Separation of tungsten from LEU fission produced 99mo solution to improve technological performance in both the processes of 99mo and 99mtc generator production

Separation of tungsten from LEU fissionproduced 99Mo solution to improve technological performance in both the processes of 99Mo and 99mTc generator production Van So Le, Cong Duc Nguyen & Minh Khoi Le Journal of Radioanalytical and Nuclear Chemistry An International Journal Dealing with All Aspects and Applications of Nuclear Chemistry ISSN 0236-5731 J Radioanal Nucl Chem DOI 10.1007/s10967-014-3426-1 23 Your article is protected by copyright and all rights are held exclusively by Akadémiai Kiadó, Budapest, Hungary This e-offprint is for personal use only and shall not be selfarchived in electronic repositories If you wish to self-archive your article, please use the accepted manuscript version for posting on your own website You may further deposit the accepted manuscript version in any repository, provided it is only made publicly available 12 months after official publication or later and provided acknowledgement is given to the original source of publication and a link is inserted to the published article on Springer's website The link must be accompanied by the following text: "The final publication is available at link.springer.com” 23 Author's personal copy J Radioanal Nucl Chem DOI 10.1007/s10967-014-3426-1 Separation of tungsten from LEU fission-produced 99Mo solution to improve technological performance in both the processes of 99Mo and 99mTc generator production Van So Le • Cong Duc Nguyen • Minh Khoi Le Received: August 2014 Ó Akade´miai Kiado´, Budapest, Hungary 2014 Abstract Method of W separation from fission-99Mo solution was studied using alumina column and H2SO4/ HNO3 eluents The distribution coefficients of WO42- and MoO42- ions and the column loading/eluting conditions were investigated to optimize the separation process 4–6 M H2SO4 solutions were successfully used to elute/separate 99 MoO42- ions from the alumina column which strongly retained WO42- ions without significant W-breakthrough The developed W/99Mo separation process is fit for being inline incorporated/integrated in the alkaline dissolution-based process of fission-99Mo recovery currently used in LEU target-based 99Mo production Keywords LEU/HEU-target Á 99Mo-production Á W-separation Á 99mTc Á Alumina Introduction 99 Mo is a parent nuclide of the 99mTc radioisotope generator which is used in nuclear medicine world-wide Today the commercial production of 99Mo is mainly based on the 235 U fission and 98Mo (n, c) 99Mo reactions [1] The targets used in 235U-fission reaction are highly enriched uranium (HEU) and/or low enriched uranium (LEU) The mass of LEU target is around five times larger than that of HEU Conversion of current technology using HEU targets ([20 % 235U) to those using LEU (\20 % 235U) requires a V S Le (&) Á M K Le Medisotec, Gymea, NSW, Australia e-mail: vansole01@gmail.com C D Nguyen ChoRay Hospital, Ho Chi Minh, Vietnam 5–6 fold increase in total uranium content to produce irradiation yields of 99Mo equivalent to current HEU targets Consequently, the W contaminant content of 99Mo solution produced from LEU targets is possibly given in approximately five times larger compared to that produced from HEU target, assuming the same purity of the both types of the targets So, among different technical solutions involved during the conversion of existing facilities using HEU targets, the reduction of W contamination in LEU target and/or the separation of W from 99Mo solution of LEU process should be addressed The large amount of W contaminant is seriously challenging the performance of the 99 Mo and 99mTc generator production technologies Particularly, the W contaminant in LEU may cause a serious problem regarding the decrease of 99Mo-retaining capacity of the ion exchange resin/sorbent columns which are currently used in both the processes of 99Mo stock solution and 99m Tc generator production This fact is due to the large amount of W element contaminant coming from a large amount of uranium and aluminium metal cladding used in manufacturing LEU targets Moreover, the high similarity in adsorption property of WO42- and MoO42- ions on the currently used sorbents is also the reason causing the reduction in adsorption capacity of MoO42- ions The above mentioned situation is also challenging the 99 Mo production using natural or 98Mo-enriched molybdenum targets Natural molybdenum targets used in 98Mo (n, c) 99Mo reactions usually contain much more W contaminant than LEU and 98Mo-enriched Mo targets So the natural Mo targets purified to reduce the W content to an extent of less than 10 ppm are needed to produce a medically useful 99mTc solution via 98Mo (n, c) 99Mo reactions The radioactive isotopes 181W (t1/2 = 121.2 days), 185 W (t1/2 = 74.8 days), 187W (t1/2 = 23.9 h), and 188W (t1/2 = 69.4 days), which are induced from neutron 123 Author's personal copy J Radioanal Nucl Chem Table Content of the elements in the simulator fission-99Mo solution (The data is based on the ICP-EOS analysis results of the 99Mo solution eluted from AG1 8/AG MP resin column which was loaded with an alkaline digestion-solution of CERCA LEU-target plates used at ANSTO’s 99Mo-production facility.) Element Concentration (ppm) Element Concentration (ppm) Element Concentration (ppm) Si 530 Cu 52 B 67 P 43 Zn 519 Ge Mn Cr Sb Ca 1098 Al 2634 Zr Fe 183 W 13 Na 114854 Mn 368 Mo 74 Zr K 916 Ni 12 Sb