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The extraction of fission product molybdenum-99 by dithiol precipitation
Experimeml
Int. J. Appl. Radiat. lsot. Vol. 36, No. 1, pp. 85-86, 1985 ~) Pergamon Press Ltd 1985. Printed in Great Britain. 0020-708X/85 $3.00 + 0.00
Reagents Dithiol. One gram of dithiol is dissolved in 200 mL of deaerated 1% KOH solution to which is added 2 mL of thioglycolic acid. (" The reagent is stored in a glass bottle under refrigeration. Filter a/d. The aid is prepared by pulping glass fibre filter pads in I tool L - ' HNO3. All rengents are of A.R. grade.
The Extraction of Fission Product Molybdenum99 by Dithiol Precipitation N. BLAGOJEVIC, R. E. BOYD and E. L. R. HETHERINGTON Radioactive Products Research Section, Australian Atomic Energy Commission, Research Establishment, Lucas Heights Research Laboratories, Private Mail Bag, Sutherlaud, N.S.W. 2232, Australia
Target preparation and dissolution of UO2 The standard target consists of 1.8% 23~U enriched uranium oxide in the form of pressed and sintered pellets each of 2.5 g. Irradiations were carried out at Lucas Heights in the A.A.E.C.'s materials testing reactor HIFAR; each target was irradiated for a nominal seven days in a neutron flux of 7 to 9 x 10t~n cm-'- s -t to yield 2500GBq ofgVMo with a specific activity of 1.85 x 10SGBqg -t after one day of cooling. (s~ The target material was dissolved in approximately 60 mL of 8 mol L - i HNO3 and made up to a volume so that the uranium content was 260 + 50 mg m L - ' and the acid concentration was approximately 1.5 tool L -s .
(Received 25 June 1984)
A method has been developed for the preliminary extraction of fission product molybdenum-99 from irradiated UOz. The method uses toluene-3,4-dithiol to precipitate carrierfree molybdenum from fission products dissolved in HNO3. The extraction efficiency is greater than 90°/0 and the radionuclidic purity is approximately 970/0.
Recommended procedure To 25 mL of fission product solution was added 0.5 mL of n-butanol. The solution was stirred vigorously until the alcohol had dissolved. Seven millilitres of dithiol reagent was then added. The resulting green molybdenum complex was filtered through a fast glass fibre filter on which 2 to 3 mm of filter aid had been deposited. The precipitate was washed with 1.5molL -t HNO~ to remove fission product traces and a small amount of air was then drawn through to remove excess liquid. The molybdenum dithiol was recovered by dissolving it in approximately 20 mL of acetone.
Introduction Molybdenum-99/technetium-99m generators manufactured from ~ M o produced in the fission of 23sU are used extensively in nuclear medicine. Uranium oxide targets, often enriched in m U, are irradiated for several days in the highest available neutron flux to yield 6.1~o **Mo. The separation of ~*Mo from highly radioactive irradiated uranium must be done in shielded hot cell facilities and its comparatively short half-life (66 h) precludes the use of protracted chemical cycles. Consequently, the extraction method must combine efficiency, specificity, speed and simplicity. A recent reviev/t~ of extraction methods indicates that most workers use tither solvent extraction or column chromatography for the preliminary separation. The majority of preliminary separations are carried out from hydrochloric or sulphuric acid solutions containing low concentrations of uranium, (2-~) using molybdenum carrier between 10 and 100 mg to aid the extraction. Solvent extraction is a difficult and complex process to perform under remote handling conditions. Although the technique can yield a pure product, a considerable quantity of radioactive aqueous and organic waste is generated. Column chromatography is a simpler technique but requires careful regulation of the column size, solution fiowrate and the number and the type of column washes. Column chromatography is used at Lucas Heights for routine WMo production. (" Experience has shown that despite all precautions, product losses and variable product quality can still occur. This note describes a preliminary extraction method suitable for routine large-scale production of WMo; it is selective, fast and compatible with waste management procedures. The method is based on the precipitation reaction between molybdenum (VI) and toluene-3, 4-dithiol (dithiol). The insoluble molybdenum complex is separated from the other fission products by filtration and then recovered by dissolution in acetone.
