Patent Application
20190043631
The present invention relates to a method for producing a radioactive substance obtained through muon irradiation and a substance to be produced therefrom. More specifically, the present invention relates to a method for producing a radioactive substance produced by causing a nuclear muon capture reaction with a radionuclide, and a substance produced therefrom.
Radiation emitted by nuclear radioactive decay and nuclear reaction has been used for various purposes by utilizing radioactive isotope (RI) or a radionuclide whose lifetime is stochastically determined in accordance with quantum mechanics. One typical example of it is nuclear medicine. In nuclear medicine, a substance containing a radionuclide as a part of the chemical structure, or a radioactive substance is used, and a radiation imaging radiation has been adopted for a living body (in vivo), such as SPECT (Single Photon Emission Computed Tomography), PET (Positron Emission Tomography), and planar images. Nuclear medicine that has been performed includes medical treatment using irradiation from medications of RI for, e.g., pain relief and in vitro nuclear medical inspection using tracers without imaging. Radioactive substances in these applications are used for nuclear medicine and related inspection, such as examination to measure metabolic capability with a tracer that is administered to a living organism (including human), as it may accumulate to a specific lesion. Such examinations include medical treatment by internal irradiation, imaging, capturing of three-dimensional images, and the like.
Conventional methods for producing radionuclides are carried out by irradiating charged particles and neutrons using a cyclotron or a nuclear reactor, or by extracting them from fission products (“nuclear fission method”). Among them, when it comes to the manufacturing method using a cyclotron, charged particles such as protons, deuterium nuclei, or alpha particles (4He nuclei) accelerated to a very high energy by a cyclotron are utilized. In contrast, in the nuclear fission method using nuclear reactors, for example, a target raw material is exposed to neutrons in a nuclear reactor, and thereafter useful nuclides are chemically separated from irradiated target materials or nuclear fission products.
The supply chain of the radionuclide produced by nuclear reactors has never been well-prepared. In particular, although it is necessary for stable production of radionuclides in the nuclear fission method to operate a nuclear reactor for a long time, institutions taking care of radionuclide production are limited to 6 research institutions (NRU reactor in Canada, HFR reactor in Netherlands, BR2 reactor in Belgium, OSIRIS reactor in France, SAFARI-1 reactor in South Africa, and OPAL reactor in Australia). Actually, Japan relies on European and Canadian reactors for supply of 99mTc (99Mo) (for simplicity, hereinafter referred to as “99Mo supply” in the background section) that is consumed domestically. High-enriched uranium (HEU) as a raw material for the nuclear reactor of the Atomic Energy of Canada Limited (NRU reactor) has been exported from the United States to Canada partly on an exemption due to medical demand of 99Mo supply; however, transportation of HEU is suspended. This is due to prevention of proliferation of nuclear related substances (hereinafter referred to as “nuclear nonproliferation”). The NRU reactor is planned to shut down in March 2018 while the Canadian government has abandoned its successor reactor plan. As for supply chain aspect, the supply of 99Mo from Europe to Japan was greatly affected by stagnation of aerial transport due to a volcanic eruption in Iceland in 2010. The situation is similar for the United States.
Against this backdrop, one of the inventors of the present application has found a method for producing a radioactive material using a nuclear muon capture reaction (nuclear muon capture reaction, abbreviated as “NMCR” in this application) with a stable nuclide that does not exhibit radioactivity as a raw material (PTL1).
