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    • 专利摘要:
    A process for producing a pure Mo-99 radioisotope containing fraction comprising a basic dissolution of uranium targets with obtaining a basic slurry, filtering said slurry to isolate a basic solution which is acidified to form an acid solution, a purification of said acid solution, a recovery of said fraction containing said pure Mo-99 radioisotope, characterized in that said purification comprises, after the adsorption step, a first elution of the molybdate by a NaOH solution with recovery of a first molybdate eluate, a passage of said first eluate on an ion exchange resin packaged in water with attachment of said Mo-99 radioisotope to said resin, and recovery of a effluent, and a nitrate elution of said Mo-99 radioisotope by addition of ammonium nitrate with recovery of the Mo-99 radioisotope.
    公开号:BE1023216B1
    申请号:E2016/5491
    申请日:2016-06-28
    公开日:2016-12-22
    发明作者:Robert GENS；Valery Host
    申请人:Institut National Des Radioéléments；
    IPC主号:
    专利说明:

    "PROCESS FOR THE PRODUCTION OF A FRACTION CONTAINING A RADIOSOTOPE OF MO-99 PUR, FRACTION AND GENERATOR CONTAINING THE SAID MO-99 PUR RADIOISOTOPE FRACTION"
    The present invention relates to a process for producing a fraction containing a pure Mo-99 radioisotope comprising the steps of: - Basic dissolution of highly enriched uranium targets to obtain a basic slurry containing aluminate and isotopes derived from the fission of highly enriched uranium and a gaseous phase of Xe-133, - Filtration of said basic slurry in order to isolate, on the one hand, a solid phase containing uranium and, on the other hand, a basic solution of molybdate and iodine salts, - Acidification of said basic solution of molybdate and iodine salts with formation of an acid solution of molybdenum salts and release of iodine in the form of gas in order to its elimination, Purification of said acid solution of molybdenum salts by adsorption of said molybdenum salts on a chromatographic column of alumina, - recovery of said fraction containing said radioisotope of pure Mo-99.
    Such a method is well known and described in the document "Reprocessing of Irradiated Uranium 235 for the production of Mo-99,1-131, Xe-133 radioisotopes. J. Salacz - review IRE tijdschrift, vol 9, No. 3 (1985) ".
    According to this document, the treatment of uranium fission products in order to produce short-lived radioisotopes involves particularly demanding working conditions.
    These particularly demanding working conditions include working in armored cells and using robotic forceps or manipulated from outside the shielded cells by the manipulators of the production line. Once the processes of processing targets containing highly enriched uranium are well established and secure in that the environmental pollution is zero or very low, the process of production of radioisotopes is clearly fixed. The slightest modification of these processes is avoided as far as possible to avoid disturbing the production scheme, insofar as the level of pollution of the environment is considered secure, each modification is considered as a new risk to be managed. in order to obtain a new design satisfying the environmental constraints. In addition, the method is carried out in cells comprising portholes of lead-shielded glass several tens of centimeters in diameter and through which the cells pass articulated arms, manipulated from outside, robotized or not.
    Several cells follow each other. In each cell, part of the process is carried out.
    A first cell is dedicated to the dissolution of highly enriched uranium targets. Once the liquid phase containing the soluble fission products of the uranium recovered by filtration, including the radioisotope Mo-99, it is transferred to a second cell in which it is acidified to deplete iodine. In the second cell, the liquid phase containing the Mo-99 radioisotope is loaded on a column of alumina after removal of the iodine.
    The liquid phase containing the Mo-99 radioisotope recovered following the elution of the alumina column is then transferred through the third cell to the fourth cell where it will be purified on an ion exchange resin and then on activated charcoal.
    The purification of the radioisotope of Mo-99 is thus carried out in the second and in the fourth cell.
    According to this document mentioned above, the alumina resin is washed before elution with a first solution of nitric acid at a concentration of one mol / l, then with water and finally with the aid of a ammonia solution at a concentration of 10 -2 mol / l.
    The alumina column is then eluted with concentrated ammonia and the Mo-99 radioisotope is recovered in the eluate. According to this document, 90% of Mo-99 is thus recovered in a volume of 2 liters of concentrated ammonia for a quantity of 3 aluminum targets containing highly enriched uranium.
