JP2017040653A - Manufacturing method of radioactive material and manufacturing device of radioactive material - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a manufacturing method of a radioactive material capable of stably supplyingCu orY with high purity by using natural zinc or zirconium without needs for isotope concentration.SOLUTION: A manufacturing method of a radioactive material including a step of manufacturingCu orY by irradiating deuteron 11 to a neutron converter 12, irradiating neutron 13 emitted from the neutron converter 12 to a member 14 containing natural zinc without conducting isotope concentration or a member 14 containing zirconium, and performing nuclear reaction of the neutron 13 and these members 14.SELECTED DRAWING: Figure 1

Description

  The present invention relates to a method for producing a radioactive substance and a production apparatus for the radioactive substance.

  In recent years, the application of neutron sources using accelerators has expanded. Among these, research related to a method for manufacturing medical RI (Radio Isotopes) (see Non-Patent Document 1) has attracted attention. With this method, various nuclides can be produced.

As nuclides often used for medical RI, positron emitting nucleus ¹⁸ F (fluorine 18) with a half-life of 2 hours, beta-emitting nucleus ⁹⁰ Y (yttrium 90) with a half-life of 2.8 days, beta with a half-life of 50 days The radiation emitting nucleus ⁸⁹ Sr (strontium 89), the gamma-ray emitting nucleus ^99m Tc (technetium 99m) having a half-life of 6 hours, and the like are well known. In so-called “PET examinations”, positron emitting nuclei are often used, and since this is a cancer diagnostic method with a very high clinical record, the number of facilities that can be examined and the number of examinations are increasing year by year. Further, in recent years, the application of gamma-ray emitting nuclei used in the “SPECT examination” has been expanded not only for examination but also for the purpose of investigating the pharmacokinetics of drugs. Furthermore, beta-emitting nuclei are used as molecularly targeted drugs for cancer pain alleviation and treatment, and their clinical application is progressing.

  Under such circumstances, PET has recently been applied to the field of “molecular imaging”. In particular, application to microdose clinical trials in drug development is expected. This is a technique in which a molecule of a drug candidate is labeled with a positron emitting nucleus, administered in a very small amount that does not exhibit a medicinal effect, and the metabolism of the molecule in vivo is examined.

However, ¹⁸ F, which is readily available at present, has a problem that it has a short half-life and can only partially observe metabolism. Therefore, in recent years, a positron emitting nucleus having a half-life of 12.7 hours, ⁶⁴ Cu, has attracted attention. Since it is an element that has a sufficient half-life for observing metabolism and is well-known in chemical properties, it is easy to handle. However, there is currently no facility that can stably produce ⁶⁴ Cu.

Currently, as a technique for producing ⁶⁴ Cu, as shown in FIG. 16, a technique of producing a highly concentrated ⁶⁴ Ni 22 by using a nuclear reaction caused by irradiating the proton beam 21 from the proton accelerator 20 is mainly used. It is. However, this method has two problems. One is that ⁶⁴ Ni, which is present only 1% in natural nickel, must be isotopically enriched to a high concentration. This is to increase the amount of ⁶⁴ Cu produced and suppress nuclides other than ⁶⁴ Cu produced as a by-product. Isotope enrichment requires high technology and cost, and is a high hurdle to stable supply. Another is the limitation of the amount of raw materials. Since all proton beams stop near the surface of the raw material, a large amount of raw material cannot be irradiated. This is also an obstacle to mass production.

On the other hand, a therapeutic agent labeled with ⁹⁰ Y that emits beta rays and labeled with ibritumomab tiuxetan is collected in the tumor site and can be irradiated with beta rays. However, depending on the patient, it may be difficult for the drug to accumulate in the tumor site, or accumulation in other normal sites may be seen, so “eligibility for treatment” must be examined in advance. However, since ⁹⁰ Y emits only beta rays, its dynamics cannot be observed from outside the body to examine “eligibility for treatment”.

Therefore, at present, a drug labeled with ¹¹¹ In with ibritumomab tiuxetane is administered to a patient in advance, and its eligibility is examined. This nuclide emits 171 keV gamma rays. Although there is no therapeutic effect by this gamma ray, the drug distribution can be examined from outside the body by SPECT examination using the high permeability. However, the pharmacokinetics obtained when ⁹⁰ Y is used and the pharmacokinetics obtained when ¹¹¹ In is used are not necessarily the same. The discrepancy between these two pharmacokinetics can lead to misjudgment of treatment eligibility.

^90m Y and ^91m Y have been proposed as radioactive yttrium that emits gamma rays that can investigate the eligibility of treatment with the same pharmacokinetics as therapeutic drugs (Patent Document 1). However, these radioactive substances remain in the body as ⁹⁰ Y (half-life 2.8 days) and ⁹¹ Y (half-life 58 days) after disintegration if they are not eligible for treatment. Have.

