EP0347625B1 - Method for separating technetium, ruthenium and palladium from solutions of nuclear fuels - Google Patents

Description

The invention relates to a process for separating the valuable substances technetium, ruthenium and palladium from material flows which arise during the reprocessing of irradiated nuclear fuel, the valuable substances being isolated by precipitation and ion exchange from a nitric acid containing the valuable substances and other fission / activation products.

When nuclear fuels are irradiated in nuclear reactors, fission and activation products are created. Activation products arise from the atoms of the nuclear fuel (uranium, plutonium) by neutron capture; they belong to the group of actinides.

Fission products are isotopes of chemical elements that arise when atoms of the nuclear fuel (uranium, plutonium) are split into two or three fragments. These isotopes can themselves be radioactive, but can also be inactive. Technetium, ruthenium and palladium belong to the group of fission products. Technetium mainly forms the isotope with the mass 99, which is weakly radioactive and therefore does not occur in nature and can therefore only be generated artificially by nuclear reactions.

The fission products ruthenium and palladium arise in nuclear reactors in large quantities, whereby both radioactive and non-radioactive isotopes are formed.

Fission ruthenium contains about 3% of the isotope Ru-106, which has a half-life of about 1 year and decays into inactive palladium (Pd-106). Fission palladium is very weakly radioactive due to its Pd-107 content. However, this radioactivity does not significantly limit the usability for technical purposes.

During reprocessing, the irradiated nuclear fuel is dissolved in boiling semi-concentrated nitric acid. The majority of the nuclear fuels and the fission and activation products go into solution.

A small remainder of the irradiated nuclear fuel remains in the undissolved form in the dissolver. This residue is known as feed sludge. It contains significant amounts of molybdenum, zirconium, technetium and precious metals.

Unused uranium and the plutonium produced are separated from the solution of the irradiated nuclear fuel by extraction. The remaining fission products are finally melted into glass.

Because of the commercial value of the fission products technetium, ruthenium and palladium, a number of proposals for the separation of these valuable materials from solutions that are produced during the reprocessing of irradiated nuclear fuels have been published (MW Davis, The Extraction of Cs, Sr and the Platinum Group Metals from Acidic High Activity Nuclear Waste, DOE / SR / 10714-T3).

These proposals relate to the use of several columns connected in series, on which the disruptive and later the desired nuclides are fixed to the columns with a retention of between 75% and 95%. This requires a high level of equipment due to the necessary use of several columns.

Furthermore, there is a considerable amount of secondary waste, since used columns are heavily contaminated and have to be disposed of separately.

A process is known from US Pat. No. 3,848,048 in which palladium, technetium, rhodium and ruthenium are separated from acidic fuel solutions by passing these solutions in succession over three carbon beds which are impregnated with different chelating agents. The elements palladium, technetium and ruthenium / rhodium are retained in chelated form on these beds.

The loaded beds are preferably incinerated and the valuable materials are isolated from the ashes.

This process has several disadvantages. The chelating agents can contaminate the solution which has been freed from the valuable substances and in this way can severely disrupt the removal of further valuable substances or the conditioning of the remaining ingredients. Because of their flammability, the carbon beds represent a significant hazard potential. They are burned after loading, whereby radioactive components are released as gases or aerosols and must be retained by an effective exhaust gas cleaning system with the help of scrubbers and filters. Washing liquids and filters must be disposed of as secondary waste.

The object of the invention is to selectively separate the valuable materials technetium, ruthenium and palladium from acidic solutions of irradiated nuclear fuel and with high efficiency. The separation of these valuable substances should not cause any additional problems in the further treatment of the radioactive substances; in particular, only substances should be used which do not contaminate the solution freed from the valuable substances or which can be removed by simple measures such. B. remove by heating or extracting the solution. Easily flammable substances should not be used. The process should be simple to carry out, with as little secondary waste as possible.

The object is achieved by the features listed in the characterizing part of the main claim.

The subclaims indicate advantageous developments of the method.

In principle, the process according to the invention can be carried out with all nitric acid solutions which are produced during the reprocessing process and which contain technetium, ruthenium and palladium.