Meanwhile, disposal of discharged spent nuclear waste remains at issue in currently operated nuclear power generation for power supply. Various methods have been proposed for the management of spent nuclear waste from nuclear power plants, among which a method of reprocessing at reprocessing plants is called nuclear fuel cycle. Generally, spent nuclear waste is divided into three types at reprocessing plants in the nuclear fuel cycle. The first one is uranium and plutonium, which are reused as nuclear fuel in the spent radioactive wastes. The second is a high-level radioactive waste including minor actinides (MAs) and fission products (FPs), which are radioactive wastes that are not recycled as nuclear fuel. The third is the remaining low level radioactive wastes. For a high-level radioactive waste, in addition to geological disposal on the premise of long-term storage for tens of thousands of years, combining “partitioning (or 4-group partitioning)” and “transmutation technology” has also been conceived for the purpose of facilitating site location and management of the disposal site (NPL1 and NPL2). In the method of partitioning, the high-level radioactive waste is separated into four groups of nuclides and processed uniquely to each group. Of these groups, a group of radioactive wastes containing MAs and FPs has a higher concentration but with reduced amount than unpartitioned one, so it is possible to reduce the amount of substances to be stored in a glass solidification form or the like only by such partitioning. However, MAs and FPs still require long-term storage. Therefore, subjecting the group of radioactive wastes containing MAs and FPs to transmutation with an accelerator or the like for the purpose of reducing the long-half-life MAs, FPs or the like is called “partitioning and transmutation” technology (NPL1 and NPL2).
PTL1: JP 2014-196997 A
NPL1: Atomic Energy Commission [of Japanese Government], “Gunbunri/Shoumetushori Gijyutu Kenkyuukaihatsu Chokikeikaku (Long-Term Research and Development Plan on Partitioning and Transmutation Technology)” (in Japanese) [online] http://www.aec.go.jp/jicst/NC/senmon/old/backend/siryo/back21/sanko2.htm [Last retrieved: Nov. 2, 2015]
NPL2: Research Organization for Information Science and Technology, Genshiryoku Hyakkajiten ATOMICA (ATOMICA, An Encyclopedia of Nuclear Energy) “fractional partitioning” (in Japanese), [online], http://www.rist.or.jp/atomica/data/dat_detail.php?Title_No=05-01-04-01 [Last retrieved: December 2 1, 2015]
NPL3: Yasuji MORITA, Kenihci MIZOGUCHI, Isoo YAMAGUCHI, Takeshi FUJIWARA and Masumitsu KUBOTA, “Gunbunrihou No Kaihatsu: Syoukibojikken Niyoru 4 Gunn Gunbunri Purosesu Ni Okeru Tekunechiumu Kyodou No Kakunin (Development of Partitioning Method: Confirmation of Behavior of Technetium in 4-Group Partitioning Process by a Small Scale Experiment)” Japan Atomic Energy Research Institute, (in Japanese) [online], http://jolissrch-inter.tokai-sc.jaea.go.jp/pdfdata/JAERI-Research-98-046.pdf
NPL4: Hisamichi YAMABAYASHI, “Kokusanka 99Mo/99mTc No Iryou Unyou Ni Mukete No Kadai-Kokusanka 99Mo/99mTc No Seizoujyou No Kadai (Problems in Clinical Practice of Domestic Supply of 99Mo/99mTc: Considerations on the Domestic Production of 99Mo/99mTc)”, RADIOISOTOPES (in Japanese), Vol. 61 (2012) No. 9 p. 489-496, Japan Radioisotope Association, doi:/10.3769/radioisotopes.61.489
NPL5: Ryohei ANDO and Hideki TAKANO, “Shiyozumi Keisuiro Nenryo No Kakushu Sosei Hyoka (Estimation of LWR Spent Fuel Composition)”, JAERI-Research (in Japanese) 99-004, (1999), Japan Atomic Energy Research Institute, [online], http://jolissrch-inter.tokai-sc.jaea.go.jp/pdfdata/JAERI-Research-99-004.pdf [Last retrieved: Dec. 24, 2015] Disclosed with detailed data at http://nsec.jaea.go.jp/ndre/ndre3/trans/sf.html
Among the above-mentioned conventional methods for producing radionuclides, the nuclear fission method has several problems inherent therein. First of all, the necessity of operating a nuclear reactor per se would become a hindrance to the stable supply. Furthermore, it requires HEU handling, therefore isolation and extraction work under high dose circumstance is unavoidable. Moreover, concerns over supply of raw materials such as HEU and over nuclear nonproliferation cannot be overcome. Furthermore, only limited facilities can handle these types of processing even if you take a look around the world. Also, when it comes to supply chain of the nuclides whose transportation time is limited due to half-life properties, the supply chain depends upon transportation circumstances by their nature so when the nuclides concerned are distributed via freight delivery. For these reasons, it is not always easy to maintain the supply chain of radionuclides for medical applications, so long as it solely relies on the nuclear fission method.
