EP3475954B1 - Verfahren zur herstellung einer iod-radioisotopenfraktion, insbesondere von i-131, iod-radioisotopenfraktion, insbesondere i-131 - Google Patents

Verfahren zur herstellung einer iod-radioisotopenfraktion, insbesondere von i-131, iod-radioisotopenfraktion, insbesondere i-131 Download PDF

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EP3475954B1
EP3475954B1 EP17732470.4A EP17732470A EP3475954B1 EP 3475954 B1 EP3475954 B1 EP 3475954B1 EP 17732470 A EP17732470 A EP 17732470A EP 3475954 B1 EP3475954 B1 EP 3475954B1
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Prior art keywords
iodine
radioisotopes
solution
fraction
radioisotope
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French (fr)
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EP3475954A1 (de
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Dominique Moyaux
Valery HOST
Caroline DECAMP
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Institut National des Radioelements IRE
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Institut National des Radioelements IRE
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21G—CONVERSION OF CHEMICAL ELEMENTS; RADIOACTIVE SOURCES
    • G21G1/00—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes
    • G21G1/001—Recovery of specific isotopes from irradiated targets
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21G—CONVERSION OF CHEMICAL ELEMENTS; RADIOACTIVE SOURCES
    • G21G1/00—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21G—CONVERSION OF CHEMICAL ELEMENTS; RADIOACTIVE SOURCES
    • G21G1/00—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes
    • G21G1/001—Recovery of specific isotopes from irradiated targets
    • G21G2001/0063—Iodine

Definitions

  • targets of highly enriched uranium are treated in order to produce radioisotopes of Molybdenum-99 and radioisotopes of iodine-131 by basic dissolution.
  • the basic slurry is then filtered and the basic liquid phase (the filtrate) is loaded onto a silver doped alumina resin.
  • a fraction containing iodine radioisotopes, in particular iodine-131, is recovered by elution from the alumina column doped with silver by sodium thiosulfate (Na 2 S 2 O 3 ).
  • the harvested fraction containing the radioisotopes of iodine, in particular iodine-131 is not sufficiently pure and must still be distilled for medical use.
  • Elution with sodium thiosulfate makes it possible to collect approximately 90% of radioisotopes of iodine, in particular of iodine-131 loaded on the column of alumina doped with silver.
  • the treatment of uranium fission products in order to produce short-lived radioisotopes implies particularly restrictive working conditions.
  • a first cell is dedicated to the dissolution of highly enriched uranium targets. Once the liquid phase containing the soluble fission products of uranium recovered by filtration, including the radioisotope of Mo-99, it is transferred to a second cell in which it is acidified to allow, during the step of exothermic acidification, release of iodine in the form of gas.
  • the solution from which the iodine is released is heated and stirred by bubbling to promote the release of the iodine in the form of a gas.
  • the gas containing the iodine radioisotopes is thus captured using a platinum asbestos trap.
  • the iodine radioisotopes, in particular I-131 are then desorbed from the platinum asbestos trap and sent to the cell to undergo chemical purification by distillation.
  • the yields of iodine radioisotopes, in particular I-131 described in this document are about 80 to 90%. 10 to 20% of the radioisotopes of iodine, in particular I-131 remain in the acidified liquid phase and contaminate the other radioisotopes.
  • the selectivity of the isolation of iodine for its production is not optimal.
  • the temperature of the acidified liquid phase increases, it is still necessary to provide additional heating and stirring by bubbling in an attempt to recover the radioisotopes of iodine, in particular I-131 maximum.
  • This heating leads to an evaporation of nitrates from the acidification which takes place with nitric acid, which contaminates the radioisotopes of iodine, in particular I-131 in the form of gas, which is problematic because it interferes with the labeling processes of subsequent biological molecules.
  • the invention aims to overcome the drawbacks of the state of the art by providing a method for improving the purity of the iodine produced by acting on the selectivity of production operations while reducing environmental risks.
  • said recovery of said fraction of radioisotopes of iodine, in particular of I-131 comprises washing the resin d alumina doped with silver by a NaOH solution at a concentration of between 0.01 and 0.1 mol / l, preferably between 0.03 and 0.07 mol / l and more preferably around 0 0.05 mol / l and an elution of radioisotopes of iodine, in particular of I-131 by a thiourea solution having a thiourea concentration of between 0.5 mol / l and 1.5 mol / l, preferably between 0.8 and 1.2 mol / l, more preferably around 1 mol / l with a collection of an eluate containing said radioisotopes of iodine, in particular of I -131 in a thiourea solution.
  • the alumina column is manufactured according to the teaching of the document “ Preparation and characterization of silver coated alumina for isolation of iodine-131 from fission products. Mushtaq et al. - Journal of Engineering and Manufacturing Technology, 2014 Except for the fact that silver is reduced by hydrazine instead of sodium sulfate.
  • the rate of impregnation of the alumina resin with silver is at least 4, preferably at least 5, preferably around 5.5% by weight of silver relative to the total weight undoped alumina.
  • the level of radioisotopes of iodine, in particular of iodine-131 eluted relative to the total content of iodine radioisotopes, in particular iodine-131 loaded on the alumina column was greater than 90%, or even approximately 95% in activity.
  • the elution using thiourea is faster, making it possible to reach a narrower elution peak, thereby increasing the selectivity of the purification of iodine radioisotopes, in particular of iodine-131, avoiding as much as possible the presence of other radioisotopes in the eluate of the column of alumina doped with silver.