    The 2-liter concentrated ammonia solution containing the Mo-99 radioisotope is then loaded onto a Dowex column, onto which the Mo-99 radioisotope binds. The column is then washed with water and eluted with a volume of 200 ml of ammonium carbonate (CO 3 (NH 4) 2).
    The 200 ml fraction of ammonium carbonate containing the Mo-99 radioisotope is then acidified with 6N sulfuric acid (3 mol / l).
    Unfortunately, during the acidification of the ammonium carbonate with 6N sulfuric acid, the resulting solution foams strongly, which makes its manipulation complicated in a constraining environment, in which the slightest overflow causes contamination of the external environment and causes potentially molybdenum losses. In addition, foaming forces the manipulator to use containers of larger capacity, which also increases the loss of molybdenum that would remain during the emptying of the container.
    The fraction containing the radioisotope of Mo-99 acidified with 6N sulfuric acid is then passed through a column of activated carbon. After fixing on the activated carbon column, the activated carbon is washed with water and the Mo-99 radioisotope is recovered after elution with 100 ml of NaOH at a concentration of 0.3 mol / l. The yield of this operation is about 85 to 90% recovery of the radioisotope for an elution of which already 10% were previously lost.
    Unfortunately, as can be seen, this process, although substantially perfected, finally yields only 70 to 80%.
    In addition, the radioisotope of Mo-99 is a substance that is used in nuclear medicine as a precursor to Tc-99 and Tc-99m radioisotopes, hereafter referred to as Te radioisotopes.
    Te radioisotopes are produced by Mo-99 / TC-99 generators in which Mo-99 is attached to an alumina resin. The molybdenum disintegrates to give Tc-99m, which is recovered by rinsing the column (elution) in physiological saline solution in the form of sodium pertechnetate (Na + TcO4-). The generator is then eluted in order to recover a solution (called eluate) of activity necessary for the preparation of the products used in nuclear medicine, typically in a sterile manner.
    Technetium 99m (Tc-99m) is a low half-life isotope, a y-ray emitter. This radioisotope is used in nuclear medicine to perform many diagnoses. Another isotope, technetium 99 (Tc-99), with a longer shelf life, is a source of ß-particles.
    Technetium is spreading faster than many radioisotopes and has low chemical toxicity. The radiological toxicity (per unit mass) depends on the isotope, the type of radiation and the half-life of this isotope.
    Technetium 99m is particularly interesting for medical applications because its radiation is similar to that of X-rays used in conventional radiography. The very short half-life of this isotope conjugated with the relatively long half-life of the Tc-99 isotope son allows it to be removed from the body before disintegrating again. This makes it possible to perform a nuclear diagnosis at the cost of introducing a relatively low dose of radiation into the body (measured in sievert).
    Metastable technetium 99 (Tc-99m) is the most widely used radioisotope in nuclear medical imaging as a marker. Its physical characteristics are almost ideal for this purpose because the half-life of 6 hours is long enough to follow the physiological processes of interest, but short enough to limit unnecessary irradiation. The eluate of Tc-99 for injection must of course have the highest possible purity and therefore the purity of Mo-99 which will then be fixed on the alumina resin in the Mo-99 / Tc-99 generator. It must also be of high purity, especially for isotopes with high activity and a long half-life that can be eluted at the same time as the Tc-99 of the generator.
    There is therefore a need to improve production yields, but also to maximize the purity of the Mo-99 radioisotope fraction. The object of the invention is to overcome the drawbacks of the state of the art by providing a method which makes it possible to maximize the purity of the fraction containing the Mo-99 radioisotope, but also to optimize the production yield of the radio -isotope of Mo-99. In order to solve this problem, it is provided according to the invention a process as indicated at the beginning, characterized in that said purification comprises: after adsorption of said molybdenum salts on said chromatographic column of alumina, a first elution of the molybdate by a solution of NaOH at a concentration between 1.5 and 2.5 mol / l with recovery of a first molybdate eluate, a passage of said first molybdate eluate on an ion exchange resin packaged in water with fixing said Mo-99 radioisotope in molybdate form on said resin, and recovering an effluent from said ion exchange resin, and nitrating eluting said Mo-99 radioisotope in the form of molybdate; the ion exchange resin by addition of ammonium nitrate with recovery of the Mo-99 radioisotope as molybdate in nitrate.