S 75299 U Fig Weight distribution coefficient of a WO42- and b) MoO42ions versus pure HNO3 solution acidity for alumina performed effectively Particularly, the natural W contains a high content of 186W which will be activated by reactor neutron during the target irradiation to produce 188W via 186 W (n, n, c) 188W reaction Consecutively, the produced 188 W will generate the radioactive 188Re (t1/2 = 16.94 h) via beta particle decay and this 188Re-radionuclide will be co-extracted with 99mTc from the 99Mo/99mTc generator and present as a radioactive contaminant in the finished product of 99mTc solution Different methods have been developed using activated charcoal and/or hydrous tin-dioxide sorbents to remove W from a macro quantity molybdate solution Some of them were described in the protocol of natural molybdenum target preparation for (n, c) reaction-based 99Mo production [2, 3] However, the developed methods not suite the separation of milligram quantities of W from the solution of a comparable amount of Mo (99Mo), which are currently processed in the LEU target-based 99Mo production plant The above mentioned methods are neither suitable to a process of fission-99Mo production in which the solutions of small volume are usually used Obviously, a suitable and technologically compatible (in-line) W/99Mo separation method should be developed to substantially overcome the disadvantages of above mentioned ones Experimental Fig Weight distribution coefficient of a WO42- and b MoO42ions versus pure H2SO4 solution acidity for alumina activation of natural abundance W element, will present in the finished 99Mo stock solution and then will be retained by the alumina column along with 99Mo in the 99Mo/99mTc generator as the radioactive contaminants, if the separation of W from the target or target solution could not be 123 A 99Mo/188W-spiked simulator of the fission-99Mo solution was used for W/Mo separation method development The simulator fission-99Mo solution contains 0.0740 mg Mo/ mL, 0.013 mg W/mL, and several given contaminant ions 99 MoO42- and 188WO42- solutions used in all experiments were prepared by eluting the used 99Mo/99mTc- and 188 W/188Re-generators, respectively, with 1.0 M NaOH solution followed by neutralization with hydrochloric acid All chemicals used were analytical grade Acidic alumina of Brockmann supplied by Sigma-Aldrich was used for Author's personal copy J Radioanal Nucl Chem Fig Effect of sorbent weight/solution volume ratio on molybdate and tungstate ions uptake from 1.0 mL sulphate solution (Molybdate and tungstate concentration are 0.074 mg Mo/mL and 0.013 mg W/mL, respectively): a W-uptake on the alumina of weights from 33.7 to 100.0 mg per mL solution; b Mo-uptake on 100.0 mg alumina per mL solution; c Mo-uptake on 67.5 mg alumina per mL solution; d Mo-uptake on 60.0 mg alumina per mL solution; e Mo-uptake on 33.7 mg alumina per mL solution; f Distribution coefficient (Kd) values of molybdate ions for alumina sorbent in the molybdate solution of variable sulphate concentration (Kd measurement was performed in the simulator fission-99Mo solution of WO42- and MoO42- ions, the element concentration of which is reported in Table 1) chromatographic column packing The element/contaminant composition of the simulator solution shown in Table is mimicked based on the ICP-EOS analysis results of a real 99Mo solution sample which was obtained from chemical processing steps in which an alkaline dissolution-processed 99Mo solution of LEU (20 % 235U) target was passed through a strong anion exchange resin (AG1 and AG MP) column to remove the majority of the fission products The related steps of basic dissolutionbased 99Mo- production process is briefly described in Fig [4] Kd values of the adsorption of molybdate and tungstate ions on the acidic alumina in the above mentioned simulator solution of variable acidity and in the pure HNO3 and H2SO4 solutions were measured using the procedure described in our previous work [5] The separation process was studied using a chromatographic column of 1.0 g acidic alumina sorbent, on which a given amount of the simulator solution was loaded The elution of 99Mo-molybdate ions was performed by sulphuric and/or nitric acid solutions of given acidity which is optimised based on the Kd measurement results Tungstate ions were striped out of the column with mL 1.0 M NH4OH solution for quantity analysis The elution fractions of mL were collected and their 99 Mo/188W radioactivity was measured using an Ortec gamma-ray spectrometer coupled with HpGe detector Results and discussion The Kd values versus acidity of HNO3 and H2SO4 solutions are shown in Figs and 2, respectively The Kd values of the tungstate and molybdate ions in both HNO3 and H2SO4 solutions have a significant difference for the solutions of [1.0 M acidity This difference of three magnitude orders offers a good opportunity to separate tungstates ions from molybdate ions using HNO3 and/or H2SO4 eluent This statement is confirmed by the separation results reported in Figs and For the reason of compatibility with other steps of the existing fission 99Mo separation process, the Mo/W separation using sulphuric acid solution is preferred The suitable acidity of sulphuric acid solutions used for the W/Mo separation is between 4–6 M The elution of alumina column with these solutions can be performed with a Mo-elution yield of [98 % and the tungsten is completely removed from the 99Mo solution 123 Author's personal copy J Radioanal Nucl Chem Fig Elution profile of WO42- (Black) and MoO42- (Grey) ions (Column: g alumina, Loading solution: mg Mo ? 