  • the volume of washing solution is provided so as to be optimized and sufficiently delayed with respect to the passage of molybdenum through the column, such as for example the presence of Mo-99 radioisotopes, which would otherwise contaminate the eluate with iodine radioisotopes, in particular iodine-131, but not too much to avoid the loss of iodine radioisotopes, in particular iodine-131.
  • the selectivity of the recovery of iodine is improved, in particular of iodine-131, but also the environmental safety, by the adsorption of the radioisotopes of iodine , in particular iodine-131 on a silver-doped alumina resin, rather than having to pass the total quantity of iodine radioisotopes, in particular iodine-131, from the solution basic molybdate and iodine radioisotope salts in the gas phase to recover all of the iodine radioisotopes, in particular iodine-131 via a gas trap.
  • said uranium targets are targets of low enriched uranium.
  • the method according to the present invention applies to all types of targets, in particular to highly enriched uranium targets, but also low enriched, the embodiment from targets of low enriched uranium is preferred.
  • HEU Highly enriched uranium
  • the method further comprises, before said filtration an addition of alkaline earth nitrate, more particularly of strontium, calcium, barium, preferably barium and sodium carbonate to said basic slurry. .
  • the filtration time of the slurry has been reduced from 4 and 6 hours to a reduced duration of between 30 minutes and two hours, depending on the number of targets engaged in the dissolution. This is however already significantly high compared to a process using targets based on highly enriched uranium (filtration time typically between 10 and 20 minutes), but represents a possibility of industrial exploitation, which otherwise would not have existed. without increasing the production price of radioisotopes from the fission of uranium 235 too significantly.
  • targets based on low enriched uranium the solid phase content in the slurry is 5 times higher.
  • these targets are based on an aluminum and uranium alloy, in particular in the form of UAl 2 , although other forms of alloy are also present (such as UAl 3 , UAl 4 , ).
  • Low enriched uranium targets contain less than 20% by weight of uranium 235 relative to the total weight of uranium present in the target.
  • Highly enriched uranium targets contain more than 90% by weight of uranium 235 relative to the total weight of uranium present in the target. Consequently, the content of enriched uranium is proportionally and significantly reduced (by a factor of around 5).
  • the contamination of the radioisotope fraction of Mo-99 by the radioisotope Sr-90 is reduced because it precipitates with the carbonate brought into the slurry.
  • This is of considerable importance since the radio-toxicity of the radioisotope Sr-90 is very high by the combination of its long physical period (radioactive period: 28.8 years), its high energy beta radiation and its long biological period (bony tropism). It is therefore very important to reduce this impurity to minimize the potential long-term side effects in the patient.
  • the filter aid used in the process according to the present invention does not disturb the fixation of the iodine on the silver alumina column, on the contrary, given the already reduced presence of contaminants at the source, it appears from the present invention that it is possible to efficiently and profitably produce, on the one hand, a radioisotope of Mo-99 from low-enriched uranium, without the fraction of radioisotope being ultimately less pure, in this way always meeting the criteria of the European pharmacopoeia, despite the massive presence of a much larger amount of waste but also of contaminants more complex to remove, such as magnesium, but also in which the risk presence of strontium in the radioisotope fraction of Mo-99 is greatly reduced, but in which approximately 90% of the iodine present in the basic slurry is collected on this alumina column doped with a rgent after filtration on the other hand.
  • the method also comprises an acidification of said eluate containing said radioisotopes of iodine, in particular of I-131 in a solution of thiourea by addition of '' a buffer solution, in particular a phosphoric acid solution at a concentration of between 0.5 and 2 mol / l, preferably between 0.8 and 1.5, and more preferably around 1 mol / l with recovery of an acidified solution of salts of radioisotopes of iodine, in particular of I-131.
  • a buffer solution in particular a phosphoric acid solution at a concentration of between 0.5 and 2 mol / l, preferably between 0.8 and 1.5, and more preferably around 1 mol / l with recovery of an acidified solution of salts of radioisotopes of iodine, in particular of I-131.
  • the radioisotopes of iodine are acidified in order to be pre-purified and separated from the majority of contaminants, including thiourea, previously used for recover the iodine from the silver alumina.
  • the term "effluent from the resin” means the mobile phase which passes through the resin and which leaves the chromatographic column.
  • the method further comprises a purification of said acidified solution of salts of radioisotopes of iodine, in particular of I-131, said purification comprising a loading of said acidified solution of salts of iodine radioisotopes, in particular of I-131 on an ion exchange column, washing of said ion exchange resin with water, elution of said ion exchange resin using NaOH at a concentration between 0.5 and 2.5 mol / l, preferably between 0.8 mol / l and 1.5 mol / l and particularly preferably around 1 mol / l with recovery of said fraction of iodine radioisotopes, in particular of I-131 in a NaOH solution.
  • said ion exchange resin is a weak anionic resin.
  • the process also comprises an acidification of the basic solution of molybdate depleted of iodine radioisotopes, in particular of I-131 passing through said silver doped alumina resin, with the formation of an acid solution of molybdenum salts and release of residual iodine radioisotopes, in particular I-131, in the form of a visible gas of his recovery.
  • the quantity of iodine radioisotopes, in particular iodine-131 which is recovered by adsorption on the column of silver doped alumina is approximately 90% active compared to the total activity of iodine radioisotopes, in particular iodine-131.