Differently from any of the above-mentioned methods for producing a radionuclide, the present invention provides a novel method for producing a radionuclide. As a result, the present invention contributes to the stable supply of radioactive substances that contains radionuclides.
The inventors of the present application conceived of adopting a radionuclide for a raw material instead of a stable nuclide conventionally adopted for the raw material in a method for producing a radionuclide utilizing NMCR by a negative muon. That is, in one embodiment of the present invention, provided is 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.
Also, in one embodiment of the present invention, provided is 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.
The inventors of the present invention focus on spent nuclear fuels associated with nuclear power generation in light water reactors. The inventors note that radionuclides that can be used for raw materials of useful nuclide production with NMCR are contained at high concentration in a high-level radioactive waste that is remains of processed spent nuclear fuel in the nuclear fuel cycle. As long as a nuclear reactor for power generation such as a light water reactor is in operation at a nuclear power plant, such radioactive nuclide demand in nuclear medicine is secured and the problem concerning stable supply does not arise.
In particular, radioactive wastes are storable raw materials with sufficiently long half-life when an LLFP (long-lived fission product) contained in FPs is adopted for the target nuclide. Therefore, supply stability of the radioactive wastes would not always be necessary, and it is unlikely that we face a raw material shortage even if nuclear reactors for power generation are stopped for any reason.
Furthermore, in one embodiment of the present invention in which 99Tc is selected for the nuclide of a raw material and then 99mTc is produced through NMCR, it is also possible to use 99Tc that would be a recycled material.
A negative muon is an elementary particle and a type of lepton. In any of the embodiments of the present invention, negative muons are made incident on a target nuclide to cause a muon nuclear capture reaction.
A descendant nuclide is a nuclide exhibiting radioactivity which has undergone one or more stages of radioactive decay. Typically, it includes not only a daughter nucleus generated by some radioactive decay from a parent nucleus, but also another descendant nucleus generated from that daughter nucleus. In any aspect of the present invention, the number of generations is not limited. Radioactive decay through which such descendant nuclide is produced also includes a series of radioactive decays (decay series) that sequentially generate a plurality of radionuclides, such as the neptunium series, the thorium series, and the actinium series.
A radionuclide is a term used to distinguish and identify atomic nuclei exhibiting radioactivity with respect to their nuclear spin states as needed. When the first radionuclide and the second radionuclide are concerned in the present application, the first radionuclide refers to a radionuclide produced directly via the muon nuclear capture reaction. In contrast, the second radionuclide is a nuclide that is determined to be different from the first radionuclide when distinction is made, in terms of nuclear spin states if necessary. The second radionuclide itself is also radioactive, and it is at least one of descendant nuclides of the first radionuclide. In accordance with the definition of descendant nuclides, a daughter nucleus obtained by radioactive decay from a nuclide that is to be classified into a second radionuclide for a certain first radionuclide must also be classified into a second radionuclide from a view point of the first radionuclide.
