METHOD FOR PREPARING RADIOACTIVE SUBSTANCE THROUGH MUON IRRADIATION, AND SUBSTANCE PREPARED USING SAID METHOD

Abstract

In order to prepare a useful radioactive substance from radionuclides included in high-level radioactive waste and the like, an embodiment of the present invention provides a method for preparing a radioactive substance including a muon irradiation step for obtaining a first radionuclide by causing negative muons to be incident onto a radioactive target nuclide and triggering a nuclear muon capture reaction. The prepared radioactive substance includes at least one of the first radionuclide and a second radionuclide that is at least one type of a descendant nuclide obtained from the first radionuclide through radioactive decay. An embodiment of the present invention also provides the radioactive substance.

Claims

1. A method for producing a radioactive substance comprising a muon irradiation step for obtaining a first radionuclide through a muon nuclear capture reaction by irradiating a target nuclide which is a radionuclide with negative muons, wherein the radioactive substance to be produced comprises at least one of the first radionuclide and a second radionuclide, the second radionuclide being a descendant nuclide obtained from the first radionuclide via radioactive decay.

2. The method for producing a radioactive substance according to claim 1 further comprising preparing a target raw material containing the target nuclide to be irradiated with negative muon prior to the muon irradiation step, wherein the target nuclide in the target raw material is any of radionuclides in long-lived fission products (LLFPs) contained in a spent nuclear fuel or a substance separated from a spent nuclear fuel.

3. The method for producing a radioactive substance according to claim 2, wherein the target nuclide is .sup.99Tc, the first radionuclide is .sup.99Mo, and the second radionuclide is .sup.99mTc.

4. The method for producing a radioactive substance according to claim 1, wherein the target nuclide is .sup.99Tc, the first radionuclide is .sup.99Mo, and the second radionuclide is .sup.99mTc, further comprising a step of preparing a target material to be irradiated with negative muons prior to the muon irradiation step, wherein the target material is a recycled raw material containing at least any of .sup.99Tc produced as a by-product in an arbitrary step until a .sup.99Mo-.sup.99mTc generator is manufactured, .sup.99Tc obtained in a left over substance of an unused pharmaceutical preparation, and .sup.99Tc produced after radioactive decay in the .sup.99Mo-.sup.99mTc generator.

5. The method for producing a radioactive substance according to claim 3, wherein the muon irradiation step is performed for an irradiation time shorter than 66 hours, which is a half-life of .sup.99Mo.

6. The method for producing a radioactive substance according to claim 3, further comprising a collection step of .sup.99Mo ion from a substance containing .sup.99Tc ion which is an ion of the target nuclide by adsorbing .sup.99Mo ion which is an ion of the first radionuclide onto an ion exchange column.

7. The method for producing a radioactive substance according to claim 2, wherein the target nuclide includes at least one nuclide selected from a group of nuclides consisting of .sup.134Cs, .sup.135Cs, and .sup.137Cs, and the first radionuclide is .sup.133Xe.

8. The method for producing a radioactive substance according to claim 2, wherein the target nuclide is .sup.90Sr, the first radionuclide is .sup.89Rb, and the second radionuclide is .sup.89Sr.

9. The method for producing a radioactive substance according to claim 1 further comprising: an unloading step of an irradiated fluid containing the first radionuclide or the second radionuclide and a fluid medium from an irradiation position of the negative muon by transferring the fluid medium, the first radionuclide or the second radionuclide having been obtained from the target nuclide in the muon irradiation step; a collecting step for selectively collecting the first radionuclide or the second radionuclide from the irradiated fluid; and a reloading step for repositioning the irradiated fluid that has undergone the collecting step into the irradiation position by transferring the fluid medium.

10. The method for producing a radioactive substance according to claim 9, wherein the unloading step, the collecting step, and the reloading step are performed in parallel while the muon irradiation step is continuously performed.

11. A radioactive substance comprising at least one of a first radionuclide and a second radionuclide, the first radionuclide obtained through a muon nuclear capture reaction by irradiating a target nuclide with negative muons, and the second radionuclide being at least one descendant nuclide obtained from the first radionuclide via radioactive decay, wherein the target nuclide is a radionuclide.

12. The radioactive substance according to claim 11, wherein the target nuclide is any of radionuclides in long-lived fission products (LLFPs) contained in a spent nuclear fuel or a substance separated from a spent nuclear fuel.

13. The radioactive substance according to claim 12, wherein the target nuclide is .sup.99Tc, the first radionuclide is .sup.99Mo, and the second radionuclide is .sup.99mTc.

14. The radioactive substance according to claim 11, wherein the target nuclide is .sup.99Tc, the first radionuclide is .sup.99Mo, and the second radionuclide is .sup.99mTc, and wherein a target material to be irradiated with negative muons is a recycled raw material containing at least any of .sup.99Tc produced as a by-product in an arbitrary step until a .sup.99Mo-.sup.99mTc generator is manufactured, .sup.99Tc obtained in a left over substance of an unused pharmaceutical preparation, and .sup.99Tc produced after radioactive decay in the .sup.99Mo-.sup.99mTc generator.

15. The radioactive substance according to claim 12, wherein the target nuclide includes at least one nuclide selected from a group of nuclides consisting of .sup.134Cs, .sup.135Cs, and .sup.137Cs, and the first radionuclide is .sup.133Xe.

16. The radioactive substance according to claim 12, wherein the target nuclide is .sup.90Sr, the first radionuclide is .sup.89Rb, and the second radionuclide is .sup.89Sr.

Description

BRIEF DESCRIPTION OF DRAWINGS

[0026] FIG. 1 is an explanatory diagram illustrating NMCRs on the chart of nuclides in an embodiment of the present invention.

[0027] FIG. 2 is an explanatory diagram illustrating a nuclear fuel cycle.

[0028] FIG. 3 is an explanatory diagram illustrating a nuclear reaction on the chart of nuclides where Mo isotope is generated by NMCR that use .sup.99Tc for a target in an embodiment of the present invention.

[0029] FIG. 4 is a decay scheme diagram among nuclides with a mass number A=99 including .sup.99mTc.

[0030] FIG. 5 is an explanatory diagram illustrating a schematic configuration of a manufacturing apparatus for manufacturing .sup.99Mo by NMCR adopting a liquid target material in an embodiment of the present invention.

[0031] FIG. 6 is an explanatory diagram illustrating an outline of a process for producing .sup.99Mo by a batch manufacturing process with NMCR in an embodiment of the present invention.

[0032] FIG. 7 is a schematic chart illustrating a process outline of ion exchange method for further processing a product obtained by batch processing in an embodiment of the present invention.

[0033] FIG. 8 is an explanatory diagram illustrating nuclear reactions on the chart of nuclides where Xe of a mass number of 133 is generated by NMCR in an embodiment of the present invention.

[0034] FIG. 9 is a decay scheme diagram between Xe and Cs of a mass number of 133.

[0035] FIG. 10 is a schematic configuration diagram illustrating an irradiation processing apparatus for manufacturing .sup.133Xe by NMCR adopting a liquid target material in an embodiment of the present invention.

[0036] FIG. 11 is a schematic configuration diagram illustrating an irradiation processing apparatus for producing .sup.133Xe by NMCR adopting a solid target material in an embodiment of the present invention.

[0037] FIG. 12 is a schematic configuration diagram illustrating a configuration of a XeCs separation apparatus in an embodiment of the present invention.

[0038] FIG. 13 is an explanatory diagram illustrating a nuclear reaction on the chart of nuclides where a Rb isotope is generated by NMCR adopting .sup.90Sr target in an embodiment of the present invention.

[0039] FIG. 14 is a decay scheme diagram between Sr and Y having a mass number of 89.

DESCRIPTION OF EMBODIMENTS

[0040] Hereinafter, embodiments related to the production of radioactive materials according to the present invention will be described with reference to the drawings. In the description common parts or elements throughout the drawings are denoted by the same reference numerals, unless otherwise mentioned. Note that materials, amounts of use, ratios, processing contents, processing procedures, elements and specific examples thereof describe in the following specific examples, application examples, nuclide-specific arguments can be appropriately changed without departing from the gist of the present invention. Therefore, the scope of the present invention is not limited to the contents of the following specific description. For the explanation, we first describe the negative muon nuclear capture reaction (NMCR), describe the radionuclide as the target nuclide, and then explain the representative nuclide.

[0041] 1. Negative Muon Atomic Nuclear Capture Reaction (NMCR)

The negative muon nuclear capture reaction (NMCR), a nuclear reaction by negative muon utilized in this embodiment, has already been disclosed by one of the inventors of the present application (PTL1). The nature and use of the muon, the nuclear reaction mechanism, the method of generating the negative muon, and the NMCR by the negative muon, all of which are described there, are also adopted in the present embodiment. That is, in NMCR by negative muon, when a negative muon is incident on a target atom (hereinafter referred to as target nuclide), the negative muon that finally arrives at 1s orbital will be annihilated by spontaneous decay of the muon or will be captured into the nucleus before the annihilation. The phenomenon of this capture into nucleus is referred to as muon nuclear capture. What is utilized in this embodiment is a nuclear reaction (nuclear muon capture reaction (NMCR)) involving nuclear transmutation of the target nuclide resulting from the muon nucleus capture. Hereinafter, muon or p represents negative muon when it is merely described, unless otherwise noted.

[0042] The muon nuclear capture reaction (NMCR) includes a nuclear reaction in which the nucleus of the target raw material nuclide captures the muon, generating another element whose atomic number is smaller by one than that of target nucleus. The expression of NMCR in a nuclear reaction mode is written as

.sup.+N(Z.sub.0,A.sub.0).fwdarw.N(Z.sub.01,A.sub.0)+v(Eq. 1).

Here, the atomic number is Z.sub.0 (i.e., the proton number is Z.sub.0), the mass number is A.sub.0 (i.e., the sum of the proton number and the neutron number is A.sub.0), N is a nucleus in general, and N is a new nucleus to be generated while atomic number Z.sub.0 and the mass number A.sub.0 are specified. The reaction scheme expressed in Eq. 1 is that muon .sup. is captured by nucleus N of the target nuclide having atomic number Z.sub.0 and mass number A.sub.0, and then isobar nucleus N is generated whose atomic number is decreased by 1 to be Z.sub.01, while a neutrino v is generated.

[0043] The actual NMCR includes several variations depending on the combination of the number of neutrons released during the reaction and the nucleon number of the generated nucleus. The first one is a reaction expressed by Eq. 1 and expressed as (.sup., v) reaction. The second one is expressed as

.sup.+N(Z.sub.0,A.sub.0).fwdarw.N(Z.sub.01,A.sub.01)+n+v(Eq. 2)

where N is a nucleus which is neither N nor N. This is a reaction in which one neutron n is released and the mass number A.sub.0 decreases by one. Moreover, another reaction, expressed as

.sup.+N(Z.sub.0,A.sub.0).fwdarw.N(Z.sub.01,A.sub.02)+2n+v(Eq. 3)

may occur in which two neutrons 2n are released and the mass number A.sub.0 is decreased by two, where N represents an atomic nucleus that is neither N, nor N, nor N. The reactions of Eqs. 1 to 3 are expressed briefly as follows: [0044] 0 neutron released: [0045] (.sup., v) reaction: N ((Z.sub.01), A.sub.0) generation, [0046] 1 neutron released: [0047] (.sup., n v) reaction: N ((Z.sub.01), (A.sub.01)) generation, and [0048] 2 neutrons released: [0049] (.sup., 2 n v) reaction: N ((Z.sub.01), (A.sub.02)) generation.
The same applies to the following. It is to be noted that which isotopes are actually produced in what proportion depends on the nucleus of the target nuclide and the structure of the nucleus generated.

[0050] The reactions of NMCR formulated in the Eqs. 1 to 3 and so on can be explained on the basis of the chart of nuclides. FIG. 1 is an explanatory diagram illustrating NMCRs on the chart of nuclides in an embodiment of the present invention, where the atomic nucleus N portion on the chart of nuclides is enlarged with the atomic number Z on the vertical axis and the neutron number on the horizontal axis. The reaction (.sup., v) according to Eq. 1 generates atomic nucleus N that is located at a cell of one right column and one below row on the chart of nuclides with relative to the cell of nucleus N, the target nuclide to which muon .sup. collides. The nuclear reaction is indicated by path T1. Reactions through which neutrons are released such as (.sup., n v) and (.sup., 2n v) reactions according to Eqs. 2 and 3 correspond to ones that produce nuclei N and N at shifted positions to the left of the cell from the one right column one below row for the nucleus N by the number of released neutrons. These nuclear reactions are indicated by paths T2 and T3, respectively. The explanation set forth herein is only for describing the positional relationship on the chart of nuclides. It does not mean intermediate nuclei under the path are generated sequentially. For example, N is generated without stopping at N.

[0051] One of useful properties of NMCR is that there are few restrictions on the kinds of producible radionuclides, that is, most radionuclides can be produced. Any sort of radionuclide can be generated so long as a relevant target nuclide for the muon irradiation can be prepared. Another useful property of NMCR is that it can be caused with a very high degree of probability as long as a muonic atom can be formed. In other words, there is an extremely high probability that the nuclear reaction occurs comparing with one in a common nuclear reaction with neutron which is governed by a reaction cross-section (unit: barn). From these properties, it can be said that the production of radionuclide by NMCR has a high degree of freedom in selection of radionuclide, and can be carried out with significant efficiency. NMCR is advantageous also in the production capacity of nuclides.

[0052] In addition, advantages of use of muon can be found also in the practically important property that muon tends to be easily captured by atoms having a large atomic number, or atoms with more protons when plural sorts of atoms are irradiated by muons. Briefly, when an element having a small atomic number such as hydrogen, helium, carbon, nitrogen, oxygen or the like and a target nuclide having a large atomic number are contained in a substance, NMCR occurs at the target nuclide having a large atomic number with a high probability. Therefore, in a material to be irradiated that contains the target nuclide (hereinafter referred to as target raw material), the target nuclide is allowed to form a compound with an element having a smaller atomic number (light element) than that of the target nuclide, or to form an association product with a light element. The target nuclide can also be get mixed with another target nuclide or other target substance consisting of only light elements, dispersed in light elements, or even diluted with a diluent containing light elements only (e.g., helium gas or water). As a result, it is easy to change the production conditions according to various manufacturing requirements. As a typical example, NMCR can be caused with a target nuclide with a high probability even if a compound of a target nuclide and an element having a smaller atomic number than that of the target nuclide is adopted as the target raw material. For another typical example, it is also easy to cause the NMCR by bringing the target material into contact with or mixing with the fluid medium for the ease of transportation. These properties greatly enhance the practicality of radioactive materials and production of radionuclides that utilize NMCR.

[0053] Furthermore, the fact that radioactivity of the radionuclide produced is determined by the half-life of the radionuclide to be produced is also an advantageous property for facilitating the radionuclide production with NMCR. This property means that a radionuclide with a short half-life can be produced in a short period of time, whereas a long time is required for a radionuclide with a long half-life, when the same amount of radioactivity is to be produced.

[0054] In addition, difference in atomic numbers between the target nuclides and the generated radionuclide is useful at the time of separation and recovery of the radionuclide after production. This is because if the physical or chemical properties change with their atomic numbers, it becomes easy to separate the target nuclide in the target raw material and the generated radionuclide by way of a physical or chemical method.

[0055] Additional advantageous properties that facilitate the production of radionuclides in NMCR may be found in the fact that it is easy to automate by utilizing an appropriate conveying device and that the amount of radioactive substances that may become impurities is small.

[0056] 2. Radionuclide for Target Nuclide

In the present embodiment, radionuclides are used for target nuclides of NMCR. The radionuclides can typically be extracted from a high level radioactive waste discharged from a reprocessing process for reprocessing spent nuclear fuel from nuclear power plants. FIG. 2 is an explanatory diagram illustrating the reprocessing system (nuclear fuel cycle) of spent nuclear fuel used in nuclear power plants. Table 1 also lists half-lives and masses of nuclides of fission products (FPs), which is a part of spent nuclear fuel, in mass content per ton. Of the FPs, radionuclides having a half-life of more than 200,000 years are also called long-lived fission products (LLFPs). Nuclides in FPs or LLFPs that can be used as raw materials for producing useful radionuclides by NMCR are described below. As indicated in the nuclear fuel cycle 100 in FIG. 2, fuel 22 to be used in the nuclear power plant 30 is uranium 12 which was mined from a uranium mine 10 and processed in a fuel processing plant 20. From a nuclear power plant 30 spent nuclear fuel 32 and a low level radioactive waste 34 are discharged. The low level radioactive waste 34 is disposed of in the low level radioactive waste disposal facility 40, whereas the spent nuclear fuel 32 is further sent to the reprocessing plant 50, where it is separated into recovered uranium or plutonium 52 and a high level radioactive waste 54. On one hand, the recovered uranium or plutonium 52 is sent to the fuel processing plant 20 again and is used for a so-called MOX fuel in power generation at the nuclear power plant 30. On the other hand, a high-level radioactive waste 54 is processed into a vitrified solidification body or the like, for example, and then sent to a high level radioactive waste storage facility 60, and finally is brought into under control at a high level radioactive waste disposal facility 70 for a long time.

TABLE-US-00001 TABLE 1 Content Nuclide Half-life (per 1 ton) .sup.79Se 295k y 6 g .sup.90Sr 28.8 y 0.6 kg .sup.93Zr 1.61M y 1 kg .sup.99Tc 211k y 1 kg .sup.107Pd 6.5M y 0.3 kg .sup.126Sn 230k y 30 g .sup.219I 15.7M y 0.2 kg .sup.135Cs 2.3M y 0.5 kg .sup.137Cs 30.1 y 1.5 kg y: year, M: 10.sup.6, k: 10.sup.3

[0057] 3. Details of Nuclides

Typical nuclides that can be used as raw materials for producing useful radionuclides through NMCR from the FPs and LLFPs are .sup.99Tc, .sup.134Cs, .sup.135Cs, and .sup.137Cs, and .sup.90Sr. As indicated in Table 1, a high-level radioactive waste contains high concentrations of .sup.90Sr, .sup.90Tc, .sup.135Cs, and .sup.137Cs. That is, .sup.99Mo is produced from .sup.99Tc, where .sup.99Mo is used for obtaining .sup.99mTc, .sup.133Xe is produced from .sup.134Cs, .sup.135Cs, and .sup.137Cs, and .sup.89Sr is produced from .sup.90Sr. Details of combinations of these raw material nuclides and radionuclide to be produced will be described.

[0058] 3-1. Production of .sup.99Mo from .sup.99Tc

According to the present embodiment, .sup.99Mo can be produced from .sup.99Tc, which is a radionuclide. To produce a .sup.99Mo-.sup.99mTc generator .sup.99Mo (half-life: 66.0 h) is used, 82.4% of which decays by .sup. decay into .sup.99mTc. By gamma decay .sup.99mTc decays into .sup.99Tc with a half-life of 6.02 hours with a property in which gamma ray of 140.5 keV is released, where .sup.99mTc is used mainly in SPECT and is used for imaging agents for various organs, including brain imaging agent, thyroid function test agent, and parathyroid disease diagnostic agents. .sup.99mTc is an important nuclide for organ scintigram, accounting for about 80% of the radionuclides consumption in nuclear medicine RI. There are some countries that rely on imports from abroad for all domestic consumption of .sup.99Mo-.sup.99mTc generators. The nuclear reaction in which Mo isotopes are generated by NMCR targeting .sup.99Tc is depicted on the nuclear diagram in FIG. 3. A decay scheme diagram regarding to .sup.99Mo-.sup.99mTc generator is illustrated in FIG. 4.