  • the remaining 10% of radioisotopes of iodine, in particular iodine-131, are still present in the basic molybdate solution previously passed through said column of silver doped alumina. Therefore recovering the residual iodine in a separate step is advantageous for two reasons.
  • the iodine thus recovered can be valued in the form of a fraction of radioisotopes of iodine, in particular iodine-131, but also because the presence of residual iodine in the basic molybdate solution. represents an environmental risk of these radioisotopes of iodine, in particular iodine-131, escaping into the ventilation system, which is also connected to the chimney.
  • the process further comprises, before said acidification of the basic solution of molybdate depleted of iodine radioisotopes, in particular of I-131 passing through of said silver doped alumina resin, cooling of the basic solution of molybdate depleted of iodine radioisotopes, in particular of I-131 passing through said alumina resin doped with silver up to a temperature less than or equal to 60 ° C, preferably less than or equal to 55 ° C, more particularly, less than or equal to 50 ° C.
  • iodine radioisotopes in particular I-131
  • the filtrate is therefore acidified with concentrated nitric acid.
  • the iodine radioisotopes are then released during acidification in much larger quantities.
  • the method further comprises, after acidification, heating the acid solution of molybdenum salts at a temperature above 93 ° C, preferably greater than or equal to 95 ° C, preferably between 96 ° C and 99 ° C, but preferably less than 100 ° C, accompanied by air bubbling in order to optimize the release of iodine in gaseous form, at a well determined time, during and after acidification.
  • said recovery of the radioisotopes of iodine, in particular I-131 when it is released is carried out by a transfer of the radioisotopes of iodine, in particular I-131 under form of gas in a tube connected at one end to an acidifier in which the acidification takes place and at another end to a closed container containing an aqueous phase and a surrounding medium, said transfer of iodine radioisotopes, in particular I-131 in the form of gas being produced so as to lead directly into the aqueous phase in which the radioisotopes of iodine, in particular I-131 in the form of gas pass through the aqueous phase and escape in the form of bubbles towards the surrounding medium of the aqueous phase, contained in the closed container.
  • the nitrates possibly present in the form of aerosols, as well as other gaseous species soluble in water, such as nitrogen oxides, are solubilized and eliminated from the radioisotopes of iodine, in particular I -131 in the form of gas.
  • said closed container is connected by tubing to a second closed container which contains an NaOH trap and in which the surrounding medium of the aqueous phase is transferred from the closed container to the second closed container containing the NaOH trap in the form of a solution at a concentration of 2 to 4, in particular approximately 3 mol / l, with discharge of the surrounding medium containing the radioisotopes of iodine, in particular I-131 from the tubing in the NaOH trap solution, with solubilization of iodine radioisotopes, in particular I-131 in the form of iodide gas of iodine radioisotopes, in particular I-131 in the aqueous solution of the NaOH trap.
  • the iodine radioisotopes are thus dissolved in the aqueous NaOH solution at an NaOH concentration of 2 to 4 mol / l, preferably 3 mol / l and forms a crude solution of iodine.
  • the aqueous solution of the NaOH trap containing the iodides of iodine radioisotopes, in particular I-131 forms a crude iodine solution, which is then purified by a second acidification to form gaseous iodine.
  • the crude solution is transferred to an iodine purification cell.
  • the crude solution is then acidified with H 2 SO 4 + H 2 O 2 to again produce gaseous iodine, which is captured in 0.2 M NaOH bubblers.
  • This solution is called “mother solution” and it is then packaged in sealed bottles, according to orders.
  • the fraction of iodine radioisotopes, in particular I-131 in a NaOH solution containing iodides of iodine radioisotopes, in particular I-131 forms a crude iodine solution and is then purified by a second acidification, preferably carried out in the presence of H 2 SO 4 and H 2 O 2 , to again produce gaseous iodine. Then, preferably, the gaseous iodine is captured in 0.2 M NaOH bubblers to form said fraction containing an iodine 131 radioisotope.
  • said fraction of iodine radioisotopes, in particular of I-131 in a NaOH solution and the aqueous solution of the NaOH trap containing the iodides of iodine radioisotopes, in particular I-131, are combined and purified together by a second acidification.
  • the subject of the invention is also a fraction of iodine radioisotopes, in particular of I-131 conditioned in a solution of NaOH having a radiochemical purity in radio-isotopes of iodine, in particular of I-131 greater than 97%, preferably at least 98%, more particularly at least 98.5% of the activity present under the iodide chemical form of said radioisotope of I-131 with respect to the total activity of said radioisotope of I-131 in all its forms in said fraction.
  • said solution of radioisotopes of iodine, in particular of I-131 is packaged in leak-proof bottles, said leak-proof bottles being enclosed in individual shielded containers.
  • the fraction of radioisotopes of iodine, in particular of I-131 has a nitrate content of less than 30 g / l.
  • the fraction of radioisotopes of iodine, in particular of I-131 is obtained by the method according to the present invention.
  • uranium 235 When uranium 235 is bombarded with neutrons, it forms fission products with a lower mass and which are themselves unstable. These products generate via a disintegration cycle of other radioisotopes. In particular, this is how the radioisotopes Mo-99, Xe-133 and I-131 are produced.
  • Low enriched uranium targets contain an aluminum alloy containing uranium.