A radioactive substance is a substance of any form including a radionuclide. Typical chemical forms of it may include an element of a radionuclide substance, a compound including a radionuclide as a part of its chemical structure irrespective of an inorganic or organic compound (radioactive compound), and an association product associated with a radionuclide or a radioactive compound, as well as their ionized cations or anions. Also, the physical form of the radioactive substance is not particularly limited and may be of any physical form including a solid, a liquid, a gas a supercritical fluid, a plasma, and a dilution thereof. In the present application, the physical form of the radioactive substance is not limited and may take any physical form, including a crystal, an amorphous solid, an ionic crystal, a molecular crystal, a powder, an aqueous solution, a non-aqueous solution, an ion, a complex, an association product, a low-molecular substance, polymer molecule, organic and inorganic compounds, and the like.
The process of producing radioactive materials may be broadly categorized into two types; the first type is to make the radionuclide generate radioactivity, to generate an artificial radionuclide or to generate artificially a natural radionuclide by some method, to increase the ratio of the target nuclide, or to reduce the proportion of nuclides that are not one to produce, whereas the second type is to make the radioactive substance containing the radionuclide (hereinafter, including the radionuclide element substance) into the one having the intended chemical structure. In the present application, any process having a nuclear reaction including the first type is referred to as production of radionuclide. It should be noted that the production of the radionuclide described in the present application can include chemical processing in addition to physical processing, in the same manner as conventional production processes of radionuclides.
In any of the embodiments of the present invention, a useful radionuclide can be produced by a muon nuclear capture reaction utilizing, for example, radioactive wastes originating from spent nuclear fuel of a nuclear power plant. This enables production of a radioactive substance containing a target nuclide through a process that has no uncertainty over the stable supply of raw materials. In addition, this enables production of 99Mo-99mTc generator by using a recycled raw material of 99Tc that is produced in 99Mo-99mTc generator manufacturing process, 99Tc found in unused chemicals after formulation, or 99Tc produced in the generator after use.
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.
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.
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
μ−+N(Z0,A0)→N′(Z0−1,A0)+v (Eq. 1).
Here, the atomic number is Z0 (i.e., the proton number is Z0), the mass number is A0 (i.e., the sum of the proton number and the neutron number is A0), N is a nucleus in general, and N′ is a new nucleus to be generated while atomic number Z0 and the mass number A0 are specified. The reaction scheme expressed in Eq. 1 is that muon ρ− is captured by nucleus N of the target nuclide having atomic number Z0 and mass number A0, and then isobar nucleus N′ is generated whose atomic number is decreased by 1 to be Z0−1, while a neutrino v is generated.
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 “(ρ−, v) reaction”. The second one is expressed as
μ−+N(Z0,A0)→N″(Z0−1,A0−1)+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 A0 decreases by one. Moreover, another reaction, expressed as
μ−+N(Z0,A0)→N′″(Z0−1,A0−2)+2n+v (Eq. 3)
may occur in which two neutrons 2n are released and the mass number A0 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:
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.
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.
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.
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.
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.
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.
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.
79Se
90Sr
93Zr
99Tc
107Pd
126Sn
219I
135Cs
137Cs
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 99Tc, 134Cs, 135Cs, and 137Cs, and 90Sr. As indicated in Table 1, a high-level radioactive waste contains high concentrations of 90Sr, 90Tc, 135Cs, and 137Cs. That is, 99Mo is produced from 99Tc, where 99Mo is used for obtaining 99mTc, 133Xe is produced from 134Cs, 135Cs, and 137Cs, and 89Sr is produced from 90Sr. Details of combinations of these raw material nuclides and radionuclide to be produced will be described.