    As can be seen, in the process according to the present invention, the elution of the alumina chromatographic column by a solution of NaOH on the one hand and the use of a column on the other hand ion exchanger eluted by ammonium nitrate has made it possible to obtain a fraction containing said isotope of pure Mo-99, while significantly improving the production yield of the Mo-99 radioisotope.
    Indeed, the elution of the ion exchange resin with ammonium nitrate has freed ammonium carbonate and thus solve the problem of foaming acidification which produced significant losses.
    In addition, the entire process according to the present invention makes it possible to reduce the contamination of the Mo-99 radioisotope fraction by significantly reducing the radioisotope content of Sr-90. This is of considerable importance since the radio-toxicity of the Sr-90 radioisotope is very high by the combination of its long physical radioactive half-life of 28.8 years, its high energy beta radiation and its long biological period (tropism). bony). It is therefore very important to reduce this impurity to minimize potential long-term side effects in the patient.
    In addition, the process according to the present invention makes it possible to obtain a production yield of Mo-99 between 85 and 98%, in particular between 85 and 95% relative to the content of Mo-99 contained in the slurry.
    Advantageously, wherein said ion exchange resin packaged in water is a strong anionic resin. Preferably, wherein said purification comprises, before said first elution of said molybdenum salts with a solution of NaOH at a concentration of between 1.5 and 2.5 mol / l, preferably between 1.7 and 2.2 mol more preferably about 2 mol / l and a washing of said column with water which is then recovered in the form of an elution effluent from said alumina chromatographic column.
    In a particular embodiment of the process according to the present invention, a second elution of said alumina chromatographic column with a solution of NaOH at a concentration of between 1.5 and 2.5 mol / l is carried out as well as a recovery. a second molybdate eluate at a time between 20 hours and 48 hours after harvesting said first molybdate eluate.
    In this way, the molybdate salts have had sufficient time to transform so that they can be eluted again and recovered as products of the process.
    Preferably, said second molybdate eluate is passed over an ion exchange resin packaged in water, with said Mo-99 radioisotope being fixed on the aforementioned resin, and recovery of an effluent from said resin exchange resin. ions, and further comprising, prior to recovering said fraction containing the pure Mo-99 radioisotope, a nitrate elution of said Mo-99 radioisotope from said second eluate of said ion exchange resin by addition of ammonium nitrate with recovery of the Mo-99 radioisotope in nitrate.
    In another advantageous embodiment of the process according to the present invention, said nitrate elution of said Mo-99 radioisotope from the ion exchange resin is preceded by a washing step after adsorption of said Mo-radioisotope. 99 to water which is recovered and mixed with said effluent of said ion exchange resin.
    More particularly, according to the present invention, said step of recovering a fraction containing the pure Mo-99 radioisotope comprises an acidification of said first molybdate eluate and / or optionally said second molybdate eluate with a sulfuric acid solution. at a concentration of between 1 and 2 mol / l, preferably equal to 1.5 mol / l, forming an acidified fraction of pure Mo-99 radioisotope in the form of molybdenum salts and purification on a carbon column active, possibly doped with silver.
    In yet another advantageous embodiment of the process according to the present invention, said purification of said acidified fraction of Mo-99 radioisotope on a column of activated carbon, optionally doped with silver, comprises a step of loading on a column. of activated charcoal, a step of washing said activated carbon column on which the Mo-99 radioisotope is attached to the water and an elution step with a 0.3 mol / l NaOH solution forming a eluate of Na299MoO4 in 0.2 mol / l NaOH solution forming the fraction containing said pure Mo-99 radioisotope. Other embodiments of the process according to the invention are indicated in the appended claims. The invention also relates to a fraction containing said pure Mo-99 radioisotope having a radiochemical purity of at least 97%, preferably at least 98%, more preferably at least 98.5%, and the most preferential way of at least 99% of the activity present in the molybdate chemical form of said Mo-99 radioisotope relative to the total activity of said Mo-99 radioisotope in all its forms in said fraction. The invention also relates to a radioisotope fraction of Mo-99 packaged in a 0.2 mol / l NaOH solution having a radioisotope radioisotope purity of Mo-99 greater than 98.5% of the activity present in the chemical form molybdate of said radioisotope of
    Mo-99 with respect to the total activity of said Mo-99 radioisotope in all its forms in said fraction.