  • the content of enriched uranium relative to the total weight of uranium is maximum 20%, and typically around 19%.
  • the targets of low enriched uranium are dissolved during a basic dissolution phase in the presence of NaOH (around 4 mol / l or more) and NaNO 3 (around 3.5 mol / l).
  • a slurry is formed as well as a gas phase of Xe-133.
  • the slurry contains a solid phase mainly formed of uranium and hydroxides of fission products and a liquid phase of molybdate (MoO 4 - ) and iodine 131 under iodine salts.
  • the volume of basic dissolution phase increases with the number of targets given the very high content of product which cannot be used after dissolution of the targets.
  • the dissolution of the target's aluminum is an exothermic reaction.
  • the Xenon gas phase is recovered by capture by means of a Xenon trap.
  • Sodium carbonate is also added at a concentration of between 1 mol / l and 1.5 mol / l, preferably around 1.2 mol / l up to 100 to 300 ml depending on the number of dissolved targets.
  • the slurry is then diluted with water to a volume of 2 to 6 liters depending on the number of targets to allow its transfer to the subsequent stage.
  • the slurry containing the solid phase and the basic liquid phase is then filtered using a glass fiber filter whose porosity is between 2 and 4 ⁇ m, preferably around 3 ⁇ m.
  • the solid phase is washed twice with a volume of water of 900 ml, recovered and optionally returned upstream of the process for a subsequent basic dissolution.
  • the filtrate (basic liquid phase recovered containing the fission products Mo-99, I-131, I-133, I-135, Cs-137, Ru-103, Sb-125 and Sb-127) but also aluminate formed by the basic dissolution of aluminum targets, which is soluble at basic pH.
  • Aluminum is soluble in basic medium as well as in acidic medium. On the other hand, it is insoluble when the pH is between 5 to 10.
  • the filtrate is loaded on a column of alumina doped with silver in order to fix the iodine therein and to recover a basic depleted filtrate in iodine-131.
  • the column of alumina doped with silver is washed using a volume of approximately 500 ml of caustic soda at a content of approximately 0.05 mol / l.
  • the degree of impregnation of the alumina resin contained in the alumina column is approximately 5.5% by weight.
  • the iodine selectively fixes by reaction with doping with silver present on the surface of the alumina to form an insoluble silver iodide.
  • the silver doped alumina column is preferably positioned between two reactors. The reactor downstream of the silver doped alumina column is placed under controlled vacuum, which allows the transfer of the liquid onto the column at a flow rate of approximately 250 ml / min.
  • the iodine capture yields are about 95%.
  • the silver doped alumina column is eluted using a thiourea solution at a concentration of between 0.5 mol / l and 1.5 mol / l, preferably around 1 mol / l.
  • the eluate then contains the iodine from the column.
  • the eluate is then brought to acid pH using an addition of a buffer mixture, in particular phosphoric acid in order to obtain an acid solution of iodine salts.
  • the acid solution of iodine salts is then loaded onto an ion exchange column, in particular onto a column of weak anionic resin previously conditioned in a non-oxidizing acid medium, in particular using phosphoric acid.
  • an ion exchange column in particular onto a column of weak anionic resin previously conditioned in a non-oxidizing acid medium, in particular using phosphoric acid.
  • the activity of iodine attached to the ion exchange resin is transferred from one cell to another in solid form.
  • the ion exchange column on which the iodine is fixed is then eluted using NaOH at a concentration of between 0.5 mol / l and 2.5 mol / l, preferably around 1.
  • the iodine radioisotopes are then transformed into iodide and dissolved in NaOH.
  • the fraction containing the iodine radioisotopes then undergoes a final purification step using the second acidification.
  • the collected filtrate must then be acidified. However, acidification also gives off heat. By therefore, before acidification, the filtrate is cooled to a temperature of about 50 ° C. Indeed, as we know from the document " Form and Stability of Aluminum Hydroxide Complexes in Dilute Solutions (JD Hem and CE Roberson - Chemistry of Aluminum in Natural Water - 1967 ), the behavior of aluminum in solution is complex and the reactions for transforming the Al 3+ ion into the precipitated hydroxide form and the soluble aluminate form are subject to certain kinetics.
  • the medium is highly radioactive and at a high temperature because of the basic dissolution but also because of the exothermic nature of the neutralization during the acidification step, the addition of acid would form localized acid over-concentrations. resulting in local heating by the neutralization reaction, and the formation of insoluble aluminum forms or with kinetics of slow re-dissolution of aluminum salts.
  • the reaction medium has a high temperature, it is highly radioactive and not very accessible, it is not possible to keep it stirring to avoid these points of aluminate concentration at high temperature.
  • the filtrate is cooled in order to avoid the precipitation of aluminum salts during the acidification at a temperature of approximately 50 ° C and in all cases below 60 ° C.
  • the filtrate is therefore acidified with concentrated nitric acid.
  • the acidified filtrate is heated to a temperature above 93 ° C, preferably greater than or equal to 95 ° C, preferably between 96 ° C and 99 ° C, but preferably less than 100 ° C and maintained under bubbling.
  • the acidification makes it possible to obtain a solution at acidic pH in order to be able to fix the radioisotope of Mo-99 on the alumina column (in the presence of an excess of acid of about 1M).
  • the acidified liquid phase depleted in iodine, is then loaded onto an alumina column, conditioned in 1 mol / l nitric acid.