3-1. Production of 99Mo from 99Tc
According to the present embodiment, 99Mo can be produced from 99Tc, which is a radionuclide. To produce a 99Mo-99mTc generator 99Mo (half-life: 66.0 h) is used, 82.4% of which decays by β− decay into 99mTc. By gamma decay 99mTc decays into 99Tc with a half-life of 6.02 hours with a property in which gamma ray of 140.5 keV is released, where 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. 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 99Mo-99mTc generators. The nuclear reaction in which Mo isotopes are generated by NMCR targeting 99Tc is depicted on the nuclear diagram in
In the present embodiment, NMCR is used in a process to produce 99Mo from target material of Tc containing 99Tc. When NMCR targeting 99Tc is performed, schemes of reaction become as follows:
These schemes are also understood through transmutation traces on the nuclear diagram in
3-1-1. Production Amount of 99Mo from 99Tc by Transmutation of NMCR
Next, the estimation of the amount of 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 NTM and is calculated by the following equation:
N
TM
=I
μ−
×R
c
×P
NC
×P
RBR,
The operating conditions of the apparatus for estimating the amount of 99Mo produced are:
Furthermore, the following assumptions were made regarding the branching ratio:
A
RI(tirradiation)=(number of muons)×(reaction coefficient)×(1−exp(−0.693/T1/2×tirradiation)).
Here, T1/2 is the half-life of the producing nucleus, and tirradiation is the muon irradiation time. Radioactivity after cooling is
A
RI(t)(tcooling)=ARI(0)exp(−0.693/T1/2×tcooling).
Here, tcooling is the cooling time, and ARI (0) is the radioactivity when the muon irradiation is terminated.
3-1-1-1. Example Estimation for Longer Irradiation Time than the Half-Life of 99Mo
Under the above assumption, 99Mo production was estimated for irradiation of 5.5 days by NMCR which is twice the half-life of 99Mo (66.0 hours). As a result, 99Mo is generated by 2.33×1012 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
The production volume per muon beam channel for generating NMCR under this condition will be described. 99Mo produced by 5.5-day irradiation with muon is 7.97×1017 atoms, and its radioactivity is 2.33×1012 Bq (2.33 TBq, 63.0 Ci). We assume that 82.4% of the 99Mo nuclei decays into 99mTc nuclei. Since the decay constant of 99mTc is 3.198×10−5 for its half-life of 6.02 hours, its radioactivity is 2.10×1013 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 99mTc that can be used as nuclear medicine RI is 1.05×1013 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 6.97×1014 Bq (18.8 kCi), which in turn corresponds to the number of administrations of about 940,000 doses. For example, the total amount of 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 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 99Tc solid target having a required size (area and thickness) for efficiently stopping the negative muon on the 99Tc solid target to produce 99Mo is about 25 g. This amount of 99Tc can be easily obtained from raw materials described below.
Next, specific radioactivity of 99Mo obtained when the irradiation time is longer than the half-life of 99Mo in this embodiment will be described. The specific radioactivity is a measure of radioactivity of 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 99Mo in the produced Mo is 5.95%. As a result, 0.0595 g of 99Mo is present in 1 g of Mo, whose number N of 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=0.0595/97.50×6.02×1023=3.67×1020/g-Mo.
Based on this value, half-life of 99Mo, T1/2=66.0 h, and the decay constant λ of the 99Mo=0.693/(66.0×3600)=2.92×106 (sec−1), the specific radioactivity R of 99Mo is calculated as follows:
The specific radioactivity values to be compared are 370 TBq/g-Mo for specific radioactivity of 99Mo obtained by the nuclear fission method and 0.074 TBq/g-Mo for the specific radioactivity of 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 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 99Mo, even using a small alumina column, as an example.
3-1-1-2. Example Estimation for Shorter Irradiation Time than the Half-Life of 99Mo
A condition under which the specific radioactivity and production amount of 99Mo are increased more efficiently than the above estimation is one that NMCR irradiation is performed for 1.0 day, which is about ⅓ of half-life (66.0 hours) of 99Mo. The resultant 99Mo amounts to 6.91×1011 Bq (0.691 TBq, 18.7 Ci). The isotope ratios of Mo after the irradiation for 1.0 day are:
99Mo Concentration, after the
99Mo Production Amount, ″
99mTc Radioactivity, ″
99mTc Available Amount, ″
99mTc Yearly Production
99Tc Raw Material Consumption
99Mo Specific Radioactivity
In other words, compared with 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 99Mo concentration of 5.5-day irradiation. With respect to the production amount per muon beam channel, 99Mo that can be manufactured for 1.0 day muon irradiation is 2.37×1017 atoms, whose radioactivity amounts to 6.91×1011 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 99mTc production of 1.14×1015 Bq (30.8 kCi) per year. The number of doses corresponding to this amount is about 1.54 million. The material consumption of 99Tc required is calculated to be 0.44 mg by calculating the corresponding amount of 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.