Results and Disemsion A number of chromophoric and precipitating chelating agents were assessed, but the majority were eliminated because of poor selectivity or sensitivity. From a literature survey it was found that chelates with sulphur as donor atoms were most selective, and of the reagents available dithiol was found to be the most suitable. Molybdenum forms a number of complexes with dithiol according to the level of acid concentration. The molybdenum (VI) species is the most dominant form if the acid concentration is above I tool L - i. Because of the reducing power provided by the two thiol groups, formation of the complex is possible even in nitric acid solutions. Once formed, the complex can be separated from the other fission products and uranium by either solvent extraction or filtration. The filtration process is the least difficult to perform by remote manipulation and has the following advantages: (i) It is fast and efficient. (ii) The precipitate is stable even if the acid concentration is increased above 6 tool L -t . (iii) Separation of ~ M o from other fission products and uranium is achieved in a single extraction step. (iv) Liquid waste is reduced to a minimum. (v) The product radionuclidic purity is better than that obtained with solvent extraction and most other preliminary extraction methods.(2-~)
The extraction efficiency is approximately 90% for the acid concentration above 1.5mol L -t (Fig. 1). The radio85
86
Technical Note Table I. Radionuclidicpurityof f.p.~ M o extractedby toluene-3,4-dithiol, Run No 9*Mo U2Te :'TSb taRa Othe# A 97.5 1.2 0.8 0.5 ND B 98.0 1.1 0.6 0.3 ND C 97.1 1.5 1.0 0.4 ND D 97.5 l.l 0.6 0.9 ND E
95.6
1.4
1.9
I. 1
ND
F 95.6 1.1 2.3 1.0 ND G 97.2 0.9 1.0 0.9 ND Mean + SD 975:1 1.2+0.2 1.2 +0.7 0.7+0.3 (A-D) Product obtained from urany! nitrate/l moi L "l H2504 solution. (E--G) Product obtained from urany[ nitratc/[.5 mol L -~ HNO; solution. 'Other nuclides not detected (ND).
100 [- *
~.
Conclusions
l-
The preliminary extraction of **Mo using dithiol chelate as the precipitating agent is a promising procedure that can be easily adapted to the large-scale production of ~'Mo. Smaller scale experiments indicate that the process is a simple time-saving extraction method that is fast, efficient and reliable.
References eo 30
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I
2
I
~
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l
3 4 5 HNO3{m~ 1)
. I
6
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Fig. l. Exmmfion efficiency of ~ M o vs acid concentration for solvent extraction and precipitation processes nuclidic purity of 9*Mo is 9 7 % + 1% (Table 1). No ,, or/~ contamination was detected and no long-lived 7 emitters other than '°3Ru and l~Sb were found.
1. Boyd R. E. Radiochem. Acta 30, 123 (1982). 2. Sivaramakrishnan C. K., Jadhav A. V., Raghuraman K., Raman S., Nair P.S. and Ramaniah M. V. BARC847 (1976). 3. Tanase M. and Motojima K. Int. J. Appl. Radiat. Isot. 32, 353 (1981). 4. Tanase M., Kase T. and Shikata E. J. NucL Sci. TeclmoL 13, 591 (1976). 5. Abroad. S., Mannan A., Ahmad I. and Qureshi I. H. Int. J. Appl. Radiat. Isot. 33, 469 (1982). 6. E1-Bayoumy S. and El-Kolaly M. Radioanal. Chem. 68, 7 (1982). 7. Hobart E. W. and Hurley E. P. Anal. Chim. Acta 27, 144 (1962). 8. Barnes R. K., Hetherington E. L. R. and Ohkubo T. Int. J. Appl. Radial. lsot. 34, 603 (1983).