  • Mo-99 is adsorbed on alumina while the majority of contaminating fission products are eliminated in the effluent from the alumina column.
  • the alumina column on which the radioisotope of Mo-99 is fixed is washed with nitric acid at a concentration of 1 mol / l, with water, sodium sulfite at a concentration of about 10 g / l and finally again with water. Washing effluent is discarded
  • the alumina column is then eluted with NaOH at a concentration of approximately 2 mol / l and then with water.
  • the eluate recovered from the alumina column forms the first eluate of the Mo-99 radioisotope in the form of molybdate.
  • the first eluate of the column is kept for a period of time between 20 and 48 h.
  • the alumina column is again eluted using NaOH at a concentration of approximately 2 mol / l and then with water before cleaning.
  • the eluate of the new elution forms the second eluate of the radioisotope of Mo-99, in the form of molybdate.
  • either the first eluate of the Mo-99 radioisotope is combined with the second eluate of the Mo-99 radioisotope and forms a single eluate which will further undergo the subsequent purification steps. Either each first and second eluate is treated separately in the subsequent purification steps, in the same way.
  • the Mo-99 radioisotope eluate from the alumina column is then loaded onto a second chromatographic column containing a strong anionic ion exchange resin previously conditioned in water.
  • the ion exchange column is then eluted with nitrate using an ammonium nitrate solution at a concentration of approximately 1 mol / l.
  • the eluate recovered therefore comprises the radioisotope of Mo-99 in a fraction containing ammonium nitrate.
  • the ammonium nitrate solution containing the radioisotope of Mo-99 is then loaded onto a column of 35-50 mesh activated carbon, which can optionally be doped with silver to recover any traces of iodine.
  • the activated carbon column on which the Mo-99 radioisotope is fixed is then washed with water and then eluted with a solution of NaOH at a concentration of approximately 0.3 mol / l.
  • the solution of Na 2 99 MoO 4 in NaOH at a preferred concentration of 0.2 mol / l is loaded onto an alumina resin in a Mo-99 / Tc generator -99 or on a titanium oxide resin to allow the generation of technetium-99 radioisotopes for nuclear medicine.
  • the acidification makes it possible to obtain a solution at acid pH in order to be able to fix the radioisotope of Mo-99 on the column of titanium oxide (in presence of excess acid 1M).
  • the acidified liquid phase, depleted in iodine, is then loaded onto a column of titanium oxide, conditioned in 1 mol / l nitric acid. Mo-99 is adsorbed on titanium oxide while the majority of contaminating fission products are eliminated in the effluent from the titanium oxide column.
  • the titanium oxide column on which the radioisotope of Mo-99 is fixed is washed with nitric acid at a concentration of 1 mol / l, with water, sodium sulfite at a concentration about 10 g / l and finally again with water. Washing effluent is discarded
  • the titanium oxide column is then eluted with NaOH at a concentration of approximately 2 mol / l and then with water.
  • the eluate recovered from the titanium oxide column forms the first eluate of the radioisotope of Mo-99 in the form of molybdate and contains approximately 90% or more of Mo-99 initially present.
  • the first eluate of the column is kept for a period of time between 20 and 48 h.
  • the elution from the titanium oxide column is continued using NaOH at a concentration of approximately 2 mol / l and forms an elution tail containing the Mo radioisotope. -99, in the form of molybdate.
  • the first molybdate eluate and / or said tail of molybdate eluate are combined or not and acidified with a sulfuric acid solution at a concentration between 1 and 2 mol / l, preferably 1.5 mol / l, thus forming an acidified fraction of radioisotope of pure Mo-99 in the form of molybdenum salts.
  • the Mo-99 radioisotope eluate from the titanium oxide column is then loaded onto a second chromatographic column containing an ion exchange resin of weak anionic type previously conditioned in water.
  • the ion exchange column is then eluted with nitrate using an ammonium nitrate solution at a concentration of approximately 1 mol / l.
  • the eluate recovered therefore comprises the radioisotope of Mo-99 in a fraction containing ammonium nitrate.
  • the ammonium nitrate solution containing the radioisotope of Mo-99 is then loaded onto a column of 35-50 mesh activated carbon, which can optionally be doped with silver to recover any traces of iodine.
  • the activated carbon column on which the Mo-99 radioisotope is fixed is then washed with water and then eluted with a solution of NaOH at a concentration of approximately 0.3 mol / l.
  • Elution from the activated carbon column makes it possible to recover a solution of Na 2 99 MoO 4 in NaOH but to keep any iodine captured on the column at a preferred concentration of 0.2 mol / l, which will then be conditioned. and packed for delivery.
  • the solution of Na 2 99 MoO 4 in NaOH at a preferred concentration of 0.2 mol / l is loaded onto an alumina resin in a Mo-99 / Tc generator -99 or on a titanium oxide resin to allow the generation of technetium-99 radioisotope for nuclear medicine
  • the sealed container includes a gas phase outlet connected to a device for recovering rare gases, isolated from the outside environment, but also an inlet for a purging gas.
  • the gas phase contains ammonia (NH 3 ) which comes from the reduction of nitrates and the main gaseous fission products which are Xe-133 and Kr-85
  • Dissolution is a very exothermic reaction, which requires two large refrigerants.
  • water vapor is present in the gas phase.
  • the gas phase is carried by a carrier gas (He) to the rare gas recovery device.
