JP5673916B2 - Radioisotope production method and apparatus - Google Patents

Description

In the present invention, radioactive isotopes used in radiodiagnostics and the like are made of a wide range of isotopes that do not use nuclear fuel material ²³⁵ U and have a high intensity and a long half-life (for example, strontium 90 to cesium 137). The present invention relates to a manufacturing method and an apparatus which can be generated efficiently and inexpensively without generating a large amount and enable stable supply.

  At present, in the medical field, radiation and radioisotopes (radioisotopes; hereinafter also referred to as RI) are indispensable for diagnosing and treating diseases. Radiation emitted from RI can be reliably detected and quantified even if the amount of the substance itself is very small, and examination and diagnosis are performed by scintigraphy using this property. The pharmaceutical used for this is called a so-called “radiopharmaceutical”, and the RI used for radiopharmaceuticals and the like is suitable to emit a gamma ray having a short half-life and a large radiation penetration.

Examples of RI used for radiopharmaceuticals and examples of use thereof include, for example, ^99m Tc for brain / thyroid / bone scintigraphy, ⁶⁷ Ga for breast cancer / lung cancer / malignant lymphoma treatment, ²⁰¹ Tl for thyroid tumor scintillator, ⁶⁰ Co Is a source for gamma knife, ³² P is leukemia treatment, ³⁵ S is DNA base sequence / gene chromosome arrangement determination, ⁵¹ Cr is measurement of circulating blood volume / red blood cell volume, ⁵⁹ Fe is measuring serum total iron binding ability (TIBC) ⁸⁹ Sr, ^153m Sm, ¹⁸⁶ Re is pain relieving agent, ⁹⁰ Y is malignant lymphoma treatment, ¹⁰³ Pd is prostate cancer treatment, ¹²⁵ I is a tumor marker, ¹³¹ I is hyperthyroidism / thyroid cancer treatment, ¹³³ Xe is local Lung ventilation function test, etc.

Among these RIs, ^99m Tc, ⁸⁹ Sr, ⁹⁰ Y, ¹³¹ I, ¹³³ Xe, etc. are currently made from highly enriched ²³⁵ U enriched with ²³⁵ U from 36% to 93%, and neutron irradiation in the reactor. Thus, it is produced by fission reaction and extracting RI from the fission product. This method of using enriched ²³⁵ U is particularly problematic from the standpoint of non-proliferation, and the International Atomic Energy Agency (IAEA) and others are working to switch to technologies that use low enriched ²³⁵ U with a ²³⁵ U enrichment of 20% or less. It is conducted in various countries, and technology development corresponding to it is being promoted all over the world. However, despite 30 years of work, most RIs in the world are still produced using highly enriched ²³⁵ U. On the other hand, when low-concentration ²³⁵ U with a ²³⁵ U enrichment of 20% or less is used as a raw material for RI production, a new problem arises that the amount of plutonium produced increases about 25 times. For this reason, a method of irradiating a raw material target with thermal neutrons (0.025 eV) of a nuclear reactor and extracting generated RI is also used. A method of generating RI by irradiating a raw material target with charged particles from a cyclotron is also used.

Some RIs are manufactured in Japan, but most of them rely on imports from overseas. In 2007, it became difficult to obtain radiopharmaceuticals because of a problem with a Canadian reactor. In August 2008, a Dutch reactor supplying approximately 26% of ⁹⁹ Mo to the global market stopped operation due to partial corrosion deformation of the bottom structure of the primary cooling system, and started operation in mid-February 2009. It was resumed. However, in May 2009, a heavy water leak was detected again in a Canadian reactor, and the operation was suspended. The restoration is considered to be early as of the end of March 2010. In this way, if most of RI is relied on by imports from other countries, it is expected that a stable supply system cannot be maintained due to domestic circumstances in other countries, aging of reactors, maintenance, troubles, etc., and stable supply of RI Has become an important and urgent issue. Especially in Canada, where Japan relies on imports of most RI, it has been decided that the nuclear reactors for supply will reach the operation deadline in 2011. There is no realistic plan from the point of view. In the United States and Europe, the stable supply of RI including ⁹⁹ Mo is eagerly awaited, but a realistic system that can cope with it has not been taken, and it is necessary to establish the system as soon as possible ( Non-patent document 1). In addition, if most of RI is dependent on imports from overseas, the price of RI used in medical treatments will rise, and this will cause a rise in overall medical costs. In 2007, the sales price of radiopharmaceuticals reached 44 billion yen (Non-patent Document 2: Page 5).

In addition, when ²³⁵ U is fissioned in a nuclear reactor, various nuclides are generated in addition to the desired RI as shown in FIG. 1 (Non-patent Document 3), and storage, management, processing, etc. of unnecessary fuel waste can be performed. It became enormous and very annoying.

