nuclear science and technology

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1 EURPEAN CMMISSIN nuclear science and technology Development and automation of chemical analytical procedures for the determination of non-gamma-emitting radionuclides in radioactive waste (DACAP) Contract N o FIKW-CT-2-22 Final report Work performed as part of the European Atomic Energy Community's R&T specific programme Nuclear Energy , key action Nuclear Fission Safety (Fifth Framework Programme) Area: Safety of the fuel cycle Directorate-General for Research 27 Euratom

2 Project coordinator J. Fachinger, S. Maischak, and K. van Hecke FZJ (Germany) Project partners J. L. Gascon and A. Espartero CIEMAT (Spain) G. Bilancia, M. Ferrando, and S. Zicari ENEA (Italy) M. Buchmeiser University of Innsbruck (Austria) II

3 Table of contents 1 Executive summary bjectives and strategic aspects Scientific and technical performance Summary of the specific project objectives Scientific and technical description of the results WP 1: Disintegration of radioactive waste matrices WP 2: Separation of radionuclides WP 3: New extracting agents WP 4: Automated separation methods WP 5: Automated measurements WP 6: Statistical estimations for total waste packages References...72 III

4 IV

5 1 Executive summary The determination of non-γ-emitting radionuclides in radioactive waste packages is divided into four main steps: sampling, disintegration of the waste matrix, separation of the radionuclides, and the measurement. Samples can only be taken from small local areas due to the large volume of the waste packages. Statistical evaluations of different local samples and empirical parameters are necessary to achieve an assumption for the content of a whole package. A development of a tool set will be started, which provides the needs for a probable estimation of the total radionuclide inventory. These tools should take into account results from non-destructive tests about the density and γ-emitting radionuclide distribution in the waste package. After sampling the waste matrix of the sample must be dissolved. Advanced disintegration methods for difficult matrices like incineration ashes, graphite or different organic material will be developed to minimise the amount of insoluble residues. Furthermore the determination of α-nuclides in fine powdered solids will be tested to determine the residual α-nuclides in the residues of the dissolved waste. The main work of the project will be related to the separation and enrichment of non-γ-emitting radionuclides. Element-group selective pre-separation precipitation or extraction will be tested to overcome analytical routine dependent on the waste matrix. Then the following separation of the single nuclides could be performed with standard analytical routines independently of the original waste composition. All newly developed methods should be examined with respect to their applicability in automated systems. Therefore new separation agents will be prepared and tested. Especially the fixation of extraction agents on solid phases will be investigated to introduce further chromatographic techniques for the radionuclide separation. This is an advantage for the proposed automation of analytical routines which is a further objective of this project. Separations with commercial robotic systems or HPLC are considered for this purpose. The last subject of the work is related to the measurement of the radionuclides itself. Especially the online detection developed further as useful method to increase the sample throughput is a basis for a fully automated separation and measurement system. However, the detection limits are not sufficient for the waste characterisation required. Therefore the detection has to be modified to achieve lower detection limits. 1

6 2 bjectives and strategic aspects A safe nuclear waste management is a major issue for public acceptance of nuclear energy in the Member States of the European Union. This includes a precise declaration of the nuclear waste. The DACAP project contributes to the control of the waste declaration, especially to the detection of radionuclides, which cannot be measured with non-destructive methods. The objective of DACAP is the development of faster and/or automated procedures for the determination of non-γ-emitting radionuclides in conditioned radioactive waste in the member states of the EU. An exact description of the inventory of radionuclides is needed information for the reliable categorisation of radioactive waste with respect to waste management regulations. Some of the radionuclides can be measured by non-destructive methods like γ- measurement. However a lot of radionuclides, especially some long-lived radionuclides, are pure α- or β-emitters, which have to be determined by destructive analysis methods. Beneath the α-emitting actinides and radium, the following β-emitting radionuclides are important for waste characterisation due to their long half-life: 129 I, 99 Tc, 93 Zr, 41 Ca, and 36 Cl. The main objectives of this project are the development of advanced methods for the destructive analysis of radioactive waste and the automation of analysis methods. Related to the self-adsorption of α- and β-irradiation, the radionuclides mentioned above have to be isolated from the waste matrix before they can be measured. Most of the actual analytical methods for such radionuclides are based on classic analytic processes. These processes vary due to the types of waste matrices, are mostly very time consuming, and need a lot of manual laboratory work. The manual handling of open radioactive material is always accompanied with the risk of dose rate and incorporation. The project aims to the development of advanced analytical methods in order to increase the reliability of methods and to minimise the risk for employees by short analytical procedures or by pre-separation of the main activity like 137 Cs or 6 Co. Developments in sample preparation and new selective extracting agents are proposed for these approaches. Furthermore, the reliability of those developments will be checked out. An optimisation is the use of automated, remote methods. Automation would reduce the dose risk for the workers to a minimum. In the Fourth Framework Programme, some analytical methods had been developed, which seem to be a good basis for automation. The development of these methods should be continued with the focus on the final aim of a complete remote procedure. Another objective of the research proposal is the analysis of insoluble residues from complex waste matrices. Beside the development of new decomposition methods, new analytic procedures will be investigated to estimate the remaining activity in insoluble residues. It is foreseeable that this can be achieved for the total α/β activity. Additionally, spectrometric methods in combination with mathematical peak analysis will be investigated to achieve semiquantitative or at least qualitative nuclide-specific information. ne further problem arises due to the inhomogeneity of the waste matrix and the activity distribution. Therefore a lot of samples have to be analysed for statistically relevant results, if the main activity areas cannot be identified by screening methods. Especially the release of 2

7 decommissioning waste as non-nuclear waste must be accompanied by a large number of samples. Automated methods are assumed as a possibility to increase the samples throughput, in order to enable the handling of this special waste stream arising from decommissioning. Finally, the DACAP project aims to the development of criteria for characterisation of inhomogeneous waste packages. 3

8 3 Scientific and technical performance 3.1 Summary of the specific project objectives Seven different work packages started at the beginning of the project and lasted for the whole three years. They are listed below explaining the specific objectives for each work package. WP 1: Disintegration of radioactive waste matrices Development of dissolution methods for complex waste matrices Determination of radionuclides in insoluble residues WP 2: Separation of radionuclides Development of group selective separation methods Development of separation methods suitable for automation Test of new extraction agents Development of faster and more selective separation methods Development of multi-element separation methods Reduction of personal dose risk WP 3: New extracting agents Development of new extracting agents for Cs and Co to minimise the dose risk during sample preparation Development of new extracting agents for actinides and long-lived fission products Development of group selective extracting agents Fixation of extraction agents on solid phases WP 4: Automated separation methods Development of automated standardised separation methods Reduction of the personal dose risk Increase of the sample throughput WP 5: Automated measurements Development of a two dimensional on-line detector unit for chromatographic methods Development of methods for an automated spectra interpretation Increase of the sample throughput WP 6: Statistical estimations for total waste packages Development of a mathematical tool to describe volume elements in a waste package Development of a mathematical model to simulate different radionuclide distributions in a waste package related to the volume elements Development of a statistical tool to estimate the reliability of destructive analysis in relation to the number of samples taken and the different distribution models WP7: Synthesis of the results Co-ordination of the project Exchange of data and information between the partners Exchange of data and information between the co-ordinator and the EC. 4

9 3.2 Scientific and technical description of the results WP 1: Disintegration of radioactive waste matrices Task 1: Dissolution methods for inorganic waste matrices Methods should be developed to dissolute all different kind of radioactive waste matrices. Pressurised microwave decomposition is a possible method for that purpose. Both partner, ENEA and FZJ, used a microwave system MILESTNE MLS-12. The MILESTNE is a microwave based sample preparation system with a 6 position rotor, allowing the rapid and complete wet-chemical decomposition of organic and inorganic sample materials in closed quartz glass or TFM reaction vessel at high pressure and high temperature. The Microwave generates electromagnetic waves at frequencies ranging from 3 MHz to 3 GHz. A built-in microwave generator (Magnetron) provides microwaves at frequency of 2.45 GHz. With this instruments dissolution test of several matrices were performed. The dissolution of different waste matrices like filter dusts, concrete, slurries and ion exchange resins with concentrated mineral acids are shown in Table 1. Table 2 shows the used microwave programmes. Nearly all kind of waste matrix can be dissolved in the microwave. Both partner found similar decomposition methods for ion exchange and evaporator concentrate samples received from a Spanish nuclear power plant. A digestion bomb (an Autoclave 3 supplied by Perkin Elmer) was also used in a glove box to dissolve the above mentioned samples received. The samples were successfully mineralised with a mixture of 8 ml HN 3 (6 %),,2 ml HCl (3 %) and,5 ml H 2 2 (3 %). As conclusion can be said that microwave decomposition is a suitable method for the dissolution of all different waste matrices. That was proved by both partners responsible for this task by the development of similar digestion methods. Table 1: Microwave decompositions of different waste matrices Concrete Filter dust Slurries Ion exchange resin Evaporator concentrate Mass.1 g 1 g 1.5 g.2 g 1 ml Reagents Microwave programme 1.) 3.5 ml HN 3 2 ml HBF 4 1 ml H ) + 1 ml H 3 B 3 1.) 1 ml HN 3 3 ml HCl 4 ml HBF 4 2 ml H ) + 1 ml H 3 B 3 1.) HN 3 HBF 4 H ) + HCl 3.) + HCl 4 6 ml HN 3 2 ml H ml HN 3 2 ml H ml H Complete dissolution Complete dissolution No complete dissolution 15 % residue containing < 5 % activity Complete dissolution Complete dissolution 5

10 Table 2: Microwave programmes Programme 3 Programme 4 time [min] power [watt] pressure [bar] temperature [ C] vent time [min] power [watt] pressure [bar] vent temperature [ C] Task 2: Disintegration of organic waste matrices The objective of this task was to develop a fast method to determine 14 C and other radionuclides in resins, evaporator concentrate and graphite. FZJ and CIEMAT examined the combustion of such samples successfully. FZJ carried out the combustion of 1 g of graphite at 8 C for 2.5 h in an oxygen flow of 6 L/h. Gas traps filled with.1 M HN 3 or 4 M NAH captured 3 H and 14 C. Figure 1 shows the apparatus of these experiments. After the combustion the Tritium can be measured directly from an aliquot of the first gas trap with an insta gel cocktail by LSC. The 14 C was also measured directly by LSC with HINIC Fluor as scintillation cocktail. Impurities of the graphite sample remain in the glass vial and are dissolved in 1 ml HCl conc. The determination of different non volatile nuclides (e.g. 9 Sr or 63 Ni) is possible out of this solution. Pressure-reducing valve (2 bar) Flow controller Three-way tap 2 Gas traps Ground unit Thermoelement Temperature controller Rotameter Furnace Lift platform xygen cylinder Figure 1: Equipment for the combustion of graphite (FZJ) The combustion of the ion exchange resin and evaporator concentrate was performed with some minor changes:.5 g wet ion exchange resin (or 2 g evaporator concentrate) was mixed with 2.5 g Cu for combustion. The sample was placed in a quartz glass tube containing a Cu bed of 6 g. The quartz glass tube was connected with the oxygen supply, which was adjusted to a flow of 6 L/h, and to the gas traps mentioned above. 6

11 CIEMAT has performed a dry method based on the combustion of the sample for the determination of 14 C in the three kinds of nuclear waste (graphite, ion exchange resins and evaporator concentrates). The sample is combusted by a stream of oxygen gas at 9 ºC and the combustion products pass through a series of catalysts at 68 ºC. The bulk of the material is converted to carbon dioxide and water. Metal salts and materials with very high melting points remain in the boat. The equipment used (Figure 2) for the combustion of the sample has two scintillation vials to trap the water and dioxide carbon produced. The fact that the water condenses at a relatively high temperature allows trapping water easily. The gaseous sample is passed through a section of teflon tubing where most of the water vapour condenses on the walls and falls straightforward into the first vial, which contains the cocktail necessary for the measurement of tritium by liquid scintillation counting (LSC). The gaseous products are then passed through the first vial where the remaining water vapour is condensed and trapped in a solution. The water droplets remaining on the walls of the Teflon tubing are then washed down into the vial with the cocktail solution. Figure 2: Separation of 14 C by the combustion of the sample in an oven (CIEMAT) Carbon dioxide passes through the first vial without reacting with the water trapping solution. The gas is then bubbled through a second vial, which contains a cocktail that reacts almost instantaneously with the carbon dioxide. The sample is then purged with nitrogen to carry off any free oxygen, which will inhibit the counting efficiency. The sample is then measured by LSC. The decontamination of β-γ emitters is higher than 99 % and there are no interferences in the spectrum of 14 C obtained by LSC (see Figure 3). To check the validity of this method (dry method) for the determination of 14 C in graphite, ion exchange resins and evaporator concentrates, the results were compared with results obtained by another method based on the collection of carbon dioxide with barium hydroxide after the mineralization of the sample (wet method). The equipment used for the mineralization method is illustrated in Figure 4. The results are shown in Table 3. 7