10 mg W): a Eluent: M H2SO4 for MoO42- elution and M NH4OH for WO42-; b Eluent: M H2SO4 for MoO42- elution and M NH4OH for WO42- Fig Elution profile of WO42- (Black) and MoO42- (Grey) ions (Column: g alumina; Eluent: M HNO3 for MoO42- elution and M NH4OH for WO42-; Loading solution: mg Mo ? 10 mg W) 123 The Mo-uptake conditions of the alumina column in the simulator solutions were investigated to determine the possible influence of different ions existing in the solution used in a real fission Mo separation process The obtained results reported in Fig shows that the W and Mo loading ca99pability of alumina is high and a small size alumina column can be used to retain almost W-content of the 99Mo solution The influence of different ions of the simulator solution on the W-adsorption is insignificant Based on the obtained results it is stated that the separation of 99Mo from W contaminant can be effectively performed using an acidic alumina column and H2SO4 eluent The elution profiles of 99Mo/W separation are shown in Figs and This process is conveniently integrated with relevant steps of the alkaline dissolution-based technology in the process of 99Mo production using LEU targets The relevant steps of this 99Mo production process are shown in Fig An additional alumina column-based Mo/W separation step following the step of AG1 and AGMP anion exchange resin column separation is proposed to eliminate or reduce W contaminant from 99Mo solution before coming into the CHELEX-100 resin column for further purification This typical design of a chromatographic column loaded with 5–10 g acidic alumina and the elution of 99Mo with 50–70 mL 4–6 M H2SO4 solutions can be effectively used to remove more than 80 % of W contaminant content (*100 mg W) from a 1.356 105 GBq activity (E.O.B) 99Mo solution (*70 mg Mo) produced using 18.4 g 235U LEU-target (8 target plates) neutron-activated for 120 h in the OPAL reactor at ANSTO (Australia) The 99Mo recovery yield will be [96 % The integration of W/99Mo separation step of above mentioned W/99Mo separation process is fit for being in-line incorporated/integrated in the alkaline dissolution processbased99Mo recovery currently used in Argentina, Australia, and Germany as described in Fig [4] This technological step incorporation is feasible with respect to the compliance with operational safety requirements and possible update of licensing procedure The advantage of alumina use for fission 99Mo separation in the field of high radiation dose is the superior radiation resistance of this sorbent compared with the organic ion exchange resins The use of alumina also conforms to an existing final purification step used the process of fission 99 Mo production, thus unnecessary to reapply for a 99Mo production licence The use of alumina sorbent for the study of W/99Mo-separation is also justified based on the compliance with an current production facilities licensed by the authority in the majority of countries of the world Author's personal copy J Radioanal Nucl Chem Fig The integration of W/99Mo separation step in the process of LEU target-based 99 Mo production is proposed based on the alkaline dissolution-based process of 99 Mo- recovery from LEU target currently used in Argentina, Australia and Germany [4] Conclusions References The process of W/99Mo separation from a simulator fission-99Mo solution was developed using acidic alumina column and H2SO4 eluent 4–6 M H2SO4 solutions were successfully used to elute/separate 99MoO42- ions from the alumina which strongly retained WO42- ions without significant W-breakthrough The use of acidic alumina sorbent for the W/99Mo-separation conforms to an existing final purification step of the fission 99Mo production process and being justified based on the compliance with a production licence for the existing fission 99Mo-production facility used in the majority of countries of the world The developed W/99Mo separation process is fit for being in-line incorporated/integrated in the alkaline dissolution–based process of fission-99Mo recovery process currently used in the LEU target-based 99Mo production Le VS (2014) Science and technology of nuclear installations, Article ID 345252, http://dx.doi.org/10.1155/2014345252 Hetherington EL, Boyd RE, Targets for the production of neutron activated molybdenum-99, IAEA-TECDOC-1065 Dadachova K, La Riviere K, Anderson P (1999) Improved processes of molybdenum-99 production J Radioanal Nucl Chem 240:935–938 Management of radioactive waste from 99Mo production, IAEATECDOC-1051 Le VS, Morcos N (2008) New SPE column packing material: retention assessment method and its application for the radionuclide chromatographic separation J Radioanal Nucl Chem 277: 651–661 123 View publication stats ... technical solutions involved during the conversion of existing facilities using HEU targets, the reduction of W contamination in LEU target and/ or the separation of W from 99Mo solution of LEU process... problem regarding the decrease of 99Mo- retaining capacity of the ion exchange resin/sorbent columns which are currently used in both the processes of 99Mo stock solution and 99m Tc generator production. .. tungsten from LEU fission- produced 99Mo solution to improve technological performance in both the processes of 99Mo and 99mTc generator production Van So Le • Cong Duc Nguyen • Minh Khoi Le Received: