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EP0170796B1 - Process for separating great amounts of uranium from small amounts of fission products which are present in aqueous basic solutions containing carbonate - Google Patents

Process for separating great amounts of uranium from small amounts of fission products which are present in aqueous basic solutions containing carbonate Download PDF

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Publication number
EP0170796B1
EP0170796B1 EP85105864A EP85105864A EP0170796B1 EP 0170796 B1 EP0170796 B1 EP 0170796B1 EP 85105864 A EP85105864 A EP 85105864A EP 85105864 A EP85105864 A EP 85105864A EP 0170796 B1 EP0170796 B1 EP 0170796B1
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concentration
uranium
fission products
aqueous solution
solution
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German (de)
French (fr)
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EP0170796A2 (en
EP0170796A3 (en
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Abdel Hadi Ali Dr. Sameh
Jürgen Haag
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Forschungszentrum Karlsruhe GmbH
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Kernforschungszentrum Karlsruhe GmbH
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/04—Treating liquids
    • G21F9/06—Processing
    • G21F9/12—Processing by absorption; by adsorption; by ion-exchange

Definitions

  • the invention relates to a process for separating large amounts of uranium from small amounts of radioactive fission products which are present in aqueous basic, carbonate-containing solutions, using an organic, basic anion exchanger.
  • nuclear reactor fuel elements for recycling irradiated nuclear fuels from compounds or alloys of highly enriched uranium have been dissolved in nitric acid and the uranium by liquid / liquid extraction, e.g. in the Purex process or in the amine extraction or in column-chromatographic separation operations, separated and worked up in nitric acid medium.
  • the elements mainly consist of aluminum-coated uranium / aluminum alloy with the approximate composition UAl3; Because of the fluctuating Al content in the compound, the designation UAl x is mostly used.
  • This type of fuel element is often used as a starting target for the production of fission product nuclides for nuclear medicine and technology; mostly smaller elements with thermal neutron fluxes of approx. Irradiated for 5 to 10 days. In order to minimize the decay losses of the desired nuclide, the targets are transported to the processing plant after a minimum cooling time of approx. 12 hours.
  • the first chemical step is usually an alkaline digestion of the target with 3 to 6 molar sodium hydroxide solution or potassium hydroxide solution;
  • the main constituent of the plate, aluminum, and the fission products soluble in this medium such as the alkali and alkaline earth cations, as well as antimony, iodine, tellurium, tin and molybdenum, go into solution, while the volatile fission products, especially xenon, go together with the Hydrogen formed from the Al solution, leave the dissolver at the upper end of the reflux condenser.
  • the hydrogen can be oxidized to water via CuO, while the xenon is preferably retained at normal temperature on activated carbon delay lines.
  • This residue is treated in a manner known per se under the action of air or an oxidizing agent, e.g. H2O2 or hypochlorite, treated with an aqueous solution containing carbonate and hydrogen carbonate ions from pH 5 to pH 11.
  • the concentration of the carbonate ions in this solution can be a maximum of 2.5 M / l, that of the hydrogen carbonate ions a maximum of approximately 1.0 M / l.
  • the oxides of uranium and the fission product species mentioned go into solution as carbonato complexes.
  • the invention is based on the object of providing a method with which uranium values present in an aqueous, basic, carbonate-containing solution, on the one hand, of fission products from the ruthenium group, Zirconium, niobium and lanthanoids, on the other hand, can be separated from one another with a relatively high degree of decontamination.
  • the process of the invention is said to be able to obtain largely decontaminated uranium or the fission products ruthenium, zirconium, niobium and lanthanoids after the alkaline digestion of a fuel element of a material test reactor (MTR).
  • MTR material test reactor
  • the process is said to be operationally reliable and low-waste and to be applicable to residues containing uranium dioxide and alkalidiuranate that have only cooled down for a few days.
  • the aqueous solution is adjusted to a ratio of the uranyl ion concentration to carbonate ions / hydrogen carbonate ion concentration of 1: 5 to 1: 8.
  • the aqueous solution is adjusted at a uranium concentration of 60 g / l to a ratio of UO2++ concentration to CO3 ⁇ / HCO3 ⁇ concentration of 1: 5.
  • the aqueous solution advantageously has a hydrogen carbonate ion concentration between 0 and 1 mol / l.
  • the CO3 ⁇ concentration in the aqueous solution is a maximum of 2.5 M / l and the pH of the aqueous solution is in the range from pH 7 to pH 11.
  • the process according to the invention can also be carried out in the absence of HCO3 ⁇ ions, but the process conditions can be set more easily if HCO3 ⁇ ions are present in the aqueous solution.
  • the application of the method spans a large concentration fluctuation range of the uranium to be decontaminated. Is the uranium concentration in the solution compared to the carbonate concentration very small, so that, for example, a free CO3 ⁇ / HCO3 ⁇ concentration is greater than 0.6 mol / l, the excess carbonate excess can either be optimized by metering in a mineral acid, preferably HNO3, or optimized to optimize the fission product retention a certain amount of carbonate ions can be trapped by adding, for example, Ca (OH) 2.
  • the uranium distribution coefficient must be minimized by adding sufficient amounts of CO3 ⁇ / HCO3 ⁇ ions so that the fission product species are not displaced by the uranium from the ion exchanger.
  • the desired separations can still be carried out at uranium concentrations of approx. 60 g U / l.
  • the limitation of the process to higher U concentrations is due to the uranium solubility in carbonate-hydrogen carbonate solutions.
  • a method for the separation of actinide ions from aqueous, basic, carbonate-containing solutions from German Offenlegungsschrift 31 44 974 was known, in which the actinide ions as carbonato complexes are adsorbed on basic ion exchangers and after separation of the loaded ion exchanger from the starting solution using an aqueous solution are desorbed and further processed by the ion exchanger, and in which a basic anion exchanger from a is used with a predominantly tertiary and to a small extent quaternary ammonium group-provided polyalkene matrix, but this method is only meaningfully applicable to aqueous, carbonate-containing waste solutions or washing solutions etc.
  • the main advantages of the method according to the invention are that the decontamination of the uranium from the fission products still present can be carried out with a relatively small amount of the anion exchanger, for example in a relatively small ion exchange column, that the ion exchanger loaded with the fission products can be used in the event that only the uranium values are to be recovered (with or without a column) without intermediate treatment directly can be given for waste treatment and disposal and, in the event that the fission product nuclides are to be obtained, can be carried out for further processing of the fission product nuclides and separation from one another.
  • the cleavage products can be eluted from the ion exchange column with an alkali or ammonium carbonate solution of higher molarity (approx. 1 to 2 M / l) or with nitric acid.
  • the method according to the invention is characterized by very reliable process control.
  • the organic anion exchanger does not have to be brought into contact with corrosive or strongly oxidizing media in any phase of the process.
  • the method according to the invention works with basic media which offer the highest possible security against the release of volatile iodine components.
  • the solution used in the process according to the invention which can contain up to a maximum of 2.5 mol / l Na2CO3 and, at a lower CO3 ⁇ concentration, up to approx. 1 mol / l NaHCO3, is chemically very easy to control and radiation-chemical resistant. There are no corrosion problems.
  • the outlay in chemicals, apparatus and working time is very low in the process according to the invention.
  • the average fission product retention during a column run under the specified loading conditions was> 97% for cerium, zirconium and niobium; With ruthenium the retention was approx. 80%.
  • Task solution Volume 100 ml U content: 1.19 g U: CO3 ⁇ / HCO3 ⁇ : 1: 7 or 1: 6 Na2CO3: 3.24 g ⁇ 90% or 2.78 g NaHCO3: 0.28 g ⁇ 10% or 0.24 g column Diameter: 15 mm Height: 130 mm Bed volume: 20 ml Feed speed: ⁇ 0.5 ml / cm2 ⁇ sec. Rinsing: 0.2 molar Na2CO3 solution *) Number of fractions: 4 x 20 *) Instead of a Na2CO3 solution, a corresponding other alkali or ammonium carbonate solution can also be used.
  • Ion exchanger moderately basic anion exchanger made of polyalkene-epoxypolyamine with tertiary and quaternary ammonium groups with the trade name Bio-Rex 5 (from Bio-Rad Laboratories, USA).