Japanese Patent No. 5522562

Y. Nagai et al, J. MoI. Phys. Soc. Jpn. 82 (2013) 064201 K. Shibata, O .; Iwamoto, T .; Nakagawa, N .; Iwamoto, A .; Ichihara, S .; Kunieda, S .; Chiba, K .; Furutaka, N .; Otuka, T .; Ohsawa, T .; Murata, H .; Matsunobu, A.M. Zukeran, S .; Kamada, and J.A. Katakura: “JENDL-4.0: A New Library for Nuclear Science and Engineering,” J. Nucl. Sci. Technol. 48 (1), 1-30 (2011) Hyunsuku Kim, Kwangsue Park, Byungdon Min, et al. , Nucl. Inst. Meth B, 217 (2004) 429-434 T.A. Sato, K .; Niita, N.A. Matsuda, S .; Hashimoto, Y. et al. Iwamoto, S .; Noda, T .; Ogawa, H .; Iwase, H .; Nakashima, T .; Fukahori, K .; Okumura, T .; Kai, S .; Chiba, T .; Furuta and L.M. Sihver, Particle and Heavy Ion Transport Code System PHITS, Version 2.52, J. MoI. Nucl. Sci. Technol. 50: 9, 913-923 (2013) Viki-Veiko Elomaa, Jori Jurttila, Johan Rajander, et al. , Applied Radiation and Isotopes, 89 (2014) 74-78. K. Abbas, J. et al. Kozempel, M.M. Bonardi, App. Rad. Iso. , 64 (2006) 1001-1005 K. Abbas, C.I. Biratari, M .; Bonardi, et al. , J .; Labeled Cpd. Radiopham, 44, Suppl. 1 (2001) Tadhiro KIN, Yasuki NAGAI, Nobuyuki IWAMOTO, J. et al. Phys. Soc. Jpn. , 82 (2013) 0342201 T.A. Kin, et al, J.A. Phys. Soc. Jpn. 82 (2013) 0342201

  This invention is made | formed in view of this situation, and it aims at providing the manufacturing method and radioactive substance manufacturing apparatus of a radioactive substance used as medical RI.

Specifically, the present invention provides a radioactive substance manufacturing method and a radioactive substance manufacturing apparatus that can stably supply high-purity ⁶⁴ Cu using natural zinc that does not require isotope enrichment. Providing is the first purpose.

In addition, the present invention provides a radioactive substance production method and a radioactive substance production apparatus capable of stably supplying high purity ⁹² Y using natural or concentrated zirconium. The second purpose.

In order to solve the above-mentioned problems, the present inventor has intensively studied and developed a method for producing ⁶⁴ Cu and ⁹² Y using a deuteron accelerator neutron source.
The present invention provides the following means.
[1] irradiating a neutron converter with deuteron, irradiating the neutron released from the neutron converter to a member containing natural zinc that has not been isotopically enriched, and causing the neutron and the zinc to undergo a nuclear reaction A method for producing a radioactive substance, characterized in that ⁶⁴ Cu is produced by
[2] The method for producing a radioactive substance according to [1], wherein the energy of the deuteron is 9 MeV or more and 20 MeV or less.
[3] Producing ⁹² Y by irradiating a neutron converter with deuteron, irradiating the neutron released from the neutron converter onto a member containing zirconium, and causing the neutron and the zirconium to undergo a nuclear reaction. A method for producing a radioactive material.
[4] The method for producing a radioactive substance as described in [3], wherein the energy of the deuteron is 15 MeV or more and 30 MeV or less.
[5] Any one of [1] to [4], wherein the neutron converter uses a deuteron that generates a (d, n) reaction that emits one neutron. The manufacturing method of radioactive substance of description.
[6] It includes at least an accelerator that accelerates deuterons, a neutron converter that emits neutrons when irradiated with the deuterons, and a member containing natural zinc that is not subjected to isotope enrichment. A device for producing a radioactive material.
[7] The radioactive substance manufacturing apparatus according to [6], wherein the accelerator includes means for applying energy of 9 MeV to 20 MeV to the deuteron.
[8] An apparatus for producing a radioactive substance, comprising at least an accelerator that accelerates deuterons, a neutron converter that emits neutrons when irradiated with deuterons, and a member containing zirconium.
[9] The radioactive substance manufacturing apparatus according to [8], wherein the accelerator has means for applying energy of 15 MeV to 30 MeV to the deuteron.
[10] Any one of [6] to [9], wherein the neutron converter is a member made of an element that generates a (d, n) reaction that receives one deuteron and emits one neutron. The manufacturing apparatus of the radioactive substance as described in one.