In consideration of such problems, the present applicants have, ²³⁵ without using a ^U, efficiently producing a highly radioactive molybdenum ^{99 Mo} which is the parent nuclide of radioactive technetium ^{99m Tc,} which is often used as a radioactive diagnostic agent The technique which performs is proposed (patent document 1). In the proposed method, an aqueous Mo solution in which a Mo compound is dissolved in water is irradiated with neutrons in an irradiation capsule installed in the core of a nuclear reactor to generate ⁹⁹ Mo by a ⁹⁸ Mo (n, γ) reaction. It is intended to produce ⁹⁹ Mo efficiently by recovering Mo aqueous solution continuously or batchwise. Likewise Patent Document 2, using a ^{98 Mo,} a technique for generating radioactive molybdenum ^{99 Mo} by thermal neutron capture reaction is proposed. However, in the case of using these thermal neutron capture reactions, since the reactor is used, the production site is limited, and in addition to being largely dependent on the operation mode of the reactor, the production cost is expensive and the reaction cross section is small. The specific activity was low and there was a problem in production efficiency. In addition, considering the safety of nuclear reactors, for example, there may be a situation where it is necessary to stop the operation for half a year, for example, during periodic inspections. In order to easily and stably supplying the ^{99 Mo} facilities such as hospitals these circumstances, it was necessary further technical devices.

  On the other hand, it is also performed to generate a RI by irradiating a raw material target with a proton or heavy ion beam using an accelerator. In the case of protons, by making the accelerator used compact, it can be easily used in facilities such as hospitals. However, when generating RI using protons generated from such a small accelerator, it is only possible to deal with light nuclide RI, and when trying to deal with heavy nuclide RI, it is possible to avoid increasing the size of the accelerator. There was a problem that I could not. That is, when generating a RI using protons, the proton has a positive charge, so to react with the target nucleus of a heavy nuclide (although it is a nucleus with many positively charged protons) Because there is a repulsive interaction between positive charges, we have to overcome this and get inside the nucleus. For that purpose, the energy of the incident protons needs to be sufficiently high. Furthermore, when protons are incident on the target material, the proton energy in the target is greatly reduced, so that the thickness of the target that can be used is limited, and as a result, the efficiency of generating sufficient RI is often not high. On the other hand, the energy loss in the target increases the temperature of the target, and the use intensity of the proton beam may be limited for a target having a low melting point. By the way, the proton beam is generated by an accelerator and transported through the vacuum pipe to a nearby location where the target is set. However, when the target is set on the atmosphere side, it is necessary to maintain the vacuum in the vacuum pipe and shut off the atmosphere side portion. The material used for blocking must be as thin as possible because it must not reduce the energy and intensity of the proton beam. However, on the other hand, since this material is continuously irradiated with a proton beam, it is destroyed as a result of radioactive damage, and it becomes difficult to use a high-intensity proton beam for a long time. In order to manufacture various RIs according to the purpose, if the target material can be set in the atmosphere, the shape and material of the target can be selected flexibly, which is actually very convenient. As described above, RI generation using a proton beam has a problem. These circumstances are similar for heavy ion beams. The problem is greater than the amount of positive charges more than protons.

JP 2008-102078 A Special Table 2002-504231

“Accelerating production of medical isotopes” Nature Vol 4 57, 29 January 2009 Science Council of Japan Basic Medical Committee / General Engineering Committee Meeting Task Review Subcommittee on Radioactive and Radioactive Use “Proposal: Stable Supply System of Radioisotopes in Japan” July 24, 2008 Nuclear Physics A462 (1987) 85-108 North-Holland, Amsterdam

The present invention solves the above-mentioned problems of the prior art, does not use the enrichment ²³⁵ U, does not use the nuclear reactor facility, and does not generate a large amount of fuel waste, thereby efficiently and inexpensively and easily. It is an object of the present invention to provide a method and an apparatus capable of realizing a stable supply of a radioisotope.

  In order to solve the above problems, the present invention provides the following technical methods or means.

[1] A power supply for applying a voltage to accelerate a charged particle beam made of deuterium or protons, an acceleration tube for applying the voltage from the power supply and accelerating the charged particle beam, and an accelerator emitting unit In addition, a target for generating fast neutrons of titanium with occluded deuterium is arranged, and a fast neutron generator for generating fast neutrons is generated by irradiation with the accelerated charged particle beam. The number obtained by multiplying the atomic number (Z) of the raw material target by 0.10 [unit: MeV / number of protons] and 0.2 [ The value obtained by adding [Unit: MeV] [Unit: MeV] as the lower limit, and the reaction cross section of the (n, p) reaction at an energy value other than the lower limit is the reaction cross section of the (n, p) reaction at the lower limit. Be equal Using a scale small accelerator enough to generate fast neutrons with the energy value in a range of the energy value and the upper limit value,
A radioactive isotope characterized by irradiating a raw material target including a solid or liquid target nucleus with fast neutrons generated by the charged particle beam irradiation to cause an (n, p) reaction to generate a radioisotope Manufacturing method.

[2] In the first invention, fast neutrons are applied to a raw material target including one target nucleus selected from the following Table 1, Table 2, Table 3, Table 4, Table 5, and Table 6 as the target nucleus. A method for producing a radioisotope characterized by irradiation.

[3] In the first or second invention, the raw material target is irradiated with fast neutrons in a state in which the raw material target is in close contact with or separated from a fast neutron generating unit provided in the accelerator emitting unit. A method for producing radioisotopes.