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Inorganic Compounds Of Heavy Metals (AREA)
  • Manufacture And Refinement Of Metals (AREA)

Description

Die Erfindung betrifft ein Verfahren zur Trennung von großen Mengen Uran von geringen Mengen von radioaktiven Spaltprodukten, die in wäßrigen basischen, karbonathaltigen Lösungen vorliegen, unter Verwendung eines organischen, basischen Anionenaustauschers.The invention relates to a process for separating large amounts of uranium from small amounts of radioactive fission products which are present in aqueous basic, carbonate-containing solutions, using an organic, basic anion exchanger.

Bisher wurden Kernreaktorbrennelemente zur Rezyklierung bestrahlter Kernbrennstoffe aus Verbindungen bzw. Legierungen von hochangereichertem Uran in Salpetersäure gelöst und das Uran durch Flüssig/Flüssig-Extraktion, wie z.B. im Purex-Prozeß oder bei der Aminextraktion oder bei säulenchromatographischen Trennoperationen, abgetrennt und in salpetersaurem Medium aufgearbeitet.So far, nuclear reactor fuel elements for recycling irradiated nuclear fuels from compounds or alloys of highly enriched uranium have been dissolved in nitric acid and the uranium by liquid / liquid extraction, e.g. in the Purex process or in the amine extraction or in column-chromatographic separation operations, separated and worked up in nitric acid medium.

Die salpetersaure Rezyklierung von Kernbrennstoffen, vor allem der Purex-Prozeß, ist eine längst bekannte und bewährte Verfahrensweise; sie ist jedoch bei der Aufarbeitung kurz abgekühlter Targets (Kühlungsdauer beispielsweise 1 - 30 Tage) äußerst problematisch. Die Nachteile sind folgende:

