A method of producing a radionuclide carbon-14
(57) Abstract:The invention relates to applied radiochemistry and relates, in particular, production facilities for extraction of the radioactive isotope carbon-14, which is widely used in the form of labeled organic compounds, as well as in the sources of radiation. The inventive mixture is purified from impurities nitrates of calcium and sodium is melted at 250-500°C. after cooling, the molten salt is crushed and loaded into the container. After irradiation in a neutron flux of the lid of the container pierce and portions served through holes nitric acid for the dissolution of irradiated substances, while stirring container 150-200°C. the Resulting gaseous compounds of carbon-14 periodically blow out of the container and sent for recycling. Advantages of the claimed invention are to increase the output of carbon-14 high isotopic purity, reducing the number of generated radioactive waste, and improving the technological process. table 1. The invention relates to the field of applied radiochemistry and relates, in particular, production to obtain a radioactive isotope of carbon14C, widely used in the form of labeled organiaed nuclear reactions14N(n, p)14C during irradiation of targets in a nuclear reactor by neutron flux. For the preparation of targets using different nitrogen compounds. A method of obtaining carbon-14 from irradiated in a nuclear reactor chetyrehokisi nitrogen /1/ Packed in containers. The disadvantage of this method is the low yield of the final product, due to the low density of the starting material. There is also known a method of producing carbon-14 from neutron-irradiated aluminum nitride /2, 3/, whereby the aluminum nitride after irradiation in the form of powder or pellets are heated in a stream of oxygen at a temperature 920-1180oC for 1-5 hours, and the aluminum nitride before irradiation is subjected to heat treatment in a stream of oxygen at a temperature of 800-850oC for 5-30 hours. The disadvantages of this method are the need to work with highly radioactive substances at high temperatures and the formation of large quantities of high level waste.The closest analogue method is a method of obtaining carbon-14 by irradiating the container with dry powdered calcium nitrate in a nuclear reactor, followed by dissolving the shells of the containers in the solution u is about the closest analogue is the low volumetric content of nitrogen in the irradiated target (bulk density dry salt is about 1 g/cm3therefore, the volumetric content of nitrogen is only 0.17 g/cm3), the formation of large quantities of liquid radioactive waste by dissolving the shells of the containers and irradiated salt.Task to be solved by the claimed method, is to increase the output of carbon-14 high isotopic purity, reducing the number of generated radioactive waste and improving the maintainability of the way.The essence of the method is that the mixture is purified from impurities of calcium nitrate and sodium before irradiation is melted at a temperature corresponding to the composition of the mixture. Another option is the melting of the pure crystalline calcium content of 1-2 mol of water per mol of salt. After cooling, the molten salt is crushed and loaded into the container. After irradiation in a neutron flux removing radionuclide carbon-14 is carried out in the container by piercing the cover and feeding it through the holes formed in the portions of nitric acid, while stirring it to a temperature of 150-200oC and periodically blowing the resulting products in the system processing flow of gas not containing carbon compounds.In the proposed method, instead of dry poroshkovaya calcium, which have a higher density of 2.0 to 2.3 g/cm3) and bulk nitrogen content of 0.3-0.4 g/cm3) that even in the free filling of the container pieces salt benefits for the life of carbon-14 is more than 1.5 times. For additional purification from carbon-12 starting material before melting acidified with nitric acid, allowing further, in the allocation of carbon-14 from irradiated in a neutron flux of targets, to obtain products with high specific radioactivity. Exception transactions dissolution of the shell of the container in the proposed method also allows to increase the specific activity of the resulting products. To implement the method proposed container with irradiated salt to use as a container for the entire process of processing of the irradiated material and excretion of carbon-14. The presence of free volume in the container formed between the pieces of salt, allows input operations reagents directly into its cavity.The method and the device are illustrated by the examples, where the sample N 1 illustrates the obtaining of carbon-14 from irradiated in a nuclear reactor chetyrehokisi nitrogen /1/, example N 2 corresponds to the method taken for the proto is miniawy container with a volume of 14 l, equipped with two tubes to fill the target and selection of the gas phase, put 1500 g camerahouse nitrogen and lowered container with a target channel of a nuclear