Int. J. Appl. Radiat. lsot. Vol. 36, No. 1, pp. 85-86, 1985 ~) Pergamon Press Ltd 1985. Printed in Great Britain. 0020-708X/85 $3.00 + 0.00
Reagents Dithiol. One gram of dithiol is dissolved in 200 mL of deaerated 1% KOH solution to which is added 2 mL of thioglycolic acid. (" The reagent is stored in a glass bottle under refrigeration. Filter a/d. The aid is prepared by pulping glass fibre filter pads in I tool L - ' HNO3. All rengents are of A.R. grade.
The Extraction of Fission Product Molybdenum99 by Dithiol Precipitation N. BLAGOJEVIC, R. E. BOYD and E. L. R. HETHERINGTON Radioactive Products Research Section, Australian Atomic Energy Commission, Research Establishment, Lucas Heights Research Laboratories, Private Mail Bag, Sutherlaud, N.S.W. 2232, Australia
Target preparation and dissolution of UO2 The standard target consists of 1.8% 23~U enriched uranium oxide in the form of pressed and sintered pellets each of 2.5 g. Irradiations were carried out at Lucas Heights in the A.A.E.C.'s materials testing reactor HIFAR; each target was irradiated for a nominal seven days in a neutron flux of 7 to 9 x 10t~n cm-'- s -t to yield 2500GBq ofgVMo with a specific activity of 1.85 x 10SGBqg -t after one day of cooling. (s~ The target material was dissolved in approximately 60 mL of 8 mol L - i HNO3 and made up to a volume so that the uranium content was 260 + 50 mg m L - ' and the acid concentration was approximately 1.5 tool L -s .
(Received 25 June 1984)
A method has been developed for the preliminary extraction of fission product molybdenum-99 from irradiated UOz. The method uses toluene-3,4-dithiol to precipitate carrierfree molybdenum from fission products dissolved in HNO3. The extraction efficiency is greater than 90°/0 and the radionuclidic purity is approximately 970/0.
Recommended procedure To 25 mL of fission product solution was added 0.5 mL of n-butanol. The solution was stirred vigorously until the alcohol had dissolved. Seven millilitres of dithiol reagent was then added. The resulting green molybdenum complex was filtered through a fast glass fibre filter on which 2 to 3 mm of filter aid had been deposited. The precipitate was washed with 1.5molL -t HNO~ to remove fission product traces and a small amount of air was then drawn through to remove excess liquid. The molybdenum dithiol was recovered by dissolving it in approximately 20 mL of acetone.
Introduction Molybdenum-99/technetium-99m generators manufactured from ~ M o produced in the fission of 23sU are used extensively in nuclear medicine. Uranium oxide targets, often enriched in m U, are irradiated for several days in the highest available neutron flux to yield 6.1~o **Mo. The separation of ~*Mo from highly radioactive irradiated uranium must be done in shielded hot cell facilities and its comparatively short half-life (66 h) precludes the use of protracted chemical cycles. Consequently, the extraction method must combine efficiency, specificity, speed and simplicity. A recent reviev/t~ of extraction methods indicates that most workers use tither solvent extraction or column chromatography for the preliminary separation. The majority of preliminary separations are carried out from hydrochloric or sulphuric acid solutions containing low concentrations of uranium, (2-~) using molybdenum carrier between 10 and 100 mg to aid the extraction. Solvent extraction is a difficult and complex process to perform under remote handling conditions. Although the technique can yield a pure product, a considerable quantity of radioactive aqueous and organic waste is generated. Column chromatography is a simpler technique but requires careful regulation of the column size, solution fiowrate and the number and the type of column washes. Column chromatography is used at Lucas Heights for routine WMo production. (" Experience has shown that despite all precautions, product losses and variable product quality can still occur. This note describes a preliminary extraction method suitable for routine large-scale production of WMo; it is selective, fast and compatible with waste management procedures. The method is based on the precipitation reaction between molybdenum (VI) and toluene-3, 4-dithiol (dithiol). The insoluble molybdenum complex is separated from the other fission products by filtration and then recovered by dissolution in acetone.
