For production of fission 99 Mo support tube, uranium target and production method

Abstract

The present invention provides a method for producing fission 99 Mo support tube, uranium target and production method. The support tube comprises: a tube body; an outer layer, the outer layer covers the outer wall of the tube body and contains enriched 235 U; and a core, the core is arranged in the inner cavity of the tube body and is configured to absorb neutrons generated by the fission reaction. The uranium target includes the support tube. In addition, the method includes: irradiating the uranium target with a CANDU heavy water nuclear reactor for a predetermined time; dissolving the outer layer of the support tube in the irradiated uranium target with an acid solution to obtain a target solution; and separating the fission neutrons obtained from the target solution. 99 The method of the present invention can utilize the existing CANDU heavy water nuclear reactor to carry out large-scale medical fission 99 The production of Mo does not affect the function of the reactor itself.

Description

Technical Field

[0001] The present invention relates to the field of nuclear engineering, in particular to a method for producing fission 99 Mo support tube, uranium target and production method. Background Art

[0002] 99m Tc nuclear properties are excellent, with a short half-life (6.02h) and appropriate gamma ray energy (140keV), which are very suitable for single photon emission tomography (SPECT) and have a low radiation dose to the human body. 99m Tc has a variety of chemical valence states and can be used to label various ligand drugs for the diagnosis of diseases of almost all organs and tissues such as the brain, myocardium, bone, thyroid, lung, liver, gallbladder, kidney, and lymph. 99m Tc is the most widely used medical radioisotope. 99m The main Tc nuclides 99 Mo decays to 99 Mo mainly from 235 U is extracted from the fission products, and the nuclear reaction is 235 U(n,f) 99 Mo. Extraction 99 Mo solution was prepared into fission 99 Mo- 99m Tc generator, for users to easily obtain when using 99m Tc.

[0003] Currently used in production internationally 99 Among the main Mo reactors, in addition to the OpenPool Australian Light Water Reactor (OPAL) in Australia, the High Flux Reactor (HFR) in the Netherlands, the Belgium BR-2 research reactor (Belgium Reactor-2), and the South African Safari-1 research reactor will be shut down before 2030. If no new reactors are built for medical use, 99 Mo production reactor, fission 99 Mo production capacity faces the risk of a sharp reduction.

[0004] At present, the existing nuclear reactors for scientific research in China do not meet the requirements for large-scale, routine production of medical 99 Mo is the most widely used isotope in nuclear medicine since 2002. 99 Mo, all rely on imports. 99 Mo raw materials— 99 Mo / 99m Tc Generator— 99mIn the complete industry chain of "Tc drug-SPECT imaging", my country is the only country that lacks 99 The process of Mo raw material production.

[0005] Therefore, it is necessary to accelerate the development and utilization of existing conditions to produce medical 99 Mo raw material technology. Summary of the invention

[0006] In view of this, the object of the present invention is to provide a method for mass production of medical 99 Mo targets and methods to fill the gap in domestic large-scale production of medical 99 Mo technology gap, reduce import dependence. To this end, the present invention provides a method for producing fission 99 Mo support tube, suitable for uranium target of CANDU heavy water nuclear reactor and fission production using CANDU heavy water nuclear reactor 99 To this end, in a first aspect, the present invention provides a method for producing fission 99 Mo support tube, characterized in that the support tube comprises: a tube body; an outer layer, the outer layer covers the outer wall of the tube body and contains enriched 235 U; and a core, the core is disposed in the inner cavity of the tube and is configured to absorb neutrons produced by the fission reaction.

[0007] In one embodiment, the outer layer 235 U enrichment is less than 20% and greater than 0.7%, or 235 U enrichment is greater than 90%; preferably, the outer layer 235 U enrichment is less than or equal to 19.75% and greater than 3%.

[0008] The outer layer is uranium dioxide formed by electroplating.

[0009] In one embodiment, the core is 235 Uranium dioxide with an enrichment of U less than 0.2%.

[0010] In one embodiment, the tube body is made of nuclear grade zirconium-based alloy or stainless steel.