ADVANTAGE OF THE INVENTION According to this invention, the manufacturing method and manufacturing apparatus of a radioactive substance which can supply ⁶⁴ Cu with high purity stably using natural zinc which does not require isotope enrichment can be provided. In addition, according to the present invention, since almost no by-product that is a problem in application, which is a concern due to the use of natural zinc, is generated, it is possible to stably supply high-purity ⁶⁴ Cu. A radioactive substance production method and a radioactive substance production apparatus can be provided.

Moreover, according to this invention, the manufacturing method and manufacturing apparatus of a radioactive substance which can supply ⁹² Y with high purity stably can be provided. ^{Since 92} Y emits 934 keV gamma rays, it can be used for treatment eligibility testing by using a SPECT test, a Compton camera, etc., and further, beta decay with a half-life of 3.5 hours, non-radioactive ^{Since it is 92} Zr, unnecessary exposure of the patient can be suppressed.

It is the figure which showed typically the structure of the manufacturing apparatus of the radioactive substance of this invention. It is a graph showing the kind of element obtained when ⁶⁴ Zn is used. It is a graph showing the kind of element obtained when ⁶⁶ Zn is used. It is a graph showing the kind of element obtained when ⁶⁸ Zn is used. (A) It is a figure which shows the simulation system of a ^nat C (d, n) neutron source. (B) It is a figure which shows the simulation system of a DT neutron source. (A)-(d) It is a graph which shows ⁶⁴ Cu production | generation distribution. (A)-(d) It is a graph which shows ⁶³ Cu production | generation distribution. (A)-(d) It is a graph which shows ⁶⁵ Cu production | generation distribution. (A)-(d) It is a graph which shows ⁶⁴ Cu isotope purity distribution. Gamma spectrum from irradiated ^nat Zn with neutrons from C (d, n) and DT reactions. It is a graph showing the kind of element obtained when ⁹⁰ Zr is used. It is a graph showing the kind of element obtained when ⁹¹ Zr is used. It is a graph showing the kind of element obtained when ⁹² Zr is used. It is a graph showing the kind of element obtained when ⁹⁴ Zr is used. It is a graph which shows the gamma ray measurement result by Y produced | generated using natural Zr. It is the figure which showed typically the structure of the manufacturing apparatus of the radioactive substance used conventionally.

  Hereinafter, a radioactive substance manufacturing method and a manufacturing apparatus according to an embodiment to which the present invention is applied will be described in detail.

  FIG. 1 schematically shows a configuration of a radioactive substance manufacturing apparatus according to the present invention. This manufacturing apparatus includes at least an accelerator that accelerates deuterons, a neutron converter that emits neutrons when irradiated with deuterons, and a raw material of a radioactive substance to be obtained.

<First Embodiment>
The manufacturing method of the radioactive substance which concerns on this embodiment using the manufacturing apparatus shown in FIG. 1 is demonstrated. In the present embodiment, a member (raw material) containing natural zinc that is not subjected to isotope enrichment is used as a radioactive material. The natural zinc here contains a plurality of types of zinc isotopes, and at least ⁶⁴ Zn.

  First, the deuteron accelerator 10 accelerates the deuteron 11 to collide with the neutron converter 12 to generate the neutron 13. As the neutron converter 12, it is preferable to use one (e.g., carbon, beryllium) that efficiently generates a (d, n) reaction that receives one deuteron 11 and emits one neutron 13. Below, this (d, n) reaction may be called C (d, n) reaction.

Subsequently, the generated neutron 13 is irradiated to natural zinc (raw material) 14, and ⁶⁴ Cu (copper 64) produced is separated and extracted as medical RI 15 for use in PET inspection. Since neutron reaction is used, ⁶⁴ Cu is generated in the whole raw material, and it is possible to irradiate a large amount of raw material. It can be said that the greatest feature of the accelerator neutron source in this embodiment is that neutron energy can be optimized. By adjusting the deuteron energy to 9 MeV or more and 20 MeV or less, more preferably 12 MeV or more and 15 MeV or less, it is possible to obtain neutrons that are extremely advantageous for the production of ⁶⁴ Cu, and even if natural zinc is used as a raw material, the isotopic concentration is 90% or more. Of ⁶⁴ Cu can be obtained.

⁶⁴ Cu is produced in natural zinc, but chemical separation is possible because the target element to be produced and the element as the raw material are different. In addition, since a large amount of heat is not generated in the neutron reaction, the raw material can be in a chemical form (for example, copper oxide powder) that facilitates chemical separation.
In recent years, an increasing number of medical institutions have introduced boron neutron capture therapy. However, since the accelerator used for this can be used as the neutron source described above, there is a possibility that a hybrid accelerator for the purpose of producing ⁶⁴ Cu may be realized.

In radiopharmaceutical production, if the target element is different from the target nuclide ( ⁶⁴ Cu in the present invention), it can be extracted by chemical separation. However, when the elements are the same (in the present invention, non-radioactive ⁶³ Cu and ⁶⁵ Cu and radioactive ⁶⁶ Cu and ⁶⁸ Cu) are chemically inseparable, these “carrier-free” nuclides are not mixed. Will be needed.