  • He carrier gas
  • the recovery of Xenon is carried out as follows:
  • the gas phase leaves the sealed container for basic dissolution and is brought to the device for recovering rare gases.
  • the gas phase containing inter alia the radioisotope Xe-133 is first passed through a molecular sieve making it possible to remove ammonia (NH 3 ) and water vapor. Then, the gas phase is passed through silica gel in order to remove any trace of residual water vapor. The gas phase is then brought to the cryogenic trap.
  • the gas phase is adsorbed on zeolite, in particular on titanosilicate or on an aluminosilicate doped with silver, preferably on Ag-ETS-10 or Ag -chabazite. It will then be either marketed directly on the zeolite, or desorbed hot and sent to a cryogenic trap.
  • the gas phase containing inter alia the radioisotope Xe-133 is therefore brought to the cryogenic trap in a U-shaped tube. immersed in liquid nitrogen (i.e. at -196 ° C) contained in an armored container, through stainless steel parings.
  • 316 stainless steel trimmings are made from 316 stainless steel rod having a diameter between 1.5 and 2 cm and a length between 10 and 20 cm, preferably between 14 and 18 cm, more particularly d '' about 16 cm using a 4-lip cutter with a diameter of 16 mm using a hydraulic vice.
  • the speed of the milling machine comprising the above cutter is 90 rpm and adjusted with a forward speed of 20 mm / min.
  • the cutting depth of the cutter is approximately 5 mm.
  • Stainless steel parings have an average weight of between 20 and 30 mg / paring, preferably between 22 and 28 mg / paring and an untapped density when formed of between 1.05 and 1.4.
  • Stainless steel trimmings have an average length of 7 mm, a diameter of about 2.5 mm and a thickness of about 1.7 mm.
  • the U-shaped tube has an amount of trimming between 90 g and 110 g.
  • the volume of 316 stainless steel parings included in the U-tube is completely immersed in liquid nitrogen.
  • the Xe-133 radioisotope from said gas phase containing the Xe-133 radioisotope is then captured by liquefaction of said Xe-133 by means of said cooled stainless steel parings, which capture the Xe-133 by condensation.
  • the liquefaction temperature of Xe-133 is around -107 ° C. Consequently, the Xe gas is condensed in liquid form on the stainless steel trimmings.
  • the pipes are purged and the injection of liquid nitrogen is cut off and the trap is brought into contact with a vacuum interrupter, the volume of which is 50 times greater than the volume of the trimmings contained in the liquid nitrogen trap.
  • the liquid nitrogen trap is then, in a closed circuit with the collection bulb, brought to room temperature. After heating 99% of the Xe-133 initially present in gaseous form is found in the bulb.
  • the radioisotopes of iodine in particular residual I-131 which are not captured by the silver-doped alumina resin before acidification, are then recovered using time of acidification of the basic slurry which allows obtaining a solution at acidic pH which allows the fixation of the radioisotope of Mo-99 on the alumina column, the acidification also makes it possible to release the radio- isotopes of iodine for recovery.
  • the recovery of iodine can then be carried out during and after the acidification of the basic filtrate previously cooled.
  • the iodine radioisotopes are released by heating the acidified filtrate to a temperature above 93 ° C, preferably greater than or equal to 95 ° C, preferably between 96 ° C and 99 ° C, but preferably less than 100 ° C and maintained under bubbling to promote the release of iodine in gaseous form.
  • the acidifier has an aqueous phase outlet tubing which plunges into a closed container of water. Another tubing comes out of this closed container. The aqueous phase therefore leaves the acidifier and is bubbled in the water contained in the closed container.
  • the part of the filtrate that has evaporated is dissolved in the water contained in the closed container, while the insoluble part, namely the radioisotopes of iodine are found above the surface of in the closed container and leave it by means of the outlet tube from the closed container to a second closed container, namely a trap containing NaOH at a concentration of 3 mol / l.
  • the iodine radioisotopes are then transformed into iodide and dissolved in the NaOH contained in the iodine trap and forms a crude iodine solution.
  • the aqueous solution of the NaOH trap containing the iodides of iodine radioisotopes, in particular I-131 is then purified by a second acidification.
  • the crude solution is transferred to an iodine purification cell.
  • the crude solution is then acidified with H 2 SO 4 + H 2 O 2 to again produce gaseous iodine, which is captured in 0.2 M NaOH bubblers.
  • This solution is called “mother solution” and is then packaged in airtight bottles contained in a shielded enclosure for shipment to the customer.