12 14 C Figure 3: 14 C obtained by liquid scintillation counting Figure 4: Separation of 14 C by collection with Ba(H) 2 (wet method) (CIEMAT) Table 3: Results obtained by the dry and wet methods for the determination of 14 C in ion exchange resins, evaporator concentrates and graphite samples Resin (Bq/g) Evaporator concentrate (Bq/mL) Graphite (Bq/g) Dry method 3.4E4±3% 1.28E1±1% 5.38E4±4% Wet method 3.45E4±4% 1.2E1±12% 5.8E4±5% 8

13 The volume of the aliquot for evaporator concentrates is.1 ml and the weight for ion exchange resins and graphite is 2 mg. The chemical yield is higher than 95 % and maximum preparation time is 6 minutes per treated sample. The detection limits for the determination of 14 C are 4.32 Bq/g for ion exchange resins and graphite, and.83 Bq/mL for evaporator concentrates. Both partners responsible for this task developed a similar and suitable dry incineration method for the disintegration of organic waste matrices. The wet method is an interesting alternative to dry method delivering the same good results. Task 3: Determination of radionuclides in residues Small residues after sample disintegration can not be avoided with any sample matrix. Therefore it is necessary to determine the residual α-activity in these residues. The problem of α-spectrometry of solid material is related to the absorption of α-particles in material within some µm. Therefore only α-emitters can be detected which are located at the surface of particles. The Figure 5 shows different possibilities for the position of surface bound α-emitters on a sample plate. nly α-particles emitted into the gas phase of a grid chamber will contribute to the counting rate. In the first case, it can be assumed, that a little bit more than 5 % of the a- particles will reach the gas phase. In the second case nearly all α-particles will be absorbed in on the sample plate. In all other case the emission rate will be somewhere in between. With respect to the distance of the residue particles and their position to the plate α-particles can be absorbed at neighbour particles (depending on the particle distance), on the sample plate (depending on the distance to the plate) or emitted into the gas phase. With the assumptions that any α-particle will be detected in an ionisation grid chamber, which reaches the gas phase above the sample, and all α-emitters are on the surface of the sample particles, an efficiency of about 25 % could be achieved. However, this will not include α-emitters, which are encapsulated in the particles. Angel of possible emission out of the material Paricle distance Figure 5: Possible position of surface-bound α-emitters on a sample plate Furthermore the a-particles will be scattered at the particles and the spectra will have pure resolutions, which causes problems for the nuclide identification. In order to quantify this problems, a cement powder, already analysed fro other purposes, was used for further investigations. The sample were obtained from drilling into a radioactive waste and investigated by dissolution in nitric acid/hydrofluoric acid and α-spectrometry after extraction of the α-nuclides with TP. Table 4 shows the results of the analysis. 9

14 Table 4: α-nuclide concentration in the cement powder sample Nuclide 238 U 235 U 234 U 239/24 Pu 238 Pu/ 241 Am Activity [Bq/g] The sample powder was mixed with a solution of Zapon lacquer in acetone. This dispersion was spread on a 2 cm stainless steel plate. The acetone was removed by heating and the lacquer remains as thin film which fixates the powder particles. The first example was a coarse crushed drilling powder. The milling was performed by hand with a normal agate mortar. The Figure 6 shows the α-spectrum of this sample with low resolution. However may be 3 up to 4 peaks could be identified but not for sure. But this spectrum could be used for an estimation of the total α-activity. The calibrated efficiency of the used grid chamber is 43 %. A total number of counts with a background of 37 counts gives a total activity of 7 Bq/g, which is a total efficiency of about 12 %. Drilling powder rough hand milling, 4 s, sample mass,5 g 4 4? Counts 2 1? Energy [kev] Figure 6: α-spectrum of coarse cement powder A α-spectrum with a better resolution can been obtained if the drilling powder was milled with agate a ball mill for 2 h (2 mg Powder sample, 3 ball with a diameter of 5 mm, inner dimensions of the milling vessel height 2 mm diameter 1 mm). Three peaks can be clearly identified in this spectrum and two further peaks can be assumed (Figure 7). An increase of the measuring time from 4 s to 1 s approves these additional 2 peaks (Figure 8). The peaks of this spectrum can be unfolded. However the standard unfolding programme Alps fro the company Westmeier is not able to handle such a wide energy region. Therefore the unfolding had been performed manually and by a self-developed folding procedure written in Excel. The following Table 5 shows the result of the peak enfolding. The error of the results 1

15 can be assumed in the range of 5 up to 1 %. Therefore the results of both methods show a sufficient consistence. The efficiency of the used grid chamber is 43 % for a thin layer standard. With an assumed further efficiency lost of about 3 % due to the solid material, the last line of Table 5 can be calculated. This result fits within the assumed error with the previous analysed inventory of the cement powder. Therefore this method can be used for a rough quantification of α-nuclides in solid material. The determination level of this method can be assumed with 1 Bg/g. The 235U with.5 Bg/g could not be measured. If only total α-activity is required, the detection limit will be reduced to.1 Bg/g with a measuring time of 1 s. 11

16 Drilling powder fine milling, 4 s, sample mass,1 g ? 25 5? 1 Counts Energy [kev] Figure 7: α-spectrum of fine milled cement powder 6 Drilling powder fine milling, 1 s, sample mass,1 g Counts Energy [kev] Figure 8: α-spectrum of fine milled cement powder with 25 times long measuring time 12

17 Table 5: Quantification of the nuclides inventory by enfolding Nuclide 238 U 235 U 234 U 239/24 Pu 238 Pu/ 241 Am Manual (cs -1 g -1 ) Execl. calc.(cs -1 g -1 ) Acitvity * (Bg/g) * Assumed total efficiency: 75 %. The identification of the nuclides is also problematic. The nuclide composition of the powder sample had been analysed after dissolution. Therefore it was possible to identify the following peaks: Pu/ 241 Am; /24 Pu; U; U. The spectrum is shifted to lower energies due to scattering effects and a recalibration of the energy scale on the base of the known nuclides leads to an acceptable energy calibration (Figure 9). A minimum of two peaks must be known to perform such a calibration. Such assumptions may be done with respect to the results of the analysis of soluble parts of the sample. If nothing would be known about the sample, reliable energy calibration could not be performed. After an energy correction with this four peaks, peak 1 could be identified as 243/244 Cm which was not analysed by extraction before because it was not requested. Drilling powder fine milling, 1 s, sample mass,1 g, new energy calibration 6 5 U-234 Pu-238 Am Pu-239 Counts 3 U-238 Cm-243 Cm Energy [kev] Figure 9: α-spectrum of the cement powder with a new energy calibration An addition of tracer radionuclides for energy calibration will not be successful. This can be seen in Figure 1. The peaks of the 244 Cm tracer and the 244 Cm from the solid particles have an energy difference of about 2 kev. There are to explanations for this energy shift. The cement powder had been mixed with 244 Cm solution and dried before milling. Thus it can be assumed that the 244 Cm atoms of the tracer will be directly on the surface of the powder whereas the 13

18 244 Cm of the cement powder is located some micrometer deeper. Further the lacquer/acetone solution can wash out the tracer and will be directly deposited on the steel plate in powder free regions. Drilling powder fine milling, Cm-244 added as tracer Cm-244 tracer Cm-244 in solid particles 12 Counts Energy [kev] Figure 1: α-spectrum of fine milled cement powder with 244 Cm as tracer Finally we can say that the described method is sufficient for a rough estimation of the α- nuclides but will lead to large uncertainties. However is seems to be the only method to analyse insoluble residues and therefore residues will be always an import uncertainty factor for destructive analysis WP 2: Separation of radionuclides Task 1: Group-selective separation of actinides The objective of this task is to study rapid methods for the decontamination of actinides from β,γ-emitters with co-precipitation. This method should be compared with group selective liquid-liquid extraction and solid phase extraction methods with special extractants. CIEMAT has chosen the co-precipitation technique. Co-precipitation by inorganic collectors (carriers) could be used to separate or to isolate actinides from a large group of elements. These procedures are commonly used when the step is preliminary to an extensive separation scheme. Co-precipitation can be used in order to do a preliminary concentration and to separate the actinide elements from β,γ-emitters, so the sample is decontaminated very fast from fission and activation products. The success of the separation process depends on both the choice of collector and precipitant medium. There are several co-precipitants, like Fe(H) 3, CaC 2 2 and BaS 4, which can be used in different media (basic or acid) to co-precipitate actinides elements. 14

19 Firstly, it was studied the effectiveness to carry out the separation from β,γ-emitters. For this experience a solution of an ion exchange resin from a nuclear power plant was used. The co-precipitation was carried out in basic medium with Fe(H) 3, and with CaC 2 4 and BaS 4 in acid medium. 1) Co-precipitation with Fe(H) 3 in basic medium. An aliquot of spent ion exchange resin dissolution was taken and 1 mg of Fe +3 were added. Iron hydroxide was precipitated by adding ammonia until ph=8. The iron hydroxide was recovered by centrifugation and the supernatant was measured by γ-spectrometry. The results obtained for the decontamination from β,γ-emitters are shown on Table 6. 2) Co-precipitation with CaC 2 4 in acid medium. An aliquot of spent ion exchange resin dissolution was taken and 5 mg of Ca +2 were added. The co-precipitation with CaC 2 4 was carried out by adding 5g of H 2 C 2 4 and adjusting the solution to ph=1.5 with ammonia. Calcium oxalate was recovered by centrifugation and the supernate was measured by γ-spectrometry. The results obtained for the decontamination from β,γ-emitters are shown on Table 6. 3) Co-precipitation with BaS 4 in acid medium. An aliquot of spent ion exchange resin dissolution was taken and 1 mg of Ba +2 were added. Barium sulphate was precipitated by adding H 2 S 4 concentrate. The precipitate was recovered by centrifugation and the supernatant was measured by γ-spectrometry. The results obtained for the decontamination from β,γ-emitters are shown on Table 6. Table 6: Results obtained for the decontamination from β,γ-emitters Carrier 54 Mn 6 Co 125 Sb 137 Cs Fe(H) CaC BaS 4 >99 >99 >99 >99 The best results for the separation of actinides from β,γ-emitters have been got by co-precipitation with BaS 4 in acidic medium. The decontamination was higher than 99 %. After the study of the decontamination from β,γ-emitters the next stage is to study the behaviour of Am, Pu and U in the co-precipitation with BaS 4. Am, Pu total (without having to adjust the valence state) and U co-precipitation with BaS 4 were studied in several media. Aliquots of Am, Pu total and U were adjusted to three different media:.1, 1 and 5 N HCl. 1 mg Ba +2 were added and BaS 4 was precipitated by adding one of the next solutions: H 2 S 4, Na 2 S 4, K 2 S 4 or K 2 S 4 +Na 2 S 4. In these conditions the valence of Am is III, Pu total is a mixture of oxidation states and U is VI. The results presented in Table 7 show that the highest yield, 1 %, is for the co-precipitation of Am(III) from.1n HCl solution in presence of sodium ions. However, the worst result, less than 1 %, is for the co-precipitation of U(VI) in any medium. In the case of Pu total, the best results were achieved for the co-precipitation from.1n HCl by adding a mixture of sodium and potassium ions. 15