reactor at 220 h, selecting the gas phase in the stainless steel tank every 24 h exposure. From this tank the gas phase was passed through a solution of sodium hydroxide. After irradiation container with the target removed from the reactor, measured by the number of remaining target material, and a solution of sodium hydroxide were combined and precipitated carbon-14 in the form of barium carbonate. The amount of carbon-14 obtained in this way was 2.4107Bq/h or 1,7103Bq/(cm3h).Example No. 2. Aluminum containers with a diameter of 36 mm and a length of 102 mm (volume of 100 cm3) fills dehydrated powdered calcium nitrate (107 g). In the irradiation process was formed to 6.8 MCI carbon-14 per 1 g of nitrogen, i.e., only 120 MCI of C-14 on the container. After irradiation for 3 of the container is loaded into the apparatus, the solvent, it was poured 3.7 M the alkaline solution was heated to 75oC and kept for 1 hour to destroy aluminum shell containers. Then the apparatus was filled with concentrated nitric acid, and gases were blown away by a stream of nitrogen in an alkaline is SUP>3h), losses could occur due to the leakage through the alkaline trap carbon dioxide, methane and other volatile carbon compounds. In this redesign process was formed 7 litres of liquid radioactive waste.Example No. 3. In an aluminum container with a volume of 1 l was loaded with 1.6 kg of a mixture of nitrates of calcium and sodium (3:1) pre-melted at a temperature of 400oC, cast in the form of mono and broken into pieces measuring up to 10 mm After sealing by welding the container was placed in the irradiation channel of a nuclear reactor. After the end of the campaign, the container was removed from the reactor, the lid with special precollege devices have been made 2 holes, through which was blowing through the internal cavity of the container by the flow of pure nitrogen. Then into the container through an opening in communication with the tube, gradually introduced 300 cm35 M nitric acid. At the bottom of the container was heated on a hotplate to a temperature of 150-200oC. the Resulting gases are passed through an absorber of nitrogen oxides, the node afterburners and alkaline traps. After entering the acid on the same line for 1 hour was applied to the stream of pure nitrogen until the termination of allocation of radioactive 1,2107Bq/h or 1,2104Bq/(cm3h).Example # 4. In an aluminum container with a volume of 1 l was downloaded 1.5 kg dihydrate calcium nitrate, pre-melted at a temperature of 200oC, cast in the form of mono and broken into pieces measuring up to 10 mm After sealing by welding the container was placed in the irradiation channel of a nuclear reactor. After the end of the campaign, the container was removed from the reactor, the lid with special precollege devices have been made 2 holes, through which was blowing through the internal cavity of the container by the flow of pure nitrogen. Then into the container through an opening in communication with the tube, gradually introduced 300 cm35 M nitric acid. At the bottom of the container was heated on a hotplate to a temperature of 150-200oC. the Resulting gases are passed through an absorber of nitrogen oxides, the node afterburners and alkaline traps. After entering the acid on the same line for 1 hour was applied to the stream of pure nitrogen until the termination of allocation of radioactive gases (control using gas radiometer). The number obtained in this way carbon-14 was 1107Bq/h or 1104Bq/(cm3h).Sravnitel> The claimed invention allows to increase the output of carbon-14 high isotopic purity, to reduce the number of generated radioactive waste and improve the manufacturability of the process.Bibliography
1. Russian Federation patent RU N 2106032 MKI. 6 G 21/06, 1/08 "Method of production of the isotope carbon-14".2. Hata, K., Shikata, E., Amano H. Release of Carbon-14 from Neutron-Irradiated Aluminium Nitride in the Dry Procedure. - Journal of Nuclear Science and Technology, 1973, v.10, N 2, p. 89-94.3. Russian Federation patent RU N 2084979 MKI.6 G 21 G 1/06 "allocation Method radionuclide carbon-14 from neutron-irradiated aluminum nitride".4. E. E. Kulish. "Some issues of radioactive isotopes in a nuclear reactor". Obtaining isotopes. Powerful gamma-ray installation. Radiometry and dosimetry. - M.: Izd-vo an SSSR, 1958, S. 23-25. The closest analogue. A method of producing a radionuclide carbon-14 by the irradiation of dehydrated calcium nitrate in a metal container with neutrons, the subsequent opening of the container, the dissolution of irradiated substances in nitric acid and feeding the extracted radioactive gaseous product containing carbon-14, in the processing system, wherein the calcium nitrate with the addition of sodium nitrate (10 - 50%) is their carbon-14 is carried out in the container, piercing the cover and feeding it through the holes formed in the portions of nitric acid, while stirring it up to 150 - 200°C and periodically blowing the resulting products in the system processing flow of gas does not contain carbon compounds.