2, 3, and 4 are graphs showing the types of elements obtained when ⁶⁴ Zn, ⁶⁶ Zn, and ⁶⁸ Zn are used as natural zinc, respectively. Both quote nuclear data “JENDL-4.0” (see Non-Patent Document 2) published by the Japan Atomic Energy Agency.

In the present invention, since there is no neutron having an energy of 12 to 15 MeV or more, ⁶⁷ Cu (a radioactive carrier) which is the most problematic is hardly generated. Other radioactive ⁶⁶ Cu and ⁶⁸ Cu have extremely short half-lives of 5 minutes and 31 seconds, respectively, and disappear from the irradiated sample after a cooling time of 1 hour (both become non-radioactive zinc). . The production of some ⁶³ Cu cannot be avoided, but the influence can be suppressed to an extremely low level of several percent when using 12 MeV deuterons and less than 10% even when using 15 MeV deuterons.

Second Embodiment
The manufacturing method of the radioactive substance which concerns on this embodiment using the manufacturing apparatus shown in FIG. 1 is demonstrated. In the present embodiment, zirconium 92 that has been subjected to isotope enrichment or a member that includes natural zirconium that has not been subjected to isotope enrichment is used as a radioactive material.

  First, the deuteron accelerator 10 accelerates the deuteron 11 to collide with the neutron converter 12 to generate the neutron 13. As the neutron converter 12, it is preferable to use one (e.g., carbon, beryllium) that efficiently generates a (d, n) reaction that receives one deuteron 11 and emits one neutron 13.

Subsequently, the generated neutron 13 is irradiated to the raw material zirconium 14 and the produced ⁹² Y (yttrium 92) is separated and extracted as medical RI 15 and used for SPECT inspection and treatment eligibility inspection using a Compton camera. . Since neutron reaction is used, ⁹² Y is generated as a whole of the raw material, and a large amount of raw material can be irradiated. It can be said that the greatest feature of the accelerator neutron source in this embodiment is that neutron energy can be optimized. The deuteron energy, or 15 MeV 30 MeV or less, preferably by adjusting to approximately 20 ^{MeV, can} be ⁹² Y produced can be obtained a very advantageous neutron obtain isotope concentration of 90% or more of the ⁹² Y.

When natural zirconium is used as a raw material, the amount of ⁸⁹ Y produced as a carrier can be kept low by setting the deuteron energy to about 15 MeV. However, in this case, ⁹⁰ Y and ⁹¹ Y are generated as carriers.

On the other hand, when concentrated zirconium 92 is used as a raw material, ⁹² Y can be obtained almost without a carrier. In this case, the production amount can be increased by increasing the deuteron energy to about 20 MeV. ^{Although 91} Y serves as a carrier, the amount thereof is extremely small compared to ⁹² Y, which is the target nuclide, and it is considered that there is almost no influence of excessive exposure on the patient.

11 to 14 are graphs showing the types of elements obtained when concentrated ⁹⁰ Zr, ⁹¹ Zr, ⁹² Zr, and ⁹⁴ Zr are used, respectively. As shown in FIG. 13, when ⁹² Zr is used as a raw material, the target ⁹² Y is produced in large quantities.

On the other hand, as shown in FIG. 11, when ⁹⁰ Zr is used as a raw material, a large amount of ⁹⁰ Y is generated. Further, as shown in FIG. 12, when ⁹¹ Zr is used as a raw material, a large amount of ⁹¹ Y is generated. As shown in FIG. 14, when ⁹⁴ Zr is used as a raw material, a large amount of ⁹⁴ Y is generated.

Natural zirconium contains not only ⁹² Zr but also ⁹⁰ Zr, ⁹¹ Zr, ⁹² Zr, and ⁹⁴ Zr, so that ⁹⁰ Y, ⁹¹ Y, and ⁹⁴ Y are generated as a by-product. become. Therefore, it can be said that it is preferable to use concentrated ⁹² Zr in order to avoid the formation of by-products.

  Hereinafter, the effects of the present invention will be made clearer by examples. In addition, this invention is not limited to a following example, In the range which does not change the summary, it can change suitably and can implement.

[Example 1]
<Calculation system>
As neutron sources (neutron converters) that can be used for actual production, the ^nat reaction (d, n) reaction due to the incidence of 12, 15, 20 MeV deuterons and the DT reaction due to the incidence of 350 keV deuterons (deuterium and tritium). (Reaction) was examined.