20 Table 7: Results of Am, Pu total and U co-precipitation with barium sulphate in different media 5N HCl 1N HCl.1N HCl Medium Am % Precipitated Pu total % Precipitated U % Precipitated H 2 S <1 Na 2 S <1 K 2 S <1 K 2 S 4 +Na 2 S <1 H 2 S <1 Na 2 S <1 K 2 S <1 K 2 S 4 +Na 2 S <1 Na 2 S <1 K 2 S <1 K 2 S 4 +Na 2 S <1 According to theses results tri- and tetravalent states are co-precipitated while oxidized forms remain in solution. So it would be interesting to know the influence of the valence state of Pu and U on the co-precipitation yield. Experiences with Pu(III), Pu(IV) and U(IV) were carried out in the same conditions than before. The results are given in Table 8. Table 8: Effect of oxidation state on co-precipitation with BaS 4 5N HCl 1N HCl.1N HCl Medium Pu(III) % Precipitated Pu(IV) % Precipitated U(IV) % Precipitated H 2 S Na 2 S K 2 S K 2 S 4 +Na 2 S H 2 S Na 2 S K 2 S K 2 S 4 +Na 2 S Na 2 S K 2 S K 2 S 4 +Na 2 S

21 In general, the co-precipitation yield increases as the concentration of hydrochloric acid decreases. The effect of Na + and K + on co-precipitation of Pu(III) and Pu total is the same, increasing the recovery when both, Na + and K +, are present (Figure 11). However, in the case of Pu(IV), at low acidities, the highest co-precipitation percentage is achieved in presence of K +. % Element Precipitated Pu(III) Pu Pu(IV) 5H 5Na 5K 5NaK 1H 1Na 1K 1NaK.1Na.1K.1NaK Figure 11: Co-precipitation of Pu(III), Pu(IV) and Pu total from different hydrochloric acid solutions in the presence of K + and Na + The behaviour of Am is like Pu(III) (Figure 12) being the highest co-precipitation recovery, higher than 95 %, at.1n HCl in the presence of Na +. n the other hand, the lowest coprecipitation percentage, less than 36.7%, is obtained at 5N HCl in the presence of Na +. % Element Precipitated Am Pu(III) 5H 5Na 5K 5NaK 1H 1Na 1K 1NaK.1Na.1K.1NaK Figure 12: Co-precipitation of Pu(III) and Am at different hydrochloric acid concentrations in the presence of K + and Na + The co-precipitation behaviour of U(IV) is more similar to Pu total, being the co-precipitation efficiency of U(IV) the higher of the two (Figure 13). The highest co-precipitation percentage for U(IV), 96.5 %, is at.1 N HCl in the presence of Na + and K +, and the lowest one, 51.5 % is at 1N HCl and when neither Na + nor K + is present. 17

22 % Element Precipitated Pu U(IV) 5H 5Na 5K 5NaK 1H 1Na 1K 1NaK.1Na.1K.1NaK Figure 13: Co-precipitation of Pu total and U(IV) at different hydrochloric acid concentrations in the presence of K + and Na + According to these data it is possible to choose the best conditions to separate Pu, Am and U. For instance, a first separation of Pu and Am could be done using conditions where the yield is the highest for both, i.e., co-precipitation of Am(III) and Pu(III) in.1n HCl by adding Na 2 S 4. After the separation of Pu and Am it is possible to separate U by reducing U(VI) to U(IV) with TiCl 3. After the reduction, U(IV) is co-precipitated with BaS 4 in.1n HCl by adding K 2 S 4 +Na 2 S 4. After the decontamination of Pu, Am and U from β,γ-emitters by co-precipitation with BaS 4, the next step is to dissolve the precipitate and to purify the elements. ne method to dissolve BaS 4 is the treatment with a saturated Na 2 C 3 solution and heating during 3 minutes. The resulting precipitate, now in the form of BaC 3, should now dissolve with a nitric acid solution. The final purification of Pu, Am and U in the solution is carried out by liquid-liquid extraction and ion exchange chromatography. FZJ made separations of actinides with actinide resin (Eichrom). This solid phase extraction resin separates all actinides from matrix components in one step. A separation scheme of this column with the partitioning steps required is shown in Figure mg ascorbic acid was added to the sample to reduce potential Fe(III) to Fe(II) so that iron was not extracted by the column. The extraction of Fe(III) would have caused an earlier breakthrough of trivalent actinides which was undesired. The liquid sample in 2 M HCl was transferred to a column with Actinide Resin. The column was washed with 5 ml 2 M HCl, 5 ml 2 M HCl /.5 M HF and 5 ml H 2. The actinides were eluted with 15 ml isopropanol together with the extracting agent DIPEX. Due to the very high distribution ratio of all actinides on this column it was not possible to elute the radionuclides with a mineral acid or even with another complexing agent containing the same functional group as the resin. Therefore the DIPEX -actinide-complex had to be destroyed for further separation steps. 18

23 Figure 14: Separation scheme for actinides on actinide resin First of all the recovery rates of different actinides were determined on Actinide Resin. Standard actinide solutions were used for these experiments and 1 ml of the isopropanol fraction was measured with LSC. The recovery rates were reproducible > 8 % for all requested radionuclides (see Table 9). The same recovery rates were obtained with concrete or ion exchange resin samples. These samples came out of the microwave decomposition with a concentration of 5 M HN 3. It was not necessary to converse into HCl to receive the same good recovery rates so that the samples could be utilised at once. Table 9: Recovery rates of actinides on actinide resin Radionuclide Recovery rate 239 Pu 91 ±5 % 241 Am 96 ± 2 % 233 U 93 ± 4 % 237 Np 86 ± 6 % 244 Cm 14 ± 5 % 232 Th 83 ± 5 % It was possible to prepare 1 ml of the isopropanol fraction directly for α-measurement to have a first look on the actinides which are present in the sample. A quantitative analysis could even be carried out when 1 µl of the fraction was prepared together with 2 ml of a 1 % Zapon lacquer solution on a stainless steel target ( 2 mm) for α-measurement in a grid chamber. A more time-consuming electrodeposition was not necessary. 19

24 As mentioned before it was necessary to destroy the extracting agent DIPEX for the next separation steps. The isoprapanol fraction was first evaporated and then heated on a hot plate with H 2 2 and NaV 3. This method led to a black residue which had to be treated the described way another three times. This destruction step was very difficult to handle because of the exothermic reaction and was thus also very time consuming. It was replaced by the destruction with HN 3 and H 2 2 using a microwave decomposition. The characteristics of this used microwave decomposition are shown in Figure 15. This step could be carried out in one hour. After evaporation a residue was obtained which contained mainly amorphous parts and methylen-diphosphonic acid but no DIPEX. This residue was dissolved in 2 M HN 3 /1 M Al(N 3 ) 3 and put into the next separation steps without any problems. It was even shown that the recovery rates after the destruction in the microwave were higher than the rates received with the other method (see Table 1). Figure 15: Microwave programme with temperature and pressure data for the DIPEX destruction Table 1: Recovery rates of the actinides after different destruction methods of DIPEX Radionuclide Recovery rate after destruction with H 2 2 and NaV 3 Recovery rate after microwave decomposition 237 Np 9 ± 3 % 85 ± 5 % 241 Am 77 ± 7 % 85 ± 5 % 233 U 65 ± 5% 9 ± 4 % 239 Pu 65 ± 6 % 9 ± 6 % 2

25 ENEA tested separation columns, appropriate and selective for the actinides (TRU and UTEVA columns. Actinides with irresolvable alpha energies, such us 241 Am, 238 Pu and 234 U, must be chemically separated to enable measurement. Use of Eichrom s columns, UTEVA and TRU Resin, permit a selective separation of these elements. 242 Cm is also separated with these techniques: it is collected together with 241 Am. Determination is carried out by α-spectrometry (deposition on hot cupel). The general scheme employed in this kind of separation is shown in the following figure: Figure 16: Scheme of separation employed in separation of actinides The partners developed and examined different kind of methods: co-precipitation and solid phase extraction for the separation of actinides. Both methods deliver good results and can be used in an equal manner. Task 2: Liquid-liquid extraction The liquid-liquid extraction described in this task is the continuation of the co-precipitation process described in Task 1 by CIEMAT. The attack of BaS 4 precipitate by a saturated Na 2 C 3 solution produces two fractions: a solution of Na 2 C 3, containing S 2-4 ions, and a precipitate of BaC 3 which is dissolved by a little amount of nitric or hydrochloric acid. In both dissolutions it is necessary to separate Pu, Am and U from each other and some elements that could interfere in the determination by alpha-spectrometry. In the case of BaC 3 fraction, Am is extracted by.45m di-(2-ethylhexyl)phosphoric acid (HDEHP) in heptane from an aqueous solution of ph = 2-3. Pu and U are extracted from a widely range of nitric or hydrochloric acid concentrations by tributyl phosphate (TBP) or HDEHP, and from weakly acidic solutions by HDEHP. Pu can be present in several oxidations states which might influence the distribution ratio, so it was checked the extraction of Pu with.45m HDEHP in n-heptane from two solutions: a 7M HN 3 solution containing Pu total and another one adjusting the oxidation state to Pu(IV). The extraction of Pu(IV) was 1 %, while it was 8 % for Pu total. The most part of U in BaC 3 fraction is U(IV), so it was studied the extraction of both oxidation states VI and IV in 1N HCl with.45m HDEHP in n-heptane. The results were 1 % for U(VI) and 55 % for U(IV). 21

26 In this fraction, Ba is also present and it could be extracted to the organic phase in some experimental conditions. If Ba is present in the final dissolution it could be interference in the preparation of the radioactivity source by a method of electrodepositon in sulphuric acid medium, so it was assayed the behaviour of Ba in the extraction with HDEHP. The result was that Ba was extracted to the organic phase from ph = 2-3 solutions and it could be stripped with.75n HCl or dissolutions of higher hydrochloric or nitric acid concentration. In the case of the saturated Na 2 C 3 solution it was necessary to study the extraction conditions for Pu, Am and U. As consequence of the HDEHP capacity to extract actinides from weak acidic solutions, this product was chosen to study the separation of actinides. A saturated Na 2 C 3 solution was adjusted to ph = 2-3 with nitric acid. Nitric acid was chosen because the sodium nitrate solubility, 86 g/l at 18 ºC, is higher than the sodium chloride solubility, 36 g/l at 18 ºC. The percentages of extraction for Pu, Am and U were 95, 88 and 1 %, respectively. Due to the tendency of actinides to form insoluble species at low acidities, the extraction from higher nitric acid concentration was studied. Concentrate nitric acid was added to the sodium carbonate solution until it was 1M HN 3, and the extraction with.45m HDEHP in n-heptane was assayed. The percentages of extraction for Pu, Am and U are shown in Table 11. Table 11: Results of Am, Pu total and U extraction with.45m HDEHP in n-heptane from 1.8M NaN 3 ph = 2-3 or 1M HN 3-1.5M NaN 3 Element % Extracted from organic phase 1.8M NaN 3 1M HN ph = M NaN 3 Pu 95 9 Am 95 1 U(VI) After the extraction of Pu, Am and U to the organic phase, by TBP or HDEHP, the next step was to study the optimum conditions for sequential stripping of actinides. The extraction of Pu from nitric acid medium yielded the presence of nitrate ions in the organic phase and it hindered the stripping of Pu by reduction to Pu(III). Successful stripping, therefore, required the use of a wash cycle with a hydrochloric acid solution to remove the nitrate prior to the back extraction of Pu with an aqueous solution of a reducing agent. Firstly, the reduction to Pu(III) by a mixture of hydrochloric acid and hydroxylamine hydrochloride was tested. The results are shown in Table

27 Table 12: Results of Pu back extraction with hydrochloric and hydroxylamine solutions Stripping solution % extracted.45m HDEHP-n-heptane 5% TBP -n-heptane 4N HCl 1 4N HCl-1N NH 2 H N HCl N HCl- 1N NH 2 H 3 91 The most successful back extraction, about 9%, is carried out by 1.5N HCl-1N NH 2 H from 5 % TBP in n-heptane and it decreases as the hydrochloric acid concentration increases. Back extraction efficiency from.45m HDEHP in n-heptane, with the mixture hydrochloric acidhydroxylamine, is very low so it was necessary to look for other experimental conditions. Further Pu back extraction tests from.45m HDEHP in n-heptane with hydrochloric acid solutions were performed. Three different concentrations of hydrochloric acid were used: 1, 8 and 11 N in a mixture with one reducing agent:.1n hydroiodic or.5n ascorbic acid. The results are shown in Table 13. Table 13: Results of Pu back extraction with hydrochloric and hydroiodic or ascorbic acid dissolutions.1hi 1N HCl 8N HCl 11N HCl.5N Ascorbic acid.1hi.5n Ascorbic acid.1hi.5n Ascorbic acid % extracted Stripping yield of Pu rises with increasing hydrochloric acid concentration to a maximum at 8N hydrochloric acid, and it decreases at higher acid concentration. However, in the case of the use of a solution of ascorbic acid in hydrochloric acid the recovery increases as the hydrochloric acid concentration does. The highest efficient stripping was achieved with the mixture.5n ascorbic acid and 11N hydrochloric acid, which extracted about 93 % of Pu. Am back extraction from the organic phase,.45m HDEHP-n-heptane, was carried out by a solution of 8N HCl or 7M HN 3. The recovery was about 9 %. 85 % of U was stripped from TBP with a solution of.5m H 2 C 2 4. For stripping U from.45m HDEHP in n-heptane, it was assayed with solutions of hydrochloric acid 1, 8 or 11N, and with a solution of.5m H 2 C 2 4. The results are shown on Table 14. The back extraction of U from.45m HDEHPi n n-heptane increases with increasing hydrochloric acid concentration. The highest recovery was about 9 % when 11N HCl was used. 23