The specific radioactivity R of 99Mo is calculated from the content of 99Mo (9.37%) in the produced Mo when irradiated with muon for 1.0 days by the same calculation as
For 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 99Mo obtained from the nuclear fission method. Even under this irradiation condition, a sufficient amount of 99Mo is supplied, so it can be said that it has high usefulness.
As described above, the method for generating 99Mo from 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 99Mo half-life, and it is preferable in the method to set the MNCR irradiation time shorter than the half-life of 99Mo.
In the following, the target material for obtaining 99Tc for production of 99Mo will be described, and the method for recovering 99Mo will also be described. Among the isotopes of Tc, 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 99Tc source. One is a high-level radioactive waste obtained by reprocessing of spent nuclear fuel, and the other is a recycled material.
3-1-2. Production of 99Mo from 99Tc in High Level Radioactive Waste
A high-level radioactive waste contains 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 UO2 fuel at the burnup rate of 45 GWd/tHM in the pressurized water reactor (PWR) and cooling thereafter for 5 years, 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 99Tc is an LLFP with a half-life of about 210,000 years. In addition, the 99Mo-99mTc generator can be produced from 99Tc, a radionuclide of Tc in the spent nuclear fuel, by way of NMCR in accordance with the above principle.
The process of producing a 99Mo-99mTc generator from a high level radioactive waste in spent nuclear fuel includes a step of extracting 99Tc from a high level radioactive waste in the first place, a step of producing 99Mo by NMCR in the second place, and a step of producing a 99Mo-99mTc generator in the third place.
In the step of extracting 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.
From the Tc and the platinum group of the precipitate obtained in the denitration precipitation step, by further acting hydrogen peroxide (H2O2) 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).
Next, 99Mo is produced from 99Tc by NMCR. For this purpose, a mixture of 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 99Tc aqueous solution (aqueous solution containing 99TcO4− ion) extracted therefrom for the target material. By irradiating the muon for a time determined based on the half-life of 99Mo, it is possible to generate 99Mo efficiently. This irradiation time is, for example, 5.5 days which is twice the half-life (66 hours) of 99Mo, or 1.0 days which is ⅓ of the half-life.
Specifically for NMCR and recovery and collection of generated 99Mo, any method by one of the present inventors and disclosed in PTL1 can be adopted. For example, ion containing 99Mo (hereinafter referred to as 99Mo ion), typically 99MoO42− produced, is absorbed onto an ion exchange column (alumina column) and separated from a substance containing the ion having 99Tc, such as 99TcO4− (99Tc ion) for recovery.
Thus, even if 99Mo is produced in the target raw material containing Tc ion (99TcO4−), 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 99Mo ion (99MoO42−). In addition, these columns themselves can be made of the same material as the column adopted for the 99mTc generator. This collection process is a totally inverted operation of an operation in the 99Mo-99mTc generator where 99Mo ions are absorbed to the ion exchange column (alumina column) and 99mTc ions generated by decay are eluted off by milking. Although stable nuclei 95Mo-98Mo can also be generated at the time of NMCR and even the Mo is mixed with 99Mo, the purity of 99mTc eluted by the milking operation of 99Mo-99mTc generator is not affected.
As another production example, a technique of a batch production process 1400 that employs a target material containing 99Mo can also be employed. When irradiating muon continuously, the half-life of 99Mo of 66 hours is not an obstacle.