FIELD: radio-chemistry; methods of production of the chromatographic generator of technetium-99m from the irradiated by neutrons molybdenum-98.
SUBSTANCE: the invention is pertaining to the field of the radio-chemistry, in particular, to the methods of production of technetium-99m for medicine. Determine the specific activity of the molybdenum and the sorptive capacity of the used aluminum oxide in molybdate-ions. The mass of the molybdenum necessary for production of the preset activity of the eluate of technetium-99m determine from the ratio:ATc= 0.867·L·m ln (m)/ln(mox·Wi), where:ATc - activity of the eluate of technetium-99m, Ki; L - the specific activity of molybdenum, Ki/g; m - mass of molybdenum, g;mox - the mass of aluminum oxide in the chromatograph column, g; Wi - the sorptive capacity of the used aluminum oxide in molybdate-ions, g/g. After making of corresponding calculations the solution of molybdenum is applied on the aluminum oxide. The technical result of the invention consists in production of the generator with the required activity of technetium-99m at usage of the minimum quantity of molybdenic raw.
EFFECT: the invention ensures production of the generator with the required activity of technetium-99m at usage of the minimum quantity of molybdenic raw.
1 ex, 2 tbl, 1 dwg
FIELD: production of radioactive isotopes.
SUBSTANCE: proposed method for producing nickel-63 radioactive isotope from target within reactor includes production of nickel-62 enriched nickel target, irradiation of the latter in reactor, and enrichment of irradiated product with nickel-63, nickel-64 content in nickel-62 enriched target being not over 2%; in the course of product enrichment with nickel-63 nickel-64 isotope is extracted from irradiated product.
EFFECT: enlarged scale of production.
1 cl, 2 tbl
FIELD: nuclear medicine.
SUBSTANCE: method of realizing of neutron-catch therapy is based upon introduction of medicinal preparation into damaged organ or tissue of human body. Preparation has isotope with high cross-section of absorption of neutrons. Then damaged organ or tissue is irradiated by neutrons of nuclear reactor. Irradiation is performed with ultra-cold neutrons with energy of 10-7 eV and higher, which neutrons are released from cryogenic converter of neutrons of nuclear reactor and are delivered to damaged organ or tissue along vacuum neutron-guide, which neutron-guide has end part to be made in form of flexible catheter. Dosage loads are reduced.
EFFECT: minimized traumatism of healthy tissues of patient.
4 cl, 1 dwg, 1 tbl
SUBSTANCE: invention concerns manufacturing of radionuclides for industry, science, nuclear medicine, especially radioimmunotherapy. Particularly it concerns method of receiving actinium -227 and thorium -228 from treated by neutrons in reactor radium-226. Method includes irradiation of target containing of metallic capsule in which there is located reaction vessel, containing radium-226 in the form of compound. Then it is implemented unsealing of target's metallic capsule, dissolving of received radium. From solution it is separated by means of precipitation, and then it is implemented regeneration, preparation to new irradiation and extraction of actinium-227 and thorium-228 from solution. At that irradiation, dissolving, radium separation, its regeneration and preparation to new irradiation are implemented in the form of its united chemical form - radium bromide, in the same reaction vessel made of platinum. Method provides reusing of the same platinum vessel for receiving of actinium-227 and thorium-228 from one portion of radium by recycling of irradiation and extraction in the same vessel. Separation of metallic capsule by means of dissolving provides saving of mechanical integrity of platinum reaction vessel for each new irradiation cycle and extraction.