28 Table 14: Results of U back extraction with different hydrochloric acid concentration or.5m oxalic acid dissolutions 1N HCl 8N HCl 11N HCl.5M H 2 C 2 4 % extracted ENEA examined the SANEX process (selective actinides extraction process). This is a selective separation of actinides from lanthanides with a trisynergistic mixture made of α-cyanodecanoic acid.9 mol/l, DMDBTDMA C14.6 mol/l and triazine TMHADPTZ.1 mol/l in TPH. Five solutions at five different acidities spiked with 35 Bq/cc of 241Am and 5,9 KBq/cc of 152Eu were tested: Table 15: Solutions used for SANEX process tests Feed label HN 3 mol/l Eu (g/l) Gd (g/l) Y (g/l) Sm (g/l) La (g/l) Ce (g/l) Pr (g/l) Nd (g/l) Mo (g/l) 1,12 1,14 1,61 3,63 4,99 8,9 21,2 9,85 34,8,16 2,46 1,5 1,41 3,28 4,34 8,1 2,8 8,92 34,2,18 3,86 1,18 1,67 3,7 5,5 8,12 21,1 1,9 34,7,18 4,23 1,1 1,5 3,35 6,34 8,36 21,84 9,33 3,,19 5,38 1,16 1,59 3,58 4,74 8,81 23, 11,7 37,9,19 The distribution coefficients are calculated as D = Conc. (rg.)/conc. (Aq.) and determinations were made by γ-spectrometry and ICP-ES spectrometry: Table 16: D for americium and europium; determinations by γ-spectrometry Concentration [H + ] mol/l D for 241 Am D for 152 Eu,12 1,18,19,46 1,24,16,86 1,12,17,23,6,11 24

29 Table 17: D for lanthanides; determinations by ICP-ES spectrometry Concentration [H + ] Distribution coefficients D mol/l HN 3 Mo Gd Sm Y La Ce Pr Nd,12,3,94,15,14,7,11,13,15,15,46,8,96,9,9,3,7,9,11,11,86,17,99,7,7,2,6,8,9,9,23,26 1,2,7,5,2,6,8,8,9 The solution with [H + ] =,38 was not analyzed because of the instability of the trisynergistic mixture at this acidity. It was possible to notice the remarkable difference between D values for americium and europium (see). 1,4 1,2 1 D values,8,6,4,2 1 2 Solution Label 3 4 Americium Europium Figure 17: SANEX process; comparison between D values for americium and europium Task 3: Solid-phase extraction and ion exchange chromatography Solid phase extraction had been examined during the 4 th Framework of the EC for the separation of actinides from radioactive wastes. These developed methods should be enhanced for the separation and enrichment of other elements like Sr, Tc, Ra or Zr. A rapid and accurate method for the determination of 9 Sr in samples from nuclear power plants (graphite, ion exchange resins and evaporator concentrates) was developed by CIEMAT. The separation of strontium from the other elements is based on the extraction chromatography technique, using a macrocyclic polyether 4,4'(5')-bis(t-butylcyclohexano)-18-crown-6 (abbreviated DtBuCH 18 C 6 ) in 1-octanol, which is immobilised on an inert polymeric substrate (acrylic ester non-ionic polymeric adsorbent, 8 to 6 µm) (Sr Resin column by Eichrom). The new procedure was developed on a standard solution containing 54 Mn, 65 Zn, 6 Co, 137 Cs, 85 Sr and 88 Y. The procedure is comprised of the following steps: 25

30 1. An aliquot of the standard solution is evaporated to dryness, 2. The residue is dissolved with 8M HN 3 -.5M Al(N 3 ) 3 and the solution is transferred onto a Sr resin column, 3. The column is rinsed with 8M HN 3, 4. The strontium is eluted from the column using.5m HN 3 and the solution is measured by γ-spectrometry. The results obtained are shown in Table 18. The chemical yield for the separation of strontium was 8 % with a high decontamination from β-γ -emitters. Table 18: Results obtained for the decontamination of strontium from β-γ-emitters in a standard solution Step 54 Mn 6 Co 65 Zn 85 Sr 88 Y 137 Cs Initial 4.68E2 5.6E2 1.1E3 6.7E2 1.26E3 3.9E2 Load 3.32E2 3.95E2 8.49E2 <1.2E 8.69E2 3.4E2 Rinse 7.43E1 7.5E1 1.73E2 5.7E1 1.96E2 6.82E1 Eluate <9.4E-1 <7.E-1 <2.5E 4.87E2 <1.5E <8.6E-1 This procedure was applied to the analysis of 89/9 Sr in wastes from nuclear power plants (graphite, ion exchange resins and evaporator concentrates). Chemical recovery was determined by gravimetric and 89/9 Sr activities were measured by LSC. 89/9 Sr activities were calculated from two measurements of 89 Sr (β max. = 1492 kev, T1/2 = 5.65 h) and 9 Sr (β max. = 546 kev, T1/2 = y) in the presence of 9 Y: the first measurement was carried out immediately after the chemical separation of strontium, and the second was undertaken eight to ten days after its separation (Figure 18). 9 Sr 89 Sr + 9 Y Figure 18: Spectrum of 89/9 Sr obtained by liquid scintillation counting, after its separation by extraction chromatography 26

31 However, when this procedure was applied to real samples, containing an activity concentration of α-emitters similar or higher than strontium, a contamination came out and produced interference in the LSC spectrum (Figure 19). In the LSC spectrum, it is shown the presence of 241 Pu and, also, there will be present the others isotopes of plutonium which are α-emitters. Moreover, the presence of both americium and curium which appear in the same region than isotopes of plutonium is possible. The presence of α-emitters in the LSC spectrum produces wrong results for the strontium determination. 9 Sr α-emitters 9 Y 241 Pu Figure 19: Liquid scintillation spectrum of strontium contaminated by α-emitters To know the behaviour of the main α-emitters on a Sr Resin column, it was done an experience with a sample spiked with 238 Pu, 241 Am and 244 Cm. The column was loaded with 8M HN 3 -.5M Al(N 3 ) 3 and was rinsed with several solutions in order to eliminate the α, β, γ emitters which could interfere in the determination of strontium: 1. 8M HN 3 -,5M H 2 C M HN M HN 3. All the fractions were measured by LSC. Table 19 summarizes the results for the study of α-emitters behaviour on a Sr Resin column. Table 19: Results obtained for the study of α emitters behaviour on a Sr resin column Step 238 Pu Am 244 Cm Initial 2.36E2 1.8E2 1.5E2 Load <2.E E2 1.47E2 Rinse 8M HN 3 -,5M H 2 C E2 <2.E-2 <2.E-2 Rinse 8M HN 3 <2.E-2 <2.E-2 <2.E-2 Rinse.5M HN 3 <2.E-2 <2.E-2 <2.E-2

32 According to the results obtained, the contamination in the spectrum (Figure 19) of strontium is debt to plutonium, because americium and curium are eliminated when the column is loaded. Therefore, to eliminate plutonium, before the elution of strontium, it is very efficient to rinse the column with a solution of 8M HN 3 -,5M H 2 C 2 4. The procedure developed for the determination of 9 Sr consists of the following steps: 1. Dissolution of the sample (graphite, ion exchange resin or evaporator concentrate) by an acid attack. 2. An aliquot of the sample solution is taken, 1 mg of Sr(N 3 ) 2 are added, and is evaporated to dryness. 3. The residue is dissolved with 8M HN 3 -.5M Al(N 3 ) 3 and the solution is transferred onto a Sr-Spec column. 4. The column is rinsed with 8M HN 3 -,5M H 2 C 2 4 and 8M HN The strontium is eluted from the column using.5m HN 3. Chemical recovery is determined by gravimetric and 89/9 Sr activity is measured by LSC. The procedure was applied to one sample that had an activity of plutonium similar to strontium and the result was very good. In Figure 2, the spectrum of strontium, obtained by LSC, after the column was rinsed with a solution of 8M HN 3 -,5M H 2 C 2 4 to eliminate the plutonium, is shown. 9 Sr 9 Y Figure 2: Liquid scintillation spectrum of strontium after the decontamination of plutonium The detection limits are 1Bq/g for the analysis of graphite, using 4.47E-2 g of sample, 4.6E-1 Bq/g for ion exchange resins, using 1.2E-1 g of sample, and 4.2E-2 Bq/mL for the analysis of evaporator concentrates, using 1.6 ml of sample. FZJ also used this method for analysing 9 Sr in nuclear waste samples and was able to solve the plutonium interference with the method developed by CIEMAT. FZJ used a solid phase extraction method for the separation of 99 Tc with TEVA resin (Eichrom). 99 Tc can be determined from dissolved samples with a commercial available extraction chromatography resin called TEVA resin from Eichrom. The sample is dissolved by a microwave composition. Before using this column it is essential to separate caesium which 28

33 can be present in the sample and will be interfere with the Technetium separation. For this purpose the sample solution in concentrated nitric acid is mixed with ammoniummolbydophosphat ((NH 4 ) x Mo(P 4 ) y ) and after 24 h this mixture is filtered. With a γ-measurement it can be assured that the sample is now free from caesium. The solution is then adjusted to 2 M HN 3 and transferred to the TEVA resin column which was conditioned with 2 M HN 3. The column is washed with.5 M HN 3 and eluted with 8 M HN 3. The separated technetium is measured by LSC. Recovery rates of % are obtained. The separation scheme is shown in Figure 21. Figure 21: Separation of 99 Tc with a TEVA resin column Task 4: HPLC HPLC is a fast easy-to-automate method for the multi-element characterisation of radioactive waste. In most of the HPLC methods developed by now, ion exchange resin is used. The main disadvantage of this resin is the easy overloading by matrix compounds, which makes pre-separation of the matrix necessary. Therefore new techniques should be found to overcome additional pre-separations and further work has to be done on the development of the HPLC as fast on-line method for waste characterisation. ENEA investigated chromatographic methods to analyse solutions spiked with radionuclides. In particular, a DINEX 5 equipped for the radio-analysis of cations, anions, transition metals and lanthanides was used. A) Anions The scheme of the apparatus used for this analysis is shown below (Figure 22), together with the operating conditions in Table 2: 29

34 Eluent Sample Regenerator (H 2 at 18 MΩ) Acquisition and elaboration data Liquid scintillator Conduct. Detector Radiochemical Detector Pre column Analytic Pump Injector Column Waste Cold Waste Hot Waste = Waste management valve Figure 22: Scheme of HPIC DINEX 5 used for analysis of anions Table 2: perating conditions for anionic analysis Eluent Na 2 C 3 /NaHC 3 = 3.5 mm/1. mm solution Type of elution Isocratic Flow 1.2 ml/min Total time of elution 12 min Column INPAC AS14A Analytical Column (4x25 mm) Pre-column INPAC AG14A Guard Column (4x5 mm) Loop 5 µl Suppressor ASRS-II (4 mm) External regenerator H2 (R> 17.8 MΩ cm) Background of conductibility < 17. µs Electrochemical detector ED-4 module (Dionex instrument) Radiochemical detector FL-NE 525TR (Packard instrument) Volume of the radiochemical cell.5 ml Flow of the liquid scintillation cocktail 2.4 ml/min In order to test the efficiency of this device on real and complex matrices, initially, tests on synthetic matrices prepared in laboratory were carried out. To operate a remarkable reduction of time work and time exposure, some standard solutions of anions with low salts content, spiked with 36 Cl, were prepared and tested. Afterwards, when the best operating conditions were found, some synthetic standard solution of anions, with high salts content, spiked with 36 Cl as standard, were prepared and tested; the interference and the effects of the matrix constituents on the separation chromatographic and on the chemical and radiochemical analysis, were shown and estimated. 3