[4] A power source for applying a voltage to accelerate a charged particle beam made of deuterium or protons, an accelerator tube for applying the voltage from the power source to accelerate the charged particle beam, and an accelerator emitting unit In addition, a target for generating fast neutrons of titanium with occluded deuterium is arranged, and a fast neutron generator for generating fast neutrons is generated by irradiation with the accelerated charged particle beam. The number obtained by multiplying the atomic number (Z) of the raw material target by 0.10 [unit: MeV / number of protons] and 0.2 [ The value obtained by adding [Unit: MeV] [Unit: MeV] as the lower limit, and the reaction cross section of the (n, p) reaction at an energy value other than the lower limit is the reaction cross section of the (n, p) reaction at the lower limit. Be equal The charged particle beam comprising: a small accelerator having a scale capable of generating fast neutrons having an energy value within a range having an energy value as an upper limit; and target support means for supporting a source target including a solid or liquid target nucleus. Radiation produced by irradiating a raw material target with fast neutrons generated by irradiation, generating a (n, p) reaction that emits one proton by irradiation with one neutron, and generating a radioisotope Isotope production equipment.

[5] In the fourth invention, the raw material target includes one target nucleus selected from the following Table 1, Table 2, Table 3, Table 4, Table 5, and Table 6 as a target nucleus. An apparatus for producing a radioisotope, characterized in that

[6] The radioisotope according to the fourth or fifth invention, wherein the raw material target is set in a state of being in close contact with or separated from a fast neutron generator provided in the accelerator emission unit. Element production equipment.

[7] In the fourth or fifth invention, the fast neutron generator provided in the accelerator emitting portion includes a cooling means, and the fast neutron generator has a partition function between the vacuum chamber and the atmosphere side, and An apparatus for producing a radioisotope, wherein the raw material target is set in close contact with the fast neutron generator.

According to the present invention, a raw material target having a solid or liquid target nucleus is irradiated with fast neutrons generated by charged particle beam irradiation from a small accelerator, and a (n, p) reaction is caused to produce a radioisotope. Therefore, it is possible to stably supply radioactive isotopes efficiently and inexpensively by reducing fuel waste with high strength and long half-life without using enrichment ²³⁵ U, without using nuclear reactor facilities. .

  In addition, the radioisotope production apparatus according to the present invention does not need to be regulated by nuclear fuel materials and can be miniaturized, and thus has an advantage that it can be easily used in facilities such as hospitals.

  Further, according to the present invention, since the RI is generated by irradiating the raw material target with neutrons, the neutron has no charge as compared with the case where the target is irradiated with a positively charged proton beam. As with a light target, it can be handled by a small-scale accelerator, and it is not bothered by energy loss in the target and the heat generated by the target. Can be irradiated at once, and the efficiency of RI generation can be increased. Further, there is an advantage that the target can be arranged in the atmosphere, and the degree of freedom of arrangement of the raw material target is increased. This is considered to give immense convenience to various users.

It is a figure which shows the production intensity distribution of the nuclide produced | generated by ²³⁵ U fission in a nuclear reactor. It is a graph which shows an example of the reaction cross-sectional area evaluation value of a raw material target and a fast neutron. It is a figure which shows typically the radioisotope manufacturing apparatus by one Embodiment of this invention. It is explanatory drawing of the sample container used when produced | generated RI is gas. It is a figure which shows typically the principal part of the radioisotope manufacturing apparatus by another embodiment of this invention. It is a block diagram which shows the manufacturing procedure of the radioisotope by this invention. The figure which shows that ⁵⁹ Fe was produced | generated by irradiating the raw material target containing the target nucleus ⁵⁹ Co with the fast neutron produced | generated by the charged particle beam irradiation from a small accelerator, (a), (b) The measurement results are 1099 keV and 1291 keV gamma rays, respectively.

  Hereinafter, the present invention will be described in detail based on embodiments.

In the present invention, a radioisotope used for a radiodiagnostic agent or the like is irradiated with fast neutrons generated by irradiation of a charged particle beam from a small accelerator on a raw material target including a solid or liquid target nucleus. Produced by initiating a (n, p) reaction that releases one proton upon irradiation. In the present invention, fast neutron means neutron having energy of 0.1 MeV or more.

When a raw material target is irradiated with fast neutrons, (n, 2n) reaction, (n, p) reaction, (n, ⁴ He) reaction, (n, 3n) reaction, (n, np) reaction, (n, n ′) γ) Various reactions such as reaction occur, but the raw material target of the present invention has a large reaction cross-section in the (n, p) reaction, and the desired radioisotope is used as in the case of using a nuclear reactor. Without producing a large amount of fuel waste, it can be manufactured with reduced radioactivity.

FIG. 2 is a graph showing, as an example, evaluation values of neutron energy and reaction cross section when a raw material target having ⁵⁹ Co as a target nucleus is irradiated with fast neutrons. As can be seen from FIG. 2, when the raw material target is irradiated with fast neutrons, the (n, p) reaction is dominant depending on the value of neutron energy, and the reaction cross section is large.

In the present invention, radioisotopes (product nuclei) are produced using (n, p) reaction using the target nuclei shown in Table 1, Table 2, Table 3, Table 4, Table 5, Table 6 below. Examples of the target to be used are shown in Table 1, Table 2, Table 3, Table 4, Table 5, and Table 6 below.