Feed clarification sludge is a preferred source of these valuable substances because it contains the valuable substances in a concentrated form.

The feed sludge can be brought into solution by treating it in a manner known per se with reducing gases such as C0 or H₂ and annealing it with carbonates. The residue on ignition is taken up with 3-7 molar nitric acid and the solution is adjusted to 1 mol HN0₃ / l. Rhodium remains behind as Rh₂0₃.

The solution freed from solid Rh₂0₃ forms the stock solution of the process according to the invention which, in addition to the valuable materials technetium, ruthenium and palladium, depending on the origin and pretreatment of the irradiated nuclear fuel, also varying amounts of the elements Pu, U, Am, Mo, Zr, Ce and other fission and may contain activation products.

Diethyl thiourea (DETH) is added to the stock solution in solid form or as an aqueous solution. The amount of DETH depends on the amount of palladium and ruthenium in the stock solution. 4 moles of DETH are added per mole of palladium and an additional 6 moles of DETH are added per mole of ruthenium. Palladium selectively forms an insoluble precipitate with the DETH reagent, which contains more than 99% of the palladium present are. Spectrophotometric studies allow the assumption that it is polymeric Pd-DETH complexes.

The precipitate containing palladium is separated off in a customary manner and roasted at about 500.degree. This forms Pd0, which can be reduced to metal by annealing at 900 ° C.

The Pd precipitate filtrate is heated to a temperature of about 70 ° C for about 30 minutes to accelerate complexation of the DETH with ruthenium.

The cooled solution is passed through a strongly acidic cation exchanger. The Ru (NO) -DETH compounds, which are only present in cationic form, are quantitatively retained on the adsorber together with the Tc0²⁺ ions present in the medium in the tetravalent state, while the accompanying impurities are only partially adsorbed and by washing the column with about 2-molar HNO₃ can be desorbed again.

The strongly acidic cation exchanger AG 50 W-X2 (Fa. BiO-RAD Laboratories GmbH, Munich) proved to be particularly efficient in terms of capacity and sorption kinetics; it consists of a macroporous co-polymer of polystyrene divinylbenzene with 2% crosslinking.

After the cation exchanger has been freed from the adsorbed undesirable substances by washing, the technetium is selectively and quantitatively eluted. The elution is preferably carried out with a solution of about 0.1 - 1 mol H₂0₂ / l and 0.1 - 1 mol HNO₃ / l. Technetium is present as pertechnetate after elution.

Ruthenium is then eluted; preferably 6-8 molar HN0₃ is used as the eluent.

An essential advantage of the process according to the invention is that, in addition to the nitric acid which is in any case necessary for the dissolution of irradiated nuclear fuel, only chemicals are used which can be removed from the solution freed from the valuable substances by simple boiling or extraction. Therefore, the further treatment of these solutions is not made difficult. The process can be integrated into the process diagram of the reprocessing without the reprocessing process having to be changed.

The chemicals used place no higher demands on the corrosion resistance of a plant for carrying out the method according to the invention than the reprocessing process. The method can be carried out simply and inexpensively. Because of the use of a single separation column which, in contrast to the process according to US Pat. No. 3,848,048, is reused, no significant amounts of secondary waste are produced.

The water solubility of the chemicals used makes the addition of flammable organic solvents unnecessary.

The method according to the invention is characterized by a high effectiveness; The high loading levels that can be achieved allow the construction of compact, easy-to-use systems.

The invention is explained in more detail below with the aid of two implementation examples.

The implementation examples are based on a stock solution which is obtained from feed sludge by carbonate digestion. Here, rhodium remains behind. An inactive simulate was used for the feed sewage sludge, which is based on published data on the composition of the feed sewage sludge (K. Naito et al, Recovery of Noble Metals from Insoluble Residue of Spent Fuel, J. Nucl. Sc. And Tech., 23 (6) , pp. 540-549 (June 1986); H. Kleykamp, composition of residues from the dissolution of irradiated LWR- (U, PU) 0₂ with recycled Pu, atomic economy, July 1982).