[0011] In a second aspect, the present invention provides a method for producing fission reactors suitable for CANDU heavy water nuclear reactors. 99 A uranium target of Mo, comprising the aforementioned support tube.

[0012] In a third aspect, the present invention also provides a method for producing fission reactors using a CANDU heavy water nuclear reactor. 99Mo method, the method comprising the following steps: using the CANDU heavy water nuclear reactor to irradiate the aforementioned uranium target for a predetermined time; using an acid solution to dissolve the outer layer of the support tube in the irradiated uranium target to obtain a target solution; and separating fissionable molybdenum from the target solution. 99 Mo.

[0013] In one embodiment, the irradiation duration is 5-7 days.

[0014] In one embodiment, the fissionable 99 Mo comprises the following steps: adjusting the pH value of the target solution to alkaline to cause precipitation in the target solution, filtering and collecting the filtrate; acidifying the filtrate to remove radioactive iodine; and separating and collecting fission iodine by alumina chromatography. 99 Mo.

[0015] In one embodiment, the method further comprises separating the fission 99 Mo is purified, the purification comprising: sequentially using anion exchange method and activated carbon chromatography to purify fission 99 Mo, and obtain the fission of molybdate 99 Mo.

[0016] In one embodiment, the fission reaction of the obtained molybdate form is further performed. 99 Mo is subjected to quality control analysis, the quality control analysis comprising: using a high-purity germanium gamma spectrometer to identify and measure the gamma impurity activity; and separately measuring the beta impurity activity and the alpha impurity activity.

[0017] The present invention provides a uranium target including a support tube, and does not require any modification of the existing nuclear reactor, so that the CANDU heavy water nuclear reactor can achieve fission while generating electricity. 99 The production of Mo can achieve multiple uses. The technical solution of the present invention can be used for medical 99 Mo can be produced on a large scale and has good economic benefits. 99 Mo can be used for the diagnosis of human bone injury, cardiovascular disease, nervous system disease and tumor. 99 Mo production method can obtain fission energy that meets medical standards 99 Mo. BRIEF DESCRIPTION OF THE DRAWINGS

[0018] In order to more clearly illustrate the embodiments of the present invention or the technical solutions in the conventional technology, the drawings required for use in the embodiments or the conventional technology descriptions are briefly introduced below. Obviously, the drawings described below are only some embodiments of the present invention. For ordinary technicians in this field, other drawings can be obtained based on these drawings without paying any creative work.

[0019] Figure 1 In the embodiment of the present invention, for producing fission 99 Schematic diagram of the structure of Mo support tube.

[0020] Figure 2 The CANDU heavy water nuclear reactor produces fission energy in the embodiment of the present invention. 99 Mo process flow chart DETAILED DESCRIPTION

[0021] To facilitate understanding of the present invention, example embodiments will now be described more fully with reference to the accompanying drawings. However, example embodiments can be implemented in a variety of forms and should not be construed as limited to the embodiments set forth herein; rather, these embodiments are provided to make the present invention comprehensive and complete and to fully convey the concepts of the example embodiments to those skilled in the art. The same reference numerals in the figures represent the same or similar structures, and thus their detailed description will be omitted.

[0022] Unless otherwise defined, all technical and scientific terms used herein have the same meaning as those commonly understood by those skilled in the art of the present invention. The terms used in the specification of the present invention herein are only for the purpose of describing specific embodiments and are not intended to limit the present invention.

[0023] When used herein, the singular forms "a", "an", and "said / the" may also include plural forms, unless the context clearly indicates otherwise. It should also be understood that the terms "include / comprise" or "have" and the like specify the presence of stated features, wholes, steps, operations, components, parts, or combinations thereof, but do not exclude the possibility of the presence or addition of one or more other features, wholes, steps, operations, components, parts, or combinations thereof. At the same time, in this specification, the term "and / or" includes any and all combinations of the relevant listed items.

[0024] In view of the lack of large-scale production of medical 99 The inventor intends to develop a method for large-scale production of medical Mo using existing nuclear reactors. 99 Mo's technology.