  • Die Anwesenheit der kürzer lebigen Spaltprodukte, vor allem Jod-131 und Xenon-133, macht den Einsatz von Rückhalte- bzw. Verzögerungsstrecken zwingend notwendig. Durch die Verwendung von Salpetersäure - andere Säuren scheiden aufgrund ihrer Korrosivität aus - und der damit verbundenen Möglichkeit einer NO₂-Entwicklung darf das wirksamste und zugleich wirtschaftlichste Filtermaterial Aktivkohle nicht eingesetzt werden, da sonst im Falle einer NO₂-Freisetzung akute Brandgefahr in den Abgasstrecken bestünde.
  • Sämtliche Flüssig/Flüssig-Extraktionsprozesse sind für hochgradig mit J-131 und Xe-133 befrachteten Systeme (wie in diesem Fall) außerordentlich schwer beherrschbar, da neben der Xe-133-Emissionsgefahr die weitere, erheblich folgenschwerere Möglichkeit der HJ- und Jod-Emission aus dem sauren System besteht.
    Ein weiterer Nachteil der Flüssig/Flüssig-Extraktion ist der notwendige erhöhte Aufwand, um die Brandgefahr zu vermeiden, verursacht durch den Extraktionsmittelverdünner. Der Einsatz von nicht brennbaren Verdünnern, wie Tetrachlorkohlenstoff, ist in diesem extrem hochaktiven System aufgrund der ausgeprägten Strahlenempfindlichkeit und der erhöhten Korrosionsgefahr durch die freiwerdende Salzsäure nicht empfehlenswert.
  • Sämtliche bisher bekannten leistungsfähigen extraktionschromatografischen Verfahren erfolgen in sauren Systemen und beinhalten neben dem bereits angeführten Nachteil der HJ- bzw. J₂-Freisetzung einen weiteren großen Nachteil und zwar die Fixierung des Urans, dem Hauptanteil im Prozeßstrom, bei verminderter Rückhaltung der Spaltprodukte. Der Nachteil dieser Vorgehensweise liegt auf der Hand: Hier müssen zur Kernbrennstoffrückhaltung unvergleichbar größere Kolonnenvolumina bereitgestellt werden.
The nitric acid recycling of nuclear fuels, especially the Purex process, is a well-known and proven procedure; however, it is extremely problematic when processing briefly cooled targets (cooling time, for example, 1-30 days). The disadvantages are as follows:
  • The presence of the shorter, lively fission products, especially iodine-131 and xenon-133, makes the use of retention or delay lines imperative. By using nitric acid - other acids differ due to their corrosiveness from - and the associated possibility of NO₂ development, the most effective and at the same time most economical filter material activated carbon must not be used, since otherwise there would be an acute fire hazard in the exhaust system in the event of NO₂ release.
  • All liquid / liquid extraction processes are extremely difficult to control for systems heavily loaded with J-131 and Xe-133 (as in this case), because in addition to the Xe-133 emission risk there is the further, considerably more serious possibility of HJ and iodine emissions consists of the acidic system.
    Another disadvantage of liquid / liquid extraction is the increased effort required to avoid the risk of fire caused by the extractant thinner. The use of non-flammable thinners, such as carbon tetrachloride, is not recommended in this extremely highly active system due to its pronounced sensitivity to radiation and the increased risk of corrosion due to the hydrochloric acid released.
  • All previously known powerful extraction chromatography processes take place in acidic systems and, in addition to the already mentioned disadvantage of HJ or J₂ release, include a further major disadvantage, namely the fixation of uranium, the main part in the process stream, with reduced retention of the fission products. The disadvantage of this procedure is obvious: Here you have to Nuclear fuel retention incomparably larger column volumes are provided.

Aufarbeitung von durch Spaltprodukte extrem verunreinigtes Urandioxid bzw. Alkalidiuranatrückständen hoher U-235-Anreicherung, wie sie nach dem alkalischen Aufschluß von Material-Test-Reaktor-Brennelementen anfallen: Die Elemente bestehen überwiegend aus mit Aluminium umhüllter Uran/Aluminiumlegierung der ungefähren Zusammensetzung UAl₃; wegen des schwankenden Al-Gehaltes in der Verbindung wird meistens die Bezeichnung UAlx verwendet. Dieser Brennelementtyp wird häufig als Ausgangstarget zur Produktion von Spaltproduktnukliden für die Nuklearmedizin und -technik eingesetzt; dazu werden meist kleinere Elemente bei thermischen Neutronenflüssen von ca.

Figure imgb0001

5 bis 10 Tage bestrahlt. Um die Zerfallsverluste des gewünschten Nuklids zu minimieren, werden die Targets nach einer Mindestabkühlzeit von ca. 12 Stunden zu der Aufarbeitungsanlage transportiert. Als erster chemischer Schritt dient in der Regel ein alkalischer Aufschluß des Targets mit 3 - 6 molarer Natronlauge bzw. Kalilauge; dabei gehen der Hauptbestandteil der Platte, das Aluminium, und die in diesem Medium löslichen Spaltprodukte, wie die Alkali- und Erdalkalikationen sowie Antimon, Jod, Tellur, Zinn und Molybdän in Lösung, während die flüchtigen Spaltprodukte, vor allem das Xenon, zusammen mit dem aus der Al-Auflösung gebildeten Wasserstoff, den Auflöser am oberen Ende des Rückflußkühlers verlassen. Der Wasserstoff kann über CuO zu Wasser oxidiert werden, während das Xenon vorzugsweise bei Normaltemperatur auf Aktivkohleverzögerungsstrecken zurückgehalten wird. Als unlöslicher Rückstand verbleiben das nicht abgebrannte Uran, in der Regel ca. 99 % der anfänglich bestrahlten Menge, als UO₂ bzw. Alkalidiuranat gemeinsam mit den unlöslichen Spaltproduktspezies, vor allem Ruthenium, Zirkonium, Niob und die Lanthanoide in Form ihrer Oxide.Processing of uranium dioxide or alkali metal diurate residues with a high concentration of U-235, which is extremely contaminated by fission products, as occurs after the alkaline digestion of material test reactor fuel elements: The elements mainly consist of aluminum-coated uranium / aluminum alloy with the approximate composition UAl₃; Because of the fluctuating Al content in the compound, the designation UAl x is mostly used. This type of fuel element is often used as a starting target for the production of fission product nuclides for nuclear medicine and technology; mostly smaller elements with thermal neutron fluxes of approx.
Figure imgb0001
Irradiated for 5 to 10 days. In order to minimize the decay losses of the desired nuclide, the targets are transported to the processing plant after a minimum cooling time of approx. 12 hours. The first chemical step is usually an alkaline digestion of the target with 3 to 6 molar sodium hydroxide solution or potassium hydroxide solution; The main constituent of the plate, aluminum, and the fission products soluble in this medium, such as the alkali and alkaline earth cations, as well as antimony, iodine, tellurium, tin and molybdenum, go into solution, while the volatile fission products, especially xenon, go together with the Hydrogen formed from the Al solution, leave the dissolver at the upper end of the reflux condenser. The hydrogen can be oxidized to water via CuO, while the xenon is preferably retained at normal temperature on activated carbon delay lines. Remain as an insoluble residue the unburned uranium, usually about 99% of the initially irradiated amount, as UO₂ or alkalidiuranate together with the insoluble fission product species, especially ruthenium, zirconium, niobium and the lanthanoids in the form of their oxides.