  Since the above-mentioned produced nuclide is different in element from the target nuclide, it can be made carrier-free.

In the present invention, in order to produce a radioisotope, a raw material target is irradiated with fast neutrons generated by charged particle beam irradiation using a small accelerator without using a nuclear reactor. In this way, compared with the case where a nuclear reactor is used, the radioactivity can be reduced without generating a large amount of fuel waste.

Small accelerator you accelerate the charged particle beam to generate a high-speed neutrons may be, for example, by using a commercially available compact accelerator, the present applicant equipment and is Japan Atomic Energy Agency nuclear fusion neutron engineering for the neutron source facility of (FNS A facility such as a DT neutron source such as) may be used.

The energy of fast neutrons used in the present invention is obtained by multiplying the threshold (Q value) of the (n, p) reaction by the atomic number (Z) of the raw material target by 0.10 [unit: MeV / number of protons]. And the value obtained by adding 0.2 [unit: MeV] [unit: MeV] as the lower limit, and the reaction cross section of the (n, p) reaction at an energy value other than the lower limit is (n, p ) Irradiation with fast neutrons having an energy value within a range having an energy value equal to the reaction cross section of the reaction as an upper limit Here, the threshold value of the (n, p) reaction is when the reaction cross section starts to take a value other than 0 (burn) from the state of 0 (burn) in the graph as shown in FIG. 2 showing the neutron energy and the reaction cross section. It is the energy value.

In the accelerator for neutron generation, for example, deuterium ( ² H) beam is irradiated to tritium ( ³ H), and fast neutrons can be generated together with helium ( ⁴ He) in the next reaction.

² H + ³ H → ⁴ He + n
The neutron energy (En) generated by this reaction is given by the following relational expression.

4 × En = Ed + 2 × {2 × Ed × En} ^1/2 × cos θ + 3Q
Here, Ed is deuterium energy, Q is the generated energy of the reaction, and Q = 17.6 MeV. θ is the angle between the generated neutrons and incident deuterium. From this equation, it can be seen that, for example, 14 MeV fast neutrons can be obtained by using low energy deuterium of 0.35 MeV. In addition, the International Fusion Materials Irradiation Facility (IFFIF), which is currently underway, will generate high-intensity fast neutrons by irradiating metallic lithium (Li) with deuterium. Furthermore, even if metal lithium (Li) or metal beryllium (Be) is irradiated with protons or deuterium, fast neutrons can be generated.

Here, ⁹⁹ Mo is considered as an example for the generation efficiency of RI according to the present invention. The amount of ⁹⁹ Mo produced by fission in the nuclear reactor (Y _furnace ) is given below.

²³⁵ U: Concentration 20%. The fission cross section due to the reaction between thermal neutrons and ²³⁵ U is 585 burns. Among them, the production ratio of ⁹⁹ Mo is 6% (see FIG. 1). From the above, the amount of ⁹⁹ Mo generated in this U = 0.20 × 585 × 0.06 = 7 burns.

¹⁰⁰ Mo: natural abundance ratio 9.6%. ⁹⁹ Mo production reaction cross section by fast neutron is 1.5 burn. From the above, ⁹⁹ Mo produced by natural Mo = 0.096 × 1.5 = 0.14 burn.

That is, the ratio of Y _{high speed} to Y _furnace = Y _{high speed} / Y _furnace = 0.14 / 7 = 0.02 Formula (1)
The amount of ⁹⁹ Mo produced by fast neutrons is 2% in the case of a nuclear reactor, excluding the amount of neutrons.

By the way, as for the amount of neutrons, the thermal neutron amount of the _reactor φ _Reactor : In the case of Japan Nuclear Energy Agency Research Reactor Facility JRR3, φ _Reactor = 10 ¹⁴ pieces / (cm ² · second)
Formula (2)
Amount of fast neutrons φ _Fast : In the case of IFMIF, φ _fast = 10 ¹⁴ pieces / (cm ² · sec)
Formula (3)
That is, the ratio of fast neutron amount to reactor amount is: φ _fast / φ _reactor = 1 formula (4)
It becomes. Above, in view of the neutron amount, the ratio of the fast neutron ^{99 Mo} production amount by utilizing a reactor ^{99 Mo} production amount by the use it is given below.

0.02 × 1 = 0.02 Formula (5)
Here, considering that a high concentration of ¹⁰⁰ Mo can be obtained relatively easily (for example, assuming 100% concentration), the ratio of equation (5) is calculated from equations (1) and (4):
0.02 ÷ 9.6 × 100 = 0.21 Formula (6)
It becomes. That is, according to the present invention, it is understood that ⁹⁹ Mo can be produced in an amount sufficiently comparable to the production amount in the nuclear reactor using fast neutrons.

  Further, when generating neutrons by irradiating metal lithium (Li) with protons, the neutron energy (En) generated by this reaction is given by the following relational expression.

En = Ep × (7/8) −Q (MeV)
Here, Ep is proton energy, Q is a reaction threshold value, and Q = 1.64 MeV. From this equation, it can be seen that, for example, when a high-energy proton of 25 MeV is used, a fast neutron of 20 MeV can be obtained.

