtained using the unfolding method.
In FIG. 18, the broken line is a spectrum that simulates the neutron slow-down irradiation apparatus set as the initial estimated spectrum, and the solid line is the neutron spectrum obtained by using the unfolding method. The number of times of calculation by unfolding is 20 times. As shown in FIG. 18, the difference in wave height distribution is sufficiently converged by 20 iterations, and the approximate shape of the neutron spectrum in the epithermal region is reproduced.

  As described above, according to the neutron detectors 1 and 2 described above, it is possible to obtain the neutron spectrum in the thermal region with high accuracy by using a combination of a plurality of materials such as the outer shell. Since the Li-containing scintillator 5 has a high light transmittance and a high light yield due to light emission, the accuracy of neutron detection can be increased. Further, in the Li-containing scintillator 5 such as lithium calcium aluminum fluoride, it is possible to discriminate gamma rays by the wave height. Moreover, even if the Li-containing scintillator 5 is made into a small piece, since a high light emission amount is obtained, the overall outer diameter can be reduced. Therefore, the disturbance that the neutron detector itself gives to the neutron field and the disturbance that the outer shell 20 gives to the neutron field can be reduced. In addition, it is possible to perform measurement in a narrow space or measurement using a plurality of neutron detectors with few obstacles, which is suitable for improving measurement accuracy and real-time measurement.

  Next, a neutron detector provided with a neutron detector will be described.

FIG. 19 is a diagram showing a main part of the neutron measurement apparatus according to the embodiment of the present invention.
As shown in FIG. 19, the neutron measurement apparatus 100 according to the present embodiment includes a plurality of neutron detectors B, a detection unit 120, a support tube 140, and a photomultiplier tube 160.

  The neutron measurement apparatus 100 is an apparatus for detecting a neutron in a neutron field and measuring a neutron spectrum. The neutron measuring apparatus 100 is used in a state where a plurality of neutron detectors B are attached and the photomultiplier tube 160 is connected to a signal processing unit (not shown) via a signal line. At least one or more, preferably two or more neutron detectors B to be attached have the same configuration as one of the neutron detector 1 and the neutron detector 2 described above.

  The detection unit 120 is a part that detects flying neutrons without decelerating with a moderator. The detection unit 120 can be formed by, for example, the Li-containing scintillator 5 described above, and the Li-containing scintillator 5 adhered to the end surface of the light guide 6 can be covered with a reflective material and installed. By providing the detection unit 120, it is possible to detect thermal neutrons or the like that have reached thermal equilibrium, but the detection unit 120 may be omitted.

  The support tube 140 is a member that supports the neutron detector B in the neutron field, and is provided for each neutron detector B. The support tube 140 connects between each of the neutron detectors B and the photomultiplier tube 160 provided for each neutron detector B. The support tube 140 is provided in a curved shape along the longitudinal direction so that the neutron detectors B connected to the photomultiplier tube 160 are arranged apart from each other in the neutron field. Such a shape can prevent the neutron detectors B from being disturbed.

  The support tube 140 has a structure in which a light guide 6 that propagates an optical signal by light emission is passed through a tube made of aluminum alloy or the like. As shown in FIGS. 1 and 2, one end of the light guide 6 is bonded to the Li-containing scintillator 5, passes through the inner shell 10 and the like, and is drawn out of the neutron detector B. The other end of the light guide 6 is connected to an incident window of a photomultiplier tube 160 connected to the support tube 140, and the Li-containing scintillator 5 and the photomultiplier tube 160 are connected by an optical waveguide.

  The photomultiplier tube 160 is a detector that photoelectrically converts light energy into an amplified current. The photomultiplier tube 160 is provided for each neutron detector B, and is individually connected to the Li-containing scintillator 5 of each neutron detector B via a light guide. For example, the photomultiplier tube 160 takes out a photocathode that performs photoelectric conversion, a focusing electrode that focuses photoelectrons, a plurality of dynodes that amplify electrons, and amplified secondary electrons inside a vacuum tube having an incident window. The high-voltage power supply and various circuits are connected and used. The light emission pulse generated by the Li-containing scintillator 5 is converted into an electric signal by the photomultiplier tube 160 and output to a signal processing unit (not shown) connected via a signal line.

  The signal processing unit is provided for performing signal processing on the electrical signal output from the photomultiplier tube 160 and estimating the neutron spectrum based on the measurement value measured by the neutron detector B. The signal processing unit includes, for example, a signal processing circuit such as a preamplifier, a shaping amplifier, and a multi-wave height analyzer, and a computer that estimates a neutron spectrum. However, the signal processing unit may be configured to perform any of counting processing, analog signal processing, digital signal processing, and the like according to the neutron field to be measured, and is configured using various amplifiers, filters, converters, etc. May be.

The signal processing unit estimates the neutron spectrum by an unfolding method based on the following determinant.
φ = R ^-1・ C
In the equation, φ is a group of neutron flux, R is a response function, and C is a group of outputs from the neutron detector. That is, a response function representing a relationship between a group of known neutron flux wave height distributions and a group of outputs from the neutron detector B for each energy is obtained in advance, and the output from the neutron detector B is converted by an inverse matrix of the response function. The neutron spectrum is derived by finding an appropriate solution by the operation. The unfolding process can be performed by an appropriate method such as a SANDII code, a Monte Carlo method, or a J1 method.