[0059] In the present embodiment, NMCR is used in a process to produce .sup.99Mo from target material of Tc containing .sup.99Tc. When NMCR targeting .sup.99Tc is performed, schemes of reaction become as follows: [0060] .sup.99Tc (.sup., v) .sup.99Mo, [0061] .sup.99Tc (.sup., n v) .sup.98Mo, [0062] .sup.99Tc (.sup., 2n v) .sup.97Mo, [0063] .sup.99Tc (.sup., 3n v) .sup.96Mo, and [0064] .sup.99Tc (.sup., 4n v) .sup.95Mo.

[0065] These schemes are also understood through transmutation traces on the nuclear diagram in FIG. 3. Among the Mo isotopes, .sup.99Mo is used for the .sup.99Mo-.sup.99mTc generator utilizing the decay illustrated in FIG. 4. In NMCR with .sup.99Tc as the target nuclide, all of .sup.95Mo-.sup.98Mo among the producible isotopes of Mo are stable nuclei, and thus any other radioactive isotopes than .sup.99Mo is not found. That is, when .sup.99Mo is manufactured from .sup.99Tc by NMCR, only the intended nuclide is produced without producing any radioactive waste.

[0066] 3-1-1. Production Amount of .sup.99Mo from .sup.99Tc by Transmutation of NMCR

Next, the estimation of the amount of .sup.99Mo produced by NMCR irradiated with muon will be described for the cases of two irradiation conditions. The amount (number count) of nuclide to be generated is called a muon transmutation rate N.sub.TM and is calculated by the following equation:

N.sub.TM=I.sub.R.sub.cP.sub.NCP.sub.RBR, [0067] where [0068] I.sub.: number of muons/sec, [0069] R.sub.c: abundance of target nuclei, [0070] P.sub.NC: muon nucleus capture rate, and [0071] P.sub.RBR: branching ratio of muon nuclear capture reaction.
The abundance R.sub.c of the target nuclei is the proportion of the target nuclei in the target material for irradiation. The muon nuclear capture rate P.sub.NC is the probability that muonic atoms are generated and muons are captured by the nucleus. The branching ratio P.sub.RBR of the muon nuclear capture reaction is a factor depending on the numbers of neutrons released. In particular, R.sub.cP.sub.NCP.sub.RBR is referred to as reaction coefficient in the present application. This reaction coefficient represents the transmutation efficiency per muon. It should be noted that the reaction coefficient and the muon transmutation efficiency do not include a reaction cross section. That is, since the muon can be stopped at the target nuclide, one muon can transmute one nucleus without fail. In other words, if a muon can be captured by the nucleus of the target material, one or more types of NMCR will necessarily occur at certain ratios. The ratios are relative proportions among the probabilities of occurrence caused by each of a plurality of NMCRs expressed as a (.sup., xn v) reaction with x being an integer of 0, 1, 2, 3, 4, 5, each of which for a target nuclide among them corresponds to the branching ratio P.sub.RBR in the above. In addition, without fail in this context means that when capturing muons in a nucleus of the target material, at least one of the NMCRs explained above occurs, and the total of occurrence probabilities of NMCRs at that time is 100%. For this reason, NMCR has high production efficiency, and is a technique that requires only a short irradiation time for RI production.

[0072] The operating conditions of the apparatus for estimating the amount of .sup.99Mo produced are: [0073] proton accelerator: 500 MeV, 5 mA, proton beam, and [0074] the number of protons: 6.210.sup.18(5/1000)=3.110.sup.16 number/sec.
Furthermore, in order to estimate the intensity of the muons irradiated from the generated protons, the following assumptions were made: [0075] Proton/negative muon conversion coefficient: 0.1 (10%), and [0076] Muon transport efficiency: 0.01 (1%)
As a result, the number of negative muons that can be irradiated is 3.110.sup.160.10.01=3.110.sup.13 counts/sec. As mentioned above, all muons are able to stop at the target material. In addition, it is assumed that nuclear absorption takes place for all negative muons from the 1 s state (P.sub.NC=1.0) and nuclear transmutation occurs according to the probability of generating nuclei (reaction coefficient) via reaction branching.

[0077] Furthermore, the following assumptions were made regarding the branching ratio: [0078] (.sup., v) reaction: 10%, [0079] (.sup., n v) reaction: 50%, [0080] (.sup., 2n v) reaction: 20%, [0081] (.sup., 3 n v) reaction: 15%, and [0082] (.sup., 4n v) reaction: 5%.
That is, 10% of the NMCR was assumed to be involved in .sup.99Mo production. Note that the actual branching ratio is determined based on experiments. Furthermore, radioactivity of the generated nuclide after muon irradiation (unit: Bq) is given by

A.sub.RI(t.sub.irradiation)=(number of muons)(reaction coefficient)(1exp(0.693/T.sub.1/2t.sub.irradiation)).

Here, T.sub.1/2 is the half-life of the producing nucleus, and t.sub.irradiation is the muon irradiation time. Radioactivity after cooling is

A.sub.RI(t)(t.sub.cooling)=A.sub.RI(0)exp(0.693/T.sub.1/2t.sub.cooling).

Here, t.sub.cooling is the cooling time, and A.sub.RI (0) is the radioactivity when the muon irradiation is terminated.

[0083] 3-1-1-1. Example Estimation for Longer Irradiation Time than the Half-Life of .sup.99Mo

Under the above assumption, .sup.99Mo production was estimated for irradiation of 5.5 days by NMCR which is twice the half-life of .sup.99Mo (66.0 hours). As a result, .sup.99Mo is generated by 2.3310.sup.12 Bq (2.33 TBq). This corresponds to 63.0 Ci in the unit once used for this purpose. Also, the isotope ratio of Mo after irradiation for 5.5 days is [0084] .sup.95Mo: 0.55%, [0085] .sup.96Mo: 16.50%, [0086] .sup.97Mo: 22.00%, [0087] .sup.98Mo: 55.01%, and [0088] .sup.99Mo: 5.95%.

[0089] The production volume per muon beam channel for generating NMCR under this condition will be described. .sup.99Mo produced by 5.5-day irradiation with muon is 7.9710.sup.17 atoms, and its radioactivity is 2.3310.sup.12 Bq (2.33 TBq, 63.0 Ci). We assume that 82.4% of the .sup.99Mo nuclei decays into .sup.99mTc nuclei. Since the decay constant of .sup.99mTc is 3.19810.sup.5 for its half-life of 6.02 hours, its radioactivity is 2.1010.sup.13 Bq (568 Ci). Assuming that the loss due to subsequent ion separation and recovery, pharmaceutical manufacture, transportation, radiation equilibrium, milking operation, etc. is 50%, the radioactivity of .sup.99mTc that can be used as nuclear medicine RI is 1.0510.sup.13 Bq (284 Ci). In this context, the amount for a dose is about 740 MBq (20 mCi). If this value is adopted, it is concluded that an amount corresponding to 14,200 doses can be manufactured for 5.5-day muon irradiation. If this process is carried out without any breaks, the amount that can be produced and used in one year is estimated by multiplying it 365/5.5=66.4 times, leading to 6.9710.sup.14 Bq (18.8 kCi), which in turn corresponds to the number of administrations of about 940,000 doses. For example, the total amount of .sup.99mTc used in nuclear medicine diagnosis in Japan is 900,000 per year (NPL4). Therefore, the demand of that scale can be covered with 1 muon beam channel. In addition, the .sup.99Tc raw material consumption amounts to 2.4 mg if it is calculated based on the consumption in 5.5-day irradiation. In reality, the amount of .sup.99Tc solid target having a required size (area and thickness) for efficiently stopping the negative muon on the .sup.99Tc solid target to produce .sup.99Mo is about 25 g. This amount of .sup.99Tc can be easily obtained from raw materials described below.

[0090] Next, specific radioactivity of .sup.99Mo obtained when the irradiation time is longer than the half-life of .sup.99Mo in this embodiment will be described. The specific radioactivity is a measure of radioactivity of .sup.99Mo in a certain amount of Mo (for example 1 g). After muon irradiation is performed for the 5.5 days as mentioned above, the content of .sup.99Mo in the produced Mo is 5.95%. As a result, 0.0595 g of .sup.99Mo is present in 1 g of Mo, whose number N of .sup.99Mo is calculated by using the mass number of Mo, where the mass number calculated from the isotope ratio of Mo produced is 97.50, in the following manner:

N=0.0595/97.506.0210.sup.23=3.6710.sup.20/g-Mo.

Based on this value, half-life of .sup.99Mo, T.sub.1/2=66.0 h, and the decay constant of the .sup.99Mo=0.693/(66.03600)=2.9210.sup.6 (sec.sup.1), the specific radioactivity R of .sup.99Mo is calculated as follows:

[00001]$\begin{array}{c}r=.Math..Math..Math.N=1.07{10}^{15}.Math..Math.\mathrm{Bq}.Math.\text{/}.Math.g.Math.\text{-}.Math.\mathrm{Mo}\\ =.Math.1.Math.\text{,}.Math.070.Math..Math.\mathrm{TBq}.Math.\text{/}.Math.g.Math.\text{-}.Math.\mathrm{Mo}.\end{array}$

The specific radioactivity values to be compared are 370 TBq/g-Mo for specific radioactivity of .sup.99Mo obtained by the nuclear fission method and 0.074 TBq/g-Mo for the specific radioactivity of .sup.99Mo obtained in a neutron activation method targeting natural Mo as another method (NPL4). In other words, specific radioactivity of about 2.9 times the value of specific radioactivity of Mo obtained from the nuclear fission method is expected for .sup.99Mo produced by NMCR by selecting a radionuclide for the target nuclide. Thus, it can be concluded that it will show high usefulness in supplying a sufficient amount of .sup.99Mo, even using a small alumina column, as an example.