Dieser Rückstand wird in an sich bekannter Weise unter Einwirkung von Luft oder eines Oxidationsmittels, wie z.B. H₂O₂ oder Hypochlorit, mit einer wäßrigen, Karbonat- und Hydrogenkarbonationen enthaltenden Lösung von pH 5 bis pH 11 behandelt. Die Konzentration der Karbonationen kann in dieser Lösung maximal 2,5 M/l, die der Hydrogenkarbonationen maximal ca. 1,0 M/l betragen. Während dieser Behandlung gehen die Oxide des Urans und der genannten Spaltprodukt-Spezies als Karbonato-Komplexe in Lösung.This residue is treated in a manner known per se under the action of air or an oxidizing agent, e.g. H₂O₂ or hypochlorite, treated with an aqueous solution containing carbonate and hydrogen carbonate ions from pH 5 to pH 11. The concentration of the carbonate ions in this solution can be a maximum of 2.5 M / l, that of the hydrogen carbonate ions a maximum of approximately 1.0 M / l. During this treatment, the oxides of uranium and the fission product species mentioned go into solution as carbonato complexes.

Aus wirtschaftlichen und sicherheitstechnischen Aspekten muß dieser kurz gekühlte, extrem kontaminierte Kernbrennstoff rezykliert, erneut targetiert und dann ausgelagert werden. Der übliche Weg mit der salpetersauren Auflösung scheidet jedoch für eine im technischen Maßstab durchführbare Aufarbeitung kurz abgekühlter Brennelemente, wie bereits ausgeführt, aus, wegen der auch nach dem Aufschluß erhöhten Jod-131-Kontamination sowie der bekannten Brandgefahr der Aktivkohle in Gegenwart von Stickoxiden.For economic and safety reasons, this briefly cooled, extremely contaminated nuclear fuel has to be recycled, re-targeted and then outsourced. However, the usual route with nitric acid dissolution is ruled out for processing of briefly cooled fuel elements that can be carried out on an industrial scale, as already stated, because of the increased iodine-131 contamination after digestion and the known risk of fire of the activated carbon in the presence of nitrogen oxides.

Der Erfindung liegt die Aufgabe zugrunde, ein Verfahren zu schaffen, mit welchem in einer wäßrigen, basischen, karbonathaltigen Lösung vorliegende Uran-Werte einerseits von Spaltprodukten aus der Gruppe Ruthenium, Zirkonium, Niob und Lanthanoiden andererseits, voneinander mit verhältnismäßig hohem Dekontaminationsgrad getrennt werden können. Insbesondere sollen mit dem erfindungsgemäßen Verfahren Uran oder die Spaltprodukte Ruthenium, Zirkonium, Niob und Lanthanoiden nach dem alkalischen Aufschluß eines Brennelementes eines Material-Test-Reaktors (MTR) weitgehend dekontaminiert erhalten werden können. Das Verfahren soll betriebssicher und abfallarm durchführbar sein und auf nur wenige Tage abgekühlte urandioxid- und alkalidiuranathaltige Rückstände anwendbar sein.The invention is based on the object of providing a method with which uranium values present in an aqueous, basic, carbonate-containing solution, on the one hand, of fission products from the ruthenium group, Zirconium, niobium and lanthanoids, on the other hand, can be separated from one another with a relatively high degree of decontamination. In particular, the process of the invention is said to be able to obtain largely decontaminated uranium or the fission products ruthenium, zirconium, niobium and lanthanoids after the alkaline digestion of a fuel element of a material test reactor (MTR). The process is said to be operationally reliable and low-waste and to be applicable to residues containing uranium dioxide and alkalidiuranate that have only cooled down for a few days.

Die Aufgabe wird in einfacher Weise erfindungsgemäß dadurch gelöst, daß

  • a) die wäßrige Lösung auf ein Verhältnis der Uranylionen-Konzentration zu Karbonationen- bzw. CO₃⁻⁻/HCO₃⁻-Konzentration von 1(UO₂⁺⁺) zu 4,5(CO₃⁻⁻ bzw. CO₃⁻⁻/HCO₃⁻) oder darunter bei einer maximalen U-Konzentration von nicht mehr als 60 g/l eingestellt wird,
  • b) die eingestellte Lösung zur Adsorption der Spaltproduktionen bzw. der Spaltprodukte enthaltenden Ionen über einen basischen Anionenaustauscher aus einer mit zu einem überwiegenden Teil tertiären und zu einem geringen Teil quarternären Ammoniumgruppen versehenen Polyalken-Matrix geleitet wird und der nicht adsorbierte Uranyl-karbonato-Komplex durch Abtrennen der uranhaltigen, verbleibenden Lösung vom Ionenaustauscher weitgehend spaltproduktfrei wiedergewonnen bzw. dekontaminiert wird und
  • c) der mit Spaltprodukten beladene Ionenaustauscher zur Spaltproduktgewinnung oder zur Abfall-Verfestigung geführt wird.
The object is achieved in a simple manner according to the invention in that
  • a) the aqueous solution to a ratio of the uranyl ion concentration to carbonate ions or CO₃⁻⁻ / HCO₃⁻ concentration of 1 (UO₂⁺⁺) to 4.5 (CO₃⁻⁻ or CO₃⁻⁻ / HCO₃⁻) or below that is set at a maximum U concentration of not more than 60 g / l,
  • b) the set solution for the adsorption of the cleavage products or the ions containing the cleavage products is passed through a basic anion exchanger from a polyalkene matrix provided with a predominantly tertiary and to a small extent quaternary ammonium groups and the non-adsorbed uranyl carbonate complex Removing the uranium-containing, remaining solution from the ion exchanger is largely recovered or decontaminated without fission product and
  • c) the ion exchanger loaded with fission products is led to fission product recovery or waste solidification.

