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Science and Technology of Nuclear Installations
Volume 2013 (2013), Article ID 932546, 9 pages
The Fission-Based 99Mo Production Process ROMOL-99 and Its Application to PINSTECH Islamabad
1GSG International GmbH, Eichenstraße 12, 8808 Pfäffikon, Switzerland
2IAF Radioökologie GmbH, Karpatenstraße 20, 01326 Dresden, Germany
3Foundation University Islamabad, 44000 Islamabad, Pakistan
4Isotope Production Division, Pakistan Institute of Nuclear Science and Technology (PINSTECH), P.O. Nilore Islamabad, 45650 Nilore, Pakistan
Received 27 June 2013; Accepted 15 August 2013
Academic Editor: Pablo Cristini
Copyright © 2013 Rudolf Muenze et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.
An innovative process for fission based 99Mo production has been developed under Isotope Technologies Dresden (ITD) GmbH (former Hans Wälischmiller GmbH (HWM), Branch Office Dresden), and its functionality has been tested and proved at the Pakistan Institute of Nuclear Science and Technology (PINSTECH), Islamabad. Targets made from uranium aluminum alloy clad with aluminum were irradiated in the core of Pakistan Research Reactor-1 (PARR-1). In the mean time more than 50 batches of fission molybdenum-99 (99Mo) have been produced meeting the international purity/pharmacopoeia specifications using this ROMOL-99 process. The process is based on alkaline dissolution of the neutron irradiated targets in presence of NaNO3, chemically extracting the 99Mo from various fission products and purifying the product by column chromatography. This ROMOL-99 process will be described in some detail.
The present sources of molybdenum-99 (99Mo; h) are research reactors by neutron-induced fission of 235U, which results in high-specific activity 99Mo, or using the (, ) nuclear reaction with 98Mo (natural Mo or enriched 98Mo = 24%), resulting in low-specific activity 99Mo. Generally, the specific activity of molybdenum produced by fission is more than 1000 times higher than that obtained by () process. The almost universal means by which technetium-99m (99mTc; h) is made available for clinical applications is from the elution of generators containing high-specific activity fission-based 99Mo.
The first chemical process for separation of fission 99Mo was described by the Brookhaven group, USA . In this process the target (93% enriched U-235 alloyed with Al) was dissolved in 6 M nitric acid catalyzed by mercuric nitrate. In the former Zentralinstitut für Kernforschung (ZfK) Rossendorf, a fission-based 99Mo separation technology became operationally ready in 1963 which was actually the basis of the first fission-based 99Mo/99mTc generator in Europe. Metallic natural uranium pellets were used as target material and the dissolution of the irradiated U-pellets was done with concentrated HCl. Quartz and glass apparatus was used in chemical processing, and yield of 99Mo was ~70% . In 1980, this process was replaced by the AMOR process (AMOR: Anlage zur Mo Production Rossendorf), developed in the same institute . The AMOR process made use of original fuel elements of the RF-reactor as qualified target which was dissolved in HNO3/Hg. Batch-wise adsorption at Al2O3 and sublimation technique were used for separation and purification of the 99Mo. This process was in operation until the shut-down of the Rossendorf Research Reactor in 1991.
Another small-scale production process for fission 99Mo was proposed by the Rossendorf group in which natural uranium as uranium oxide was used as target material . This procedure was particularly interesting for those which do not dispose of enriched nuclear fuel material. Approximately 400 g of uranium oxide enclosed in irradiation cans are dissolved in nitric acid after irradiation for 100 hrs at a neutron flux of in a research reactor. The separation of 99Mo from the fuel-fission product solution is performed by ion exchange with alumina in a chromatography column. Final purification includes the repeated chromatography separation and subsequently a sublimation stage.
Based on their own long-term experiences and considering international achievements in 99Mo production, scientists of the Radio-Isotope department of the former Rossendorf institute ZfK designed a new process for fission-based 99Mo production named ROMOL-99 [5–7]. The basic principles of this process are as follows (see also the flow scheme, Figure 1).(i)The dissolution of the UAlx/Al-clad targets shall be performed in a mixture of NaOH/NaNO3 without H2 generation, under reduced pressure conditions.(ii)The Xe shall be trapped cryogenically after passing a gas treatment line.(iii)The NH3 generated in the dissolving process shall be separated prior to Mo separation.(iv)The radioiodine shall be separated prior to 99Mo-separation as well.(v)During dissolving process nitrite is generated which shall be eliminated prior to the 99Mo separation.