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Claims (15)

  1. Verfahren zur Herstellung einer Fraktion von lod-Radioisotopen, insbesondere von I-131, umfassend die folgenden Schritte:
    (i) Basische Auflösung von Targets aus angereichertem Uran unter Erhalt einer basischen Aufschlämmung, die Salze von Aluminium, Uran und Isotopen aus der Spaltung von angereichertem Uran und einer Gasphase von Xe-133 enthält,
    (ii) Filtration dieser basischen Aufschlämmung, um einerseits eine feste uranhaltige Phase und andererseits eine basische Lösung von Molybdat und Salzen von Iod-Radioisotopen zu isolieren,
    (iii) Adsorption der Salze von Iod-Radioisotopen an einem silberdotierten Aluminiumoxidharz und Rückgewinnung der basischen Molybdatlösung, die an Iod-Radioisotopen abgereichert ist, insbesondere an I-131, das durch das silberdotierte Aluminiumoxidharz hindurchgeht, und
    (iv) Rückgewinnung der Fraktion von Iod-Radioisotopen, insbesondere von I-131,
    dadurch gekennzeichnet, dass die Rückgewinnung der Fraktion von Iod-Radioisotopen, insbesondere von I-131, ein Waschen des silberdotierten Aluminiumoxidharzes mit einer NaOH-Lösung mit einer Konzentration zwischen 0,2 und 1,5 mol/l, bevorzugt zwischen 0,3 und 1 mol/l und bevorzugter von etwa 0,5 mol/l, und ein Eluieren von Iod-Radioisotopen, insbesondere von I-131, durch eine Thioharnstofflösung mit einer Thioharnstoffkonzentration zwischen 0,5 mol/l und 1,5 mol/l, bevorzugt zwischen 0,8 und 1,2 mol/l, bevorzugter um 1 mol/l, mit einer Sammlung eines Eluats, das die Iod-Radioisotope, insbesondere I-131, enthält, in einer Thioharnstofflösung, umfasst.
  2. Verfahren zur Herstellung einer Fraktion von Iod-Radioisotopen, insbesondere von I-131, nach Anspruch 1, wobei es sich bei den Uran-Targets um Targets aus schwach angereichertem Uran handelt.
  3. Verfahren zur Herstellung einer Fraktion von Iod-Radioisotopen, insbesondere von I-131, nach Anspruch 2, weiter umfassend, vor der Filtration, eine Zugabe von Erdalkalinitrat, insbesondere von Strontium, Calcium, Barium, bevorzugt Barium und Natriumcarbonat, zu der basischen Aufschlämmung.
  4. Verfahren zur Herstellung einer Fraktion von lod-Radioisotopen, insbesondere von I-131, nach einem der Ansprüche 1 bis 3, weiter umfassend eine Säuerung des Eluats, das die Iod-Radioisotope, insbesondere I-131, enthält in einer Lösung von Thioharnstoff durch Zugabe einer Pufferlösung, insbesondere einer Phosphorsäurelösung mit einer Konzentration zwischen 0,5 und 2 mol/l, bevorzugt zwischen 0,8 und 1,5, und bevorzugter etwa 1 mol/l, mit einer Rückgewinnung einer gesäuerten Lösung von Salzen von Iod-Radioisotopen, insbesondere von I-131.
  5. Verfahren zur Herstellung einer Fraktion von Iod-Radioisotopen, insbesondere von I-131, nach Anspruch 4, weiter umfassend eine Reinigung der gesäuerten Lösung von Salzen von Iod-Radioisotopen, insbesondere von I-131, wobei die Reinigung eine Beladung der gesäuerten Lösung mit Salzen von lod-Radioisotopen, insbesondere von I-131, auf eine Ionenaustauschersäule, ein Waschen des lonenaustauscherharzes mit Hilfe von Wasser, ein Eluieren des lonenaustauscherharzes mit Hilfe von NaOH bei einer Konzentration zwischen 0,5 und 2,5 mol/l, bevorzugt zwischen 0,8 mol/l und 1,5 mol/l und besonders bevorzugt etwa 1 mol/l mit einer Rückgewinnung der Fraktion von Iod-Radioisotopen, insbesondere von I-131, in einer NaOH-Lösung, umfasst.
  6. Verfahren zur Herstellung einer Fraktion, die ein Radioisotop von lod-131 enthält, nach Anspruch 5, wobei es sich bei dem lonenaustauscherharz um ein schwach anionisches Harz handelt.
  7. Verfahren zur Herstellung einer Fraktion, die ein Radioisotop von lod 131 enthält, nach einem der vorstehenden Ansprüche, weiter umfassend eine Säuerung der basischen Molybdatlösung, die an Iod-Radioisotopen abgereichert ist, insbesondere an I-131, das durch das silberdotierte Aluminiumoxidharz hindurchgeht, unter Bildung einer sauren Lösung von Molybdänsalzen und Freisetzung von Restradioisotopen von lod, insbesondere I-131, in Form eines Gases zu dessen Rückgewinnung.
  8. Verfahren zur Herstellung einer Fraktion, die ein Radioisotop von lod-131 enthält, nach Anspruch 7, weiter umfassend, vor der Säuerung der basischen Molybdatlösung, die an Iod-Radioisotopen abgereichert ist, insbesondere an I-131, das durch das silberdotierte Aluminiumoxidharz hindurchgeht, eine Abkühlung der basischen Molybdatlösung, die an lod-Radioisotopen abgereichert ist, insbesondere an I-131, das durch das silberdotierte Aluminiumoxidharz hindurchgeht, auf eine Temperatur von weniger als oder gleich 60 °C, bevorzugt weniger als oder gleich 55 °C, insbesondere weniger als oder gleich 50 °C.
  9. Verfahren zur Herstellung von Iod-Radioisotopen, insbesondere von I-131, nach Anspruch 7 oder Anspruch 8, weiter umfassend, nach der Säuerung, ein Erhitzen der sauren Lösung von Molybdänsalzen auf eine Temperatur über 93 °C, bevorzugt über oder gleich 95 °C, bevorzugt zwischen 96 °C und 99 °C, aber bevorzugt unter 100 °C, begleitet von einer Luftblasenbildung.