FIG. 5A shows a calculation system for neutrons by the ^nat C (d, n) reaction. The energy and angular distribution of incident neutrons were evaluated using TTNY (neutron yield from a thick target). The sample has a relatively high neutron yield, 45 deg. Arranged more inside. Since the time of actual production ^{of nat} ZnO powder to make it easier to separate and extract the ⁶⁴ Cu after irradiation is used, in this calculation was ^nat ZnO and samples. As the density of the powder ^nat ZnO, the value disclosed in Non-Patent Document 3 was used.

  FIG. 5B shows a calculation system for neutrons by the DT reaction. Since DT neutrons are isotropic in intensity, they can be activated in all directions from a neutron source. However, because there are facilities such as an accelerator beam line and cooling water pipes behind the beam, it is difficult to place the irradiated sample, so the sample was placed only in front of the neutron source.

  Here, as for by-products, it was shown that radioactive impurities are small in a neutron source having an energy of 15 MeV or lower. Since the neutron source used in the present invention has very few neutrons with higher energy than DT neutrons, it is considered that there are also few radioactive impurities.

Here, the production nuclides that may cause a problem in actual production are ⁶³ Cu and ⁶⁵ Cu which are stable isotopes of ⁶⁴ Cu. Since these nuclides are difficult to separate from ⁶⁴ Cu, the labeling rate of the drug may be lowered. Therefore, in this simulation, the amount of production of ⁶³ Cu and ⁶⁵ Cu as the by-products of ⁶⁴ Cu and the carrier was determined.

<Calculation result>
( ⁶⁴ Cu production amount distribution)
The distribution of the amount of ⁶⁴ Cu produced in the cross section passing through the center of the ^nat ZnO sample in each irradiation is shown in FIGS. FIG. 6A shows the ⁶⁴ Cu production distribution by the ^nat C (d, n) reaction, Ed = 12 MeV. FIG. 6B shows a ⁶⁴ Cu production distribution by a ^nat C (d, n) reaction, Ed = 15 MeV. FIG. 6C shows a ⁶⁴ Cu production distribution by a ^nat C (d, n) reaction, Ed = 20 MeV. FIG. 6D shows a ⁶⁴ Cu production distribution by DT reaction, Ed = 350 keV.

^It can be seen that neutrons from the ^nat C (d, n) reaction have a forward property, and the higher the energy, the more ⁶⁴ Cu is generated. A region in which 80% or more of ⁶⁴ Cu is generated is indicated by a dotted line. Here, the ^nat neutron (C, d, n) reaction has a truncated cone shape, and the DT neutron has a hemispherical shape. Assuming that sufficient ⁶⁴ Cu can be recovered in this region, ⁶⁴ Cu and by-products generated in this region will be described below.

^{64 Cu} generation amount contained in the foregoing areas (PHITS calculation result (see Non-Patent Document 4)), and the region size, and the ^{64 Cu} production amount per hour in this region are shown in Table 1.

However, for this production amount, deuteron irradiation at 2 mA was assumed assuming that a small cyclotron HR-30 was used for the ^nat C (d, n) neutron source. In the case of DT neutrons, irradiation at 25 mA was assumed on the assumption that a JAEA FNS 0-degree beamline was used. ^{When the nat} C (d, n) reaction was used, it was found that a production amount of 200 MBq, which is a standard for one diagnosis, can be achieved by irradiation for 1 hour at any incident energy.

(Carrier production amount distribution)
· ⁶³ Cu generation distribution generated distribution of stable nuclide ⁶³ Cu to produce as a by-product, shown in FIG. 7 (a) ~ (d) . FIG. 7A shows a ⁶³ Cu production distribution by a ^nat C (d, n) reaction, Ed = 12 MeV. FIG. 7B shows a ⁶³ Cu production distribution by a ^nat C (d, n) reaction, Ed = 15 MeV. FIG. 7 (c) shows a ⁶³ Cu production distribution by a ^nat C (d, n) reaction, Ed = 20 MeV. FIG. 7D shows a ⁶³ Cu production distribution by DT reaction, Ed = 350 keV. Intended for 12MeV not been producing little energy it can be seen that many of ^{63 Cu} higher neutron source is produced.

· ⁶⁵ Cu generation distribution generated distribution of stable nuclide ⁶⁵ Cu to produce as a by-product, shown in Figure 8 (a) ~ (d) . FIG. 8A shows a ⁶⁵ Cu production distribution by a ^nat C (d, n) reaction, Ed = 12 MeV. FIG. 8B shows a ⁶⁵ Cu production distribution by a ^nat C (d, n) reaction, Ed = 15 MeV. FIG. 8C shows a ⁶⁵ Cu production distribution by a ^nat C (d, n) reaction, Ed = 20 MeV. FIG. 8D shows a ⁶⁵ Cu production distribution by DT reaction, Ed = 350 keV. It can be seen that 12 MeV and 15 MeV are hardly generated, and that a higher energy neutron source generates more ⁶⁵ Cu.