Results and Disemsion A number of chromophoric and precipitating chelating agents were assessed, but the majority were eliminated because of poor selectivity or sensitivity. From a literature survey it was found that chelates with sulphur as donor atoms were most selective, and of the reagents available dithiol was found to be the most suitable. Molybdenum forms a number of complexes with dithiol according to the level of acid concentration. The molybdenum (VI) species is the most dominant form if the acid concentration is above I tool L - i. Because of the reducing power provided by the two thiol groups, formation of the complex is possible even in nitric acid solutions. Once formed, the complex can be separated from the other fission products and uranium by either solvent extraction or filtration. The filtration process is the least difficult to perform by remote manipulation and has the following advantages: (i) It is fast and efficient. (ii) The precipitate is stable even if the acid concentration is increased above 6 tool L -t . (iii) Separation of ~ M o from other fission products and uranium is achieved in a single extraction step. (iv) Liquid waste is reduced to a minimum. (v) The product radionuclidic purity is better than that obtained with solvent extraction and most other preliminary extraction methods.(2-~)
The extraction efficiency is approximately 90% for the acid concentration above 1.5mol L -t (Fig. 1). The radio85
86
Technical Note Table I. Radionuclidicpurityof f.p.~ M o extractedby toluene-3,4-dithiol, Run No 9*Mo U2Te :'TSb taRa Othe# A 97.5 1.2 0.8 0.5 ND B 98.0 1.1 0.6 0.3 ND C 97.1 1.5 1.0 0.4 ND D 97.5 l.l 0.6 0.9 ND E
95.6
1.4
1.9
I. 1
ND
F 95.6 1.1 2.3 1.0 ND G 97.2 0.9 1.0 0.9 ND Mean + SD 975:1 1.2+0.2 1.2 +0.7 0.7+0.3 (A-D) Product obtained from urany! nitrate/l moi L "l H2504 solution. (E--G) Product obtained from urany[ nitratc/[.5 mol L -~ HNO; solution. 'Other nuclides not detected (ND).
100 [- *
~.
Conclusions
l-
The preliminary extraction of **Mo using dithiol chelate as the precipitating agent is a promising procedure that can be easily adapted to the large-scale production of ~'Mo. Smaller scale experiments indicate that the process is a simple time-saving extraction method that is fast, efficient and reliable.
References eo 30
• =~ecIDitClhOn
20
* $obvent exm3¢hon
10 0
I
1
I
2
I
~
.
I
l
3 4 5 HNO3{m~ 1)
. I
6
I *
?
~
8
Fig. l. Exmmfion efficiency of ~ M o vs acid concentration for solvent extraction and precipitation processes nuclidic purity of 9*Mo is 9 7 % + 1% (Table 1). No ,, or/~ contamination was detected and no long-lived 7 emitters other than '°3Ru and l~Sb were found.
1. Boyd R. E. Radiochem. Acta 30, 123 (1982). 2. Sivaramakrishnan C. K., Jadhav A. V., Raghuraman K., Raman S., Nair P.S. and Ramaniah M. V. BARC847 (1976). 3. Tanase M. and Motojima K. Int. J. Appl. Radiat. Isot. 32, 353 (1981). 4. Tanase M., Kase T. and Shikata E. J. NucL Sci. TeclmoL 13, 591 (1976). 5. Abroad. S., Mannan A., Ahmad I. and Qureshi I. H. Int. J. Appl. Radiat. Isot. 33, 469 (1982). 6. E1-Bayoumy S. and El-Kolaly M. Radioanal. Chem. 68, 7 (1982). 7. Hobart E. W. and Hurley E. P. Anal. Chim. Acta 27, 144 (1962). 8. Barnes R. K., Hetherington E. L. R. and Ohkubo T. Int. J. Appl. Radial. lsot. 34, 603 (1983).