35 The conductivity detector in the HPIC-LSC device was calibrated on three levels with three standard solutions at low salts content and with a concentration of anions shown in the following Table 21. These three standard solutions held a constant quantity of 36 Cl standard. Table 21: Concentrations of the anions in the three standard solutions with low salts content Element Conc. Std 1 (mg/l) Conc. Std 2 (mg/l) Conc. Std 3 (mg/l) F Cl -* 3.5* 5.5* 9.5* N Br N 3 - HP 4 2- S * The total concentration of the Cl - refers at the sum of cold Cl and 36 Cl radioisotope concentrations. In the following figures, the chemical and radiochemical chromatograms of the anions present in the three standard solutions at low salts content are reported; these solutions were used for the calibration of the conductivity detector and for the contemporaneous radiochemical detection of the radioisotope 36 Cl, which it is present in the three standard solutions. Figure 23: Chemical and radiochemical chromatograms of the anions present in the Std 1 solution obtained with conditions illustrated in Table 2 31

36 Figure 24: Chemical and radiochemical chromatograms of the anions present in the Std 2 solution, obtained in the same conditions of Std 1 Figure 25: Chemical and radiochemical chromatograms of the anions present in the Std 3 solution, obtained in the same conditions of Std 1 and 2 Also the liquid scintillation counter was calibrated, using solutions with low salt contents and increasing quantity of the standard 36 Cl. The results are reported in the following Table

37 Table 22: Bq injected and average values of Bq detected activity of the 36 Cl present in the three standard solutions with low salts content used for the calibration of the LS counter Element 36 Cl - Activity injected (Bq) Activity detected (Bq). 2.8 ± ± ± ±.16 Then, the HPIC-LSC device was tested with solutions at high salts content. These solutions are different from previous solutions at low salt content because they also contain all the elements present in the cement (alkali and earth-alkali metals, transitions metals, silicate, etc.), at given concentrations. So there is a high matrix effect. Also in this case, the conductivity detector of the HPIC-LSC device, was calibrated on three levels, using three solutions at high salts content (these solutions were prepared dissolving the cement by acid digestion in microwave oven, and then carrying out an appropriate dilution) with the anions at the same standard concentrations reported in the previous Table 21, and with a constant quantity of standard 36 Cl. In the three following figures, the chemical and radiochemical chromatograms of the anions present in the three standard solutions at high salts content are reported: Figure 26: Chemical and radiochemical chromatograms of the anions present in the Std 1 solution 33

38 Figure 27: Chemical and radiochemical chromatograms of the anions present in the Std 2 solution, obtained in the same conditions of Std 1 Figure 28: Chemical and radiochemical chromatograms of the anions present in the Std 3 solution, obtained in the same conditions of Std 1 and 2 Also for the liquid scintillation counter a calibration with three levels was carried out, using the solutions at high salts content. In the following figures the radiochemical chromatograms with increasing quantities of 36 Cl are reported. 34

39 Figure 29a-b-c-d-e: Radiochemical chromatograms of the anions present in the solutions with high salt content and with increasing quantity of 36 Cl 35

40 Table 23: Injected and detected activity of the 36 Cl present in the standard solutions with high salt content used for the calibration of the LS counter Element 36 Cl Activity injected (Bq) Activity detected (Bq) Injections I II III IV V Average value ± ± ± ± 7.9 B) Lanthanides In the following Figure 3, the scheme of the HPIC-LSC device employed and the relative operating conditions are shown in Table 24 and 25 Table 25: Figure 3: Scheme of HPIC DINEX 5 used for analysis of lanthanides 36

41 Table 24: perating conditions for lanthanide analysis Eluent Type of elution Flow rate Total time of elution Column Pre-column A) H 2 C 2 4 /(CH 3 ) 4 N + H - /KH = 24 mm/3 mm/5.6 mm/15 mm B) Diglicolic acid 16 mm/kh 19 mm C) 18 MΩ deionised water Gradient 1.2 ml/min 12 min Loop 5 µl Absorbance detector Post-column reagent Background of absorbance detector Radiochemical detector Volume of the radiochemical cell INPAC CS5A Analytical Column (4 x 25 mm) INPAC CG5A Guard Column (4 x 5 mm) AD2 (Dionex instrument) [4-(2-pyridil-azo)-resorcinol].5 mm.1 AU FL-NE 525 TR (Packard Instrument).5 ml Flow of the liquid scintillation cocktail 2.4 ml/min Table 25: Gradient conditions for lanthanide analysis Time %A %B %C Init The chemical and radio-chemical chromatograms obtained in this way are reported below: 37

42 Figure 31: Chromatogram of a standard solution of lanthanides Figure 32: Radio-chromatogram of 151 Sm and 152 Eu C) Transition metals The scheme of the device was already shown in Figure 3. In Table 26 the operating conditions relative to this analysis are reported: 38

43 Table 26: perating conditions for transition metals analysis Eluent Type of elution Flow rate Total time of elution Column Pre-column PDCA/KH/K 2 S 4 /HCH = 7. mm/66 mm/ 5.6 mm/74 mm Isocratic 1.2 ml/min 15 min Loop 5 µl Absorbance detector Post-column reagent Background of absorbance detector Radiochemical detector Volume of the radiochemical cell INPAC CS5A Analytical Column (4x25 mm) INPAC CG5A Guard Column (4x5 mm) AD2 (Dionex instrument) [4-(2-pyridil-azo)-resorcinol].5 mm.1 AU FL-NE 525 TR (Packard Instrument).5 ml Flow of the liquid scintillation cocktail 2.4 ml/min In this case standard solutions containing 55 Fe and 63 Ni were analyzed and the results are presented in the following figures. Figure 33: Chemical-chromatogram of a standard solution containing 6 ppm of Fe and 8 ppm of Ni 39

44 Figure 34: Radio-chromatogram of a standard solution containing 12.5 ppm of Fe and 25 ppm of Ni spiked with 8 Bq/cc of 55 Fe and 14 Bq/cc of 63 Ni The results are summarized in Table 27. Table 27: Experimental results for radio-analysis of transition metals Radionuclide Bq injected Cps detected Efficiency (%) Detection limits 55 Fe ± Bq 63 Ni ± Bq 55 Fe ± Bq 63 Ni ± Bq Blank solution. 1.9±.3 Fe.9±.1 Ni - 3σ FZJ focused on the development of a reversed phase HPLC for this project. With the help of a suitable organic ion-pairing or complexing agent, inorganic ions can also be separated on a reversed phase HPLC column. If the chosen organic agent favours to complex the analyte ions above the matrix ions, the problem of column overloading, can be overcome. The initial idea was to develop a reversed phase HPLC method (RP-HPLC) for the separation of actinides. During previous research with solid phase extraction, DIPEX complexes have been proved to be extremely stable and DIPEX is an actinide selective complexing agent. Due to both of these characteristics, DIPEX seems to be the ideal agent to complex the actinides and separate the complexes by RP-HPLC. The matrix separation and complex building has been performed in one step by solid phase extraction on Actinide Resin (Eichrom). After the actinides were extracted by the DIPEX and the SPE-column was washed to remove the matrix, the whole actinide-dipex complexes were eluted from the column with methanol or isopropanol in order to separate these complexes on a RP-HPLC column. The HPLC device used for this work consists of a gradient pump with gradient mixer and an HPLC column coupled online with a UV/VIS detector and LSC detector. A schematic overview of the device can be seen in Figure 35. 4

45 Eluent Gradient mixer Eluent pump Helium LSC detector Waste UV/VIS detector HPLC column Data acquisition and evaluation Figure 35: Experimental arrangement of the HPLC system with online LSC detector The first RP-HPLC column tested with different eluents was a Nucleosil C18 (DS) column. The operating conditions which were held constant during the experiments are listed in Table 28. n this column, U-DIPEX and Cm-DIPEX complexes could be eluted with different eluent strength. For the 233 U-DIPEX complex a peak on the online LSC detector was observed after 5 min. of elution with 75 % ethanol and 25 % water. A 244 Cm-DIPEX peak could not be observed with this eluent within a reasonable amount of time (Figure 36). Table 28: Chromatographic conditions held constant during all experiments eluent flow rate 1 ml/min sample loop 1 µl column temperature 3 C volume LSC cell LSC cocktail flow rate.5 ml 2 ml/min 41

46 35 3 C ount rate / CPM U Cm T im e / m in Figure 36: LSC-Chromatograms of 233 U- and 244 Cm-DIPEX with eluent: 75% ethanol, 25% water With a stronger, more apolar, eluent, a mixture of 85% ethanol and 15% water, the 244 Cm- DIPEX complex could be eluted. However, problems with the recovery occurred. It was found that the 244 Cm remained on the column could be displaced by the injection, by way of the sample loop, of a solution containing the complexing agent DIPEX. A chromatogram of the 244 Cm-DIPEX complex can be seen in Figure C o u n t ra te / C P M Cm Injection DIPEX Time / min Figure 37: LSC chromatogram of 244 Cm-DIPEX with eluent: 85 % ethanol, 15 % water. Injection of DIPEX solution necessary to remove all Cm from the column. 42

47 In literature [1], there were indications that the actinide-dipex complexes are polynuclear, which would make a separation impossible. This polynuclear nature of the complexes could not been proved with HPLC experiments. A mixture of 244 Cm- and 233 U-DIPEX complexes has been prepared on a SPE column, eluted with methanol and separated on the HPLC column. The eluate was collected in several vials to make nuclide identification with α-spectroscopy possible. It was expected that it would be possible to separate the 2 complexes with a gradient elution. A linear gradient form ethanol-water to 1 % ethanol within 3 min., followed by 3 min. of elution with 1 % ethanol, has been chosen. nly U peaks (possibly 2 different oxidation states), with a small retention time, could be observed in the chromatogram. The 244 Cm could not be detected, so it was still on the column. After injection of the ligand DIPEX, the Cm activity could be detected on the online LSC detector. In Figure 38, a chromatogram of the separation of the 244 Cm-DIPEX U-DIPEX mixture can be observed. Several injections of DIPEX were necessary to remove the 244 Cm from the column. During repeated measurements, it was observed that the recovery of both nuclides drops significantly. Another problem was clogging of the column because of the high viscosity of the DIPEX. Due to the several problems arisen, the experiments with this column have been terminated Cm 5 Count rate / CPM U Injection DIPEX Time / min Figure 38: LSC-Chromatogram of 233 U- and 244 Cm-DIPEX with linear gradient from 75% ethanol, 25% water to 1% ethanol in 3 min, followed by 3 min 1% ethanol Because of the difficulties in eluting the actinide-dipex complexes from the nucleosil C18 column, it was thought that a column with a shorter C-chain and with a better endcapping could give better results. Two RP-HPLC columns (Inertsil C8 and Inertsil C4) were tested with different eluents. The actinide-dipex complexes stuck even more on both of these columns than on the nucleosil C18 column. Even with such a strong eluent as 1% isopropanol, it has not been possible to elute them from these columns. nly after injection of a DIPEX solution by way of the sample loop, radioactivity could be detected with the LSC detector. Because the DIPEX seems to be able to displace the actinide-dipex complexes from the column it was tried whether the addition of a small amount of DIPEX to the eluent could give better results. The actinide-dipex complexes could be eluted this time, but all showed the same retention time, so their separation was impossible. Besides, the addition of the DIPEX to the eluent caused the clogging of the column. 43