・⁶⁴ Cu / ^total Cu production ratio
⁶³ Cu and ⁶⁵ Cu may lower specific radioactivity when ⁶⁴ Cu is used as a drug. ^Therefore, the proportion of ⁶⁴ Cu occupying the isotope of the generated Cu is important in producing a 64 Cu. Regarding this isotope purity, the proportion of ⁶⁴ Cu in the Cu isotopes contained in the irradiated sample is shown in FIGS.

FIG. 9 (a) shows the ⁶⁴ Cu isotope purity distribution by the ^nat C (d, n) reaction, Ed = 12 MeV. FIG. 9B shows the ⁶⁴ Cu isotope purity distribution by the ^nat C (d, n) reaction, Ed = 15 MeV. FIG. 9 (c) shows the ⁶⁴ Cu isotope purity distribution according to the ^nat C (d, n) reaction, Ed = 20 MeV. FIG. 9 (d) shows the ⁶⁴ Cu isotope purity distribution according to the DT reaction, Ed = 350 keV. It can be seen that the lower the neutron energy is, the higher the ratio of ⁶⁴ Cu is, and in particular, that with Ed = 12,15 MeV, very high purity ⁶⁴ Cu can be obtained. It was also found that DT neutrons have a very low purity of ⁶⁴ Cu.

Table 2 ^shows the 64 Cu efficiently produced can shape, ⁶⁴ Cu production amount per hour possible with accelerator HM-30 in the case of using the shape, and the isotopic purity of ⁶⁴ Cu at that time.

It was found that DT neutrons are not suitable for producing ⁶⁴ Cu, and the stability of impurities, ⁶³ Cu, is not suitable because it significantly reduces the purity.

In the neutron by the ^nat C (d, n) reaction using 20 MeV deuterons, a very large amount of ⁶⁴ Cu was produced, which was 4.45 GBq per hour. Assuming that 200 MBq of ⁶⁴ Cu is used for one inspection, about 100 cases of diagnostic ⁶⁴ Cu can be manufactured by operating for 5 hours. That is, instead of installing an accelerator in a hospital, it is considered that a large accelerator can be installed at a manufacturing base and transported to a local hospital. Further, ⁶⁴ Cu and inseparable ⁶³ Cu is produced as a ^by-product, 64 Cu purity ⁽⁶⁴ Cu ratio) was found to be 71.7%. The microdose test requires attention because the labeled compound is a trace amount and ⁶⁴ Cu having high isotopic purity is required.

In the neutron by the ^nat C (d, n) reaction using 15 MeV deuteron, 1.72 GBq per hour and a large amount of ⁶⁴ Cu production were obtained. Since it is possible to obtain about 2 to 19 ⁶⁴ Cu per hour using an existing medical accelerator, it is considered very useful. In addition, the purity of ⁶⁴ Cu is high and can be used for a micro dose test. Furthermore, since the existing medical accelerator can be used, there is an advantage that the introduction cost is relatively low.

With neutrons by the ^nat C (d, n) reaction using 12 MeV deuterons, a large amount of ⁶⁴ Cu was produced, 0.875 GBq per hour. Since it is possible to obtain about 1 to 9 cases of ⁶⁴ Cu per hour using a small medical cyclotron HM-30 manufactured by Sumitomo Heavy Industries for the purpose of BNCT, it can be said that the yield is sufficient for practical use. Moreover, the purity of ⁶⁴ Cu is very high and can be used for a micro dose test. Furthermore, since the energy of the incident deuteron is low and the activation of the facility is reduced, it can be said that this is a neutron source with less burden on management and maintenance of radioactive materials in the production facility.

In addition, the purity of ⁶⁴ Cu was 99.7% in the neutron by the ^nat C (d, n) reaction using 9 MeV deuteron. In this case, a production amount of ⁶⁴ Cu of about 30% when 12 MeV deuteron was used was obtained.

⁶⁴ Cu The currently produced ⁶⁴ Ni (p, n) (see Non-Patent Document 5) reaction is used. In an alternative method using an accelerator, a ⁶⁴ Zn (d, 2p) reaction (see Non-Patent Document 6), a ^nat Zn (d, x) reaction (see Non-Patent Document 7), and ⁶⁴ Zn (n, p) Reaction is listed as a candidate. Furthermore, with respect to the ⁶⁴ Zn (n, p) reaction, production experiments and simulations by neutron irradiation by a ^nat C (d, n) reaction using 40 MeV deuterons have been carried out (see Non-Patent Document 8). The literature values for these production volume predictions are shown in Table 3. In addition, the isotope purity of ⁶⁴ Cu in these production methods was calculated using PHITS. The results are shown together in Table 3.