The inactive simulate was carried with radioactive isotopes of the corresponding elements. Isotope 239 was used for plutonium; the rare earths are represented by the element cerium.

Table 1 shows the average composition of the feed sludge in% by weight.

Tables 2 and 3 show the molar concentrations of the individual elements in the stock solution used. Table 2 (Experiment 1) element Concentration (mol / l) Ru / external nuclide ratio Decontamination factor DF Ru 5 x 10⁻³ - 3 x 10³ U 5 x 10⁻⁴ 10th 4.6 x 10² Pu 1.25 x 10⁻⁵ 400 > 1 x 10³ At the 3 x 10⁻⁶ 333 > 6 x 10² Mon 1.87 x 10⁻³ 2.67 > 1 x 10⁶ Tc 3.75 x 10⁻⁴ 13.33 1 x 10⁴ Pd 1 x 10⁻³ 5 > 99 Zr 3.75 x 10⁻⁵ 133.33 6.25 x 10² Ce 1.25 x 10⁻⁴ 40 5 x 10⁴ (Experiment 2) element Concentration (mol / l) Ru / external nuclide ratio Decontamination factor DF Ru 1 x 10⁻² - 2.8 x 10³ U 1 x 10⁻³ 10th 5.1 x 10² Pu 2.5 x 10⁻⁵ 400 > 1 x 10³ At the 6 x 10⁻⁶ 333 > 6 x 10² Mon 3.7 x 10⁻³ 2.67 > 1 x 10⁶ Tc 7.5 x 10⁻⁴ 13.33 1.3 x 10⁴ Pd 2 x 10⁻³ 5 > 99 Zr 7.5 x 10⁻⁵ 133.33 5.8 x 10² Ce 2.5 x 10⁻⁴ 40 4.8 x 10⁴

The decontamination factor DF indicates the proportion of the elements removed when carrying out the method according to the invention.

The valuable materials ruthenium, technetium and palladium are separated with decontamination factors from 2800 - 3000 or 10000 to 13000 or> 99.

The separated materials are only very slightly contaminated by the undesired fission products and actinides, as can be seen from the decontamination factors of these elements.

Implementation examples (experiment 1, experiment 2)

A 1 molar nitric acid solution was used as the stock solution. The element concentrations for test 1 are shown in table 2, for test 2 in table 3.

With constant stirring, the organic complexing agent N, N′-diethylthiourea was added to the stock solutions at room temperature. 4 moles of DETH were added per mole of palladium present in the stock solution and 6 moles of DETH were added per mole of ruthenium present in the stock solution. After about 15 minutes, more than 99% of the palladium has precipitated as a Pd-DETH complex. The residue was separated and the solution freed from the residue was heated in a thermostated water bath at 70 ° C. for 30 minutes. Under these conditions, the ruthenium nitrosyl nitrate complexes, which are present in different Ru values, are quantitatively converted into the double and triple positively charged ruthenium nitrosyl diethyl urea complexes, while technetium, which originally existed as pertechnetrate, is reduced to Tc0²⁺.

The solution pretreated in this way was applied to a column loaded with AG50W-X2, with ruthenium and technetium being completely retained.

The column was washed with 4-5 column volumes of 2-molar HNO₃, the interfering substances (Zr, Ce, U, Pu, Am, Mo) partially retained on the column being removed.

The column was then oxidized with 4 column volumes of an aqueous solution, each containing HNO₃ and H₂0₂ in a concentration of 0.5 mol / l, the technetium being released as the pertechnetrate.

The column was then eluted to release Ru with 14 column volumes of a 7 molar HNO₃.

The results of experiments 1 and 2 are shown in Tables 2 and 3.