In einer besonderen Ausbildung des erfindungsgemäßen Verfahrens wird die wäßrige Lösung auf ein Verhältnis der Uranylionen-Konzentration zu Karbonationen/Hydrogenkarbonationen-Konzentration von 1 : 5 bis 1 : 8 eingestellt. Vorteilhafterweise wird die wäßrige Lösung bei einer Uran-Konzentration von 60 g/l auf ein Verhältnis UO₂⁺⁺ -Konzentration zu CO₃⁻⁻ /HCO₃⁻ -Konzentration von 1 : 5 eingestellt.In a special embodiment of the method according to the invention, the aqueous solution is adjusted to a ratio of the uranyl ion concentration to carbonate ions / hydrogen carbonate ion concentration of 1: 5 to 1: 8. Advantageously, the aqueous solution is adjusted at a uranium concentration of 60 g / l to a ratio of UO₂⁺⁺ concentration to CO₃⁻⁻ / HCO₃⁻ concentration of 1: 5.

Als basischer Anionenaustauscher wird ein solcher aus Polyalken-epoxypolyamin mit tertiären und quarternären Ammoniumgruppen des chemischen Aufbaus R-N⁺H(CH₃)₂Cl⁻ und R-N⁺(CH₃)₂(C₂H₄OH)Cl⁻, wobei R das Molekül ohne Aminogruppen bedeutet, verwendet.As a basic anion exchanger, such a polyalkene-epoxypolyamine with tertiary and quaternary ammonium groups of chemical structure R-N⁺H (CH₃) ₂Cl⁻ and R-N⁺ (CH₃) ₂ (C₂H₄OH) Cl⁻, where R is the molecule without amino groups, is used.

Vorteilhafterweise weist die wäßrige Lösung eine Hydrogenkarbonationen-Konzentration zwischen 0 und 1 Mol/l auf. Die CO₃⁻⁻-Konzentration in der wäßrigen Lösung beträgt maximal 2,5 M/l und der pH-Wert der wäßrigen Lösung liegt im Bereich von pH 7 bis pH 11.The aqueous solution advantageously has a hydrogen carbonate ion concentration between 0 and 1 mol / l. The CO₃⁻⁻ concentration in the aqueous solution is a maximum of 2.5 M / l and the pH of the aqueous solution is in the range from pH 7 to pH 11.

Das erfindungsgemäße Verfahren ist auch in Abwesenheit von HCO₃⁻-Ionen durchführbar, doch sind die Verfahrensbedingungen leichter einstellbar, wenn in der wäßrigen Lösung HCO₃⁻-Ionen vorhanden sind.The process according to the invention can also be carried out in the absence of HCO₃⁻ ions, but the process conditions can be set more easily if HCO₃⁻ ions are present in the aqueous solution.

Der Einsatzbereich des Verfahrens umspannt einen großen Konzentrationsschwankungsbereich des zu dekontaminierenden Urans. Ist die Urankonzentration in der Lösung gegenüber der Karbonatkonzentration sehr klein, so daß beispielsweise eine freie CO₃⁻⁻/HCO₃⁻-Konzentration höher als 0,6 Mol/l vorliegt, so kann zur Optimierung der Spaltproduktrückhaltungen der zu große Karbonatüberschuß entweder durch Zudosierung einer Mineralsäure, vorzugsweise HNO₃, zerstört oder durch Zugabe von z.B. Ca(OH)₂ eine bestimmte Menge an Karbonationen weggefangen werden.The application of the method spans a large concentration fluctuation range of the uranium to be decontaminated. Is the uranium concentration in the solution compared to the carbonate concentration very small, so that, for example, a free CO₃⁻⁻ / HCO₃⁻ concentration is greater than 0.6 mol / l, the excess carbonate excess can either be optimized by metering in a mineral acid, preferably HNO₃, or optimized to optimize the fission product retention a certain amount of carbonate ions can be trapped by adding, for example, Ca (OH) ₂.

Im umgekehrten Fall jedoch, d.h. liegen höhere Urankonzentrationen vor, so muß durch Zugabe von ausreichenden Mengen an CO₃⁻⁻ /HCO₃⁻-Ionen der Uranverteilungskoeffizient so weit minimiert werden, daß die Spaltprodukt-Spezies nicht durch das Uran vom Ionenaustauscher verdrängt werden. Die gewünschten Trennungen lassen sich noch bei Urankonzentrationen von ca. 60 g U/l durchführen. Die Begrenzung des Verfahrens zu höheren U-Konzentrationen hin ist durch die Uranlöslichkeit in Karbonat-Hydrogenkarbonat-Lösungen begründet.In the opposite case, however, i.e. if higher uranium concentrations are present, the uranium distribution coefficient must be minimized by adding sufficient amounts of CO₃⁻⁻ / HCO₃⁻ ions so that the fission product species are not displaced by the uranium from the ion exchanger. The desired separations can still be carried out at uranium concentrations of approx. 60 g U / l. The limitation of the process to higher U concentrations is due to the uranium solubility in carbonate-hydrogen carbonate solutions.