48 No useful separation of the actinide-dipex complexes could be developed. Besides, if the complexes have a polynuclear character like suspected in literature, the separation in nuclide specific complexes would theoretically be impossible. For these reasons the experiments with the DIPEX as complexing agent were broken off. The next complexing agent investigated for actinide separation was CMP, commonly used in solid phase and liquid liquid extraction. CMP shows affinity for actinides and lanthanides. The actinide-cmp complexes were prepared in the same way as the actinide-dipex complexes on a self made SPE resin. Two RP-HPLC columns were tested, a Nucleosil C18 and an Inertsil C4 column. n both of these columns the free ligand CMP showed already a high retention. After injection of the actinide-cmp sample, no radioactivity could be eluted with different eluents. nly with such a strong eluent as cyclohexane-isopropanol 6-4 %, radioactivity was eluted within the column death time. So the only way to get a signal on the radioactivity detector was to make the eluent that apolar that the actinide-cmp complex didn t show affinity to the stationary phase of the column any more. As soon as the mobile phase was polar enough to allow the complex to interact with the stationary phase, it became impossible to elute the radioactivity. Also reversed phase thin layer chromatography (RP-TLC) has been used to speed up the eluent optimisation. The results of the separation on HPLC and RP-TLC suggested there could be problem with the actinide-cmp complex stability. UV/VIS spectroscopy of different lanthanide and lanthanide-cmp solutions showed that in the sample solution the complex was still not dissociated. In order to avoid radioactivity, for the UV experiments lanthanides were chosen to represent the actinides, because they have a similar chemical behaviour. In Figure 39 an example of a UV/VIS spectrum of different Pr and Pr- CMP solutions can be seen. Because with auto-radiographic detection, only γ-radiation could be detected, some RP-TLC experiments with 152 Eu-CMP complexes were performed. These experiments showed that the radioactivity was scattered on the TLC-plate and only 5% of the radioactivity was on the same place where the CMP was visible with UV detection. This indicates that the complexes probably dissociate on the stationary phase.,6 Pr in ACN Pr + CMP in ACN Pr in Water Pr + CMP in xylene Pr in IPA Pr + CMP in IPA,5,4 A,3,2,1 -, λ (nm) Figure 39: Detail of UV/VIS spectra of Pr and Pr-CMP complexes in different solvents 44

49 Alternatively, a RP column on the base of styrene-divinylbenzene copolymer has been loaded with the complexing agent CMP. Based on the principles of high-performance chelation ion chromatography (HPCIC), actinides will be separated with aqueous eluents. HPCIC combines the high efficiency and robustness of RP-HPLC with the selectivity of the chelating agent. This chromatography method is used to be relative insensitive to high ionic strengths. The separation of a mixture of Am and U could be achieved on our self modified column with a two step gradient from.5 M HN 3 to water and from water to.25 M Na 2 C 3. A chromatogram can be observed in Figure Cm Radioactivity / CPM U Time / min Figure 4: Chromatogram of the separation of a Cm-U mixture on the self-modified column The separation of an Am-U and of an Am-Pu mixture has also been successful. The valence of Pu has been adjusted to the IV+ state by NaN 2. Because of the extremely high partition coefficients of the tetra- and hexavalent actinides on our column, the separation of Pu and U has turned out to be very difficult. The tetra- and hexavalent actinides could only been eluted as negative carbonate complexes. These negative carbonate complexes could then be separated on an anion exchange column connected with our CMP modified column. The tetravalent Pu could be eluted from the anion exchanger with a Na 2 C 3 solution of a higher concentration as has been used to elute the tetra and hexavaltent actinides from the CMP modified column. For the elution of U a mixture of Na 2 C 3 and NaCl or NaN 3 is necessary. With the same elution programme, Th could be separated from U on these 2 connected columns. An overview of the HPLC arrangement with the two column approach can be seen in Figure 41. The parameters which were held constant during all experiments are listed in Table

50 Eluent Gradient mixer HPLC pump Sample loop Helium HPLC pump Waste Reduction agent 5 ml loop LSC Cocktail LSC detector Waste HPCIC column Waste UV/VIS detector IE column Data acquisition and evaluation Figure 41: Experimental arrangement of the HPLC system with two columns Table 29: Chromatographic conditions held constant during all experiments eluent flow rate 1 ml/min sample loop 1 µl column temperature 3 C volume LSC cell.5 ml LSC cocktail flow rate 2 ml/min 46

51 In order to separate both tetravalent actinides, Th and Pu, the reduction of Pu(IV) to Pu(III) on the CMP modified column has been applied since trivalent actinides are almost not retained on this column. Different reduction agents have been tested. Due to the high complexation of Pu(IV) by the CMP, this reduction was difficult to achieve. Strong reduction agents were not able to reduce the Pu(IV) without the presence of Fe(II). In a solution of ascorbic acid, the Fe(II) concentration could be kept as low as possible. This is important because a next separation on the same column can only be performed after the Fe is removed from the column. The higher the Fe(II) concentration in the reduction agent, the higher the required column recovery time. The final elution programme for the separation of the actinides Am, Pu, Th, and U can be observed in Tables 3 and 31. A chromatogram obtained with this elution programme is illustrated in Figure Am Pu 233U 5 CPM 4 229Th Time [min] Figure 42: Chromatogram of a standard solution containing 19. Bq 241 Am, 15.9 Bq 238 Pu, 17.1 Bq 229Th, and 32.9 Bq 233 U Table 3: Eluents for actinide elements Eluent A.5M HN 3 +.1M NaN 2 Eluent B 2M Na 2 C 3 Eluent C.25M Na 2 C 3 +.5M NaN 3 Reduction agent.5m Fe(II)S4 in 1 mg/l ascorbic acid 47

52 Table 31: Separation conditions for actinide elements Time [min] Eluent Eluent A 5 Injection of reduction agent through the 5mL loop 1 Closing of the 5mL sample loop 12 85% eluent A + 15% eluent B 15 Eluent B 2-3 Eluent C The online LSC detector can provide for a direct determination of the radioactivity, provided the detector efficiency has been determined previously. This detection efficiency depends on the radionuclide and the matrix (eluent). The detector efficiencies for each radionuclide are described in Table 32. The detection limits, which are also represented in Table 32, have been determined using the 3σ-criterion. Table 32: Detector efficiency and detection limits Nuclide Injected Bq Measured CPS Efficiency ε % Detection limits 241 Am ± x Pu ± x Th ± U ± Bq ng Due to their similar chemical characteristics, Am and Cm can not be separated with this method. The trivalent Am has almost no retention on the CMP modified column, even if greater concentrations of N 3 - are present in the eluent. This is the reason that the Am peak is interfered if other radionuclides are present in the sample like for example fission products or the radioactive daughters of 229 Th. Therefore, for the quantitative experiments the 229 Th standard has been purified. Np, which has been stabilised in the pentavalent state by NaN 2, is not retained on the CMP modified column as has been expected, because the complexation of pentavalent actinides by CMP is very small. The influence of the sample matrix on the separation has been tested with simulated waste samples. A salt solution and a cement solution have been spiked with Am, Pu, Th and U. The separation method seems to be insensitive to high ionic strengths in the sample matrix and worked well with these simulated samples. 48

53 3.2.3 WP 3: New extracting agents Task 1: Design and preparation of new extracting agents New extracting agents should be designed and prepared for the selective extraction of radionuclides to analyse or to isolate radionuclides, which cause a high dose rate. A variety of ligands for complexing and separating long-lived actinide and fission products were synthesised in the course of the project mainly by UIBK and the subcontractor of FZJ. A variety of ligands for complexing and separating long-lived actinide and fission products were synthesized in the course of the project by the group of Dr R. Ludwig. Two separation tasks were concentrated on: (i) The selective complexation of Cs by a macrocyclic ligand which is anchored onto a polymer. The monomeric ligand, which is characterized by a high Cs/M-selectivity (M = Na, K, Mg, Ca), has been synthesized. (ii) The selective complexation of trivalent actinides in the presence of a high excess of concrete matrix components is the second task. Water-insoluble macrocyclic ligands which can be applied in extraction chromatography were synthesized for that purpose. They are based on the concept of co operation of mixed functional groups within the ligand and the use of relativistic effects for separation, which was previously established by the group of Dr R. Ludwig. The analysis of fission products or minor actinides contained in a complicated matrix such as concrete requires their separation in order to avoid matrix effects such as high background in counting or absorption. The chemical separation from small amounts of matrix preferably applies stationary methods such as columns. Therefore the design of the ligand was so as to allow physical or chemical immobilization on a carrier such as silica gel or an organic polymer. The design has to take into account that (i) a high selectivity and a good extraction are required, (ii) the selectivity must be retained when the ligand is immobilized, (iii) the chemical properties of the ligand do not deteriorate upon contact with radiation and mineral acids, (iv) the separation efficiency remains during several cycles (the ligand becomes not poisoned by metals), (v) the loaded isotopes can be eluted quantitatively under mild chemical conditions. Results for caesium Conventional methods: The complexation of Cs+ by conventional ligands is too weak to be applied in practice. Crown ethers can be used for complexation, but they have two disadvantages: Their selectivity is insufficient because their molecular structure is twodimensional and also flexible. That causes interfering ions such as K, Na, Mg and Ca to be complexed because these matrix components are in very high excess over Cs. ne can imagine their complexation such as the ligand can adjust its structure by tilting and subsequently accommodate and complex smaller ions. Since Cs is a large ion, the crown ethers are rather flexible and consequently their selectivity over smaller ions is limited. This limitation leaves the inappropriate for this separation task. Macrocyclic ligands based on calixarenes: Ludwig et al described previously the selective complexation of Cs by calixarenes [2] and apply this type of ligand for the separation task. The structure is shown in Figure 43 together with a three-dimensional view. The molecule consists of a calixarene backbone which holds a crown moiety. Both parts form a cavity which exactly matches the size of unhydrated Cs+. The flexibility of the crown unit is restricted because it is attached to the calixarene which itself cannot tilt in the plane of the crown. The phenyl groups are π-basic and contribute to the selectivity by interacting only with Cs. The ligand contains two H-groups for further derivatization by project partners. This derivatization will then allow incorporating the ligand into a polymer by means of ring-opening metathesis. In the course of these experiments it turned out to be important to control the ligand structure during this derivatization with 7-oxanorborn-2-enecarboxylic chloride: Without a template the ligand 49

54 structure can become fixed in a cone conformation. It appears necessary to add a template such as Cs-salt to fix an alternating conformation, because the monomeric ligand adopts this conformation when it binds Cs+, but a cone conformation when it binds Na+. H H H H Figure 43: Ligands for Cs Trivalent actinides Conventional methods: There are no satisfactorily separation methods when trivalent actinides have to be separated from a matrix which contains an excess of major actinides, FPs, alkali and alkaline earth salts. The most difficult separation in that matrix is the lanthanide/actinide separation. Macrocyclic ligands: In order to achieve a selectivity for Am3+ and Cm3+ over M+ (M=Na, K, Cs), M2+ (M= Mg, Ca, Sr, Fe), M3+ (M=Ln, Fe), Th(IV), Np(V), U(VI) and anions such as silicate, chloride and nitrate we apply calixarenes with the strategy as described recently by Ludwig et al [2-1]. A macrocyclic backbone of the calixarene type provides an enlarged cavity. As the ligand consists of six repeating phenyl rings it is called calix[6]arene in difference to the beforementioned calix[4]arene. We add several functional groups with donor atoms such as oxygen to this backbone, because the actinide ions have high coordination numbers. Due to their charge of = 3, we choose 3 acid groups. Thus the metal ion charge is neutralized and an excess of ionizable groups is avoided. Such an excess would reduce the selectivity as mono- and divalent cations can be complexed outside the ligand cavity. The donor atoms are arranged so as to match the geometry of the metal ion coordination sphere. The cavity is π-basic and allows the interaction with actinide ions in terms of back-donation. Due to the relativistic effect this interaction between molecular orbitals is larger for actinides than for lanthanides and thus contributes to the separation. The structure of the ligands synthesized is shown in Figure 44. Tertiary octyl groups were introduced in order to make the ligand more hydrophobic, although all of them are completely insoluble in water. 5

55 C C C H 2 H 2 H 2 CH 2 R CH 2 R C C H H C H 2 CH 2 C H C H 2 R C H 2 R = H, CH 3, C()NEt 2 C C C H 2 H 2 H 2 CX 2 H CX 2 H C H C H C H 2 CX 2 C H C H 2 H C H 2 X = F, Cl C C C H 2 H 2 H 2 CH 2 R CH 2 C C H H R C H 2 CH 2 C H C H 2 R C H 2 C C C H 2 H 2 H 2 CX 2 H CX 2 H C H C H C H 2 CX 2 C H C H 2 H C H 2 Figure 44: Ligands for actinide(iii) The main part of synthesizing new extracting agents was done by UIBK: The successful accomplishment of this task was a permanent process, since preliminary data from extraction experiments such as stability, selectivity, capacity, etc. were permanently incorporated into the design and modified synthesis of the ligands. This was particularly true for preliminary data on liquid-liquid (l-l) extraction of the metal ions of interest using monomeric calix[4]arene, calix[6]arene and terpyridine ligands. A. Ligands for Co extraction 1) Terpyridine-based ligands Two different monomers containing these systems were synthesized and fully characterized (Figure 45). N N N N N Figure 45: Structure of polymerizable terpy-1 and bipy-1 ligands based on ter- and bipyridines The polymerization chemistry of both monomers was studied. From a synthetic point of view, only the terpyridine ligand was accessible in suitable quantities (i.e. on a multi-gram scale). Polymerization results are summarized in Figure 46. As can be deduced there from, high molecular weight block copolymers form norborn-2-ene (NBE) and terpy-1 can be prepared using Mo-initiated RMP. Though major amounts of the Mo-based polymerization catalyst can be removed, residual amounts remain, yet do not interfere with the extraction of Co. 51