  FIG. 3 schematically shows an apparatus for producing a radioisotope according to an embodiment of the present invention.

  In the figure, 1 is a high voltage power source, 2 is a power cable, 3 is an accelerator terminal, 4 is an acceleration tube, 5 is a deuteron transport line, 6 is a fast neutron generator, 7 is a cooling tube, 8 is a cooling system, 9 is A raw material target, 10 is a target support frame or sample container, 11 is a target support, and 12 is an RI container (target storage) with radiation shielding.

  The generation efficiency of RI increases when the raw material target 9 is brought into close contact with the fast neutron generator 6 as shown in FIG. In this case, the raw material target 9 or the sample container 10 containing it is brought into close contact with the fast neutron generator 6 and the tip of the cooling tube 7 via a cooling member made of, for example, Cu. In this case, the cooling member provided in the fast neutron generator 6 has a partition function between the vacuum chamber of the deuteron transport line 5 and the atmosphere side where the raw material target 9 is located. In some cases, as shown in FIG. 3B, the raw material target 9 may be separated from the fast neutron generator 6 by about 10 mm, and this distance is not limited.

  In order to generate a large amount of neutrons by the neutron generation reaction, the high voltage power source 1 outputs a high voltage for making the deuterium beam energy of about 0.35 MeV, for example. The power cable 2 is for applying the high voltage of the high voltage power supply 1 to the acceleration tube 4 in contact with the accelerator terminal 3. The fast neutron generator 6 is provided with a metal plate with excellent thermal conductivity such as Cu provided with a vapor deposition film of titanium or the like occluded with deuterium. It plays a role in generating a large amount of neutrons by causing a neutron generation reaction. The cooling system 8 serves to cool the metal plate by the cooling pipe 7 in order to prevent the deuterium in the deposited film irradiated with the deuterium beam from thermally diffusing. Cooling is performed by water cooling or the like. The metal plate may be a fixed type or a rotary type.

  As the raw material target 9, a natural target element or a target element concentrated to a natural abundance ratio or higher is used as an oxide powder or the like, or a powder obtained by compressing and molding this powder into a high density ( (Bulk density of 60% or more) can be used (Japanese Patent Laid-Open No. 55-22102). Moreover, when using the concentrated target element, it is necessary to perform an electromagnetic separation collection method etc. as the pretreatment. In the case of using a powder of a target element oxide or the like, it is necessary to seal it in a quartz tube and then seal it in an aluminum metal irradiation container. When using pelletized powder of target element oxide or the like, it is directly sealed in a metal irradiation container. This metal irradiation container is the sample container 10. Furthermore, a target element metal can also be used as a target. However, in this case, dissolution with nitric acid or the like is required for extraction of the target element metal.

When using a pelletized material as the raw material target 9, for example, a size having a diameter of 10 mm and a thickness of 0.5 mm can be used, but of course, this is an example and is not limited thereto. Considering the irradiation energy, yield and the like of fast neutrons, the shape and size can be appropriately set. In that case, if the raw material target 9 is too thick, a problem of neutron scattering occurs and the generation efficiency is lowered. Therefore, it is necessary to consider this point. Neutrons are emitted from the fast neutron generator 6 in all directions, and the neutron flux (pieces / cm ² · sec) decreases at 1 / r ² . For this reason, when the raw material target 9 is in close contact with the fast neutron generator 6, the RI generation efficiency is maximized. Note that r is the distance from the fast neutron generator 6 to the raw material target 9.

  The raw material target 9 is fixed or accommodated in the target support frame or the sample container 10. The target support 11 serves to fix the target support frame or the sample container 10. The RI container 12 includes a radiation shield, and the generated RI is put in this, taken out of the laboratory, and transported and moved to a required place. Each member other than the RI container 12 also shields radiation as necessary.

The raw material target is irradiated with fast neutrons by the apparatus configured as described above, and the irradiation time can be set in consideration of the half-life of the produced nuclide. In addition, when the half-life is longer than 5 days, a desired amount of RI can be obtained if the irradiation time is about 5 days. In this case, since fast neutrons generated by charged particle irradiation from a small accelerator are used, fission is not used, so a large amount of fuel waste is not generated, and (n, p) reaction with a relatively large reaction cross section. Therefore, a desired RI can be stably supplied with high efficiency. Furthermore, since the apparatus configuration can be very downsized using a commercially available accelerator, RI can be manufactured and used easily and stably in a facility such as a hospital.

  When the generated RI is a gas, a sample container 10 as shown in FIG. In this case, as the sample container 10, a highly airtight container such as stainless steel is used. The sample container 10 is provided with a vacuum valve 10A, and the generated RI can be discharged. Such a sample container 10 is used when a gas RI is generated by irradiating a solid target nucleus with fast neutrons.

  When the gas RI is generated, an alkali solution such as sodium hydroxide or an acid solution such as hydrochloric acid is placed in the sample container 10 and stirred to dissolve the gas RI generated by the nuclear reaction. Thereafter, the vacuum valve 10A is connected to the air collection system 13 as shown in FIGS. Here, the sample container 10 is heated, and the dissolved gas RI is discharged into the vacuum and introduced into the air collection system 13. In the air collection system 13, for example, the gas RI is adsorbed by an adsorbent such as molecular sieve. At this time, the air collection system 13 is cooled with liquid nitrogen or the like as necessary. The remaining solid target can be reused.