    In an advantageous variant, the radioisotope fraction of Mo-99 is packaged in a 0.2 mol / l NaOH solution in sealed flasks which are enclosed in individual shielded containers. Other embodiments of the fraction according to the invention are indicated in the appended claims. The subject of the invention is also a radioisotope fraction of Mo-99 packaged in a 2 mol / l NaOH solution having a radioisotope purity of Mo-99 greater than 98.5% of the present activity. in the molybdate chemical form of said Mo-99 radioisotope relative to the total activity of said Mo-99 radioisotope in all its forms in said fraction.
    In an advantageous variant, the radioisotope fraction of Mo-99 is conditioned in a 2 mol / l NaOH solution in sealed flasks which are enclosed in individual shielded containers.
    In a fraction according to the invention, additives may be present such as, for example, ammonium nitrate, NaOCI, sodium nitrate and the like. Other embodiments of the fraction according to the invention are indicated in the appended claims.
    The present invention also relates to a radioisotope moiety of Mo-99 which is obtained by the process of the present invention. Other embodiments of the fraction obtained by the process according to the invention are indicated in the appended claims.
    The present invention also relates to a Tc-99 generator derived from a stock solution comprising a Mo-99 radioisotope fraction according to the present invention. Other embodiments of the generator according to the invention are indicated in the appended claims. Other features, details and advantages of the invention will emerge from the description given below, without limitation and with reference to examples.
    In the context of the present invention, the term "resin effluent" means the mobile phase which passes through the resin and leaves the chromatographic column.
    When uranium 235 is bombarded with neutrons, it forms fission products of a lower mass which are themselves unstable. These products generate via a fission cycle other radioisotopes. This is how the radioisotopes of Mo-99, Xe-133 and 1-131 are produced.
    Aluminum targets containing highly enriched uranium are dissolved during a basic dissolution phase in the presence of NaOH (3 mol / l) and NaNO 3 (4 mol / l). During the dissolution, a slurry is formed as well as a gaseous phase of Xe-133. The slurry contains a solid phase consisting mainly of uranium and fission product hydroxides and a liquid phase of molybdate (MoO4) and iodine-131 as iodine salts.
    The slurry is then diluted with water to a volume of about 1.8 liters for a number of targets ranging from 3 to 9 to allow its transfer to the subsequent step.
    The gaseous phase of Xenon is recovered separately.
    The dissolution of the aluminum of the target is an exothermic reaction. The slurry containing the solid phase and the basic liquid phase is then filtered. 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 acidic dissolution. The filtrate (recovered basic liquid phase containing the fission products of Mo-99, ΙΊ-131, ΓΙ-133, ΓΙ-135, Cs-137, Ru-103, Sb-125 and Sb- 127) but also aluminate formed by the basic dissolution of the aluminum targets, which is soluble at basic pH. Aluminum is soluble in basic medium as well as in acid medium. On the other hand, it is insoluble when the pH is between 5 to 10
    The filtrate collected must then be acidified. However, the acidification also causes a release of heat. Therefore, before acidification, the filtrate is cooled to a temperature of about 50 ° C.
    Indeed, as is known from the "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 The conversion of the Al 3+ ion to the precipitated form of hydroxide and the soluble form of aluminate are subject to some kinetics.
    The formation of metastable solids is known and equilibrium conditions are sometimes difficult to achieve even with large reaction times. Oxides and hydroxides of aluminum form different crystalline structures (bayerite, gibbsite, ...) which are chemically similar but different in their solubility. The experimental conditions of temperature, concentration and also the rate of addition of the reagents strongly influence the results obtained.
    The reaction that governs the balance between different forms of aluminum is as follows at the time of acidification
    Since 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 overconcentrations at localized locations. acid causing local heating by the neutralization reaction, and the formation of insoluble aluminum forms or kinetics of slow redissolution of aluminum salts. However, given the constraints of the method described in the state of the art, the reaction medium has a high temperature, is highly radioactive and difficult to access, it is not possible to maintain stirring to avoid these points of high temperature aluminate concentration.