[0025] At the end of the last century, my country's Qinshan Nuclear Power Station imported two CANDU-6 heavy water reactors (CANDU reactors for short) from Canada. The CANDU reactor has a high core neutron injection rate, which can reach an average of 2.0×10 14 n / cm 2 ·s. High neutron flux rate is achieved 99Mo mass production provides the necessary premise. Secondly, the CANDU reactor is operated in a non-stop refueling mode, using a refueling machine to load and unload uranium targets. Therefore, the CANDU reactor has a long annual operation time, more than 300 days per year. Based on the above characteristics, the CANDU reactor has the ability to be used for large-scale and conventional medical fission reactors. 99 The basic conditions for Mo production are expected to be used to irradiate uranium targets to produce fission 99 Mo.

[0026] The CANDU reactor uses heavy water as a moderator and natural uranium (UO2 sintered pellets) as nuclear fuel. The fuel bundle of the CANDU reactor (such as the CANDU-6 fuel bundle) includes seven components: high-density natural uranium UO2 fuel pellets, cladding tubes, graphite coatings (applied to the inner wall of the cladding tubes), end plugs, support pads, isolation blocks and end plates. The UO2 pellets are loaded into Zr-4 alloy cladding tubes, and the two ends are sealed and welded by end plugs to form a single rod. Each CANDU fuel bundle is composed of several single rods (for example, there are 37 single rods in the CANDU-6 fuel bundle). The single rods are arranged in a ring at a fixed position and are fixed on both sides by welding with end plates to form a fuel bundle. Unlike the structure of ordinary research reactors, there is no spacer in the CANDU reactor for placing 99 The irradiation channel of the uranium target for Mo production. In the fuel bundle, the natural uranium UO2 fuel pellets are produced by fission. 99 Mo has low enrichment and is difficult to extract, so it cannot be used directly.

[0027] To this end, the present invention proposes to use a support tube containing low-enriched uranium or high-enriched uranium to replace the high-density natural uranium UO2 fuel pellets in a single rod, perform scientific structural design, and thereby obtain a suitable production 99 Therefore, there is no need to make other modifications to the CANDU reactor. The uranium target of the present invention is used to replace the fuel rod bundle therein, and the fission energy of the required activity can be obtained after irradiation in a heavy water reactor. 99 The uranium target of the present invention is suitable for the operating environment of the heavy water reactor, is safe and reliable, has no influence on the power generation function of the reactor itself, and thus improves the economic value of the reactor.

[0028] Support tube

[0029] In order to solve the above problems, the embodiment of the present invention provides a method for producing fission 99 Mo support tube, which will be combined with Figure 1 The support tube structure is described in detail. 99 The Mo support tube comprises: a tube body (1), an outer layer (2) and a core (3).

[0030] The material forming the tube body (1) may be a nuclear-grade zirconium-based alloy or a stainless steel material. In the present embodiment, the material of the tube body is preferably a nuclear-grade Zr-4 alloy.

[0031] The outer layer (2) covers the outer wall of the tube body (1) and contains 235 U. 235 U pass 235 U(n,f) 99 Mo reaction produces fission 99 In an embodiment of the present invention, the enriched 235 U is formed by electroplating and exists in the chemical form of UO2 with low or high enrichment. Specifically, uranyl nitrate is dissolved in a certain solvent, added to the electroplating tank of the electroplating device, and the power is turned on. After a period of electroplating, a UO2 coating attached to the outer wall of the support tube is obtained. However, since highly enriched uranium can be used to make nuclear weapons, in order to prevent nuclear proliferation, low-enriched uranium is used to produce fissionable uranium. 99 Mo will be the future development trend. In one embodiment, 235 The enrichment of U is less than 20% and greater than 0.7%. In another embodiment, 235 The enrichment of U is greater than 90%. In an embodiment of the present invention, preferably, 235 The enrichment of U is less than or equal to 19.75% and greater than 3%. More preferably, 235 The enrichment of U is 19.75%.

[0032] The core (3) is arranged in the inner cavity of the tube body (1). 235 The enrichment of U is higher than that of natural uranium pellets. 235 U has a high enrichment rate, and a large number of neutrons are produced in the fission reaction. In order to ensure the balance and safety of the reactor, the core is configured to absorb the neutrons produced by the fission reaction. In this embodiment, the core is 235 Uranium dioxide with a U enrichment of less than 0.2% is called depleted uranium.