EFFECT: increasing of radiationally-environmental safety of process, excluding operations of increased radiation hazard.
2 cl, 2 ex
SUBSTANCE: limiting specific weight of acid mL A, required for complete termination of its reaction with aluminium oxide is determined. The amount of acid mHCl required for treating aluminium oxide with mass mox is calculated using the relationship: mHCl=mL A·mOX. After making the corresponding calculations, aluminium oxide is treated with acid, put into a chromatographic column and a molybdenum solution is added.
EFFECT: more reliable operation of a technetium-99m generator in terms of prevention molybdenum from falling into the eluate owing to achieving maximum sorption capacity of the oxide used.
3 dwg, 1 ex
SUBSTANCE: method of making a chromatographic technetium-99m generator from neutron-irradiated molybdenum-98 involves depositing a predetermined mass of molybdenum into a chromatographic column with aluminium oxide. For this purpose, eluate output of technetium-99m from the generators with different adsorbed molybdenum mass is determined. Through extrapolation from the obtained calibration curve, the mass of molybdenum which corresponds to maximum output of technetium-99m from the generator Be=1 is found as mi=exp[(1-a)/b], where a and b are coefficients of the calibration curve Bi=a+b·ln mi, where Bi is the eluate output of technetium-99m from the generator for the adsorbed mass of molybdenum mi.
EFFECT: obtaining a generator based on neutron-irradiated molybdenum-98 with a narrow eluate profile for extracting technetium-99m.
1 cl, 3 dwg
SUBSTANCE: method for neutron doping of a substance involves slowing down fast source neutrons with a retarder substance, forming a stream of slow neutrons in a selected region and irradiating the substance to be doped with the slow neutrons. During the slowing down process, the fast source neutrons are separated according to propagation angles thereof; streams thereof moving a direction selected by the structure of the retarder substance are selected; streams selected by the structure are summed up, formed into a narrow band and directed onto the substance to be doped, which is controllably moved in the focal region of the neutron streams.
EFFECT: high efficiency of the doping process and forming regions with high degree of doping in given areas of the doped substance.
5 cl, 3 dwg, 3 ex
FIELD: physics, atomic power.
SUBSTANCE: invention relates to nuclear engineering, particularly to production of stable isotopes using neutron beams, and can be used in the electronic industry when producing semiconductor silicon structures using ion implantation techniques, as well as nuclear engineering when designing neutron retarding elements. The disclosed method includes making a starting target from a substance which contains a mixture of boron-10 and boron-11 isotopes, irradiating the target with neutron flux to the required or complete burn-off of the boron-10 isotope and extracting the 11B isotope from the substance.
EFFECT: obtaining boron and compounds thereof with high, more than 99,9%, enrichment on the 11B isotope and high degree of purity.
SUBSTANCE: invention relates to a method of producing radionuclides. The disclosed method includes irradiating a target medium containing at least a target nuclide material in a neutron radiation zone. Formation of radionuclides is carried out in the target radionuclide material as a result of irradiation, and at least some of the formed radionuclides are extracted from the target nuclide material. The extracted radionuclides are then captured and collected using carbon-based recoil particle capturing material which is free of an empty mesh structure at the crystallographic level.
EFFECT: obtaining radionuclides with high specific activity and soft radiation using the Szilard-Chalmers effect.
16 cl, 4 tbl
SUBSTANCE: in the disclosed method, target material containing a starting nickel-62 isotope, is given the shape and function of a structural component of a nuclear reactor core and then loaded for irradiation in place of said element. After achieving a given degree of irradiation, the material is unloaded and initial and newly formed nickel isotopes are extracted during chemical treatment.
EFFECT: improved utilisation of neutrons without affecting the reactivity margin of the nuclear reactor.