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 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 99Tc, 99Mo after generation, and 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. 99mTc ions generated during transport can be easily chemically separated from 99Mo ions. The target material 1402 inside the containers 1404A to 1404D in
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 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.
Once a necessary amount of 99Mo is obtained for 99Mo-99mTc generator from a high level radioactive waste, it is no longer necessary to operate the nuclear reactor using HEU to obtain 99Mo for the sake of 99Mo-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 99Mo-99mTc generator.
3-1-3. Manufacture of 99Mo from Recycled Raw Material of 99Tc
In the process of actually producing 99mTc and 99Mo, which amount to the most part of the nuclides used in nuclear medicine applications, 99Tc can be easily obtained from a by-product in the manufacturing process of the generator for milking, unused chemicals after formulating 99mTc, and used generators per se. There is no particular difficulty in using such 99Tc for a target nuclide in the present embodiment. A recycled raw material containing 99Tc in the present embodiment means, among substances containing 99Tc, 99Tc that is produced as a by-product in an arbitrary step up until 99Mo-99mTc generator is produced, 99Tc obtained by leaving unused drugs after formulation, or 99Tc produced by radioactive decay in the 99Mo-99mTc generator. Since 99Tc of the nuclide as a raw material needs to be artificially obtained by some method, the substance containing 99Tc should be substantially one manufactured in connection with the 99Mo-99mTc generator, except for the radioactive waste mentioned above. The 99Tc in the case of a nuclide of the recycled raw material means 99Tc as explained above, while excluding 99Tc that has been produced via 99mTc once produced for 99Mo-99mTc generators and then administered to the human body or the like for their inherent purpose. The manufacturing process for producing 99mTc that lead to a 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 99Tc after passing through the ion separation column illustrated in
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 99Mo-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 99Mo-99mTc generators.
3-2. Production of 133Xe from 134Cs, 135Cs, and 137Cs
It is possible to manufacture 133Xe from Cs raw material containing 135Cs, 137Cs, which are LLFPs contained in a high level radioactive waste.
In the situation after UO2 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
The reaction modes of NMCR with 137Cs as the target nuclide are as follows:
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 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 133Xe only. Gas containing 133Xe is separated and recovered for use as nuclear medicine RI.
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 133Xe are:
3-2-1. Process (Outline)
The process of producing 133Xe by NMCR from 134Cs, 135Cs, and 137Cs in a high level radioactive waste is carried out by the following three steps:
In Step 2, as the first cooling, 1 hour cooling is carried out after muon irradiation. At this time, the half-life of 137Xe is 3.83 minutes, thus the 1 hour cooling period corresponds to 15.7 half-lives. With so much time, the majority of 137Xe decays by β− decays into 137Cs (LLFP). The 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:
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 135Xe. Since the half-life of 135Xe is 9.10 hours, most part of it decays by β− decay into 135Cs (LLFP) in the end. The 135Cs can be separated and recovered in an aqueous solution. The ratio among the number of isotopic atoms on completion of Step 3 is:
Using the mass number (133.00) calculated from the isotopic distribution of generated Xe, the number of 133Xe in 1 g of generated Xe is calculated to be 2.31×1020/g-Xe. Furthermore, using the half-life of 133Xe (T112=5.25 days), the specific radioactivity of 133Xe becomes 353 TBq/g-Xe. The specific radioactivity to be compared is 370 TBq/g-Mo of the specific radioactivity of 99Mo obtained by the nuclear fission method (NPL4).
The production volume of 133Xe per muon channel is 9.23×1011 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.
3-2-2. Details of Process
For implementing the process of producing 133Xe by NMCR from 134Cs, 135Cs, 137Cs, two candidates seem promising: the same method as the batch production process 1400 for 99Mo illustrated in
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
In the case of the on-line production method, muon irradiation for NMCR is carried out using a flow path for gas and liquid.
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.