[0091] 3-1-1-2. Example Estimation for Shorter Irradiation Time than the Half-Life of .sup.99Mo

A condition under which the specific radioactivity and production amount of .sup.99Mo are increased more efficiently than the above estimation is one that NMCR irradiation is performed for 1.0 day, which is about of half-life (66.0 hours) of .sup.99Mo. The resultant .sup.99Mo amounts to 6.9110.sup.11 Bq (0.691 TBq, 18.7 Ci). The isotope ratios of Mo after the irradiation for 1.0 day are: [0092] .sup.95Mo: 0.53%, [0093] .sup.96Mo: 15.90%, [0094] .sup.97Mo: 21.20%, [0095] .sup.98Mo: 53.00%, and [0096] .sup.99Mo: 9.37%.
Even when the irradiation time is set to 1.0 days, the same calculation as in the case of irradiation for 5.5 days is carried out to estimate specific radioactivity and production amount. The results are indicated in Table 2 in comparison with the irradiation value of 5.5 days.

TABLE-US-00002 TABLE 2 NMCR Irradiation Time 5.5 days 1.0 day .sup.99Mo Concentration, after the 5.95% 9.37% Irradiation Time, per a Muon Beam Channel .sup.99Mo Production Amount, 2.33 10.sup.12Bq (62.2 Ci) 6.91 10.sup.11Bq (18.7 Ci) 7.97 10.sup.17 atoms 2.37 10.sup.17 atoms .sup.99mTc Radioactivity, 2.10 10.sup.13Bq (568 Ci) 6.24 10.sup.12Bq (169 Ci) .sup.99mTc Available Amount, 1.05 10.sup.13Bq (284 Ci) 3.12 10.sup.12Bq (84 Ci) 14,200 doses 4,200 doses .sup.99mTc Yearly Production 6.97 10.sup.14Bq/y (18.8 kCi/y) 1.14 10.sup.15Bq/y (30.8 kCi/y) 940k doses/y 1.54M doses/y .sup.99Tc Raw Material Consumption 2.4 mg 0.44 mg .sup.99Mo Specific Radioactivity 1,070TBq/g-Mo 1,690TBq/g-Mo (2.9 times Nuclear Fission (4.6 times Nuclear Fission Method) Method) y: year, T: 10.sup.12, M: 10.sup.6, k: 10.sup.3

[0097] In other words, compared with .sup.99Mo concentration of 5.95% for 5.5-day NMCR irradiation as described above, it is 9.37% for 1.0-day, or 1.57 times the .sup.99Mo concentration of 5.5-day irradiation. With respect to the production amount per muon beam channel, .sup.99Mo that can be manufactured for 1.0 day muon irradiation is 2.3710.sup.17 atoms, whose radioactivity amounts to 6.9110.sup.11 Bq (0.691 TBq, 18.7 Ci). To reflect the fact that the number of manufacturing processes in one year will increase for repetitive manufacturing processes than in the case of 5.5-day irradiation, the production volume in the case of 1.0-day irradiation is multiplied by 5.5, which yields generated radioactivity increase of about 1.64 times. After all, 1.0-day irradiation enables .sup.99mTc production of 1.1410.sup.15 Bq (30.8 kCi) per year. The number of doses corresponding to this amount is about 1.54 million. The material consumption of .sup.99Tc required is calculated to be 0.44 mg by calculating the corresponding amount of .sup.99Tc for 1.0-day irradiation. This amount can be easily obtained from raw materials described below as in the case of irradiation for 5.5 days.

[0098] The specific radioactivity R of .sup.99Mo is calculated from the content of .sup.99Mo (9.37%) in the produced Mo when irradiated with muon for 1.0 days by the same calculation as

[00002]$\begin{array}{c}R=.Math..Math..Math.N=1.69{10}^{15}.Math..Math.\mathrm{Bq}.Math.\text{/}.Math.g.Math.\text{-}.Math.\mathrm{Mo}\\ =.Math.1.Math.\text{,}.Math.690.Math..Math.\mathrm{TBq}.Math.\text{/}.Math.g.Math.\text{-}.Math.\mathrm{Mo}.\end{array}$

For .sup.99Mo produced by NMCR for 1.0 days by selecting a radionuclide for the target nuclide, we can expect specific radioactivity of about 4.6 times the specific radioactivity value of .sup.99Mo obtained from the nuclear fission method. Even under this irradiation condition, a sufficient amount of .sup.99Mo is supplied, so it can be said that it has high usefulness.

[0099] As described above, the method for generating .sup.99Mo from .sup.99Tc, a radionuclide, by nuclear transmutation through NMCR in the present embodiment shows sufficient practicability in the cases of shorter- and longer-MNCR irradiation time in comparison with .sup.99Mo half-life, and it is preferable in the method to set the MNCR irradiation time shorter than the half-life of .sup.99Mo.

[0100] In the following, the target material for obtaining .sup.99Tc for production of .sup.99Mo will be described, and the method for recovering .sup.99Mo will also be described. Among the isotopes of Tc, .sup.99Tc, which is useful as a target nuclide, is an artificial radionuclide and thus it is necessary to artificially manufacture it. There are two promising candidates for the .sup.99Tc source. One is a high-level radioactive waste obtained by reprocessing of spent nuclear fuel, and the other is a recycled material.

[0101] 3-1-2. Production of .sup.99Mo from .sup.99Tc in High Level Radioactive Waste

A high-level radioactive waste contains .sup.99Tc at a certain rate of 1 kg per ton, and it is easy to isolate Tc from other metallic elements after the partitioning mentioned above followed by an additional processing. Also, in the situation after using UO.sub.2 fuel at the burnup rate of 45 GWd/tHM in the pressurized water reactor (PWR) and cooling thereafter for 5 years, .sup.99Tc concentration in Tc isotopes contained in the spent nuclear fuel is 100 percent, i.e., Tc isotope of the other mass number is not included (NPL5). The .sup.99Tc is an LLFP with a half-life of about 210,000 years. In addition, the .sup.99Mo-.sup.99mTc generator can be produced from .sup.99Tc, a radionuclide of Tc in the spent nuclear fuel, by way of NMCR in accordance with the above principle.

[0102] The process of producing a .sup.99Mo-.sup.99mTc generator from a high level radioactive waste in spent nuclear fuel includes a step of extracting .sup.99Tc from a high level radioactive waste in the first place, a step of producing .sup.99Mo by NMCR in the second place, and a step of producing a .sup.99Mo-.sup.99mTc generator in the third place.

[0103] In the step of extracting .sup.99Tc from a high level radioactive waste, any chemical treatment and physical treatment can be adopted for implementing the present embodiment. One example of a method for separating nuclides currently being studied for a high-level radioactive wastes is a method called partitioning. For the partitioning, wet method (method using nitric acid) and some dry methods can be adopted. Here, for an example of the wet method, specific description will be set forth below based on the 4-group partitioning process (NPL1 and NPL2). A high level radioactive liquid waste, which is a high level radioactive waste, already contains nitric acid. Pretreatment is carried out by acting formic acid on it (denitration). Solvent extraction is carried out by applying a DIDPA (diisodecylphosphoric acid) solvent to the solution from which the precipitate has been removed. In the solvent extraction, when a solvent and an aqueous solution are placed in an identical container, elements which migrate from the aqueous solution layer to the solvent layer and which do not migrate can be separated, allowing the extraction. Among them, raffinate, a component remaining in the aqueous solution layer and does not migrate to the solvent layer, contains Tc. The raffinate is further reacted with formic acid and heated to precipitate (denitration precipitation). This precipitate is a group of partitioned 4 groups and contains Tc and platinum group. The other groups may be contained in each component separated so far or will be separated by additional operation. There is no particular obstacle to implementation by those skilled in the art.

[0104] From the Tc and the platinum group of the precipitate obtained in the denitration precipitation step, by further acting hydrogen peroxide (H.sub.2O.sub.2) on the dissolution, Tc can be dissolved in the aqueous solution with high yield and can be separated from the platinum group elements (Ru, Pd, Rh) (NPL3).

[0105] Next, .sup.99Mo is produced from .sup.99Tc by NMCR. For this purpose, a mixture of .sup.99Mo and Mo having stable nuclei can be produced by NMCR according to the reaction mode described above. In the case of using a high-level radioactive waste as a raw material, a process of elution with the nitric acid, the partitioning mentioned above, and the like allows us to employ .sup.99Tc aqueous solution (aqueous solution containing .sup.99TcO.sub.4.sup. ion) extracted therefrom for the target material. By irradiating the muon for a time determined based on the half-life of .sup.99Mo, it is possible to generate .sup.99Mo efficiently. This irradiation time is, for example, 5.5 days which is twice the half-life (66 hours) of .sup.99Mo, or 1.0 days which is of the half-life.