[0033] This structural design ensures that low-enriched uranium or highly enriched uranium will undergo fission reaction. 99 Mo product quality or output at the same time, the core 235 Uranium dioxide with U enrichment less than 0.2% absorbs a large number of neutrons produced by the reaction, balancing the neutrons produced by the fuel. 235 The impact of increased U enrichment on the reactor is to ensure the balance and safety of the reactor.

[0034] Uranium Target

[0035] The embodiment of the present invention further provides a uranium target, wherein the uranium target comprises the aforementioned support tube. In the embodiment of the present invention, the uranium target is suitable for a CANDU heavy water nuclear reactor.

[0036] The fuel rod bundle in the CANDU stack includes seven components: high-density natural uranium UO2 fuel pellets, cladding tubes, graphite coating (applied on the inner wall of the cladding tubes), end plugs, support pads, isolation blocks, and end plates. Taking the CANDU-6 type fuel rod bundle as an example, each CANDU-6 type fuel rod bundle is composed of 37 single rods. The UO2 fuel pellets are loaded into the cladding tube, and the cladding tube material includes a nuclear-grade zirconium-based alloy, preferably a Zr-4 alloy, and its two ends are sealed and welded by end plugs to form a single rod. The 37 single rods are arranged in a ring according to a fixed position, and the two sides are welded and fixed with end plates to form a fuel rod bundle.

[0037] In the uranium target provided in this embodiment, the support tube is located in the cladding tube of the fuel rod bundle, that is, it is obtained by replacing the high-density natural uranium UO2 fuel pellets in a single rod of the fuel rod bundle with the support tube.

[0038] The fuel bundle in the CANDU stack adopts a modular design. The reactor contains multiple fuel bundles. The continuous operation can be achieved by adopting a non-stop stacking and refueling operation mode. In the embodiment of the present invention, according to specific production requirements, one or more single rods in the CANDU stack, or all single rods in one or more fuel bundles, are replaced with the fuel rods provided by the present invention for producing fission fuel. 99 Mo uranium target, other fuel single rods / rod bundles maintain normal production functions. The uranium target provided by the present invention undergoes fission 99 Mo production enables CANDU heavy water reactor to achieve multiple uses, ensuring normal production while mass production of fission 99 Mo.

[0039] 99 Production method of Mo

[0040] The embodiment of the present invention also provides a method for producing fission products by using a CANDU heavy water nuclear reactor. 99 Mo's method. Figure 2 As shown, the method comprises the following steps: S001, irradiating the uranium target with the heavy water nuclear reactor for a predetermined time; S002, dissolving the outer layer of the outer wall of the support tube in the irradiated uranium target with an acid solution to obtain a target solution; and S003, separating fissionable 99 Mo.

[0041] In step S001, when the uranium target is irradiated in the CANDU heavy water reactor, 235 U(n,f) 99 Mo reaction produces fission 99 Mo, irradiation time depends on the uranium target 235 U content and required 99 The production capacity of Mo is adjusted accordingly. In an embodiment of the present invention, the irradiation time is preferably 5-7 days.

[0042] After irradiation, the uranium targets are removed from the pile, put into lead cans and transferred to the isotope production line. The irradiated targets are placed in a target dissolving device for dissolving, and then the uranium targets are transferred to a target dissolving device.

[0043] In step S002, an acid solution is used to dissolve the UO2 coating in the uranium target after irradiation to obtain a target solution. In one embodiment of the present invention, dilute nitric acid or dilute sulfuric acid is used to dissolve the UO2 coating.

[0044] In step S003, after obtaining the target solution, the fission 99 Specifically, the fission molybdenum is separated from the target solution. 99 Mo comprises the following steps: adjusting the pH value of the target solution to alkaline to cause precipitation in the target solution, filtering and collecting the filtrate; acidifying the filtrate to remove radioactive iodine; and separating by alumina chromatography 99 Mo.