Zwar wurde ein Verfahren zur Abtrennung von Aktinoidenionen aus wäßrigen, basischen, karbonathaltigen Lösungen aus der deutschen Offenlegungsschrift 31 44 974 bekannt, bei welchem die Aktinoidenionen als Karbonato-Komplexe an basischen Ionenaustauschern adsorbiert und nach Abtrennen des beladenen Ionenaustauschers von der Ausgangslösung mit Hilfe einer wäßrigen Lösung wieder vom Ionenaustauscher desorbiert und weiterverarbeitet werden, und bei welchem zur Adsorption der Aktinoidenionen ebenfalls ein basischer Anionenaustauscher aus einer mit zu einem überwiegenden Teil tertiären und zu einem geringen Teil quarternären Ammoniumgruppen versehenen Polyalken-Matrix verwendet wird, doch ist dieses Verfahren nur sinnvoll anwendbar auf wäßrige, karbonathaltige Abfall-Lösungen oder Waschlösungen etc. Für entsprechende Lösungen mit einem relativ hohen Gehalt an Uranyl-Ionen würde der apparative Aufwand zu groß werden und die genaue Einhaltung der Karbonationen-Konzentration im Bereich der Verhältnisse UO₂⁺⁺-Konzentration zu CO₃⁻⁻-Konzentration zwischen 1 : 3 und 1 : 4 nicht in jedem Falle problemlos sein. Außerdem wäre das Verfahren nach der DE-OS 31 44 974 für größere Uran-Konzentrationen in der Lösung zu umständlich, da die Uranylionen, im Gegensatz zum erfindungsgemäßen Verfahren, vom Anionenaustauscher adsorbiert werden, wobei die Spaltproduktionen mit der verbleibenden Lösung durch den Anionenaustauscher hindurch laufen, und das Uran vom Ionenaustauscher wieder eluiert werden muß. Dem gegenüber werden die Uranylionen im erfindungsgemäßen Verfahren an der gleichen Anionenaustauscherart nicht festgehalten, sondern nur die noch vorhandenen Spaltprodukt-Spezies.A method for the separation of actinide ions from aqueous, basic, carbonate-containing solutions from German Offenlegungsschrift 31 44 974 was known, in which the actinide ions as carbonato complexes are adsorbed on basic ion exchangers and after separation of the loaded ion exchanger from the starting solution using an aqueous solution are desorbed and further processed by the ion exchanger, and in which a basic anion exchanger from a is used with a predominantly tertiary and to a small extent quaternary ammonium group-provided polyalkene matrix, but this method is only meaningfully applicable to aqueous, carbonate-containing waste solutions or washing solutions etc. For corresponding solutions with a relatively high uranyl ion content the expenditure on equipment would be too great and the exact observance of the carbonate ion concentration in the range of the ratios UO₂⁺⁺ concentration to CO₃⁻⁻ concentration between 1: 3 and 1: 4 would not be problem-free in every case. In addition, the method according to DE-OS 31 44 974 would be too cumbersome for larger uranium concentrations in the solution since, in contrast to the method according to the invention, the uranyl ions are adsorbed by the anion exchanger, the cleavage products with the remaining solution passing through the anion exchanger , and the uranium must be eluted again from the ion exchanger. In contrast, the uranyl ions in the process according to the invention are not held to the same type of anion exchanger, but only the fission product species still present.

Die wesentlichen Vorteile des erfindungsgemäßen Verfahrens liegen darin, daß die Dekontamination des Urans von den noch vorhandenen Spaltprodukten mit einer verhältnismäßig kleinen Menge des Anionenaustauschers, z.B. in einer verhältnismäßig kleinen Ionenaustauschersäule, durchgeführt werden kann, daß der mit den Spaltprodukten beladene Ionenaustauscher in dem Falle, daß nur die Uranwerte wiedergewonnen werden sollen, (mit oder ohne Säule) ohne Zwischenbehandlung direkt zur Abfallbehandlung und -beseitigung gegeben werden kann und in dem Falle, daß die Spaltprodukt-Nuklide gewonnen werden sollen, zur Weiterverarbeitung der Spaltprodukt-Nuklide und Trennung voneinander geführt werden kann. Die Spaltprodukte können mit einer Alkali- oder Ammonium-Karbonatlösung höherer Molarität (ca. 1 bis 2 M/l) oder mit Salpetersäure aus der Ionenaustauschersäule eluiert werden. Durch ein- oder mehrmaliges Wiederholen des erfindungsgemäßen Verfahrens an weiteren kleinen Anionenaustauscherchargen erhält man einen hohen Reinheitsgrad des wiederzugewinnenden Urans.The main advantages of the method according to the invention are that the decontamination of the uranium from the fission products still present can be carried out with a relatively small amount of the anion exchanger, for example in a relatively small ion exchange column, that the ion exchanger loaded with the fission products can be used in the event that only the uranium values are to be recovered (with or without a column) without intermediate treatment directly can be given for waste treatment and disposal and, in the event that the fission product nuclides are to be obtained, can be carried out for further processing of the fission product nuclides and separation from one another. The cleavage products can be eluted from the ion exchange column with an alkali or ammonium carbonate solution of higher molarity (approx. 1 to 2 M / l) or with nitric acid. By repeating the process according to the invention one or more times on further small batches of anion exchangers, a high degree of purity of the uranium to be recovered is obtained.