[0106] Specifically for NMCR and recovery and collection of generated .sup.99Mo, any method by one of the present inventors and disclosed in PTL1 can be adopted. For example, ion containing .sup.99Mo (hereinafter referred to as .sup.99Mo ion), typically .sup.99MoO.sub.4.sup.2 produced, is absorbed onto an ion exchange column (alumina column) and separated from a substance containing the ion having .sup.99Tc, such as .sup.99TcO.sub.4.sup. (.sup.99Tc ion) for recovery. FIG. 5 is an explanatory diagram illustrating a schematic configuration of a manufacturing apparatus 1200 for manufacturing .sup.99Mo by NMCR adopting a liquid raw material. In this method that employs a liquid material, Mo containing .sup.99Mo produced by NMCR is collected in the form of MoO.sub.4.sup.2 ion onto a column 1212A in a line 1210A or a column 1212B in a line 1210B. The columns at this time are selected to be an adsorption column or an ion-exchange column. Onto the columns 1212A and 1212B, MoO.sub.4.sup.2 is collected but .sup.99TcO.sub.4.sup. is not captured due to the difference in electric charge. The alumina column adsorbs ions by electrostatic action, where Mo ion (MoO.sub.4.sup.2) is more significantly adsorbed in comparison with .sup.99mTc nuclide or .sup.99Tc nuclide, which takes form of .sup.99mTcO.sub.4.sup. or .sup.99TcO.sub.4.sup. respectively. Therefore, the generated Mo can be efficiently collected while preventing .sup.99Tc from being mixed. That is, if the liquid target material 1202, which is an aqueous solution containing Mo ions at the irradiation position of the muon, is irradiated with the muon beam MB while circulating the liquid flow LS in the circulation path by an appropriate pump 1220, generated Mo ion is collected from the substance of the liquid flow LS discharged from the irradiation position due to the flow (the irradiated fluid) when it passes through the columns 1212A and 1212B. Furthermore, it is preferable that the .sup.99Tc ion will be reloaded, or repositioned into the irradiation position of the muon beam for the target nuclide of the muon due to the significant utilization efficiency of the raw material, even if there remains .sup.99Tc ion in the liquid flow LS that is not collected by the columns 1212A and 1212B. In this method adopting a liquid material, it is useful to arrange a plurality of columns such as Tc ion collecting columns 1212A and 1212B to make each other a standby system. It is to be noted that the frequency of replacing the columns 1212A and 1212B and recovering the Mo ions adsorbed thereon corresponds to the NMCR irradiation time in the case that .sup.99Mo is produced in the form of Mo ions by the manufacturing apparatus 1200. Estimates for this frequency to be once for every 5.5 days and for 1.0 days have been described in the section 3-1-1.

[0107] Thus, even if .sup.99Mo is produced in the target raw material containing Tc ion (.sup.99TcO.sub.4.sup.), .sup.99Mo can be easily separated and collected from the target raw material by letting the columns 1212 A and 1212B such as an alumina column absorb .sup.99Mo ion (.sup.99MoO.sub.4.sup.2). In addition, these columns themselves can be made of the same material as the column adopted for the .sup.99mTc generator. This collection process is a totally inverted operation of an operation in the .sup.99Mo-.sup.99mTc generator where .sup.99Mo ions are absorbed to the ion exchange column (alumina column) and .sup.99mTc ions generated by decay are eluted off by milking. Although stable nuclei .sup.95Mo-.sup.98Mo can also be generated at the time of NMCR and even the Mo is mixed with .sup.99Mo, the purity of .sup.99mTc eluted by the milking operation of .sup.99Mo-.sup.99mTc generator is not affected.

[0108] As another production example, a technique of a batch production process 1400 that employs a target material containing .sup.99Mo can also be employed. When irradiating muon continuously, the half-life of .sup.99Mo of 66 hours is not an obstacle. FIG. 6 is an explanatory diagram illustrating the outline of the process of producing .sup.99Mo by the batch manufacturing process 1400 by NMCR. Target raw materials that can be adopted this time are Tc.sub.2O.sub.7 solids or pertechnetate aqueous solutions, a certain amount of which is contained in one of appropriate containers 1404A to 1404D. The process of irradiating a unit amount of the target material 1402 with a predetermined dose of the muon beam MB for the batch processing is not only suitable for sequential processing by replacing the target material 1402 in each container, such as the containers 1404A to 1404D, but also able to be automated by using a transfer apparatus. For a unit amount of irradiated solid or liquid, necessary steps for separation process, formulation or the like can be performed thereafter. In the actual process, muonic atom X-rays and -e decay electrons can be measured from the outside, and muon incident energy can be optimized.

[0109] Contamination of the external environment is less likely to occur in batch processing, and there is an advantage that the outer container becomes a protective container for transport without any modification. It is also useful from a practical point of view as a method for manufacturing radioactive materials to process by NMCR while a substance having radioactivity is encapsulated in a container. For example, it is possible to transport .sup.99Mo after NMCR while sealing it as much as possible up until it is recovered or separated from the target material. In this embodiment in which all of the target nuclide .sup.99Tc, .sup.99Mo after generation, and .sup.99mTc after generation are radioactive substances, the practicality of the production of radioactive substances by NMCR in batch processing is high from the viewpoint of radiation protection. .sup.99mTc ions generated during transport can be easily chemically separated from .sup.99Mo ions. The target material 1402 inside the containers 1404A to 1404D in FIG. 6 can be either solid or liquid. Furthermore, in the batch manufacturing process 1400, although only a single container (here, the container 1404B) is indicated for a target of NMCR at a time, it is possible to make various modifications such as simultaneous irradiation to a plurality of containers 1404 by distributing muon beams according to implementation requirements.

[0110] FIG. 7 is a schematic chart illustrating outline of a process 1600 for further processing the product obtained by the batch processing, indicating cases where a solid .sup.99Tc target is adopted and a liquid .sup.99Tc target is adopted, based on the example of a product containing Mo ions. In the processing process 1600, when a solid .sup.99Tc target 1612 is used for the muon beam MB irradiation, an aqueous solution 1620 is prepared by dissolving the solid .sup.99Tc target after the irradiation in which .sup.99Mo is generated with an appropriate acid or the like to generate .sup.99Mo-.sup.99Tc ion. For ease of dissolution, it is preferable that the solid .sup.99Tc target 1612 has been made into a fine powder in advance. In the case the liquid .sup.99Tc target 1614 is used for the muon beam MB irradiation, an aqueous solution corresponding to that aqueous solution has been used from the muon irradiation step and is adopted as it is. In order to separate .sup.99Tc ions and .sup.99Mo ions from each other in the aqueous solution 1620 and the liquid .sup.99Tc target 1614, it is convenient to adopt an ion separation column 1630 such as an alumina column similar to FIG. 5. As a result, .sup.99Mo ion is collected onto the ion separation column 1630 and .sup.99Tc ion passes through it while staying in the aqueous solution. The aqueous solution 1640 containing the .sup.99Tc can be recycled as a liquid .sup.99Tc target, or a solid .sup.99Tc target can be produced therefrom by appropriate chemical treatment or physical treatment for reuse. Since .sup.99Mo is adsorbed onto the ion separation column 1630 at this stage, it is useful to utilize itself for a .sup.99Mo-.sup.99mTc generator. In addition, when it is necessary to release .sup.99Mo from the ion separation column, it is possible to elute .sup.99Mo from the ion separation column to obtain an aqueous solution 1660 in a form of .sup.99MoO.sub.4.sup.2 ion, or in another form containing .sup.99Mo, by letting an eluent 1650 (e.g., aqueous sodium hydroxide solution) pass through an ion separation column. it is possible to make the obtained .sup.99Mo into another chemical form suitable for further processing, or it is possible to employ any known chemical manipulation or physical manipulation method already known in the art.

[0111] The product containing Mo ions obtained by the batch processing can be separated by a precipitation method or coprecipitation method in addition to the ion exchange method. The precipitation method (and coprecipitation method) that can be adopted is similar to the method used for ordinary chemical separation. It is possible to make the obtained .sup.99Mo into another chemical form suitable for further processing, or it is possible to adopt any known chemical manipulation or physical manipulation method already known in the art.

[0112] Once a necessary amount of .sup.99Mo is obtained for .sup.99Mo-.sup.99mTc generator from a high level radioactive waste, it is no longer necessary to operate the nuclear reactor using HEU to obtain .sup.99Mo for the sake of .sup.99Mo-.sup.99mTc generator. In this regard, the present method of using a high-level radioactive waste as a raw material greatly contributes to the establishment and maintenance of the supply chain of .sup.99Mo-.sup.99mTc generator.

[0113] 3-1-3. Manufacture of .sup.99Mo from Recycled Raw Material of .sup.99Tc

In the process of actually producing .sup.99mTc and .sup.99Mo, which amount to the most part of the nuclides used in nuclear medicine applications, .sup.99Tc can be easily obtained from a by-product in the manufacturing process of the generator for milking, unused chemicals after formulating .sup.99mTc, and used generators per se. There is no particular difficulty in using such .sup.99Tc for a target nuclide in the present embodiment. A recycled raw material containing .sup.99Tc in the present embodiment means, among substances containing .sup.99Tc, .sup.99Tc that is produced as a by-product in an arbitrary step up until .sup.99Mo-.sup.99mTc generator is produced, .sup.99Tc obtained by leaving unused drugs after formulation, or .sup.99Tc produced by radioactive decay in the .sup.99Mo-.sup.99mTc generator. Since .sup.99Tc of the nuclide as a raw material needs to be artificially obtained by some method, the substance containing .sup.99Tc should be substantially one manufactured in connection with the .sup.99Mo-.sup.99mTc generator, except for the radioactive waste mentioned above. The .sup.99Tc in the case of a nuclide of the recycled raw material means .sup.99Tc as explained above, while excluding .sup.99Tc that has been produced via .sup.99mTc once produced for .sup.99Mo-.sup.99mTc generators and then administered to the human body or the like for their inherent purpose. The manufacturing process for producing .sup.99mTc that lead to a .sup.99Tc nuclide used for the recycled material can be carried out not only by the conventional method but also by the method of any of the embodiments, though it may not be limited specifically. For example, the aqueous solution 1640 containing .sup.99Tc after passing through the ion separation column illustrated in FIG. 7 is an example of the recycled material.