[0045] The pH value of the target solution is adjusted to alkaline. In one embodiment, a concentrated alkali, such as a concentrated KOH or NaOH solution, is preferably used. The greater the concentration of the alkaline concentrated solution, the more conducive it is to ensure that the total volume of the target solution will not increase significantly during the process of adjusting the pH value of the target solution. For example, in an embodiment of the present invention, a concentrated NaOH solution is used to adjust the pH value of the target solution until it is alkaline, and the alkaline solution is slowly added until the precipitation no longer increases. In the alkaline solution, Sr, U, Pu and other α impurities are precipitated. 99 Mo exists in the solution in the form of water-soluble molybdate ions. The solution is then filtered, the filtrate is collected, and the impurities are initially removed in the form of precipitation. 99 Mo remains in the filtrate.

[0046] Further, the filtrate is acidified to remove radioactive iodine, for example, concentrated acid, such as concentrated nitric acid, is generally used for acidification. In a specific embodiment, concentrated nitric acid solution is added to the filtrate for acidification, and heated to about 105° C., and the radioactive iodine in the solution escapes in the form of gas to remove most of the radioactive iodine in the solution.

[0047] Further, the fission 99 For example, acidic Al2O3 can be used as the stationary phase and adsorbed in a 0.1-1.0 mol / L HNO3 solution. 99 Mo, after the column is fully washed with deionized water and saline, 1.0-3.0 mol / L NH4OH solution is used to 99 Mo desorption. Alumina chromatography can 99 More than 90% of Mo solution 131I. 103 Such as 132 Te, 95 Zr, 89 / 90 Sr and other impurities are further removed separately. 99 The Mo recovery rate is greater than 90%.

[0048] Further, in step S004, the separated fission 99 Mo purification: Anion exchange and activated carbon chromatography were used in sequence. 99 Mo, and obtain the fission of molybdate 99 Mo.

[0049] Strongly basic anion exchange resin 99 Mo is initially purified. For example, domestic 717 resin or AG1-×8 resin is used as the stationary phase, and NH4OH solution is used for adsorption. 99 Mo, after being fully eluted with deionized water and dilute malonic acid solution, is desorbed with ammonium carbonate solution or ammonium sulfate solution 99 Mo, and obtain the anion exchange column desorption liquid. The anion exchange method further removes 99 Mo solution such as 131 I. 103 Such as 132 Te, 95 Zr, 89 / 90 Sr and other impurities, the removal rate of various impurities is greater than 99%, 99 The Mo recovery rate is higher than 95%.

[0050] Then it was further purified by activated carbon chromatography 99 Mo. Activated carbon can further remove radioactive iodine by adsorbing various elements through adsorption. Preferably, compared with conventional activated carbon, silver-coated activated carbon has a better iodine removal effect. At the same time, organic impurities introduced during the separation process can be removed by the adsorption of activated carbon. For example, specifically, the anion exchange column desorption liquid is loaded onto the activated carbon chromatography column, and after the adsorption is completed, the chromatography column is washed with dilute nitric acid solution and bromine water (such as about 3%), and then desorbed with sodium hydroxide solution (such as 0.2M). 99 Mo, in solution 99 Mo is converted to Na2 99 MoO4 chemical form, finally get to meet the medical requirements 99 Mo. In this process, 99 The Mo recovery rate is close to 90%.

[0051] at present, 99Mo production processes mainly include the American Cintichem process, the Belgian IRE process and the South African AEC process. The American Cintichem process uses strong acid to dissolve the target and then uses α-benzoin oxime precipitation to separate and purify. 99 Mo, the target dissolving solution used in this process is small in volume, and there is no large amount of impurity aluminum in the target dissolving solution. However, the α impurity content in the target dissolving solution is difficult to control. The Belgian IRE process uses concentrated alkaline solution to dissolve the target, and a large amount of fission products produced by the reaction are removed in the form of precipitation. However, the target dissolving solution has a large volume, and the subsequent iodine removal process is complex and time-consuming, and a large amount of radioactive waste liquid is generated in the process. The South African AEC process uses sodium hydroxide solution to dissolve the target, and adds an oxidant to speed up the dissolution rate. Subsequently, ion exchange resins and chelating resins are used to purify the product. Uranium and major fission products are filtered out in the form of precipitation, but this process cannot be comprehensively utilized. 131 I and 133 Xe.