56 N Fc CMe 2 Ph N Terpy-1 N m n N N N Poly-Terpy-1- co-nbe catalyst: Mo(N-2,6-i-Pr 2 -C 6 H 3 )(CHCMe 2 Ph)(CMe(CF 3 ) 2 ) 2 M w = 117,1 Da block-size NBE = 89 block-size Terpy = 95 Figure 46: Block-copolymerisation of terpy-1 with norborn-2-ene (NBE) and physical properties of the resulting block-copolymer 2) ther ligands for Co extraction As an alternative to pyridine-based ligands, a polymeric amino-phenol-based system has been developed. A similar, non-polymer-based ligand was previously reported by other groups to exhibit selective extraction behavior for Co [11-12]. Though the material itself will not be investigated in course of this project, the synthesis of the sorbent shall be outlined briefly (Figure 47) since collaboration with FZJ exceeding the present project is planned. Br H Br [HCl] Br Br 1 % NaN 3 DMS N 3 N 3 1 % 1. LiAlH 4 2. NaH/H 2 /ethanol 3. HCl H H 3 N NH 3 5 % (BC) 2 /NaH H 2 /t-buh NaHC 3 H 2 N H Cl HCl (2 M in diethyl eher) BC HN NH BC BC NH HN BC 83 % NEt 3 /CH 2 Cl 2 82 % MEH H 3 N NH 3 95 % H H H H N N H H Br H H NaCNBH 3 NH 2 H NH HN H Br Br Br Br [Ru] H H NH HN H H n silica-supported ligand Br Br Figure 47: Synthesis of polymerisable ligand for the selective Co extraction 52

57 B. Ligands for Cs extraction A selective extraction of Cs as a group IA metal is generally hard to perform since it hardly forms any stable complexes. ne exception is the formation of mixed metal complexes using the Co(CN) 6 or the Fe(CN) 6 dianion [13-15]. Nevertheless, there exist some reports on the extraction by phenol-based resins (presumably via ion-exchange) [16]. Consequently, silicabased polymer-coated resins were also prepared from a phenol-containing monomer (Figure 48). Additionally, we also elaborated synthetic routes for the synthesis of polymerizable calix[4]arenes and calix[6]arenes. Polymerization data were obtained for both calix[n]arenebased monomers. A summary of the structures and the polymer data are also given in Figure 48. N H 5 H Calix-1 N 5 H H ligo-calix[6]-2 ligo-calix-1 catalyst: Cl 2 Ru(PCy 3 ) 2 (CHPh) M w = 12,4 Da PDI = 1.39 DP = 8 catalyst: Cl 2 Ru(PCy 3 ) 2 (CHPh) M w = 9, Da PDI = 1.54 DP = 12 Figure 48: Monomers (and the resulting polymers) for Cs extraction Task 2: Fixation of extracting agents A. Fixation of ligands for Co extraction According to the deliverables list, provision of a first set of sorbents was foreseen for month 9 of the project (i.e. June 21). Two materials for Cs-extraction as well as one material for Co extraction were sent to FZJ on 2 June 21. Sorbents based on these polymers were synthesized according to previously elaborated procedures. An illustration for the synthesis of the terpy-1-based material is given in Figure

58 R S I L I C A Si Fc Si Si Si m R = CMe 2 Ph n N N N T N N S I L I C A Si R Si R Si R R m CMe 2 Ph CMe 2 Ph x R CMe 2 Ph y R R CMe 2 Ph z N Figure 49: Coating procedure and structure of silica 6 coated with terpy-1-co-nbe Vinylated silica G-6 (2-4 mesh, 5 m 2 /g) was provided with a 7.5 % coating of the above-mentioned copolymer. The use of terpy-1-co-nbe block-copolymer provided enhanced chemical stability, in particular versus low or high ph. Calixarene-based supports were synthesized in a similar way. For the exact chemistry of such coating procedures see literature [17]. B. Fixation of ligands for Cs (I) extraction Two coated materials, a oligo-(calix[4]-1) and a oligo-(calix[4]-2)-coated silica (1 % coating) have been prepared. These have been used in standard SPE experiments. Task 3: Test of the extracting agents Cobalt extraction UV-Vis experiments Preliminary liquid-liquid (l-l) extraction data were carried out with the monomeric terpyridine ligand. The results are summarized in Table 33. As can be deduced from Table 33, the extraction ranges from ph 1 to 6 for most ions. Table 33: Preliminary l-l extraction data (H 2 -CH 2 Cl 2 ) for transition metal ions carried out with monomeric terpy-1 ligand Metal ion (nitrates) λ max [nm]* Metal ion (nitrates) λ max [nm]* Zn (II) 37, 319 Cu (II) 314,329 Cd (II) 37, 319 Ni (II) 295, 3 Hg (II) 311 Pd (II) 313, 327 Co (III) 36 Pt (IV) 315, 327 Rh (III) 313,327 Fe (III) 29, 3 Ir (III) 315, 327 Ru (III) 295 Mn (II) 37, 32 Re (II/III) 3 Ligand (ph = 6): 283, 289, 3 (sh); (ph = 1): 289, 3 (sh). All data were obtained at ph = 6. *First maximum at ca. 288 nm. 54

59 Due to the selectivity of the terpyridine ligand no main group elements were extracted. In addition, these data were not comparable to those expected for the supported catalyst. Thus, octahedral, bis-ligated complexes are formed in solution (M-L 2 ), in contrast, mono-ligated species had to be expected on the support. In order to retrieve information on the extraction behavior of the sorbent in the presence of several transition metals, these experiments with silica-based terpy-1 sorbent were carried out under competitive conditions. These investigations revealed high selectivity for Au, Ag, Pd, Pt, Hg, and Re at ph = < 1. A first set of data is shown in Figures Quantification was achieved by means of ICP-ES Pt Pd Rh Ir Au Ag µg Mn µg M n+ Figure 5: Extraction data for Pt, Pd, Rh, (Re), Ir, Au, Ag under competitive ph =.7 (left) ph = 3.5 (right). Values for Ag (ph = 6.9, left) are added from a single extraction experiment Re Pt Ir Rh Au Pd µg M n+ Zn Co Ni Cu Cd Hg µg M n+ Figure 51: Extraction data for Zn, Co, Ni, Cu, (Cd), (Ag), Hg under competitive ph =.3 (left) ph = 3.5 (right) Zn Ni Co Hg Cu Ag Fe Mn Re Ru Cr 75 5 Cd Cr Mn µg M n µg M n+ Figure 52: Extraction data for (Fe), Mn, (Re), (Cu), (Cd), Cr under competitive ph =.6 (left) ph = 3.5 (right) 55

60 In order to distinguish between those elements which are strongly bound to the sorbent (i.e. Au, Ag, Pd, Pt, Hg), a mix containing Au, Hg, and Pd was passed over the ph = 3.5 (Figure 53) Au Hg Pd µg M n+ Figure 53: Extraction data for Au, Hg, Pd under competitive ph = 3.5 As can be deduced from these data, the extraction efficiency at ph <1 decreases in the order Au > Hg > Pd. Since Au exists solely in form of AuCl3, Ag cannot be included into these investigations (formation of AgCl). Complementary, extraction was carried ph = 3.5 with single metal ion standards (Figure 54). Quantitative extraction of Ag, Hg, and Au was accomplished at this ph Ag Pd Hg Co Au µg Mn+ Figure 54: Extraction data (single metal ion standards) for Au, Hg, Pd, Co, and ph = 3.5 From these data, the following conclusion can be drawn: Poly(NBE-b-terpy-1)-coated silica is capable of extracting Co (II)/Co(III). At ph < 1, the material is selective for Au, Ag, Pd, Pt, Hg, and Re. At this ph, these metal ions can be removed. At ph < 1, the extraction order is Au > Hg > Pd. At ph 3.5, the material is selective for Au, Hg, and Ag. In case Au, Ag, Pd, Pt, Hg, and Re have been removed at ph < 1, Co can be extracted together with Zn, Cd, Ni, Cu, Cr, and Mn. For extraction purposes, the following materials have been prepared and sent to the corresponding partners: 56

61 Material Element Partner Calix-6 grafted silica Cs CIEMAT Hydroxyphenyl grafted silica Cs ENEA Thiocalix 4 grafted silica Cs CIEMAT Calix 4 grafted silica Cs FZJ Calix-4 grafted silica Cs FZJ Hydroxyphenyl grafted silica Cs FZJ Extraction experiments for 6 Co carried out via l-l extraction and SPE With monomeric terpy-1, an aqueous solution of Co-6 containing 2,5 cpm (l/l, H 2 /CH 2 Cl 2 ) resulted in 84 % extraction. In order to obtain a complete set of data, FZJ, CIEMAT as well as ENEA were provided with this type of sorbent. SPE-experiments with the sorbent were carried out at various phs (aqueous solution of 15 % methanol spiked with 6 Co) and are shown in Figure 55. The distribution ratio and the percentage of extraction was calculated after γ-measurement of the aqueous phase. The following formula was used for the calculation of the distribution ratio D: c cs Vs D = c m c = activity of 6 Co in aqueous phase before extraction c s = activity of 6 Co in aqueous phase after extraction V s = volume of aqueous phase m = mass of dry material As can be deduced from Figure 55, results obtained from l-l extraction were confirmed. In preliminary experiments, maximum extraction (> 9 %) was obtained within a ph rage of 2-5. Low extraction at ph > 6 resulted from hydroxide formation. Extraction experiments carried out at CIEMAT indicated that neither actinides nor lanthanides were extracted by this sorbent. In addition, in extraction experiments carried out at FZJ, Co was extracted with efficiencies > 98 % from an artificial mix onto the sorbent (see Table 34). This artificial mix was a simulated high active waste (HAW) solution and the various elements were measured by ICP- MS. The results are in good agreement with the results of the l-l extraction. s D [ml/g] ph Extraction in % D [ml/g] ph Extraction in % Figure 55: ph-dependant SPE of 6 Co on a poly(terpy-1-co-nbe)-coated silica; results of two different charges of the sorbent obtained by γ-measurement 57

62 Table 34: Extraction of various elements with a HAW simulate solution; ph = 1,96, D( 6 Co) = 192 ml/g; extraction of 6 Co = 99 % Ag Ba Cd Ce Cr D in ml/g 159,48 19,11 91,32,1,7 Extraction in % 95,22 7,49 91,94,14,83 Cs Cu Eu Fe Gd D in ml/g,16 675,43,8,58,91 Extraction in % 2,2 98,83,97 6,77 1,23 La Na Nd Ni Pd D in ml/g,1 7,99,16 4,26 337,75 Extraction in %,1 49,98 2,7 34,74 97,69 Pr Rb Rh Ru Sb D in ml/g,21,56,7 1,1 9,19 Extraction in % 2,75 6,53,89 12,4 53,45 Se Sm Sr Te Y D in ml/g 7,3,37,11 3,53,1 Extraction in % 47,7 4,46 1,4 3,64 1,25 ENEA obtained also good results for the Co extraction over a wide ph-range (see Figure 56). The measurements were made by ICP-ES. D [ml/g] ph Figure 56: ph-dependant SPE of cobalt on a poly(terpy-1-co-nbe)-coated silica; results obtained by ICP-ES Extraction% l-l extraction data for Cs using monomeric Calix[6]-2 for Cs-137 An aqueous solution containing 2 cpm of Cs-137 was subject to a l/l extraction using monomeric calix[6] disssolved in methylene chloride. With one single l-l extraction, 79 % of the initial amount of Cs-137 was extracted. This underlines the well-known selectivity of t-bucalix[6]areneesters in Cs-binding as described in lit. [18]. Additional results are presented by 58