In the processing apparatus 2400 for targeting the solid Cs target of
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 133Xe, continuous irradiation can be performed to increase the production rate of 133Xe.
It is useful to recover 133Cs, 135Cs, 137Cs produced by decay in an aqueous solution during at least one of the two cooling periods.
3-3. Production of 89Rb-89Sr from 90Sr
89Sr can be produced from Sr raw material containing 90Sr which is an LLFP contained in a high level radioactive waste. In nuclear medical applications, 89Sr is used as an internal therapeutic agent for pain relief in the case of painful bone metastasis, where it releases β− ray of maximum energy of about 1.49 MeV. It is a nuclide with a physical half-life of 50.5 days.
In the situation after using UO2 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
3-3-1. Process (Outline)
The process for producing 89Sr from the Sr target material containing 90Sr includes the following three steps:
3-3-2. Details of Process
In Step 1, muon irradiation is performed on the target material of Sr containing 90Sr for 90 minutes. Thereafter, after irradiating the muon, the Rb ion is separated and recovered from the Sr ion.
The reaction modes of NMCR using 90Sr as the target nuclide are as follows:
Assuming identical beam conditions and identical reaction branching ratio to those for 99Tc and 133Xe described above, the reaction coefficients from Sr isotopic proportion of spent nuclear fuel to each Rb are calculated to be:
In Step 1, muon irradiation is performed for 90 minutes toward Sr solid or aqueous solution target containing 90Sr. This irradiation time of 90 minutes is six times the 89Rb half-life (15.2 minutes). After muon irradiation, Rb ions are separated and recovered from Sr ions. Then the radioactivity of 89Rb becomes about 8.84×1012 Bq.
Next, in Step 2, Rb ions are cooled for 25 minutes. This period is ten times of the half-life of 90Rb half-life (2.6 min). As a result, 90Rb decays by β− decay into 90Sr. 90Sr is an LLFP. At this point, 89Sr and 88Sr, which are daughter nuclei of 89Rb and 88Rb respectively, are also mixed together. The ratio of radioisotopes (in atomic fraction) of Sr at this time is:
Furthermore, in step 3, Rb ions separated from Sr are cooled for 150 minutes. This period is 10 times the half-life of 90Rb half-life (15.2 minutes). In addition, 88Rb, 86Rb, 84Rb are included. Among these, 88Rb decays by β− decay and becomes stable nucleus 88Sr. 86Rb and 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 89Sr can be used for nuclear medicine RI.
The isotope ratio (in atomic fraction) of Sr at this point is:
At the time when muon irradiation is carried out for 90 minutes and step 3 is completed, number N of 89Sr in 1 g of produced Sr is given by
N=0.567/88.46×6.02×1023=3.86×1021/g-Sr,
where the mass number (88.46) calculated from the isotopic distribution of the generated Sr is used. If the half-life of 89Sr: T1/2=50.5 days and decay constant of 89Sr: λ=0.693/(50.5×24×3600)=1.58×10−7 (sec−1) are adopted for this calculation, the specific radioactivity R of 89Sr is calculated as:
R=λN=6.10×1014Bq/g-Sr
=610TBq/g-Sr.
This specific radioactivity of the 89Sr is about 1.6 times 370 TBq/g-Mo, a specific radioactivity of 99Mo obtained by the nuclear fission method (NPL4).
The production volume of 89Sr per day per muon channel is 9.43×109 Bq (255 mCi). Since the amount for a dose to a patient weighing 70 kg is 1.4×108 Bq (3.8 mCi), the above-mentioned production amount per day corresponds to about 67 doses.
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 99Tc and 99Mo described in
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.
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.
| Number | Date | Country | Kind |
|---|---|---|---|
| 2016-019334 | Feb 2016 | JP | national |
| Filing Document | Filing Date | Country | Kind |
|---|---|---|---|
| PCT/JP2017/003226 | 1/30/2017 | WO | 00 |