The basic parameters of this process has been developed with modern nonradioactive analytical techniques by the IAF-Radioökologie GmbH Dresden, while the active testing and optimization of the process has been carried out at PINSTECH Islamabad under supervision of the German scientists. In this paper the chemical process of the ROMOL-99 technology will be described in some detail.
2. Materials and Methods
All chemicals were purchased from E. Merck (Germany) and were of guaranteed reagent grade (GR) or analytical reagent (AR) grade. Al2O3 (90 active acidic for column chromatography, 70–230 mesh ASTM) was used. Silver-coated alumina was freshly prepared at institute. Organic anion-exchange resin was purchased from BioRad, USA.
The non-radioactive development work was performed using uranium-free Al-plates (purchased from PINSTECH) having the same composition as the material used for the original targets.
Tracer experiments were performed using 131I tracer activities which were taken from the PINSTECH routine 131Iodine production, and the 99Mo tracer was taken from the routine PAKGEN 99mTc generator production (99Mo imported from South Africa).
2.1. Irradiation of Target
Qualified HEU/Al alloy clad with high purity aluminum target plates  were irradiated for 12–18 h at a neutron flux of ~1.5 × 1014 cm−2 s−1 inside the core of the Pakistan Research Reactor-1 (PARR-1). After 24 h cooling, the irradiated target plates were transferred to the 99Molybdenum Production Facility (MPF) for separation of 99Mo from the uranium, actinides, and fission products. For the warm test runs, targets were irradiated for short times at lower flux density, but the target composition was identical with those for production runs. The irradiation conditions were chosen in a way that the total activity inventory for the development work was of the order of 4 GBq.
2.2. Process Control and Quality Control
Gamma ray spectroscopy high-purity Ge detector (Canberra Series 85 multichannel analyzer) was used to determine the activity balance during all process steps and for the determination of radionuclide impurities in the final 99Mo product. This concerns mainly 131I and 103Ru, 132Te. Beta counting of 89Sr and 90Sr was done by a liquid scintillation analyzer (Tri-Carb 1900 TR, Packard Canberra Company) after separation by ion-exchange and precipitation with the aid of carrier. Contamination of alpha emitters was done with the -counter (UMF-200). Radiochemical purity of [99Mo] molybdate was determined by means of paper chromatography with a mixture of hydrochloric acid, water, ether, and methanol (5 : 15 : 50 : 30) as mobile phase. The chemical purity was occasionally determined after decay by optical emission spectrometry (Optima 3300XL, Perkin Elmer). It was used for the determination of toxic elements such as Cr, Co, As, Sn, Cd, Pb, and U; the detection limits in ppm were 2, 5, 5, 5, 1, 5, and 2, respectively.
The final 99Mo product was dispensed and assayed by means of a calibrated ionization chamber. Radioactivity concentration (MBq/cm3) was calculated by dividing the total product activity by the final volume of the product solution. All required nuclear data were taken from NuDat 2.5 .
3. Results and Discussion
3.1. The Dissolving Process
When dissolving the target plates in the solvent, 3 M NaOH/4 M NaNO3, 3 reactions must be considered leading to different reaction products as follows: The most important is reaction (1), where the Al reduces the nitrate ion down to NH3. Close to the end of the dissolving process the nitrate is reduced only to nitrite (2). This fraction is of the order of 10 to 15%. The theoretically possible reduction (3) generating hydrogen is nearly suppressed. Gas chromatographic determination of hydrogen in the off-gas from the dissolving process did not show any signal for H2, meaning, the upper limit for H2 generation is <2% and is therefore without any danger.
Under the conditions that represents the mass unit of the target matrix that undergoes to nitrite formation and consequently represents the mass units of the target matrix that undergoes under NH3 formation, we obtain the “master equation” for dissolving the targets following: The value , representing the fraction of the aluminum that is dissolved under nitrite formation, ranges between .
The solvent volume needed for the process is determined by the solubility of the sodium aluminate (NaAlO2) which is 2.1 M/L corresponding to 57 g/L Al. Furthermore, the Na concentration should be kept as high as possible, in order to reach safely the saturation concentration for the precipitate Na2U2O7. A high nitrate concentration is needed for avoiding the formation of hydrogen, while the viscosity of the solution should be suitable for easy filtration. We found a composition of 3 M NaOH/4 M NaNO3 as most suitable for the dissolving process.