    The effects of overconcentrations in acid should be avoided to avoid two main reasons. On the one hand, the formation of precipitates of aluminum salts causes a significant risk of clogging of the installation, which affects the yield of production, but also adds a health risk in view of the high radioactivity of the reaction mixture. It is indeed not simple, or even unthinkable to intervene manually to unclog the installation, but in addition, this could only be done at the expense of production efficiency.
    Therefore, the filtrate is cooled to avoid precipitation of aluminum salts during acidification at a temperature of about 50 ° C and in all cases less than 60 ° C. The filtrate is thus acidified with concentrated nitric acid (65%). The radioisotopes of iodine are then released during acidification. To promote the release of iodine, 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. Acidification makes it possible to obtain an acidic pH solution in order to be able to fix the Mo-99 radioisotope on the alumina column (and to release the radioisotopes of iodine with a view to their subsequent recovery. .
    The acidified liquid phase, depleted in iodine, is then loaded onto a column of alumina, conditioned in 1 mol / l nitric acid. The
    Mo-99 is adsorbed on alumina while the majority of the contaminating fission products are removed in the alumina column effluent.
    The column of alumina on which the Mo-99 radioisotope is attached is washed with nitric acid at a concentration of one mol / l, with water, sodium sulphite at a concentration of about 10 g / l and finally again to the water. Wash effluent is discarded
    The alumina column is then eluted with NaOH at a concentration of about 2 mol / l and then with water. The eluate recovered from the alumina column forms the first eluate of the Mo-99 radioisotope.
    In a preferred mode of the process according to the present invention, the first eluate of the column is stored for a period of time between 20 hours and 48 hours. After this predetermined period of time, the alumina column is again eluted with NaOH at a concentration of about 2 mol / l and then with water, before cleaning. The eluate of the new elution forms the second eluate of the Mo-99 radioisotope. At this stage, the first eluate of the Mo-99 radioisotope is combined with the second eluate of the Mo-99 radioisotope and forms a single eluate that will still undergo the subsequent purification steps. Either each first and second eluates are treated separately in subsequent purification steps in the same manner. For simplicity, the radioisotope eluate of Mo-99 will now be referred to as the first eluate of the Mo-99 radioisotope or the second eluate of the Mo-99 radioisotope. or two together. The radioisotope eluate of Mo-99 from the alumina column is then loaded onto a second chromatographic column containing a strong anionic ion exchange resin pre-conditioned in water.
    The ion exchange column is then eluted with a solution of ammonium nitrate at a concentration of about 1 mol / l. The recovered eluate thus comprises the Mo-99 radioisotope in a fraction containing ammonium nitrate.
    The ammonium nitrate solution containing the Mo-99 radioisotope is then loaded onto a 35-50 mesh activated carbon column, which may be optionally doped with silver to recover any traces of iodine. The activated carbon column on which the Mo-99 radioisotope is attached is then washed with water and then eluted with a solution of NaOH at a concentration of about 0.3 mol / l. Elution of the activated carbon column makes it possible to recover a solution of Na299MoO4 in NaOH but to keep any iodine captured on the column at a preferred concentration of 0.2 mol / l, which will then be packaged and packaged for delivery. .
    In a particular embodiment of the invention, the solution of Na299MoO4 in NaOH at a preferred concentration of 0.2 mol / l is loaded onto an alumina resin in a Mo-99 / Tc-99 generator or on a titanium oxide resin to enable the generation of technetium 99 radioisotope for nuclear medicine.
    During the formation of the mud, fission products of uranium are released, some in soluble form, others in the form of gas. It is among others the case of xenon and krypton which are therefore in a gaseous phase. The gaseous phase leaves the liquid medium and remains confined in the sealed container in which the dissolution takes place. The sealed container comprises a gas phase outlet connected to a rare gas recovery device, isolated from the external environment, but also an inlet for a purge gas.
    The gaseous phase contains ammonia (NH3) which is derived 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 imposes two large refrigerants. Nevertheless, water vapor is present in the gas phase. The gaseous phase is carried by a carrier gas (He) to the noble gas recovery device.
    The aqueous phase leaves the basic dissolution tight container and is fed to the noble gas recovery device. The gaseous phase containing among others the Xe-133 radioisotope is first passed through a molecular sieve to remove ammonia (NH3) and water vapor. Then, the gaseous phase is passed through silica gel in order to eliminate any trace of residual water vapor
    The gaseous phase containing inter alia the radioisotope Xe-133 is then brought into a U-shaped tube immersed in liquid nitrogen (ie at -196 ° C) contained in a shielded container, through clippings. stainless steel.