  Next, another embodiment of the present invention will be described.

  FIG. 5 is a diagram schematically showing a main part of an RI manufacturing apparatus according to another embodiment of the present invention, in which (a) is a schematic diagram viewed from a direction perpendicular to the traveling direction of the deuterium beam. (B) is a schematic view seen from the deuterium beam traveling direction, (b) is a case where the neutron generator and the raw material target are separated from each other.

  In the figure, 21 is a deuterium beam, 22 is a rectangular parallelepiped vacuum beam tube, 23 is a copper plate having a titanium film adsorbing tritium (tritium), 24 is a target sample, and 25 is a cooling member. The cooling member 25 may be integrated with the copper plate 23. In this case, the copper plate 23 has an inner wall and an outer wall, and a titanium film adsorbing deuterium on the surface of the copper plate 23 on the inner wall (vacuum chamber side). The target sample 24 is disposed in close contact with or spaced from the surface of the copper plate on the outer wall (atmosphere side) so that a cooling medium such as water passes through the space between the inner wall and the outer wall. Yes.

The fast neutrons generated by irradiating the deuterium ( ² H ⁺ ) beam 21 to the deuterium ( ³ H) isotropically in almost the entire space regardless of the incident direction of the deuterium beam 21. Has the property of being released. For this reason, in order to make maximum use of neutrons generated in a limited neutron utilization time, the target sample 24 is arranged as follows. As the target sample 24, for example, a pellet-like material obtained by compressing and solidifying and sintering a powder of a natural target element or a concentrated oxide of a target element may be used. Metals can also be used.

3 for tritium deuterium beam 21 of high strength by increasing the temperature of the titanium deuterium containing titanium film is irradiated to set only was the copper plate 23 prevent the thermal diffusion, the tritium-containing titanium film The copper plate 23 is cooled by the cooling member 25. In order to use a higher-intensity deuterium beam 21 within a given cooling capacity range, it is conceivable to reduce the heat load per unit area provided by the deuterium beam 21. Therefore, the size of the deuterium beam 21 is changed from the usual 5 mm diameter to, for example, 10 mm diameter by changing the beam transport system of the accelerator. As a result, the heat load per unit area can be reduced to ¼ and increased up to four times the intensity of the conventional deuterium beam, and the resulting neutrons can be used four times as much. Since fast neutrons are isotropically emitted into the entire space, the target sample 24 is set not only in front of the deuterium beam 21 but also on the side surface as shown in FIG.

The deuterium beam 21 is irradiated to a copper plate 23 having a tritium-containing titanium film through a beam transport system in which a vacuum is maintained by a vacuum beam tube 22 which is a rectangular parallelepiped. Then ^{2 H + + 3 H → 4} He + n fast neutrons generated by the reaction, placed (closest distance) on the atmosphere side of the cooling member 25 (the copper plate 23) is irradiated to the target sample 24. On the other hand, in order to efficiently use fast neutrons emitted backward with respect to the deuterium beam 21 incident at a right angle to the copper plate 23 having a tritium-containing titanium film, a vacuum beam tube (a rectangular parallelepiped) close to the fast neutron generation site. ) Four surfaces of 22 are processed and the target sample 24 is embedded as shown. With such a configuration, fast neutrons are isotropically emitted into the entire space, so that RI generation is performed with higher efficiency.

  Moreover, when using a faster neutron, as mentioned above, a proton etc. are irradiated to Li or Be. In this case, most of the neutron flux is emitted along the direction of the proton beam. Therefore, the arrangement of the target sample 24 is in front of the neutron generator as shown in FIGS. 5 (d) and 5 (e). Reference numeral 26 denotes a proton beam. FIG. 5D shows the case where the target sample 24 is brought into close contact with the neutron generator, and FIG. 5E shows the case where the target sample 24 is spaced from the neutron generator. As described above, according to the present invention, since the target sample 24 can be arranged on the atmosphere side instead of in a vacuum, there is an advantage that the degree of freedom of the shape and arrangement of the target sample is increased.

  Next, a method for manufacturing RI according to the present invention will be described.

The RI manufacturing method according to the present invention basically irradiates a raw material target including a target nucleus with fast neutrons generated by charged particle beam irradiation from a small accelerator, and irradiates one proton with one neutron. (N, p) reaction is generated to generate RI, and RI is generated.

  Hereinafter, an example of an RI manufacturing method according to the present invention will be described with reference to the block diagram of the manufacturing procedure of FIG.

  First, for example, using a natural target element, a powder such as an oxide is compressed, molded, and sintered to produce a pellet-shaped raw material target (step S1).

  Next, a raw material target is put in a sample container and set at a neutron irradiation position (step S2).

Next, a deuterium beam of, for example, 0.35 MeV is irradiated from a neutron generator to a titanium film on which a deuterium is occluded provided on a cooling copper plate. As a result, 14 MeV fast neutrons are generated (step S3) .
In the raw material target of the present invention, the (n, p) reaction predominates due to the irradiation of fast neutrons, and RI is generated (step S4).