  The neutron measuring apparatus 100 described above has a plurality of neutron detectors B attached thereto, and a photomultiplier tube 160 is connected to a signal processing unit (not shown) via a signal line in the neutron field to be measured. Installed. The number of series of the neutron detector B, the support tube 140 and the photomultiplier tube 160 to be attached is not limited to three as shown in FIG. 19, but the neutron detector B, the support tube 140 and the photomultiplier tube 160. The number of sequences can be any number.

  As the neutron detector B attached to the neutron measuring apparatus 100, a neutron detector including the inner shell 10 composed of the Li-containing scintillator 5 and the moderator and not including the outer shell 20, the neutron detector 1 and the neutron detector described above. It is also possible to use together with any of the vessels 2. However, from the viewpoint of obtaining a neutron spectrum in a predetermined energy region in the outer region with high accuracy, in order to obtain a peculiar response in the energy region, two or more, more preferably three or more, and still more preferably 4 It is preferable that the number is the same as that of any one of the neutron detector 1 and the neutron detector 2 described above.

  Each of the plurality of neutron detectors B included in the neutron measurement apparatus 100 is used in a combination in which the material of the outer shell (second shell) 20 has a main peak of the cross-sectional area of the neutron resonance absorption reaction in a different range. It is preferred that When the neutron detectors B having different materials for the outer shell 20 are used in combination, it is possible to obtain a unique response for each of a plurality of energy regions that cause a resonance absorption reaction even in a single measurement. Therefore, a neutron spectrum in a plurality of energy regions in the outer region can be obtained with high accuracy by one measurement or real-time measurement.

  Each of the plurality of neutron detectors B included in the neutron measurement apparatus 100 is configured to include not only the material of the outer shell 20 but also the neutron detectors 1 and 2 having different thicknesses of the inner shell 10 and the outer shell 20. May be. When the neutron detectors B having different thicknesses of the inner shell 10 and the outer shell 20 are used in combination, it is possible to obtain different responses including those other than the energy region in which the resonance absorption reaction occurs. Therefore, a neutron spectrum in a wider energy range can be obtained with high accuracy.

  Each of the plurality of neutron detectors B included in the neutron measurement apparatus 100 is preferably made of two or more, more preferably three of the materials of the outer shell 20 among gold, indium, silver, manganese, tungsten, and nickel. As described above, it is more preferable that the combination is more than four kinds. When the material of the outer shell 20 is used in combination, it is possible to obtain a unique response for each energy region that causes a resonance absorption reaction even in a single measurement. Therefore, a neutron spectrum in a wide energy range in the outside region can be obtained with high accuracy by one measurement or real-time measurement.

  As mentioned above, although embodiment of this invention was described, this invention is not limited to the said form, A various change is possible in the range which does not deviate from the meaning of this invention. For example, the present invention is not necessarily limited to the one having all the configurations included in the embodiment. A part of the configuration of the embodiment may be replaced with another configuration, or a part of the configuration of the embodiment may be omitted.

  For example, the neutron detectors 1 and 2 are Bonner sphere type neutron detectors. However, as long as the neutron detector includes the first shell 10 that surrounds the Li-containing scintillator and the second shell 20 that covers the outside of the first shell 10, the neutron detector may have other forms and shapes such as a neutron position detector. Good.

  The neutron detectors 1 and 2 include a Li-containing scintillator 5 bonded to the light guide 6. However, the Li-containing scintillator 5 may be connected to a daylighting function imparting optical fiber, a light guide, or the like at an appropriate contact surface. As the optical fiber, an appropriate one such as a wavelength shift fiber may be used.

  The neutron measurement apparatus 100 includes a photomultiplier tube 160, and the Li-containing scintillator 5 is connected to the photomultiplier tube 160. However, the neutron measurement apparatus 100 may include a photodiode, a photoconductive element, or the like instead of the photomultiplier tube 160. The neutron measurement apparatus 100 may support the neutron detectors 1 and 2 with a support material having an appropriate shape and structure.

1, 2 Neutron detector 5 Li-containing scintillator 6 Light guide 10 Inner shell (first shell)
20 Outer shell (second shell)
30 Thermal neutron shielding shell (third shell)
100 Neutron Measuring Device 120 Detector 140 Support Tube 160 Photomultiplier Tube B Neutron Detector

Claims (5)

A Li-containing scintillator that generates light upon receiving flying neutrons;
A first shell surrounding the Li-containing scintillator;
A second shell covering the outside of the first shell,
The first shell is composed of a moderator that decelerates neutrons,
The second shell is a neutron detector made of a material having a main peak of a cross-sectional area of a resonance absorption reaction in a range of 1 eV to 10 keV.   The neutron detector according to claim 1, wherein the material of the second shell is gold, indium, silver, manganese, tungsten, or nickel. A third shell covering the outside of the second shell;
The neutron detector according to claim 1, wherein the third shell is made of a material that absorbs thermal neutrons. A Li-containing scintillator that generates light upon receiving flying neutrons, a first shell that surrounds the Li-containing scintillator, and a second shell that covers the outside of the first shell, the first shell being a deceleration that decelerates neutrons A neutron detector made of a material having a main peak of a cross-sectional area of a resonance absorption reaction in a range of 1 eV or more and 10 keV or less;
A neutron measurement apparatus comprising: a photomultiplier tube connected to the Li-containing scintillator via a light guide. Comprising a plurality of said neutron detectors,
5. The neutron measurement apparatus according to claim 4, wherein each of the plurality of neutron detectors has a main peak of a cross-sectional area of a resonance absorption reaction of neutrons in a range of energy different from each other in the material of the second shell.

Images