    The stainless steel trimmings are made from 316 stainless steel rod having a diameter of between 1.5 and 2 cm and a length of between 10 and 20 cm, preferably between 14 and 18 cm, more particularly of about 16 cm using a 4-lip burr with a diameter of 16 mm using a hydraulic vice. The speed of the milling machine comprising the aforementioned milling cutter is 90 rpm and set with a forward speed of 20 mm / min. The depth of the cutter is about 5 mm.
    The stainless steel shavings have an average weight of between 20 and 30 mg / claw, preferably between 22 and 28 mg / claw and a non-typed density when shaped between 1.05 and 1.4.
    The U-tube has a clipping amount of between 90 g and 110 g. The volume of 316 stainless steel shavings included in the U-tube is fully immersed in liquid nitrogen.
    The radioisotope Xe-133 from said gaseous phase containing the radioisotope Xe-133 is then captured by liquefaction of said Xe-133 by means of said cooled stainless steel clippings, which capture the Xe-133 by condensation.
    The liquefaction temperature of Xe-133 is around -107 ° C. Therefore, Xe gas is condensed in liquid form on stainless steel shavings.
    On the other hand, since the liquefaction temperature of Kr-85 is around -152 ° C, there is a much smaller amount of Kr that is trapped in the liquid nitrogen trap and the residual Kr is collected in specific traps with the gases resulting from the process described here, namely, inter alia the gaseous phase substantially depleted in Xe-133.
    Once the Xe-133 has been captured by the liquid nitrogen trap, the lines are purged and the liquid nitrogen injection is cut off and the trap is brought into contact with a vacuum bulb whose volume is 50 times more large than the volume of clippings 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 ampoule.
    As mentioned above, the acidification of the basic slurry makes it possible to obtain an acidic pH solution which allows the binding of the radioisotope of Mo-99 to the alumina column but also to release the radioisotopes of the iodine for recovery.
    The recovery of the iodine is carried out during and after the acidification of the previously cooled basic filtrate.
    The radioisotopes of iodine are evolved 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 below 100 ° C and maintained under bubbling to promote the release of iodine in gaseous form.
    During the heating of the acidified filtrate, a gas phase is formed which contains the radioisotopes of iodine but also a part of the filtrate which has evaporated. The acidifier has an aqueous phase outlet tubing which dips into a closed container containing 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. In this way, the part of the filtrate which 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 the container. water in the closed container and leave it by means of the outlet pipe of the closed container to a second closed container, namely a trap containing NaOH at a concentration of 3 mol / l. The radioisotopes of iodine are then converted into iodide iodine salts and solubilized in the NaOH contained in the iodine trap.
    The fraction containing the radioisotopes of iodine is then packaged in hermetic bottles contained in an armored enclosure for shipment to the customer.
    EXAMPLES
    Example 1 according to the invention 3 aluminum targets containing highly enriched uranium are dissolved during a basic dissolution phase in the presence of 1.5 liters of NaOH (3 mol / l) and NaNO 3 (4 mol / l) . The slurry is then diluted with water to a volume of about 1.8 liters.
    The slurry containing the solid phase and the basic liquid phase is then filtered. The solid phase is washed twice with a volume of water of 900 ml, recovered and discarded.
    The filtrate is then cooled before acidification to a temperature of about 50 ° C. To promote the release of iodine, the acidified filtrate by adding 1.5 liters of concentrated nitric acid is heated to a temperature above 95 ° C and kept bubbling.
    The acidified liquid phase, depleted in iodine, is then loaded onto a column of alumina, conditioned in 1 liter of 1 mol / l nitric acid for a volume of alumina of approximately 500 ml.
    The column of alumina on which the Mo-99 radioisotope is attached is then washed successively with 1250 ml of 1M HN0 3; 1500 ml of H 2 O; 1000 ml of Na2SO3; 1000 ml H20 Wash effluent is discarded
    The alumina column is then eluted with 1 liter of NaOH at a concentration of about 2 mol / l and then with 500 ml of water. The eluate recovered from the alumina column forms the first eluate of the Mo-99 radioisotope which is purified on the strong anionic and activated carbon ion exchange resin as described below for elution of the day. next.