Legend for the figures:

• Fig. 1:
  Flow diagram of the recovery of valuable materials from a stock solution, which is obtained by carbonate digestion and subsequent dissolution in nitric acid.
• Fig. 2:
  Dynamic retention of technetium R. Ordinate: R as a percentage of the amount of technetium added. Abscissa: amount of technetium given in mg / g of cation exchanger. Cation exchanger: AG 50 W-X2 (50 - 100 mesh). Temperature: 24 ± 0.1 ° C.
• Fig. 3:
  Technetium concentration C _Tc in the eluate depending on the eluate volume V. Cation exchanger: AG 50 W-X2. Eluent: 0.5 M H₂0₂ / 1 M HNO₃.
• Fig. 4:
  Dynamic Ru-DETH retention R. Ordinate: R in% of the amount of ruthenium added. Abscissa: amount of ruthenium given m _Ru in mg Ru / g cation exchanger. Cation exchanger AG 50 W-X2 (200 - 400 mesh) 500 mg; Feed speed 0.76 ml / min. Temperature: 22 ° C.
• Fig. 5:
  Dependence of the ruthenium concentration C _Ru in the eluate from a column loaded with 28 mg Ru / g cation exchanger. Ordinate: C _Ru in mol / l. Abscissa: eluate volume; V as a multiple of the column volume. Cation exchanger: AG-50W-X2 (200 - 400 mesh). Column volume 5 ml. Eluent 6 m HNO₃.
• Fig. 6:
  Ruthenium elution yield (A _Ru ) in percent of the fixed ruthenium as a function of the eluate volume (V: multiple of the column volume).
  The following figures relate to the elution yield A of adsorbed undesirable substances during the washing of the cation exchanger. A is given as a percentage of the fixed undesirable substance.
  The multiples of the column volume are plotted on the abscissa. AG 50W-X2 was used as the cation exchanger; the eluent (washing medium) consisted of 1-molar HN0₃ for uranium, for the other undesirable substances of 2-molar HNO₃.
• Fig. 7: Elution yield A _U for uranium
• Fig. 8: Elution yield A _Pu for plutonium
• Fig. 9: Elution yield A _Am for americium
• Fig. 10: Elution yield A _Zr for zirconium
• Fig. 11: Elution yield A _Ce for cerium.

Claims (7)

1. Method of separating the valuable substances technetium, ruthenium and palladium from substance streams, which are produced during the reprocessing of irradiated nuclear fuel, the valuable substances being isolated from a nitrate stock solution, which contains the valuable substances and other fission/activation products, by means of precipitation and ion exchange, characterised by the following features:

   a) adding diethyl thiourea (DETH) to the stock solution, whereby palladium is selectively precipitated, and separating the precipitate;

   b) introducing the solution, which has been freed of palladium, into a bed of a strongly acidic cation exchanger for the separation of technetium and ruthenium;

   c) washing the cation exchanger with dilute nitric acid, preferably substantially 2-molar nitric acid;

   d) selectively and oxidisingly elutriating the technetium with a dilute nitric acid solution which contains an oxidising agent;

   e) selectively elutriating the ruthenium by means of concentrated nitric acid, preferably 6-8-molar nitric acid.
2. Method according to claim 1, characterised in that substantially 4 mol DETH are added per mol palladium present in the stock solution, and additionally substantially 6 mol DETH are added per mol ruthenium present in the stock solution.
3. Method according to claim 1, characterised in that a macroporous copolymer of polystyrene divinyl benzene is used as the strongly acidic cation exchanger, with 2-8 % cross-linking, preferably with 2 % cross-linking.
4. Method according to claim 1, characterised in that, for the selective, oxidising elutriation of the technetium, a nitric acid solution is used, which contains hydrogen peroxide and contains both hydrogen peroxide and nitric acid in the concentration range 0.05 - 3 mol/l, preferably 0.1 - 1 mol/l.
5. Method according to claim 1, characterised in that the stock solution is produced from feed clarification sludge, whereby the main quantity of the feed clarification sludge is brought into solution by a decomposition of carbonate and an addition of nitric acid, and the solution is separated from insoluble rhodium oxide.
6. Method according to claim 1, characterised in that the precipitate, being formed from DETH and palladium, is converted into palladium oxide at substantially 500° C and is produced from the palladium oxide by heating palladium metal at substantially 900° C.
7. Method according to claim 1, characterised in that the solution, freed of palladium, for accelerating the complex formation of ruthenium with DETH is heated for substantially 30 min. to a temperature of substantially 70° C.

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