  When neutron irradiation is performed for an appropriate time, the irradiation is stopped, the sample container containing the RI is taken out, and a desired RI is obtained (step S5).

  Although an example of the manufacturing method has been described above, of course, the manufacturing method of the present invention is not limited to this example, and each step can be performed using the various methods described above.

  As described above, the same method is applied to the raw material target of the present invention even when fast neutrons generated by irradiating a metal Li (lithium) or metal Be (beryllium) with a proton beam or a deuterium beam are used. By doing so, it is possible to efficiently produce a desired radioisotope without producing a large amount of fuel waste.

  In addition, when the production nucleus is a long-life RI relative to the short-life RI of the daughter nuclide, the milking can be performed by a system called a cow milking system or a generator system.

The present inventors conducted the following experiment in order to confirm that a raw material target can be irradiated with a neutron beam generated by charged particle beam irradiation from a small accelerator to generate RI.
<Experimental purpose>
* Confirmation that ⁹⁹ Mo is generated with 14 MeV fast neutrons using a natural Mo sample with the expected reaction cross section * Generated with 14 MeV neutrons using the ⁹³ Nb sample used to determine the absolute value of the above reaction cross section Measurement of ⁹² Nb radiation and confirmation that it can be used to determine the cross-sectional area * ⁹⁹ Quantitative evaluation of residual radioisotope (fuel waste) produced in association with Mo production reaction * ² H + ³ H → ⁴ He + n Verifying 14MeV neutrons generated by the reaction is stably generated by the neutron intensity expected ^{(3 H} target performance evaluation)
* Confirmation that ² H (deuterium beam) for inducing the above reaction is supplied stably (Verification that small accelerator operates stably)
* Confirmation that Mo and Nb targets can be easily and flexibly installed and removed from the neutron irradiation site <Experiment location> Neutron Source Facility for Fusion Neutron Engineering, Japan Atomic Energy Agency (FNS)
<Experiment date>
Neutron irradiation experiment: January 27 to January 30, 2009 (6 hours of irradiation per day)
Measurement of generated Mo radioactivity: January 27 to February 5, 2009 <Sample> Natural Mo
Sample 1: Diameter: about 10 mm, thickness: 50 microns (0.05 mm), weight 40.214 mg
Sample 2: Diameter: about 10 mm, thickness: 5 microns (0.005 mm), weight 3.663 mg
Sample 1 was measured after 6 hours of irradiation.
Sample 2 was measured after neutron irradiation until the last day.
<Sample> ⁹³ Nb
Sample 3: diameter 10 mm, thickness: 0.1 mm, weight 67.196 mg
<Mo target installation location>
² Location 10 cm away from the neutron generation site in the direction of extension of the H beam axis.
<Neutral irradiation conditions>
* 14 MeV neutron production reaction
² H + ³ H → ⁴ He + n: ² H beam energy: 0.35 MeV
* Neutron generation amount 1.8 × 10 ¹¹ n / cm ² · second [January 27] to 1.5 × 10 ¹¹ n / cm ² · second [January 30]
^<99 Mo production reaction>
¹⁰⁰ Mo + n (R) ⁹⁹ Mo + 2n
< ⁹² Nb production reaction>
⁹³ Nb + n (R) ⁹² Nb + 2n
<Measurement of ⁹⁹ Mo and residual radioactivity (measured with FNS)>
* Measurement conditions: Start measurement after cooling time of approximately 1 hour after neutron irradiation * Measurer: Ge semiconductor detector * ⁹⁹ Mo sample and ⁹² Nb sample placement: Set 5 cm away from Ge detector <Result>
* Confirm that ⁹⁹ Mo is produced in the amount expected from the initial prediction * Confirm that ⁹³ Nb sample can be used to determine ⁹⁹ Mo cross-sectional area * ⁹⁹ Residual radioisotope (fuel) generated in association with Mo generation reaction Quantitative evaluation of waste) (Check that it is a small amount compared to the amount of ⁹⁹ Mo).
* Confirmed that 14 MeV neutrons generated by ² H + ³ H → ⁴ He + n reaction are stably generated with the expected neutron intensity ( ³ H target could be evaluated to have high performance)
* It was confirmed that ² H (deuterium beam) for inducing the above reaction was stably supplied.
(Verification of stable operation of small accelerators)
* It was confirmed that the Mo target and Nb target can be easily and flexibly installed and removed from the neutron irradiation site.