63 FZJ. SPE of Cs-137 using sorbents based on Calix-1 and Calix[6]-2 These experiments have been carried out at FZJ as described above and are summarized in Figures D [ml/g] ph Extraction in % D [ml/g] Extraction in % D [ml/g] Extraction in % ph ph Figure 57: Extraction of 137 Cs onto poly(calix[4]-1)-coated silica (three different sorbents) D [ml/g] ph Extraction in % Figure 58: Extraction of 137 Cs onto poly(calix[6]-2)-coated silica As can be deduced there from, similar results were obtained with both sorbents. The optimum extraction ph was around 8 (> 65 % extraction). In view of the comparably low extraction data for calix[4]-1, polymerization must change the geometric situation of the ligand in a way that Cs can be incorporated. In case similar extraction data were obtained with high 137 Cs loadings for both supports, this means that template-assisted syntheses was not necessary at all. 59

64 SPE data for Cs using polymer phenol ligand Unfortunately, only low extraction efficiencies were observed at FZJ (Figure 59). D [ml/g] ph Extraction in % Figure 59: Extraction of 137 Cs onto polymer phenol ligand Additionally, CIEMAT tested extracting agents for actinides. In order to improve the selectivity and the capacity for separation of actinides by liquid-liquid extraction or extraction chromatography it has been studied new products or mixtures. In the case of extraction chromatographic the experimental conditions for the separation of these elements are usually studied for the mobile phase and they are seldom studied for the stationary phase. The products studied have been: octyl(phenyl)-n,n-diisobutylcarbamoylmethyphosphine oxide (CMP), N,N -dimethyl-n,n -dioctylhexylethoxy malonamide (DMDHEMA), bis(2,4,4-trimethylpentyl) dithiophosphinic acid (Cyanex-31) and a mixture of four trialkyl phosphine oxides (Cyanex-923). Until now, these products have been used for the separation of actinides from fission product in the nuclear fuel reprocessing. Data about actinides extraction by these compounds have been obtained by liquid-liquid extraction. Moreover, synergistic solvent extraction systems have been studied. Extraction experiments were performed using equal volumes of organic and aqueous phases. After equilibration and centrifugation, each phase was sampled and analysed. Firstly, it has been carried out Pu, Am and U extractions from nitric or hydrochloric acid solutions by.2m CMP using dodecane as diluent. The effectiveness of CMP as a neutral extractant for the trivalent actinides and lanthanides from strong acid media has been well established, and also, it would be interesting to study CMP as synergistic effect in combination with TBP or HDEHP to separate actinides from acidic solutions. Another extraction have been carried out from aqueous media which result from the barium sulfate dissolution. The dissolution of this precipitate by a sodium carbonate solution produces two fractions: a solution of Na 2 C 3, containing S -2 4 ions, and a precipitate of BaC 3, which is dissolved by a little amount of HN 3 or HCl. The sodium carbonate solution is neutralized by nitric acid and it is produced a 1.8M NaN 3 -HN 3 (ph = 2) solution or, if more acid is added, a 1.5M NaN 3-1M HN 3 solution is got. In both fractions it is necessary to separate Pu, Am and U from each other and some elements that could interfere in the determination by alpha-spectrometry. There is an increase in U extraction (Figure 6) with an increase in nitric acid concentration. However, when sodium nitrate is present, the behaviour is the opposite. In the case of U 6

65 extraction from hydrochloric acid solutions, it decreases as hydrochloric acid concentration increases, being the highest extraction, about 1 %, at 1N hydrochloric acid. The best results so far have been achieved with the synergistic combination of TBP or HDEHP and CMP. The effect is increasing the U extraction at any media studied, being U extraction higher than 96 %. U-CMP U-CMP-TBP U-CMP-HDEHP % Element Extracted M NaN3 -ph=2 1M HN3-1.5NaN3 1M HN3 7M HN3 1N HCl 7N HCl 11N HCl Figure 6: U extraction by.2m CMP,.2M CMP-1.2M TBP, and.2m CMP- 1.2M HDEHP in dodecane from several aqueous solutions Pu extraction behaviour (Figure 61) is similar to U for all the conditions studied. Also for Am (Figure 62) except that Am is not extracted at 1N HCl. Pu-CMP Pu-CMP-TBP Pu-CMP-HDEHP % Element Extracted M NaN3 -ph=2 1M HN3-1.5NaN3 1M HN3 7M HN3 1N HCl 7N HCl 11N HCl Figure 61: Pu extraction by.2m CMP,.2M CMP-1.2M TBP, and.2m CMP- 1.2M HDEHP in dodecane from several aqueous solutions 61

66 Am-CMP Am-CMP-TBP Am-CMP-HDEHP % Element Extracted M NaN3 - ph=2 1M HN3-1.5NaN3 1M HN3 7M HN3 1N HCl 7N HCl 11N HCl Figure 62: Am extraction by.2m CMP,.2M CMP-1.2M TBP, and.2m CMP- 1.2M HDEHP in dodecane from several aqueous solutions Cyanex 923 is a mixture of four trialkyl phosphine oxides and has the advantage of being a liquid and is completely miscible with all commonly used hydrocarbons. Cyanex 923 has been suggested as a potential extractant for the extraction and separation of yttrium from trivalent lanthanides and for the separation of iron ant titanium from waste chloride liquors of the titanium minerals processing industry. The highest extraction is for Pu (Figure 63), more than 65 % is extracted in any condition studied. U is extracted about 1 % from low hydrochloric acid concentration or any nitric acid concentration, except at 1.5M NaN 3-1M HN 3 where it is not extracted at all. The best conditions for Am extraction are 1.8M NaN 3 (ph = 2) where Am is extracted about 1 %, while it is not extracted from any hydrochloric acid solutions. % Element extracted M NaN3 -ph=2 U-Cyanex-923 Pu-Cyanex-923 Am-Cyanex-923 1M HN3-1.5NaN3 1M HN3 7M HN3 1N HCl 7N HCl 11N HCl Figure 63: U, Pu and Am extraction by.15m Cyanex-923 in dodecane from several aqueous solutions Cyanex 31 has been used for lanthanide/actinide separations in the nuclear fuel cycle. It contains a thio derivative of a diakylphosphonic acid which is susceptible to oxidation and decomposition at high nitric acid concentration, so a concentration higher than 2.M must be used to avoid. 62

67 More than 97 % of U (Figure 64) is extracted from any of the conditions studied. Am is only extracted at ph > 2 and Pu extraction also increases with decreasing acid concentration. U-Cyanex-31 Pu-Cyanex-31 Am-Cyanex-31 %Element Extracted M HN3 1N HCl ph=1 ph=2 ph=3 Figure 64: U, Pu and Am extraction by.5m Cyanex-31 in toluene from several aqueous solutions Diamide extractants, like DMDHEMA, have been investigated for the application to the chemical treatment of nuclear wastes. The behaviour of U and Pu in the extraction from nitric and hydrochloric acid solutions, by.5m DMDHEMA in dodecane, is very similar (Figure 65). The extraction percentages are higher than 8 %, except for Pu extractions from both 1N HCl and 1.8M NaN 3 (ph = 2) solutions. The best conditions for Am extraction are from 7M HN 3 ; it is about 98 %. These conditions are also suitable for the extractions of Pu and U. U-DMDHEMA Pu-DMDHEMA Am-DMDHEMA % Element Extracted M NaN3 -ph=2 1M HN3-1.5NaN3 1M HN3 7M HN3 1N HCl 7N HCl 11N HCl Figure 65: U, Pu and Am extraction by.5m Cyanex-31 in dodecane from several aqueous solutions 63

68 3.2.4 WP 4: Automated separation methods Task 1: Solid phase extraction Solid phase extraction with commercial available resins (Eichrom) is a method which can be automised for the separation of radionuclides. In the Fourth Framework Programme (project Qualification of radio-analytical routines for the determination of alpha-emitting nuclides in conditioned radioactive waste) it could be shown that with solid phase extraction all different kinds of actinides can be separated. For this project a laboratory robot system ASPEC XL (Gilson) was bought and was first optimised to our demands. That means that new racks and collect zones were developed. The collect zone can now handle 2 ml Packard vials. The original collect zones included 1 ml tubes, which is for some separations too little. The modified laboratory robot system is shown in Figure 66. Figure 66: Modified laboratory system ASPEC XL The first programmes were written for the separation of uranium, plutonium and americium on a TRU resin column. These tests were successfully, the same good results and recovery rates were obtained compared with hand made separations. Problems occurred for the separation of neptunium and thorium with TEVA resin. No neptunium was found in the neptunium fraction. Nearly 1 % of the neptunium was found in the thorium fraction. This is in contrary to the hand made separations. We could find out that the stainless steel needle that transfers the minerals acids to the column caused this problem. For the separation of neptunium and thorium a 9 M HCl is used which let the needle corrode. Fe(III) is transferred to the column and influences the oxidation state of neptunium. Neptunium was reduced with Fe(II) to Np(IV) at the beginning of the separation. TEVA can only extract tetravalent actinides. Fe(III) now oxidises Np(IV) to Np(V) which is not extracted by the resin and causes therefore an early breakthrough of the neptunium. We manufactured titanium needles for that reason. This kind of needle should not be corroded by HCl and has been tested. The separation of thorium and neptunium is now possible with the robot system. But it has to be mentioned that the titanium needle breaks very easily what happened during several test runs. The material properties of the stainless steel needle are in this case better for the use in a robot system. Therefore we coated a stainless steel needle with MeltiLex a low molecular weight aromatic polymer and an 64

69 ethylene vinyl acetate copolymer together with paraffin. It was only possible to coat the inner surface of the needle. There was no corrosion protection on the outer surface. As expected the results showed that the separation of neptunium and thorium was still influenced by Fe(III) coming from corrosion of the outer surface. It was not possible to buy fully coated needle for our experiments. Task 2: HPLC ENEA assembled an ionic exchange chromatograph in a fume hood. This chromatograph equipped with an autosampler, a gradient pump column section, an electrochemical detector, a spectrophotometric detector and a liquid scintillation counter, coupled on-line with the whole HPIC system. All modules are combined and controlled by a central computer with dedicated software. This is comparable to the system of FZJ and both systems can be seen as automated systems WP 5: Automated measurements Task 1: nline measurements As already mentioned in the previous work packages both partners ENEA and FZJ coupled their chromatographic system with the liquid scintillation counter. A separation and measurement system completely automatic is available. Such systems allow the determination of the total concentration of a nuclide (in mg/l), through the electrochemicalspectrophotometric detector, and the activity of his radioactive isotope (in Cps) through the liquid scintillation counter, with a single analysis WP 6: Statistical estimations for total waste packages Task 1: Development of tools for the description of statistical distribution of inhomogeneous samples The destructive analysis of radioactive waste packages is restricted to small aliquots of the waste package. Therefore it is difficult to estimate the total activity of such a waste package due to a possible inhomogeneity. The aim of this task is the development of a statistical method to estimate the total activity of a waste package from destructive analysis of small volume elements of the waste package. The first attempt is the development of models for the distribution of the content of a waste package. Therefore the waste package will be divided into small volume elements and each volume element will be filled with data for the activity. Three different types of distribution for the activity have been implemented in the model yet. Homogenous filling Random filling (Figure 67) Gradient filling from a central point (Figure 62). 65

70 Figure 67: Gradient filling Figure 68: Random filling The first case of a total homogenous filling has not been considered for the further work because it is obvious, that this will always lead to 1 % recognition of the inventory. The two other distribution models are written in C++ and have been implemented on a central AIX UNIX computer. The programme listings are added at the end of this chapter. The gradient filling programme Zentren requires the following input parameter: 1. Diameter of the waste package in cm 2. Height of the waste package in cm 3. X-position of the activity spot 4. Y-position of the activity spot 5. Z- position of the activity spot 6. Gradient G z of the activity in % 7. Input of further activity spots (1 = yes, any other key = no). The programme repeats the steps 3 to 7 until the input is not equal to 1. This allows the input of an unlimited number of activity spots. 8. Flag, whether the results should be saved (1 = yes) 9. Name of the output file for the results The activity (a) of the central spot will be assumed as 1 % and the activity of the other volume elements is calculated by a = d * g a: activity, d: distance from the spot, g: gradient 66