  10. Verfahren zur Herstellung von Iod-Radioisotopen, insbesondere I-131, nach einem der Ansprüche 7 bis 9, wobei die Rückgewinnung der lod-Radioisotope, insbesondere I-131, bei dessen Freisetzung durch eine Überführung der Iod-Radioisotope, insbesondere I-131, in Form von Gas in ein Rohr, das an einem Ende mit einem Säuerungsmittel, in dem die Säuerung stattfindet, und am anderen Ende mit einem geschlossenen Behälter, der eine wässrige Phase und ein Umgebungsmedium enthält, verbunden ist, durchgeführt wird, wobei die Überführung der Iod-Radioisotope, insbesondere I-131, in Form von Gas, so durchgeführt wird, dass sie direkt in die wässrige Phase führt, in der die lod-Radioisotope, insbesondere I-131, in Form von Gas die wässrige Phase durchlaufen und in Form von Blasen in das Umgebungsmedium der wässrigen Phase entweichen, das in dem geschlossenen Behälter enthalten ist.
  11. Verfahren zur Herstellung von Iod-Radioisotopen, insbesondere I-131, nach Anspruch 10, wobei der geschlossene Behälter durch ein Rohr mit einem zweiten geschlossenen Behälter verbunden ist, der eine NaOH-Falle enthält, und wobei das Umgebungsmedium der wässrigen Phase aus dem geschlossenen Behälter in den zweiten geschlossenen Behälter überführt wird, der die NaOH-Falle in Form einer Lösung in einer Konzentration von 2 bis 4, insbesondere etwa 3 mol/l, enthält, wobei das Umgebungsmedium, das die Iod-Radioisotope, insbesondere I-131, enthält, aus dem Rohr in die NaOH-Fallen-Lösung entladen wird, wobei die Iod-Radioisotope, insbesondere I-131, in Form von Iodidgas von lod-Radioisotopen, insbesondere I-131, in der wässrigen NaOH-Fallen-Lösung gelöst werden.
  12. Verfahren zur Herstellung einer Fraktion, die ein Radioisotop von lod 131 enthält, nach Anspruch 11, wobei die wässrige NaOH-Fallen-Lösung, die die lodide von Iod-Radioisotopen, insbesondere I-131, enthält, eine Rohlösung von lod bildet, die dann durch eine zweite Säuerung gereinigt wird, um gasförmiges lod zu bilden.
  13. Verfahren zur Herstellung einer Fraktion, die ein Radioisotop von lod 131 enthält, nach einem der Ansprüche 5 bis 7, wobei die Fraktion von lod-Radioisotopen, insbesondere I-131, in einer NaOH-Lösung, die lodide von lod-Radioisotopen, insbesondere von I-131, enthält, eine Rohlösung von lod bildet und dann durch eine zweite Säuerung gereinigt wird.
  14. Verfahren zur Herstellung einer Fraktion, die ein Radioisotop von lod 131 enthält, nach Anspruch 12 und Anspruch 13, wobei die Fraktion von lod-Radioisotopen, insbesondere von I-131, in einer NaOH-Lösung und die wässrige NaOH-Fallen-Lösung, die die lodide von Iod-Radioisotopen, insbesondere I-131, enthält, zusammengeführt und gemeinsam durch eine zweite Säuerung gereinigt werden.
  15. Fraktion von Iod-Radioisotopen von I-131, konditioniert in einer NaOH-Lösung mit einer radiochemischen Reinheit der Iod-Radioisotope von I-131 von mehr als 97 %, bevorzugt von mindestens 98 %, insbesondere von mindestens 98,5 % der in der chemischen Iodidform des Radioisotops von I-131 vorhandenen Aktivität, bezogen auf die Gesamtaktivität des Radioisotops von I-131 in allen seinen Formen in dieser Fraktion.
EP17732470.4A 2016-06-28 2017-06-28 Verfahren zur herstellung einer iod-radioisotopenfraktion, insbesondere von i-131, iod-radioisotopenfraktion, insbesondere i-131 Active EP3475954B1 (de)

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PCT/EP2017/065974 WO2018002127A1 (fr) 2016-06-28 2017-06-28 Procede de production d'une fraction de radio-isotopes d'iode, en particulier d'i-131, fraction de radio-isotopes d'iode, en particulier d'i-131

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CN110444310B (zh) * 2019-07-17 2021-03-09 中国原子能科学研究院 一种放射性碘废物的处理方法
CN110444312B (zh) * 2019-09-03 2020-12-29 中国科学院近代物理研究所 利用干馏法从铀裂变产物中分离医用同位素131i的方法
CN112403032A (zh) * 2020-11-18 2021-02-26 中国核动力研究设计院 一种均匀性水溶液核反应堆燃料溶液中99Mo、131I共提取的方法
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EP3475954A1 (de) 2019-05-01
KR20190021251A (ko) 2019-03-05
KR102416164B1 (ko) 2022-07-04
CN109416952B (zh) 2023-12-29
US11017910B2 (en) 2021-05-25
US20190228870A1 (en) 2019-07-25
RU2018145516A3 (de) 2020-07-28
ZA201808651B (en) 2020-05-27
RU2018145516A (ru) 2020-07-28
CN109416952A (zh) 2019-03-01
HUE050258T2 (hu) 2020-11-30
RU2745524C2 (ru) 2021-03-25
BE1023851B1 (fr) 2017-08-14
CA3028852A1 (fr) 2018-01-04

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