[0052] Compared with the above 99 The production process of Mo, the present invention proposes a separation process method of dissolving the target with dilute acid, removing impurities with concentrated alkali, and then acidifying to remove iodine, and further separating and purifying by alumina chromatography, anion exchange method, and activated carbon chromatography 99 Mo. UO2 coating target solution has a small volume, which saves costs. By adjusting the pH value of the target solution with concentrated alkali, a large amount of fission products in the target solution are removed in the form of precipitation, which improves the impurity removal efficiency. In addition, the radioactive iodine in the solution is removed by acidification with concentrated acid, which avoids the time-consuming evaporation of iodine removal and improves the iodine removal efficiency. Therefore, the present invention reasonably selects the target dissolution method, combined with the precipitation method, to effectively remove the impurity elements in the target nuclide, thereby obtaining a fission nuclide that meets medical requirements. 99 In addition, in the method, the acid and alkali solution used to adjust the pH value are concentrated acid and concentrated alkali, and the required volume is small, so the overall volume of the target solution will not increase significantly during the separation process.

[0053] In addition, the present invention further comprises step S005: 99 Mo is subjected to quality control analysis. The quality control analysis includes using a high-purity germanium gamma spectrometer to identify and measure the activity of gamma impurities; and separately measuring the activity of beta impurities and the activity of alpha impurities.

[0054] Among them, the fission 99 Mo emits gamma rays. Other substances that emit gamma rays may include 131 I. 103 Such as 132 Te, 95 Zr and other impurities, so a high-purity germanium gamma spectrometer was used to analyze the fission 99 The γ impurities in Mo are screened and their activity is determined. 131 I.103 Such as 132 Te, 95 Zr, etc., the content should reach the pharmaceutical standard, that is, the radioactivity of these impurities should be 99 The ratio of Mo activity is less than 5.0×10 -3 %. Then, the β impurities were determined by liquid scintillation counter. The β impurities mainly include 89 / 90 Sr. was measured using a liquid scintillation counter and its activity was 99 The ratio of Mo activity should be less than 6.0×10 -5 % The α impurities were measured by α spectrometer to determine the fission 99 Radioactive purity of Mo, α impurity activity and 99 The ratio of Mo activity should be less than 1.0×10 -7 %.