    The next day, the column of alumina is eluted again under the same conditions before cleaning. The eluate of the new elution forms the second eluate of the Mo-99 radioisotope. The radioisotope eluate of Mo-99 from the alumina column is then loaded onto a second column of strong anionic ion exchange resin previously conditioned in water.
    The ion exchange column is then washed with 400 ml of water and eluted with 120 ml of a solution of ammonium nitrate at a concentration of about 1 mol / l.
    Recovered eluates comprising the Mo-99 radioisotope in a fraction containing ammonium nitrate from different elutions of the alumina column can be pooled. The recovered eluate thus comprises the Mo-99 radioisotope in a fraction containing ammonium nitrate which is then acidified with 15 ml of 1.5 mol / liter sulfuric acid.
    The ammonium nitrate solution containing the Mo-99 radioisotope is then loaded onto a column of active carbon with a volume of 30 ml. The activated carbon column on which the Mo-99 radioisotope is attached is then washed with water (400 ml) and then eluted with a solution of NaOH at a concentration of about 0.3 mol / l (110 ml). ml). Elution of the activated carbon column makes it possible to recover a solution of Na299MoO4 in NaOH at a preferred concentration of 0.2 mol / l.
    The yield of radioisotope fraction of Mo-99 was about 88% for the two pooled fractions.
    The characteristics of the Mo-99 radioisotope fraction of the pooled fractions had the characteristics shown in Table 1.
    Table 1.-
    Example 2 according to the invention
    A series of tests was carried out on aluminum targets treated according to Example 1.
    The results are shown in Table IB below and indicate the radiochemical purity of said fraction containing said radioisotope of pure Mo-99, expressed as a percentage of the activity present in the molybdate chemical form of said Mo-99 radioisotope. relative to the total activity of said Mo-99 radioisotope in all its forms in said fraction.
    Table Ibis-.
    Comparative Example.-
    The process according to the document "Reprocessing of Irradiated Uranium 235 for the production of Mo-99, 1-131, Xe-133 radioisotopes. J.
    Salacz - review IRE tijdschrift, vol 9, No. 3 (1985) "was reproduced from 3 aluminum targets containing highly enriched uranium.
    The production yield of radioisotope fraction of Mo-99 was about 72%.
    The characteristics of the radioisotope fraction of Mo-99 had the characteristics shown in Table 2.
    Table 2.-
    As can be seen, compared to the process according to the invention, in the comparative example, the activity content (a total) / Mo-99 and (total β, γ) / Mo-99 is much lower, on the other hand, the Sr content is reduced in the Mo-99 sample, which is particularly advantageous in view of its radio-toxicity is very high by the combination of its long physical radioactive half-life of 28.8 years). high energy beta radiation and its long biological period
    It is understood that the present invention is in no way limited to the embodiments described above and that many modifications can be made without departing from the scope of the appended claims.
    权利要求:
    Claims (16)
    [1]
    1. Process for producing a fraction containing a pure Mo-99 radioisotope comprising the steps of: - Basic dissolution of highly enriched uranium targets to obtain a basic slurry containing aluminate and isotopes derived from fission of highly enriched uranium and a gaseous phase of Xe-133, - Filtration of said basic slurry in order to isolate on the one hand a solid phase containing uranium and on the other hand a basic solution of molybdate and iodine salts, - Acidification of said basic solution of molybdate and iodine salts with formation of an acid solution of molybdenum salts and release of iodine as a gas for the purpose of its removal, Purification of said acid solution of molybdenum salts by adsorption of said molybdenum salts on an alumina chromatographic column, - recovery of said fraction containing said pure Mo-99 radioisotope, characterized in that said purification further comprises: after adsorption of said molybdenum salts on said chromatographic column of alumina, a first elution of the molybdate by a solution of NaOH at a concentration of between 1.5 and 2.5 mol / l with recovery of a first molybdate eluate, a passage of said first molybdate eluate on an ion exchange resin packaged in water with binding of said Mo-99 radioisotope in molybdate form to said resin, and recovering an effluent from said resin exchange resin; ions, and - a nitrate elution of said Mo-99 radioisotope in molybdate form of the ion exchange resin by addition of ammonium nitrate with recovery of the Mo-99 radioisotope in the form of molybdate in nitrate.