[0114] FIG. 4 is a decay scheme diagram among nuclides with a mass number A=99 including .sup.99mTc. FIG. 4 indicates .sup.99Tc (ground state with nuclear spin J=9/2+) obtained through .sup.99mTc. Since .sup.99mTc becomes .sup.99Tc with a half-life of 6.02 hours, .sup.99Tc is inevitably generated when handling .sup.99Mo-.sup.99mTc generator. Therefore, .sup.99Tc for the recycled raw material can be obtained at any process for producing .sup.99Mo-.sup.99mTc generator, at the site where .sup.99mTc is actually used, or from the used .sup.99Mo-.sup.99mTc generators that have been returned for storage. Both .sup.99Mo and .sup.99mTc manufactured for .sup.99Mo-.sup.99mTc generators for medical purposes are usually radioactive materials that are subject to radioactive control, though .sup.99Tc derived from these nuclides exhibits low radioactivity. For this reason, control of .sup.99Tc manufactured for medical purpose is maintained and most of it is returned. In this embodiment, since .sup.99Tc exhibiting radioactivity is used for a target nuclide for NMCR, any material including such .sup.99Tc can be adopted as a target material.

[0115] In carrying out the present method using recycled raw materials, there is no need to newly obtain nuclides by the nuclear fission method or nuclides from the high-level waste as in the present embodiment. Therefore, looking at the entire supply chain of .sup.99Mo-.sup.99mTc generators for medical purposes, if this method of using recycled raw materials is implemented, the necessity of implementing a method to handle a high level radioactive waste is mitigated, though it cannot be totally removed. In this regard, the present method using recycled materials greatly contributes to the establishment and maintenance of the supply chain of .sup.99Mo-.sup.99mTc generators.

[0116] 3-2. Production of .sup.133Xe from .sup.134Cs, .sup.135Cs, and .sup.137Cs

It is possible to manufacture .sup.133Xe from Cs raw material containing .sup.135Cs, .sup.137Cs, which are LLFPs contained in a high level radioactive waste. FIG. 8 is an explanatory diagram illustrating nuclear reactions on the chart of nuclides in which Xe having a mass number of 133 is generated by NMCR. In the production of .sup.133Xe, Xe gas containing .sup.133Xe is separated and recovered for use as nuclear medicine RI. It should be noted that .sup.133Xe is used for pulmonary function test and cerebral blood flow test. FIG. 9 is a decay scheme diagram between Xe and Cs of a mass number of A=133. In a typical nuclear medicine application of .sup.133Xe, gamma rays at 81 keV are measured with SPECT. The amount for a dose is about 370 MBq (10 mCi). Also, as an example, Japan's demand of .sup.133Xe is covered by imports. .sup.135Cs has a half-life of 2.310.sup.6 years, and .sup.137Cs has a half-life of 30.08 years. We also considered .sup.134Cs which is not an LLFP but has a half-life of 2.06 years. In particular, .sup.135Cs, .sup.137Cs are contained, 0.5 kg and 1.5 kg respectively in 1 ton of a high-level radioactive waste.

[0117] In the situation after UO.sub.2 fuel is used at a burnup of 45 GWd/tHM in a pressurized water reactor (PWR) and after cooling for 5 years, the ratio of isotopes of Cs contained in spent nuclear fuel is [0118] .sup.133Cs: 42.1%, [0119] .sup.134Cs: 1.02%, [0120] .sup.135Cs: 14.8%, [0121] .sup.136Cs: 0.0%, and [0122] .sup.137Cs: 42.1%
(NPL5). Note that the natural abundance ratio of Cs is 100%.sup.133Cs.

[0123] The reaction modes of NMCR with .sup.137Cs as the target nuclide are as follows: [0124] .sup.137Cs (.sup., v) .sup.137Xe, [0125] .sup.137Cs (.sup., n v) .sup.136Xe, [0126] .sup.137Cs (.sup., 2n v) .sup.135Xe, [0127] .sup.137Cs (.sup., 3n v) .sup.134Xe, and [0128] .sup.137Cs (.sup., 4n v) .sup.133Xe.
Also, those with .sup.135Cs as the target nuclide are as follows. [0129] .sup.135Cs (.sup., v) .sup.135Xe, [0130] .sup.135Cs (.sup., n v) .sup.134Xe, [0131] .sup.135Cs (.sup., 2n v) .sup.133Xe, [0132] .sup.135Cs (.sup., 3n v) .sup.132Xe, and [0133] .sup.135Cs (p, 4n v) .sup.131Xe.
Of generated Xe Isotopes .sup.136Xe, .sup.134Xe, .sup.132Xe, .sup.131Xe are stable nuclei, but .sup.137Xe decays by .sup. decay into .sup.137Cs with a half-life of 3.83 minutes, .sup.135Xe decays by .sup. decay into .sup.135Cs with a half-life of 9.10 hours, and .sup.133Xe decays by .sup. decay into .sup.133Cs (stable) with a half-life of 5.25 days. Since a neutron emission level is found for .sup.137Xe and a phenomenon that the neutron absorption cross section becomes huge (phenomenon known as xenon override in power control of the nuclear reactor) may occur for .sup.135Xe, it is possible that probability of generation of .sup.136Xe increases.

[0134] The reaction coefficient of each Cs isotope generated by NMCR was calculated based on the values for the reaction coefficient mentioned above. The beam condition and reaction branching ratio at that time were assumed to be identical to those for .sup.99Tc. The produced Xe isotopes have mass numbers ranging from 129 to 137, and Xe of each mass number is generated from Cs having a different mass number. The radioactive Xe nuclide having a relatively long half-life included in the remaining Xe gas is .sup.133Xe only. Gas containing .sup.133Xe is separated and recovered for use as nuclear medicine RI.

[0135] The reaction coefficients were calculated using all combinations of the Cs isotopes and Xe isotopes mentioned above. For example, the reaction modes leading to the target .sup.133Xe are: [0136] .sup.133Cs (.sup., v) .sup.133Xe, [0137] .sup.134Cs, (.sup., n v) .sup.133Xe, [0138] .sup.135Cs (p, 2n v) .sup.133Xe, and [0139] .sup.137Cs (p, 4n v) .sup.133Xe.
The nuclear reaction diagram is illustrated in FIG. 8. Note that we omit those other than of a mass number of 133. Distribution among isotopes of the reaction coefficients for Xe of each mass number estimated based on the reaction branching ratio and abundance ratio was obtained as follows: [0140] .sup.129Xe: 0.0211, [0141] .sup.136Xe: 0.0637, [0142] .sup.131Xe: 0.0931, [0143] .sup.132Xe: 0.2347, [0144] .sup.133Xe: 0.0978, [0145] .sup.134Xe: 0.1382, [0146] .sup.135Xe: 0.0990, [0147] .sup.136Xe: 0.2105, and [0148] .sup.137Xe: 0.0421.

[0149] 3-2-1. Process (Outline)

The process of producing .sup.133Xe by NMCR from .sup.134Cs, .sup.135Cs, and .sup.137Cs in a high level radioactive waste is carried out by the following three steps: [0150] Step 1: muon irradiation, [0151] Step 2: cooling (first), and [0152] Step 3: cooling (second).
In Step 1, muon irradiation is performed on the target material of Cs containing .sup.134Cs, .sup.135Cs, and .sup.137Cs for 5.5 days (about 1 half-life of .sup.133Xe). The radioactivity at that time is estimated as: [0153] .sup.133Xe (5.25 days): 1.5710.sup.12 Bq, [0154] .sup.135Xe (9.10 hours): 3.0710.sup.12 Bq, and [0155] .sup.137Xe (3.83 min): 1.3110.sup.12 Bq.
Note that Xe of other mass numbers are stable, exhibiting no radioactivity, though they are generated according to their reaction coefficients.

[0156] In Step 2, as the first cooling, 1 hour cooling is carried out after muon irradiation. At this time, the half-life of .sup.137Xe is 3.83 minutes, thus the 1 hour cooling period corresponds to 15.7 half-lives. With so much time, the majority of .sup.137Xe decays by .sup. decays into .sup.137Cs (LLFP). The .sup.137Cs can be separated and recovered in an aqueous solution. The ratio of the number of isotopic atoms of Cs on completion of Step 2 is: [0157] .sup.133Cs: 33.7%, [0158] .sup.134Cs: 0.0%, [0159] .sup.135Cs: 63.5%, [0160] .sup.136Cs: 0.0%, and [0161] .sup.137Cs: 2.6%.
In terms of the radioactivity ratio, .sup.137Cs accounts for 100%.

[0162] In Step 3, the second cooling is performed for a longer period (for example, 4 days). The period of 4 days corresponds to 10.5 half-lives of .sup.135Xe. Since the half-life of .sup.135Xe is 9.10 hours, most part of it decays by .sup. decay into .sup.135Cs (LLFP) in the end. The .sup.135Cs can be separated and recovered in an aqueous solution. The ratio among the number of isotopic atoms on completion of Step 3 is: [0163] .sup.133Cs: 74.3%, [0164] .sup.134Cs: 0.0%, [0165] .sup.135Cs: 25.7%, [0166] .sup.136Cs. 0.0%, and [0167] .sup.137Cs: 0.0%.
In terms of the radioactivity ratio, .sup.135Cs accounts for 100%. Likewise, the ratio of the number of isotopic atoms of Xe on completion of Step 3 is: [0168] .sup.129Xe: 2.63%, [0169] .sup.130Xe: 7.94%, [0170] .sup.131Xe: 11.60%, [0171] .sup.132Xe: 29.25%, [0172] .sup.133Xe: 5.11%, [0173] .sup.134Xe: 17.23%, [0174] .sup.135Xe: 0.00%, and [0175] .sup.136Xe: 26.24%.
.sup.133Xe content in Xe gas is 5.11%. Furthermore, the radioactivity ratio of .sup.133Xe accounts for 99.8%, and the radioactivity is 9.2310.sup.11 Bq (24.9 Ci).