[0055] Example 1

[0056] A support tube is installed in the cladding tube of a single rod in a CANDU fuel bundle. The support tube has a nuclear grade Zr-4 alloy as the tube body and is electroplated with low-enriched uranium dioxide (UO2) on the outer wall of the tube body. 235 U enrichment is 19.75%). Depleted uranium is placed in the tube to maintain the balance of the reactor. The ends of the cladding tube are sealed with end plugs and welded to form a single rod. 37 single rods are arranged in a ring according to the same fixed position as the CANDU reactor fuel rod bundle, and the two sides are welded and fixed with end plates to form a single rod for producing fission fuel. 99 The uranium target was installed in the CANDU reactor of Qinshan Nuclear Power Plant at a neutron flux rate of 2.0×10 14 n / s·cm 2 The uranium target is irradiated for 5 days under the condition of 10000 mL / min, and then put into a lead can after being taken out of the reactor and transferred to the isotope production line. The uranium target is placed in a target dissolution device to dissolve the target, and then the uranium target is transferred to a target dissolution device. 2000 mL of 1.0 mol / L nitric acid solution is added to dissolve the coating UO2. The tail gas purification system connected to the target dissolution device absorbs the radioactive substances released during the target dissolution. 132 / 1331 I. 133 Xe, 85 Kr, etc. Use 5 mol / L concentrated NaOH solution to adjust the pH value of the target solution to alkaline. Sr, U, Pu and other α impurities in the solution will precipitate. Slowly add alkali solution until the precipitate no longer increases. Use a filter to filter the precipitate, collect the filtrate, and treat the precipitate as radioactive waste. Transfer the filtrate to an acidification device, acidify it with commercially available 63% concentrated nitric acid, and heat it to about 105°C for 1 hour. The radioactive iodine in the solution escapes in the form of gas to further remove the radioactive iodine. Separate by alumina chromatography 99Mo. Acidic Al2O3 was selected as the stationary phase and adsorbed in 0.1-1.0 mol / L HNO3 solution. 99 After the adsorption was complete, the column was washed with deionized water and saline, and then desorbed with 1.0 mol / L ammonia solution. 99 Mo, collect the desorption liquid, and achieve 99 Separation of Mo from other fission products. The collected alumina chromatography column desorption liquid was loaded onto the AG1-×8 anion exchange column pre-equilibrated with ammonia water. After complete adsorption, the anion exchange column was washed with deionized water and malonic acid solution, respectively, and then desorbed with 1.0 mol / L ammonium carbonate solution. 99 Mo, collect the desorption liquid. Use concentrated sulfuric acid to adjust the acidity of the desorption liquid to 2.5, load it onto the activated carbon column for adsorption purification, and after adsorption, wash the column with dilute nitric acid solution and 3% bromine water, respectively, and then desorb with 0.2 mol / L sodium hydroxide solution. 99 Mo, obtained in the form of molybdate 99 Mo products. The obtained products were measured by high purity germanium gamma spectrometer. 99 Mo products 131 I. 103 Such as 132 Activity of gamma impurities such as Te; measured by liquid scintillation counter 89 / 90 Sr activity; α impurity activity is measured using an α spectrometer. Calculation 99 in Mo solution 131 I. 103 Such as 132 Te, 89 / 90 The content of Sr and α impurities and the purity of radioactive nuclei. 99 The ratio of Mo activity is as follows: 131 I = 2.83 × 10 -3 %, 103 Ru=3.21×10 -3 %, 132 Te=2.36×10 -3 %and 89 / 90 Sr and α impurities are both below the detection limit. 99 The purity of Mo radionuclides is greater than 99.0%, and both the impurity content and the purity of radionuclides meet the requirements of the pharmacopoeia.

Claims

1. A method of producing fission using a CANDU heavy water reactor 99 Mo's method, the method comprising the following steps: The uranium target is irradiated by the CANDU heavy water reactor for 5 to 7 days, wherein the uranium target comprises a support tube, the support tube comprises: a tube body, an outer layer and a core, the outer layer covers the outer wall of the tube body and contains enriched 235 U, where 235 U enrichment is greater than 0.7% and less than 20%, or 235 The enrichment of U is greater than 90%; the core is arranged in the inner cavity of the tube body and is configured to absorb neutrons generated by the fission reaction; dissolving the outer layer of the support tube in the irradiated uranium target with an acid solution to obtain a target dissolving solution; Separate the fission from the target solution 99 Mo, wherein the fission 99 Mo includes the following steps: Adjusting the pH value of the target solution to alkaline to precipitate the matrix U and fission metal impurities in the target solution, filtering and collecting the filtrate, wherein the fission metal impurities include Sr, Ru, and Zr; acidifying the filtrate to remove radioactive iodine; and Alumina chromatography was used to separate and collect fission 99 Mo; The fission precipitate was purified by anion exchange and activated carbon chromatography. 99 Mo, and obtain the fission of molybdate 99 Mo; The fission of the resulting molybdate form 99 Mo performs quality control analysis, which includes: Use high purity germanium gamma spectrometer to identify and measure the activity of gamma impurities; and The activity of β impurity was determined by liquid scintillation counter and α impurity was determined by α spectrometer.

2. The method according to claim 1, wherein: The outer layer 235 U enrichment is less than or equal to 19.75% and greater than 3%.

3. The method according to claim 1 or 2, wherein: The outer layer is uranium dioxide formed by electroplating.

4. The method according to claim 1 or 2, wherein: The core is 235 Uranium dioxide with an enrichment of U less than 0.2%.

5. The method according to claim 1 or 2, wherein: The tube body is made of nuclear grade zirconium-based alloy.

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