Next, the present inventors irradiate a raw material target with a fast neutron beam generated by irradiation with a charged particle beam from a small accelerator, and emit one proton by irradiation with one neutron (n, p). The following experiment was performed to confirm that the reaction was caused and RI could be generated.
<Experiment location> Japan Atomic Energy Agency, Neutron Source Facility for Fusion Neutron Engineering (FNS)
<Sample> Natural Co
Diameter: about 10mm, thickness: 1mm, weight 727.5mg
<Mo target installation location>
² Location 10 cm away from the neutron generation site in the direction of extension of the H beam axis.
<Neutral irradiation conditions>
* 14 MeV neutron production reaction
² H + ³ H → ⁴ He + n: ² H beam energy: 0.35 MeV
* Neutron generation amount 1.8 × 10 ¹¹ n / cm ² · seconds < ⁵⁹ Fe and residual radioactivity (measured with FNS)>
* Measurement condition: Start measurement 1 hour after neutron irradiation is completed * Measurer: Ge semiconductor detector * ⁵⁹ Fe sample placement: set at a position 5cm away from Ge detector * ⁵⁹ Fe released due to beta decay 1099 keV and 1291 keV gamma rays measured with a Ge semiconductor detector are shown in FIG. FIGS. 7A and 7B show data of 1099 keV and 1291 keV, respectively.
<Result>
The present inventors further irradiate a raw material target having ⁹⁹ Tc ( ⁹⁹ Tc ground state) as a target nucleus with fast neutrons generated by charged particle beam irradiation from a small accelerator, and irradiating with one neutron. It was confirmed that ⁹⁹ Mo can be generated by initiating a (n, p) reaction that releases one proton. According to this method, it is expected to contribute to the establishment of a stable supply system of ⁹⁹ Mo as a radiopharmaceutical for which a stable supply is currently desired.

  A similar method is applied to the raw material target of the present invention using fast neutrons generated by irradiating a metal beam Li (lithium) or metal Be (beryllium) with a proton beam, deuterium beam or tritium beam. Thus, the desired radioisotope can be efficiently produced without producing a large amount of radioactive waste.

DESCRIPTION OF SYMBOLS 1 High voltage power supply 2 Power supply cable 3 Accelerator terminal 4 Acceleration tube 5 Deuteron transport line 6 Fast neutron accelerator 7 Cooling tube 8 Cooling system 9 Raw material target 10 Target support frame or sample container 10A Vacuum valve 11 Target support stand 12 RI container ( Target storage)
DESCRIPTION OF SYMBOLS 13 Air collection system 21 Deuterium beam 22 Vacuum beam tube 23 Copper plate which has a trideuterium containing titanium film 24 Target sample 25 Cooling member 26 Proton beam

Claims (7)

A power source for applying a voltage for accelerating a charged particle beam made of deuterium or protons, an accelerator tube for accelerating the charged particle beam to which the voltage from the power source is applied, and an accelerator emitting unit, and 3 A target for generating fast neutrons of titanium with occluded deuterium is disposed, and a fast neutron generator for generating fast neutrons is generated by irradiation with the accelerated charged particle beam. One target is irradiated with one neutron. And the number obtained by multiplying the atomic number (Z) of the raw material target by 0.10 [unit: MeV / number of protons] and 0.2 [unit: MeV] ] was added value [unit: MeV] the lower limit value, (n, p) in the energy values other than the lower limit the reaction cross-sectional area of the reaction equal to (n, p) reaction cross section of the reaction in the lower limit energy Using a scale small accelerator enough to generate fast neutrons with the energy values within the range of the over value and the upper limit value,
A radioactive isotope characterized by irradiating a raw material target including a solid or liquid target nucleus with fast neutrons generated by the charged particle beam irradiation to cause an (n, p) reaction to generate a radioisotope Manufacturing method. 2. A raw material target including one target nucleus selected from the following Table 1, Table 2, Table 3, Table 4, Table 5, and Table 6 as a target nucleus is irradiated with fast neutrons. A method for producing the radioisotope described in 1.

3. The radioactive isotope according to claim 1 or 2, wherein the raw material target is irradiated with fast neutrons in a state in which the raw material target is closely attached to or separated from a fast neutron generator provided in the accelerator emitting portion . Production method. A power source for applying a voltage for accelerating a charged particle beam made of deuterium or protons, an accelerator tube for accelerating the charged particle beam to which the voltage from the power source is applied, and an accelerator emitting unit, and 3 A target for generating fast neutrons of titanium with occluded deuterium is disposed, and a fast neutron generator for generating fast neutrons is generated by irradiation with the accelerated charged particle beam. One target is irradiated with one neutron. And the number obtained by multiplying the atomic number (Z) of the raw material target by 0.10 [unit: MeV / number of protons] and 0.2 [unit: MeV] ] was added value [unit: MeV] the lower limit value, (n, p) in the energy values other than the lower limit the reaction cross-sectional area of the reaction equal to (n, p) reaction cross section of the reaction in the lower limit energy And small accelerator scale degree for generating fast neutrons having an energy value within the range of the over value and the upper limit value,
Comprising a target support means for supporting a raw material target including a solid or liquid target nucleus,
Irradiating a source target with fast neutrons generated by the charged particle beam irradiation, generating a (n, p) reaction that emits one proton by irradiation of one neutron, and generating a radioisotope Characteristic radioisotope production equipment. The raw material target is a raw material target containing one type of target nucleus selected from the following Table 1, Table 2, Table 3, Table 4, Table 5, and Table 6 as a target nucleus. The radioisotope production apparatus described in 1.

The radioisotope manufacturing apparatus according to claim 4 or 5, wherein the raw material target is set in a state of being in close contact with or separated from a fast neutron generator provided in the accelerator emission unit . A state in which the fast neutron generator provided in the accelerator emission part has a cooling means, the fast neutron generator has a partition function on the vacuum chamber and the atmosphere side, and the raw material target is in close contact with the fast neutron generator The radioisotope production apparatus according to claim 4, wherein the radioisotope production apparatus is set.

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