[0176] Using the mass number (133.00) calculated from the isotopic distribution of generated Xe, the number of .sup.133Xe in 1 g of generated Xe is calculated to be 2.3110.sup.20/g-Xe. Furthermore, using the half-life of .sup.133Xe (T.sub.112=5.25 days), the specific radioactivity of .sup.133Xe becomes 353 TBq/g-Xe. The specific radioactivity to be compared is 370 TBq/g-Mo of the specific radioactivity of .sup.99Mo obtained by the nuclear fission method (NPL4).

[0177] The production volume of .sup.133Xe per muon channel is 9.2310.sup.11 Bq (24.9 Ci). Since the amount for a dose to a patient is 370 MBq (10 mCi), the production volume corresponds to about 2,500 doses.

[0178] 3-2-2. Details of Process

For implementing the process of producing .sup.133Xe by NMCR from .sup.134Cs, .sup.135Cs, .sup.137Cs, two candidates seem promising: the same method as the batch production process 1400 for .sup.99Mo illustrated in FIG. 6, and an on-line manufacturing method. For both cases, liquid targets and solid targets containing Cs are adopted. Solids that can be adopted as the solid target are listed with brief annotations on characters and properties: [0179] cesium hydroxide (CsOH, colorless, hygroscopic), [0180] cesium carbonate (Cs.sub.2CO.sub.3, white powder), [0181] cesium nitrate (CsNO.sub.3, white solid, water insoluble), and [0182] cesium chloride (CsCl, solid).
In the form of these simple substances or mixtures, solids containing .sup.134Cs, .sup.135Cs, .sup.137Cs can be extracted from a high level radioactive waste. On the other hand, typical liquid targets are listed with solubility in the following: [0183] cesium hydroxide (CsOH, solubility 395 g/100 cm.sup.3, 15 C.), [0184] cesium carbonate (Cs.sub.2CO.sub.3, solubility 260.5 g/100 cm.sup.3, 15 C.), and [0185] cesium chloride (CsCl, solubility 162 g/100 ml).
Liquid targets can also be extracted from a high level radioactive waste as simple substances or mixtures in the form of an aqueous solution containing .sup.134Cs, .sup.135 Cs, and .sup.137Cs ions.

[0186] In the case of the batch processing, a typical solid target or liquid target is irradiated with muons as a target material. Apparatus configuration for this process is almost the same as that of FIG. 6. The target raw material of cesium nitrate solid or cesium hydroxide aqueous solution is stored in a container (inner sealed container, not shown in FIG. 6). At this time, the remaining internal volume of the inner sealed container is replaced with high-purity helium gas. The inner sealed container is stored in a container 1404 (FIG. 6) which is an outer container, and muon irradiation is performed from the outside. As a result, the target .sup.133Xe gas can be obtained in the next step by separating the Cs ions and the rare gas of Xe. Muon incident energy can be optimized by measuring muonic atom X-rays and -e decay electrons. In this method, the outer container can be transported as it is to the next process while being used for a protective transportation container, which substantially prevents contamination of the external environment. In the method using the target material container, it is possible to irradiate sequentially by using a large number of target material containers, which is advantageous in that automation can be easy realized.

[0187] In the case of the on-line production method, muon irradiation for NMCR is carried out using a flow path for gas and liquid. FIGS. 10 and 11 are schematic configuration diagrams illustrating a processing apparatus for manufacturing .sup.133Xe by NMCR to be implemented in the present embodiment. FIG. 10 illustrates an irradiation processing apparatus 2200 for a liquid target material, and FIG. 11 illustrates an irradiation processing apparatus 2400 for a solid target material. On the liquid target 2210 in FIG. 10, the muon is irradiated by the same process as indicated for .sup.99Tc in FIG. 5. At that time, it is the sealed target container 2212 that corresponds to the liquid target material 1202. The sealed target container 2212 is filled with a liquid target 2210 together with helium gas to be a target of irradiation. During the muon beam MB irradiation, the valves V2 and V3 are closed and the valve V1 is kept open. A gas line 2214 is connected to the upper space of the sealed target container 2212 with its one end open, and the Xe gas liberated from the liquid target 2210 is collected from the headroom above the liquid level. The other end of the gas line 2214 is connected to the buffer tank 2220. The gas in the buffer tank 2220 is, via the gas line 2222, bubbled into the solution of the sealed aqueous solution trap 2240 by the gas line 2232, with the aid of the gas circulation pump 2230. From the above space of the liquid surface of the aqueous solution trap 2240, a path for bubbling into the liquid of the sealed target container 2210 through the gas line 2242 is established. In the aqueous solution trap 2240, an aqueous solution is stored from which Cs generated from Xe gas is to be recovered. As a result, Cs generated due to radioactive decay during circulation and Cs generated in the buffer tank 2220 is collected in the aqueous solution trap 2240. If the muon irradiation is continued with the gas circulation pump operated, the concentration of the Xe gas generated as a result of NMCR in the liquid target 2210 is increased in the helium gas while the recovery of Cs is continued in the aqueous solution trap 2240.

[0188] After the irradiation is completed, a suitable trap such as a liquid nitrogen trap 2280 is connected at the appropriate position along the path of the gas, then the valves V2 to V5 are opened while valve V1 is closed. Thereafter, by operating the gas circulation pump 2230, the Xe gas contained in the helium gas is collected into the liquid nitrogen trap 2280.

[0189] In the processing apparatus 2400 for targeting the solid Cs target of FIG. 11, an inner container 2414 that houses the solid Cs target 2410 is also positioned inside the sealed target container 2412 and is also filled with helium gas. A solid Cs target 2410 containing, .sup.134Cs, .sup.135Cs, .sup.137Cs and so on, which is a fine power, is contained in the inner container. This inner container 2414 is opened in the internal space of the sealed target container 2412, and the Xe gas liberated upon irradiation with muons is released to the inside of the sealed target container 2412. It is also preferred to have a temperature controller (e.g., heater 2416) for appropriately controlling the temperature of the solid Cs target to facilitate Xe gas release. The Cs generated by decay in the released Xe gas is collected in the aqueous solution trap 2240 according to the same method as in the case of the liquid Cs target.

[0190] In either of the liquid target and the solid target, it is not necessary to interrupt the irradiation of the muon, and the liquid nitrogen trap 2280 can be connected to recover the Xe gas in a timely manner. Thus, even when the muon beam intensity may become the rate-determining factor of the generation rate of .sup.133Xe, continuous irradiation can be performed to increase the production rate of .sup.133Xe.

[0191] It is useful to recover .sup.133Cs, .sup.135Cs, .sup.137Cs produced by decay in an aqueous solution during at least one of the two cooling periods. FIG. 12 is a schematic configuration diagram illustrating the configuration of the XeCs separation device 2800. In the first place, the liquid nitrogen trap 2280 used in the irradiation treatment apparatuses 2200 and 2400 (FIGS. 10 and 11) is connected to the XeCs separation device 2800. Then, the liquid nitrogen is removed and the temperature of the liquid nitrogen trap 2280 is raised, whereby the Xe gas trapped in the liquid nitrogen trap 2280 is evaporated. The Xe gas is then circulated in a path provided with a suitable buffer tank 2820 and a gas circulation pump 2830 using helium as a circulating gas. A Cs ion trap 2810 has been inserted in the path. Since the gas blown into the Cs ion trap 2810 via the gas line 2832, the gas circulation pump 2830, and the gas line 2834 contains Cs that has been generated by decay of radioactive Xe, dissolving that gas into the aqueous solution of the Cs ion trap 2810 enables the mechanism of the separation and recovery at issue to function. By continuing circulation along the path from the Cs ion trap 2810 and back to the liquid nitrogen trap 2280 via the gas line 2822, the buffer tank 2820, and the gas line 2824, Cs generated by decay will be removed during the cooling period. After recovering the Cs ions, Xe gas containing .sup.133Xe can be recovered by injecting liquid nitrogen again into the liquid nitrogen trap 2280.

[0192] 3-3. Production of .sup.89Rb-.sup.89Sr from .sup.90Sr

.sup.89Sr can be produced from Sr raw material containing .sup.90Sr which is an LLFP contained in a high level radioactive waste. In nuclear medical applications, .sup.89Sr is used as an internal therapeutic agent for pain relief in the case of painful bone metastasis, where it releases .sup. ray of maximum energy of about 1.49 MeV. It is a nuclide with a physical half-life of 50.5 days. FIG. 13 is an explanatory diagram illustrating the nuclear reaction on the chart of nuclides where Rb isotope is generated by NMCR adopting .sup.90Sr target. In addition, FIG. 14 is a decay scheme diagram between Sr and Y having a mass number A=89. .sup.89Sr is administered in the form of strontium chloride .sup.89SrCl.sub.2 and the like, and it is intravenously administered to an adult 2.0 MBq/kg for a dose (in the case of 70 kg patient: 1.410.sup.8 Bq (3.8 mCi)). However, it is up to 141 MBq. For example, Japan imports 100% of the demand for .sup.89Sr. In this embodiment, about 0.6 kg of .sup.90Sr, which is a target nuclide, is contained in 1 ton of the high-level radioactive waste.

