EP0170796B1 - Verfahren zur Trennung von grossen Mengen Uran von geringen Mengen von radioaktiven Spaltprodukten, die in wässrigen, basischen, karbonathaltigen Lösungen vorliegen - Google Patents

Verfahren zur Trennung von grossen Mengen Uran von geringen Mengen von radioaktiven Spaltprodukten, die in wässrigen, basischen, karbonathaltigen Lösungen vorliegen 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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EP
European Patent Office
Prior art keywords
concentration
uranium
fission products
aqueous solution
solution
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Expired - Lifetime
Application number
EP85105864A
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German (de)
English (en)
French (fr)
Other versions
EP0170796A3 (en
EP0170796A2 (de
Inventor
Abdel Hadi Ali Dr. Sameh
Jürgen Haag
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Forschungszentrum Karlsruhe GmbH
Original Assignee
Kernforschungszentrum Karlsruhe GmbH
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Publication of EP0170796A3 publication Critical patent/EP0170796A3/de
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Classifications

    • 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)
EP85105864A 1984-08-04 1985-05-13 Verfahren zur Trennung von grossen Mengen Uran von geringen Mengen von radioaktiven Spaltprodukten, die in wässrigen, basischen, karbonathaltigen Lösungen vorliegen Expired - Lifetime EP0170796B1 (de)

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
DE3428877 1984-08-04
DE19843428877 DE3428877A1 (de) 1984-08-04 1984-08-04 Verfahren zur trennung von grossen mengen uran von geringen mengen von radioaktiven spaltprodukten, die in waessrigen basischen, karbonathaltigen loesungen vorliegen

Publications (3)

Publication Number Publication Date
EP0170796A2 EP0170796A2 (de) 1986-02-12
EP0170796A3 EP0170796A3 (en) 1989-02-22
EP0170796B1 true EP0170796B1 (de) 1993-04-14

Family

ID=6242417

Family Applications (1)

Application Number Title Priority Date Filing Date
EP85105864A Expired - Lifetime EP0170796B1 (de) 1984-08-04 1985-05-13 Verfahren zur Trennung von grossen Mengen Uran von geringen Mengen von radioaktiven Spaltprodukten, die in wässrigen, basischen, karbonathaltigen Lösungen vorliegen

Country Status (4)

Country Link
US (1) US4696768A (GUID-C5D7CC26-194C-43D0-91A1-9AE8C70A9BFF.html)
EP (1) EP0170796B1 (GUID-C5D7CC26-194C-43D0-91A1-9AE8C70A9BFF.html)
CA (1) CA1239799A (GUID-C5D7CC26-194C-43D0-91A1-9AE8C70A9BFF.html)
DE (1) DE3428877A1 (GUID-C5D7CC26-194C-43D0-91A1-9AE8C70A9BFF.html)

Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US9238212B2 (en) 2011-01-12 2016-01-19 Mallinckrodt Llc Process and apparatus for treating a gas stream

Families Citing this family (9)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
DE3428878A1 (de) * 1984-08-04 1986-02-13 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe Verfahren zur rueckgewinnung von uran-werten in einem extraktiven wiederaufarbeitungsprozess fuer bestrahlte kernbrennstoffe
DE3428877A1 (de) * 1984-08-04 1986-02-13 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe Verfahren zur trennung von grossen mengen uran von geringen mengen von radioaktiven spaltprodukten, die in waessrigen basischen, karbonathaltigen loesungen vorliegen
JPS63239128A (ja) * 1986-12-26 1988-10-05 Unitika Ltd 酸化ウランの製法
DE3708751C2 (de) * 1987-03-18 1994-12-15 Kernforschungsz Karlsruhe Verfahren zur nassen Auflösung von Uran-Plutonium-Mischoxid-Kernbrennstoffen
GB2326268A (en) * 1997-06-12 1998-12-16 British Nuclear Fuels Plc Recovery of uranium carbonato complex by ion flotation
US6329563B1 (en) 1999-07-16 2001-12-11 Westinghouse Savannah River Company Vitrification of ion exchange resins
DE102004022705B4 (de) * 2004-05-05 2012-05-31 Atc-Advanced Technologies Dr. Mann Gmbh Verfahren zur Abtrennung von Uranspecies aus Wasser und Verwendung eines schwachbasischen Anionenaustauschers hierfür
WO2009076629A2 (en) * 2007-12-12 2009-06-18 The Regents Of The University Of Michigan Compositions and methods for treating cancer
KR100961832B1 (ko) * 2008-04-25 2010-06-08 한국원자력연구원 고 알카리 탄산염 용액 계를 사용하는 사용후핵연료의우라늄 분리회수방법과 그 장치

Family Cites Families (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US2811412A (en) * 1952-03-31 1957-10-29 Robert H Poirier Method of recovering uranium compounds
US2864667A (en) * 1953-06-16 1958-12-16 Richard H Bailes Anionic exchange process for the recovery of uranium and vanadium from carbonate solutions
US3155455A (en) * 1960-12-12 1964-11-03 Phillips Petroleum Co Removal of vanadium from aqueous solutions
US3835044A (en) * 1972-10-16 1974-09-10 Atomic Energy Commission Process for separating neptunium from thorium
US3922231A (en) * 1972-11-24 1975-11-25 Ppg Industries Inc Process for the recovery of fission products from waste solutions utilizing controlled cathodic potential electrolysis
US4280985A (en) * 1979-03-16 1981-07-28 Mobil Oil Corporation Process for the elution of ion exchange resins in uranium recovery
DE3144974C2 (de) * 1981-11-12 1986-01-09 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe Verfahren zur Abtrennung von Aktinoidenionen aus wäßrigen, basischen, carbonathaltigen Lösungen
DE3428877A1 (de) * 1984-08-04 1986-02-13 Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe Verfahren zur trennung von grossen mengen uran von geringen mengen von radioaktiven spaltprodukten, die in waessrigen basischen, karbonathaltigen loesungen vorliegen

Cited By (1)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US9238212B2 (en) 2011-01-12 2016-01-19 Mallinckrodt Llc Process and apparatus for treating a gas stream

Also Published As

Publication number Publication date
DE3428877C2 (GUID-C5D7CC26-194C-43D0-91A1-9AE8C70A9BFF.html) 1990-10-25
EP0170796A3 (en) 1989-02-22
DE3428877A1 (de) 1986-02-13
EP0170796A2 (de) 1986-02-12
CA1239799A (en) 1988-08-02
US4696768A (en) 1987-09-29

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