Aufgrund der raschen Durchführbarkeit des erfindungsgemäßen Verfahrens fällt für die Wiederaufarbeitung und Rückführung des Urans in den Kernbrennstoffkreislauf eine nachteilige Bildung von Degradationsprodukten (wie z.B. bei den Extraktionsverfahren eine solche des Extraktionsmittels oder des Verdünnungsmittels) weg. Das erfindungsgemäße Verfahren zeichnet sich durch eine sehr sichere Prozeßführung aus. Beispielsweise muß der organische Anionenaustauscher in keiner Phase des Verfahrens mit korrosiven oder stark oxidierenden Medien in Kontakt gebracht werden.Due to the rapid feasibility of the process according to the invention, there is no disadvantageous formation of degradation products for the reprocessing and recycling of the uranium in the nuclear fuel cycle (such as, for example, that of the extraction agent or of the diluent in the extraction processes). The method according to the invention is characterized by very reliable process control. For example, the organic anion exchanger does not have to be brought into contact with corrosive or strongly oxidizing media in any phase of the process.

Das erfindungsgemäße Verfahren arbeitet mit basischen Medien, die die höchstmögliche Sicherheit gegen Freisetzung flüchtiger Jodkomponenten bieten. Die im erfindungsgemäßen Verfahren verwendete Lösung, welche bis zu maximal 2,5 Mol/l Na₂CO₃ und bei geringerer CO₃⁻⁻-Konzentration bis zu ca. 1 Mol/l NaHCO₃ enthalten kann, ist chemisch denkbar einfach zu beherrschen und strahlenchemisch resistent. Korrosionsprobleme sind nicht vorhanden. Außerdem ist bei dem erfindungsgemäßen Verfahren der Aufwand an Chemikalien, Apparaturen und Arbeitszeit sehr niedrig.The method according to the invention works with basic media which offer the highest possible security against the release of volatile iodine components. The solution used in the process according to the invention, which can contain up to a maximum of 2.5 mol / l Na₂CO₃ and, at a lower CO₃⁻⁻ concentration, up to approx. 1 mol / l NaHCO₃, is chemically very easy to control and radiation-chemical resistant. There are no corrosion problems. In addition, the outlay in chemicals, apparatus and working time is very low in the process according to the invention.

Im folgenden wird die Erfindung anhand zweier beispielhafter Versuche näher erläutert.The invention is explained in more detail below on the basis of two exemplary experiments.

In zwei dynamischen Kolonnendurchlaufexperimenten wurden bei verschiedenen Uran zu Karbonat-/Hydrogenkarbonat-Verhältnissen die Wirksamkeit des erfindungsgemäßen Verfahrens untersucht.In two dynamic column run-through experiments, the effectiveness of the method according to the invention was investigated at different uranium to carbonate / hydrogen carbonate ratios.

Die durchschnittliche Spaltprodukt-Rückhaltung lag bei einem Kolonnendurchlauf unter den angegebenen Beladebedingungen bei > 97 % für Cer, Zirkonium und Niob; bei Ruthenium lag die Rückhaltung bei ca. 80 %.The average fission product retention during a column run under the specified loading conditions was> 97% for cerium, zirconium and niobium; With ruthenium the retention was approx. 80%.

Nachfolgend sind die Ergebnisse im einzelnen angegeben: Aufgabelösung Volumen: 100 ml U-Gehalt: 1,19 g U: CO₃⁻⁻/HCO₃⁻: 1 : 7 bzw. 1 : 6 Na₂CO₃: 3,24 g ≙ 90 % bzw. 2,78 g NaHCO₃: 0,28 g ≙ 10 % bzw. 0,24 g Kolonne Durchmesser: 15 mm Höhe: 130 mm Bettvolumen: 20 ml Aufgabegeschwindigkeit: ∼0,5 ml/cm² · sec. Nachwaschlsg.: 0,2 molare Na₂CO₃-Lösung *) Anzahl der Fraktionen: 4 x 20 *) anstelle einer Na₂CO₃-Lösung kann auch eine entsprechende andere Alkali- oder Ammonium-Karbonatlösung eingesetzt werden. The results are detailed below: Task solution Volume: 100 ml U content: 1.19 g U: CO₃⁻⁻ / HCO₃⁻: 1: 7 or 1: 6 Na₂CO₃: 3.24 g ≙ 90% or 2.78 g NaHCO₃: 0.28 g ≙ 10% or 0.24 g column Diameter: 15 mm Height: 130 mm Bed volume: 20 ml Feed speed: ∼0.5 ml / cm² · sec. Rinsing: 0.2 molar Na₂CO₃ solution *) Number of fractions: 4 x 20 *) Instead of a Na₂CO₃ solution, a corresponding other alkali or ammonium carbonate solution can also be used.

Ionenaustauscher:
mäßig basischer Anionenaustauscher aus Polyalken-epoxypolyamin mit tertiären und quarternären Ammoniumgruppen mit der Handelsbezeichnung Bio-Rex 5 (der Firma Bio-Rad Laboratories, USA).Ion exchanger:
moderately basic anion exchanger made of polyalkene-epoxypolyamine with tertiary and quaternary ammonium groups with the trade name Bio-Rex 5 (from Bio-Rad Laboratories, USA).