It can be seen that the ⁶⁴ Ni (p, n) ⁶⁴ Cu reaction can produce much more ⁶⁴ Cu and higher isotope purity than other production methods. However, ⁶⁴ Ni as a raw material is very expensive because it is only 0.9% in nature. ^Nat Zn (d, x) ⁶⁴ Cu can obtain a relatively large amount of ⁶⁴ Cu. However, a large amount of ⁶¹ Cu is generated compared to ⁶⁴ Cu. ⁶¹ Cu is a radioactive isotope that is difficult to separate from ⁶⁴ Cu with a half-life of 3.3 hours and may cause extra exposure to the patient. For this reason, it is necessary to provide a cooling time in which ⁶¹ Cu collapses to reduce the influence, but assuming that this is an 8 half-life, ⁶⁴ Cu is also attenuated by about 76%. ⁶⁴ Zn (d, 2p) ⁶⁴ Cu can obtain a relatively large amount of ⁶⁴ Cu, and the formation of ⁶¹ Cu can be suppressed by using concentrated ⁶⁴ Zn. ^Since much ⁶⁵ Cu is produced, the isotopic purity is as low as 28.3%.

Although ⁶⁴ Zn (n, p) ⁶⁴ Cu using ^nat C (d, n) by 40 MeV incidence can obtain a relatively large amount of ⁶⁴ Cu, the isotope purity is as low as 25.2%. The neutron source is also considered to be used for ⁹⁹ Mo production and can be used in combination.

Based on the above results, the following two types of production methods can be used in combination for a future ⁶⁴ Cu production system.

^{· Nat C (d, n)} , Ed = by ^{40MeV 64 Zn (n, p)} 64 Cu prepared first using 64 Cu, use the large accelerator at manufacturing facilities, with the reaction ⁶⁴ Cu Is to mass-produce. Compared to the current production of PET nuclides for diagnosis in hospitals, the burden of radioactive material management in hospitals can be greatly reduced, and the cost of PET examinations can be greatly reduced. In addition, ⁹⁹ Mo can be produced using this neutron source, and a large-scale accelerator neutron source construction plan for medical RI production is being actively pursued as a GRAND project. The purpose of this method is to produce ⁶⁴ Cu for PET diagnosis, and since the isotope purity of ⁶⁴ Cu is low, it is considered that attention should be paid to its use for microdose inspection.

・ Manufacture of ⁶⁴ Cu using ^nat Zn (n, x) ⁶⁴ Cu with ^nat C (d, n), Ed = 15 MeV Second, using a small accelerator in a hospital, this reaction is used to ^{convert 64} Cu. Is to manufacture. Since ⁶⁴ Cu having high isotope purity can be produced, it is possible to produce ⁶⁴ Cu for microdose tests and diagnostics adapted to a small number of patients. Using an existing HM-30 that can be installed in a hospital, it is possible to produce a sufficient amount of ⁶⁴ Cu for one test in one hour, which can be realized earlier than the above method. It can be said that this is a manufacturing method with low introduction cost. HM-30 can also be applied to medical RI production.

[Example 2]
Medical RI production experiment using neutrons from C (d, n) reaction using Kyushu University tandem accelerator was conducted. Furthermore, the neutron yield distribution using foil activation was obtained. As a result, the advantages of the C (d, n) neutron source and the future production amount of ⁶⁴ Cu, etc. were shown. Details of the experiment will be described below.

Here, ⁶⁴ Cu or the like, which is a new RI candidate, capable of molecular imaging using positive emission tomography (PET) was used. The energy distribution of the neutron source is greatly related to the nuclide production and by-product production. Therefore, neutrons from the C (d, n) reaction, which can adjust the neutron energy distribution by changing the incident deuteron energy, were used. First, it aimed at test manufacture of ⁶⁴ Cu etc. using the natural sample by this neutron source. This reaction has few experimental data of double differential thick target neutron yield (DDTTNY), and cannot evaluate the validity of a manufacturing experiment result. Thus, another object was to derive DDTNY using foil activation.

The experiment was conducted at the Tandem Accelerator Facility (KUTL), Faculty of Science, Kyushu University. Accelerated deuterons (Ed = 9 or 12 MeV) are incident on a thick carbon target that becomes a neutron converter. Subsequently, a C (d, n) reaction occurs and accelerator neutrons are obtained.
(1) First, a raw material of ^RI, irradiated etc. In this neutron ^nat Zn and ^nat Ni, were RI production experiments.
(2) Furthermore, when measuring DDT TNY, ¹⁹⁷ Au, ⁵⁹ Co, ²⁷ Al, ⁹³ Nb, ^nat Fe, and ^nat Ni are installed every 10 degrees in the range of 0 to 90 degrees, and radiation by neutron irradiation Made.