[0193] In the situation after using UO.sub.2 fuel at a burnup of 45 GWd/tHM in a pressurized water reactor (PWR) and after cooling for 5 years, the ratio of isotopes of Sr contained in spent nuclear fuel is [0194] .sup.84Sr: 0.00%, [0195] .sup.85Sr: 0.00%, [0196] .sup.86Sr: 0.08%, [0197] .sup.87Sr: 0.00%, [0198] .sup.88Sr: 41.95%, [0199] .sup.89Sr: 0.00%, and [0200] .sup.90Sr: 57.97%
(NPL5). Incidentally, the natural abundance ratio of Sr is [0201] .sup.84Sr: 0.56%, [0202] .sup.85Sr: 0.00%, [0203] .sup.86Sr: 9.86%, [0204] .sup.87Sr: 7.00%, [0205] .sup.88Sr: 82.58%, [0206] .sup.89Sr: 0.00%, and [0207] .sup.90Sr: 0.00%.

[0208] 3-3-1. Process (Outline)

The process for producing .sup.89Sr from the Sr target material containing .sup.90Sr includes the following three steps: [0209] Step 1: muon irradiation of the target material of Sr containing [0210] Step 2: cooling of the separated and recovered Rb ions for 25 minutes, and [0211] Step 3: cooling of Rb ions for another 150 minutes.

[0212] 3-3-2. Details of Process

In Step 1, muon irradiation is performed on the target material of Sr containing .sup.90Sr for 90 minutes. Thereafter, after irradiating the muon, the Rb ion is separated and recovered from the Sr ion.

[0213] The reaction modes of NMCR using .sup.90Sr as the target nuclide are as follows: [0214] .sup.90Sr (.sup., v) .sup.90Rb (.sup. decays into .sup.90Sr with a half-life of 2.6 minutes), [0215] .sup.90Sr (.sup., n v) .sup.89Rb (.sup. decays into .sup.89Sr with a half-life of 15.2 minutes, [0216] .sup.89Sr then decays by .sup. decay into .sup.89Y with a half-life of 50.5 days), [0217] .sup.90Sr.sup. (.sup., 2n v) .sup.88Rb (.sup. decay into .sup.88Sr with a half-life of 17.8 minutes), [0218] .sup.90Sr (.sup., 3 n v) .sup.87Rb (stable, 4.810.sup.10 years), and [0219] .sup.90Sr (.sup., 4n v) .sup.86Rb (.sup. decays into .sup.86Sr with a half-life of 18.7 days).
The modes of nuclear reactions are understood from the chart of nuclides on FIG. 13. It should be noted that a Rb isotope is generated in the first place from each isotope of Sr and that, in the case NMCR occurs by targeting .sup.90Sr as a target nuclide and .sup.89Rb is generated thereafter, .sup.89Rb decays by .sup. decay into .sup.89Sr in a short time (half-life of 15.2 minutes), and the half-life of generated .sup.89Sr becomes about 50.5 days, as indiated by a dashed line on FIG. 13.

[0220] Assuming identical beam conditions and identical reaction branching ratio to those for .sup.99Tc and .sup.133Xe described above, the reaction coefficients from Sr isotopic proportion of spent nuclear fuel to each Rb are calculated to be: [0221] .sup.84Rb: 0.0210, [0222] .sup.85Rb: 0.0629, [0223] .sup.86Rb: 0.1129, [0224] .sup.87Rb: 0.2967, [0225] .sup.88Rb: 0.1579, [0226] .sup.89Rb: 0.2899, and [0227] .sup.90Rb: 0.05797.

[0228] In Step 1, muon irradiation is performed for 90 minutes toward Sr solid or aqueous solution target containing .sup.90Sr. This irradiation time of 90 minutes is six times the .sup.89Rb half-life (15.2 minutes). After muon irradiation, Rb ions are separated and recovered from Sr ions. Then the radioactivity of .sup.89Rb becomes about 8.8410.sup.12 Bq.

[0229] Next, in Step 2, Rb ions are cooled for 25 minutes. This period is ten times of the half-life of .sup.90Rb half-life (2.6 min). As a result, .sup.90Rb decays by .sup. decay into .sup.90Sr. .sup.90Sr is an LLFP. At this point, .sup.89Sr and .sup.88Sr, which are daughter nuclei of .sup.89Rb and .sup.88Rb respectively, are also mixed together. The ratio of radioisotopes (in atomic fraction) of Sr at this time is: [0230] .sup.84Sr: 0.0%, [0231] .sup.85Sr: 0.0%, [0232] .sup.86Sr: 0.09%, [0233] .sup.87Sr: 0.0%, [0234] .sup.88Sr: 35.3%, [0235] .sup.89Sr: 61.4%, and [0236] .sup.90Sr: 3.1%.
The radioactivity ratio is: [0237] .sup.89Sr: 100.0% and [0238] .sup.90Sr: 0.02%.

[0239] Furthermore, in step 3, Rb ions separated from Sr are cooled for 150 minutes. This period is 10 times the half-life of .sup.90Rb half-life (15.2 minutes). In addition, .sup.88Rb, .sup.86Rb, .sup.84Rb are included. Among these, .sup.88Rb decays by .sup. decay and becomes stable nucleus .sup.88Sr. .sup.86Rb and .sup.84Rb have a half-life of 18.7 days and 32.8 days respectively, and the numbers of decay are very small during cooling for 150 minutes. Sr ions are separated and recovered from the cooled Rb ions. The resulting .sup.89Sr can be used for nuclear medicine RI.

[0240] The isotope ratio (in atomic fraction) of Sr at this point is: [0241] .sup.86Sr: 1.1%, [0242] .sup.88Sr: 42.2%, [0243] .sup.89Sr: 56.7%, and [0244] .sup.90Sr: 0.008%.
The radioactivity ratio is [0245] .sup.89Sr: 100.0%, and [0246] .sup.90Sr: 0.00007%.
The radioactivity of .sup.89Sr generated according to the 90-minute irradiation is 5.9010.sup.8 Bq (15.9 mCi). At 1 day (24 hours), it becomes 9.4310.sup.9 Bq (255 mCi).

[0247] At the time when muon irradiation is carried out for 90 minutes and step 3 is completed, number N of .sup.89Sr in 1 g of produced Sr is given by

N=0.567/88.466.0210.sup.23=3.8610.sup.21/g-Sr,

where the mass number (88.46) calculated from the isotopic distribution of the generated Sr is used. If the half-life of .sup.89Sr: T.sub.1/2=50.5 days and decay constant of .sup.89Sr: =0.693/(50.5243600)=1.5810.sup.7 (sec.sup.1) are adopted for this calculation, the specific radioactivity R of .sup.89Sr is calculated as:

R=N=6.1010.sup.14Bq/g-Sr

=610TBq/g-Sr.

This specific radioactivity of the .sup.89Sr is about 1.6 times 370 TBq/g-Mo, a specific radioactivity of .sup.99Mo obtained by the nuclear fission method (NPL4).

[0248] The production volume of .sup.89Sr per day per muon channel is 9.4310.sup.9 Bq (255 mCi). Since the amount for a dose to a patient weighing 70 kg is 1.410.sup.8 Bq (3.8 mCi), the above-mentioned production amount per day corresponds to about 67 doses.

[0249] In relation to each step, a method for separating Sr ions from the Rb ions mentioned above will be described further. The separation method can be carried out by an ion exchange method and a precipitation method (or coprecipitation method). The ion exchange method is the same as the method for separating .sup.99Tc and .sup.99Mo described in FIG. 7. Since the Rb ion is a monovalent ion of an alkali metal and the Sr ion is a divalent ion of an alkaline earth metal, the same processing can be carried out by using an ion separation column utilizing the difference in ion valence and chemical properties. This also applies to the precipitation method.

[0250] The embodiments of the present invention have been concretely described above. Each of the above-described embodiments, specific examples, application examples, each theory and method of manufacturing for each nuclide are described for the purpose of explaining the invention, and the scope of the invention of the present application should be determined based on the claims. Also, modifications within the scope of the present invention including other combinations of the respective embodiments are also included in the scope of the claims.

INDUSTRIAL APPLICABILITY

[0251] The method for producing the radioactive substance of the present invention and the substance to be produced can be used for any test, apparatus, diagnostic and analytical method using a radioactive substance, and nuclear medicine application.

REFERENCE SIGNS LIST

[0252] 100 nuclear fuel cycle [0253] 10 uranium mine [0254] 12 uranium [0255] 20 fuel processing plant [0256] 22 fuel [0257] 30 nuclear power plant [0258] 32 spent nuclear fuel [0259] 34 low level radioactive waste [0260] 40 low level radioactive waste disposal facility [0261] 50 reprocessing plant [0262] 52 recovered uranium and plutonium [0263] 54 high level radioactive waste [0264] 60 high-level radioactive waste storage facility [0265] 70 high-level radioactive waste disposal facility [0266] 1200 manufacturing apparatus [0267] 1202 liquid target raw material [0268] 1210A, B system [0269] 1212A, B column [0270] 1220 pump [0271] 1400 batch manufacturing process [0272] 1402 target material [0273] 1404 container [0274] 1600 process of ion exchange processing [0275] 1612 solid .sup.99Tc Target [0276] 1620, 1640, 1650, 1660 aqueous solution [0277] 1614 liquid .sup.99Tc target [0278] 1630 ion separation column [0279] 2200, 2400 irradiation treatment device [0280] 2210 liquid target [0281] 2212 sealed target container [0282] 2214, 2222, 2232, 2242 gas line [0283] 2220, 2820 buffer tank [0284] 2230, 2830 gas circulation pump [0285] 2280 liquid nitrogen trap [0286] 2410 solid Cs target [0287] 2412 sealed target container [0288] 2414 inner container [0289] 2416 heater (temperature controller) [0290] 2800 XeCs separation device [0291] 2240, 2810 Cs ion trap [0292] 2822, 2824, 2832, 2834 gas line [0293] MB muon beam [0294] LS liquid flow