Versuch 1 %-Anteile in den durchgelaufenen LösungenTest 1% shares in the solutions run through

Uranuranium Cercerium RutheniumRuthenium Zirkoniumzirconium Niobniobium 100 ml D.L.100 ml D.L. 81,781.7 1,661.66 13,4313.43 1,361.36 1,061.06 20 ml W.l20 ml W.l 14,814.8 0,320.32 4,064.06 0,260.26 0,190.19 20 ml W.220 ml W.2 2,12.1 0,270.27 1,311.31 0,180.18 0,130.13 20 ml W.320 ml W.3 0,80.8 0,140.14 0,550.55 0,090.09 0,060.06 20 ml W.420 ml W.4 0,40.4 0,100.10 0,290.29 0,070.07 0,050.05 Summe:Total: 99,899.8 2,492.49 19,6419.64 1,961.96 1,491.49 D.L. = Durchlauf der Speiselösung (100 ml).
W. = WaschlösungenDL = passage of the food solution (100 ml).
W. = washing solutions

Versuch 2 %-Anteile in den durchgelaufenen LösungenTrial 2% shares in the solutions run through

100 ml D.L.100 ml D.L. 80,580.5 1,841.84 13,5113.51 1,381.38 1,311.31 20 ml W.120 ml W.1 15,015.0 0,350.35 4,204.20 0,270.27 0,220.22 20 ml W.220 ml W.2 2,62.6 0,250.25 1,201.20 0,240.24 0,160.16 20 ml W.320 ml W.3 1,01.0 0,150.15 0,430.43 0,080.08 0,060.06 20 ml W.420 ml W.4 0,60.6 0,100.10 0,310.31 0,060.06 0,050.05 Summe:Total: 99,799.7 2,692.69 19,6519.65 2,032.03 1,801.80 D.L. = Durchlauf der Speiselösung (100 ml).
W. = WaschlösungenDL = passage of the food solution (100 ml).
W. = washing solutions

Claims (7)

  1. Method of separating large quantities of uranium from small quantities of radioactive fission products, which are present in aqueous basic solutions containing carbonate, by utilising an organic, basic anion exchanger, characterised in that
    a) the aqueous solution is adjusted to a ratio of the uranyl ion concentration relative to the carbonate ion concentration, i.e. the CO₃⁻⁻/HCO₃⁻ concentration of 1(UO₂⁺⁺) to 4.5(CO₃⁻⁻), i.e. CO₃⁻⁻/HCO₃⁻) or less with a maximum U concentration of no more than 60 g/l;
    b) the adjusted solution for the adsorption of the fission product ions, i.e. the ions containing fission products, is conducted, via a basic anion exchanger, from a polyalkene matrix, which is provided with a large portion of tertiary ammonium groups and with a small portion of quaternary ammonium groups, and the non-adsorbed uranyl carbonato complex is recovered in a state largely free of fission products, i.e. decontaminated, by separating the remaining, uranium-containing solution from the ion exchanger; and
    c) the ion exchanger, which is charged with fission products, is passed-on for the recovery of fission products or for the solidification of waste.
  2. Method according to claim 1, characterised in that the aqueous solution is adjusted to a ratio of the uranyl ion concentration to the carbonate ion/hydrogen carbonate ion concentration of l : 5 to 1 : 8.
  3. Method according to claim 1, characterised in that the aqueous solution, with a U concentration of 60 g/l, is adjusted to a ratio of the UO₂⁺⁺ concentration to the CO₃⁻⁻/HCO₃⁻ concentration of 1 : 5.
  4. Method according to claim 1, characterised in that the basic anion exchanger used is one formed from polyalkene-epoxypolyamine with tertiary and quaternary ammonium groups of the chemical structure R-N⁺H(CH₃)₂Cl and R-N⁺(CH₃)₂(C₂H₄OH)Cl⁻, R representing the molecule without amino groups.
  5. Method according to claim 1, characterised in that the aqueous solution has a hydrogen carbonate ion concentration of between 0 and 1 mol/l.
  6. Method according to claim 1, characterised in that the maximum CO₃⁻⁻ concentration in the aqueous solution is 2.5 M/l.
  7. Method according to claim 1, characterised in that the pH value of the aqueous solution lies in the range of between pH 7 and pH 11.
EP85105864A 1984-08-04 1985-05-13 Process for separating great amounts of uranium from small amounts of fission products which are present in aqueous basic solutions containing carbonate Expired - Lifetime EP0170796B1 (en)

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DE3428877 1984-08-04
DE19843428877 DE3428877A1 (en) 1984-08-04 1984-08-04 METHOD FOR SEPARATING LARGE AMOUNTS OF URANIUM FROM LITTLE AMOUNTS OF RADIOACTIVE FUSE PRODUCTS EXISTING IN AQUEOUS BASIC SOLUTIONS CONTAINING CARBONATE

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DE3428877A1 (en) * 1984-08-04 1986-02-13 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe METHOD FOR SEPARATING LARGE AMOUNTS OF URANIUM FROM LITTLE AMOUNTS OF RADIOACTIVE FUSE PRODUCTS EXISTING IN AQUEOUS BASIC SOLUTIONS CONTAINING CARBONATE
DE3428878A1 (en) * 1984-08-04 1986-02-13 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe METHOD FOR RECOVERY OF URAN VALUES IN AN EXTRACTIVE REPROCESSING PROCESS FOR IRRADIATED FUELS
JPS63239128A (en) * 1986-12-26 1988-10-05 Unitika Ltd Production of uranium oxide
DE3708751C2 (en) * 1987-03-18 1994-12-15 Kernforschungsz Karlsruhe Process for the wet dissolution of uranium-plutonium mixed oxide nuclear fuels
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US8088749B2 (en) * 2007-12-12 2012-01-03 The Regents Of The University Of Michigan Compositions and methods for treating cancer
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DE3428877A1 (en) * 1984-08-04 1986-02-13 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe METHOD FOR SEPARATING LARGE AMOUNTS OF URANIUM FROM LITTLE AMOUNTS OF RADIOACTIVE FUSE PRODUCTS EXISTING IN AQUEOUS BASIC SOLUTIONS CONTAINING CARBONATE

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US9238212B2 (en) 2011-01-12 2016-01-19 Mallinckrodt Llc Process and apparatus for treating a gas stream

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EP0170796A2 (en) 1986-02-12
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DE3428877A1 (en) 1986-02-13
US4696768A (en) 1987-09-29
EP0170796A3 (en) 1989-02-22

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