As an example of experiment (1), FIG. 10 shows a spectrum obtained by measuring decay gamma rays from an irradiated ^nat Zn sample in an experiment with Ed = 12 MeV using a Ge detector. For comparison, an experimental result (see Non-Patent Document 9) using neutrons (14 MeV quasi-monochromatic) obtained by the DT reaction is also shown. Both are standardized so that the integral value is 10. From this result, it can be said that ⁶⁴ Cu can be produced with a low ratio of by-products in the C (d, n) reaction adjusted to a low energy neutron distribution as compared with DT neutrons.
RI Production Experiment for Ed = 12 MeV, Comparison of production amount calculated using DDTTNY result derived from Experiment (2) and actual generated radioactivity, 2015 Atomic Energy Society (Spring Annual Meeting) It is reported in.

The conclusion of Example 2 is shown below. We examined medical RI manufacturing using accelerator neutrons. It was found that by using a neutron source having a low neutron energy distribution, ⁶⁴ Cu and the like can be produced with a low by-product amount even when a natural sample is used as a raw material. This can be said to be an advantage of the C (d, n) neutron source having a variable neutron energy distribution. In addition, the production amount of ⁶⁴ Cu and the like for practical use could be predicted.

[Example 3]
A neutron converter was irradiated with 20 MeV deuterons, and the generated neutron was irradiated onto a sample (raw material) of concentrated ⁹² Zr to produce ⁹² Y. As the sample, a cylindrical one having a diameter of 50 mm and a length of 50 mm was used. This sample was placed 50 mm away from the deuteron accelerator when viewed from the neutron converter.

The production amount of ⁹² Y was 0.274 GBq per hour. In the produced ⁹² Y, it was observed that the purity increased from just after the production to about 3.5 hours. This is because ^91m Y (50 minutes) having a short half-life generated as a by-product is decreased by decay. The final purity (isotopic purity) was 94%.

[Example 4]
Using the AVF930 accelerator of Tohoku University Cyclotron RI Center, an experiment was conducted to evaluate the production amount according to the method of the present invention. Deuterons were accelerated to 20 MeV and made incident on a thick carbon member that was a neutron converter. As a result, a C (d, n) reaction occurred, and the generated accelerator neutrons were irradiated onto the natural Zr foil to generate ⁹² Y inside. The amount of ⁹² Y produced was determined by gamma ray measurement using a germanium semiconductor detector. The spectrum of gamma rays obtained by the measurement is shown in FIG.

From FIG. 15, it can be seen that ⁹⁰ Y and ^91m Y are generated as by-products. Most of ⁹⁰ Y is generated by (n, p) reaction of ⁹⁰ Zr, and most of ^91m Y is generated by (n, p) reaction of ⁹¹ Zr. ^When using an isotope-enriched sample of ⁹² Zr as a raw material, it is presumed that such a by-product is not generated, and it is better to use ⁹² Zr that is more concentrated than natural ⁹² Zr as a raw material. It is considered preferable.

10 ... deuteron accelerator, 11 ... deuteron, 12 ... neutron converter,
13 ... Neutron, 14 ... Raw material (member), 15 ... Medical RI,
20 ... Proton accelerator, 21 ... Proton, 22 ... Raw material.

Claims (10)

^64. By irradiating a neutron converter with deuteron, irradiating the neutron emitted from the neutron converter onto a member containing natural zinc that has not been subjected to isotope enrichment, and causing the neutron and the zinc to undergo a nuclear reaction. A method for producing a radioactive substance, comprising producing Cu.   2. The method for producing a radioactive substance according to claim 1, wherein the energy of the deuteron is 9 MeV or more and 20 MeV or less. ⁹² Y is produced by irradiating a neutron converter with deuterons, irradiating a neutron released from the neutron converter onto a member containing zirconium, and causing a nuclear reaction between the neutron and the zirconium. Production method of radioactive material.   The method for producing a radioactive substance according to claim 3, wherein the energy of the deuteron is 15 MeV or more and 30 MeV or less.   The radioactive material according to any one of claims 1 to 4, wherein the neutron converter is one that undergoes a (d, n) reaction that receives one deuteron and emits one neutron. Production method. An accelerator that accelerates deuterons,
A neutron converter that emits neutrons when irradiated with the deuterons;
An apparatus for producing a radioactive substance, comprising at least a member containing natural zinc that has not been subjected to isotope enrichment.   The said accelerator is a manufacturing apparatus of the radioactive substance of Claim 6 which has a means to give the energy of 9 MeV or more and 20 MeV or less with respect to the said deuteron. An accelerator that accelerates deuterons,
A neutron converter that emits neutrons when irradiated with the deuterons;
An apparatus for producing a radioactive substance, comprising at least a member containing zirconium.   The radioactive accelerator manufacturing apparatus according to claim 8, wherein the accelerator includes means for applying energy of 15 MeV to 30 MeV to the deuteron.   10. The member according to claim 6, wherein the neutron converter is a member made of an element that undergoes a (d, n) reaction that receives one deuteron and emits one neutron. 10. Radioactive material production equipment.