    [2]
    The method of claim 1, wherein said ion exchange resin packaged in water is a strong anionic resin.
    [3]
    The process according to claim 1 or claim 2, wherein said purification comprises, prior to said first elution of said molybdenum salts by said NaOH solution at a concentration of between 1.5 and 2.5 mol / l, a wash of said column with water which is then recovered in the form of an elution effluent from said chromatographic column of alumina.
    [4]
    4. Process according to any one of the preceding claims, comprising a second elution of the alumina chromatographic column with a solution of NaOH at a concentration of between 1.5 and 2.5 mol / l and recovery of a second eluate. molybdate at a period of time between 20 hours and 48 hours after harvest of said first molybdate eluate.
    [5]
    The process according to claim 4, wherein said second molybdate eluate is passed over an ion exchange resin packaged in water, with said Mo-99 radioisotope being fixed to the above-mentioned resin, and recovery of an effluent of said ion exchange resin, and further comprising, prior to recovery of said fraction containing the pure Mo-99 radioisotope, a nitrate elution of said Mo-99 radioisotope from said second eluate of said resin ion exchange by addition of ammonium nitrate with recovery of the Mo-99 radioisotope in nitrate.
    [6]
    The method according to any one of the preceding claims, wherein said nitrate elution of said Mo-99 radioisotope from the ion exchange resin is preceded by a washing step after adsorption of said Mo-radioisotope. 99 to water which is recovered and mixed with said effluent of said ion exchange resin.
    [7]
    A process according to any one of the preceding claims, wherein said step of recovering a fraction containing the pure Mo-99 radioisotope comprises acidification of said first molybdate eluate and / or optionally said second molybdate eluate, with a sulfuric acid solution at a concentration of between 1 and 2 mol / l, preferably at 1.5 mol / l, forming an acidified fraction of pure Mo-99 radioisotope in the form of molybdenum salts and a purification on a column of activated carbon, optionally doped with silver.
    [8]
    The process according to claim 7, wherein said purifying said acidified Mo-99 radioisotope fraction on an activated carbon column comprises a step of charging on a column of activated carbon, optionally doped with silver, a washing step of said activated carbon column on which the Mo-99 radioisotope is attached to water and an elution step with a solution of NaOH at 0.3 mol / l forming a Na299MoO4 eluate in a 0.2 mol / l NaOH solution forming the fraction containing said pure Mo-99 radioisotope.
    [9]
    The process according to any one of claims 1 to 8, wherein said fraction containing said pure Mo-99 radioisotope has a radiochemical purity of at least 97%, preferably at least 98%, more particularly at least 98.5%, and most preferably at least 99% of the activity present in the molybdate chemical form of said Mo-99 radioisotope relative to the total activity of said radio-isotope of Mo-99; isotope of Mo-99 in all its forms in said fraction.
    [10]
    10. Fraction containing a radioisotope of Mo-99 packaged in a 0.2 mol / l solution of NaOH with a radiochemical purity in Mo-99 greater than 98.5% of the activity present in the chemical molybdate form of said radioisotope of Mo-99 relative to the total activity of said Mo-99 radioisotope in all its forms in said fraction.
    [11]
    11. Fraction containing a Mo-99 radioisotope packaged in a 0.2 mol / l solution of NaOH in sealed flasks that are enclosed in individual shielded containers.
    [12]
    12. Fraction containing a radioisotope of Mo-99 conditioned in a 2 mol / l solution of NaOH with a radiochemical purity in Mo-99 of greater than 98.5% of the activity present in the molybdate chemical form of said radio- isotope of Mo-99 with respect to the total activity of said Mo-99 radioisotope in all its forms in said fraction.
    [13]
    13. Fraction containing a radioisotope of Mo-99 packaged in 2 mol / l NaOH solution in sealed flasks which are enclosed in individual shielded containers.
    [14]
    14. A fraction according to any one of claims 10 to 13 containing additives such as ammonium nitrate, NaOCI, sodium nitrate and the like.
    [15]
    15. Fraction containing a radioisotope of pure Mo-99 obtained by the process according to any one of claims 1 to 9.
    [16]
    16. A Tc-99 generator derived from a stock solution comprising a Mo-99 radioisotope fraction according to any one of claims 10 to 15.
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