The dissolving process is strong exothermic (close to 600 kcal are generated for dissolving 100 g Al), and in addition the dissolving speed increases with the second power of the temperature. Thus, the reaction is self-accelerating. Following the experiences collected in Dresden (IAF) and PINSTECH, the control of the dissolving process is easy and safely possible by short heating and cooling pulses. With these techniques one can easily adjust the dissolving temperature at around 70–80°C. Furthermore, the process can be performed at slightly reduced pressure conditions (see Figure 2). The dissolving process is performed in a special, dissolving vessel, equipped with heater and cooling jacket.
3.2. NH3 Distillation
Since the iodine shall be removed from the process solution using a silver-coated column material, the NH3 is recommended to be eliminated because it has potential to influence the efficiency of the iodine removal at the Ag-coated column. The simplest way to separate the NH3 is the distillation from strong basic solution. Preliminary experiments have shown that 150–200 mL distilled volume is sufficient. This volume can be distilled off from the target solution within about 20 minutes. In the production runs, the distilled NH3 is trapped in 5 N H2SO4 solution.
The precipitate that is formed during the dissolving process is composed of mainly 2 components: the Na-diuranate and in addition the nonsoluble hydroxides, oxides or carbonates of several alloying metals of the Al-matrix that are coprecipitated together with the Na2U2O7. Based on analytical data of the Al-matrix material used for the target preparation, the following quantities for the precipitate should be expected (Table 1).
Assuming a density of the precipitate of 4.4 g/cm3 (based on ~30% porosity) and the uranium in the form of Na2U2O7 × 6 H2O, one would obtain a precipitate volume of ~2.37 cm3, which corresponds to a filter bed thickness of mm.
The target element uranium after dissolution must be present exclusively in the chemical form of Na2U2O7 because it is well known that uranium species of lower oxidation stage absorb 99Mo and consequently lower the production yield. Dissolving the same targets alone in NaOH or KOH (without NaNO3) , an additional oxidation process (usually H2O2) is required to reach the oxidation stage of +6 for both of the U and the Mo.
As shown from the crystallographic analysis, the target element uranium was found after our ROMOL-99 dissolving process straight as sodium diuranate (Na2U2O7) in the precipitate (Figure 3) without any further treatment.
The time needed for filtration is mainly determined by the surface area and the porosity of the used filter plate and the filter cake, the viscosity of the solution and the filtration pressure. The filter plate consists of a 3 mm thick metallic (INOX) sinter plate with a porosity of ~30 μm. The cold and hot runs showed that first of all the precipitate can be filtrated from the target solution with the above given alloying components, and in addition sufficient filtration speed (200–300 mL/min) is achieved in 10–20 min at a temperature of around 50°C.
3.4. Iodine Removal
In order to minimize the risk of iodine release in later production steps and waste, the adsorption on silver is the most promising approach for trapping the radioiodine before the 99Mo is separated. During the production process, we have to deal with 132I (2.3 h half-life, daughter of 132Te), 133I (20.8 h half-life), and 131I (8.02 d half-life). Freshly prepared silver-coated Al2O3 material has shown to be the most appropriate material; this material has been prepared according to the Wilkinson et al. method . The iodine removal process is performed by controlled filtration of the filtrated target solution over a column filled with this material. The flow rate needs to be controlled. Since the optimal flow rate for high iodine trapping efficiency is identical with the speed for introducing the basic target solution into the strong acid reaction vessel (next process step); there is no technological separation of both steps, thus, while transferring the basic solution into the nitric acid for acidification the radioiodine is removed simultaneously. The transfer process lasts for about 60–90 minutes. The efficiency for iodine trapping has been determined to be >98%. The Ag-column also traps a good fraction of the Ru (see Figure 4); 99Mo could not be detected within a detection limit of 3%.
3.5. Acidification and Nitrite Destruction
For the main separation step—the separation of the 99Mo from the process solution after iodine removal—Al2O3 column chromatography has been selected. Molybdate is adsorbed from weak HNO3-acid media at Al2O3 (this principle is used in the 99Mo/99mTc-generator technology). Thus, the strong basic process solution needs to be acidified. This is not an easy step, since the Al-concentration is high. An anion-exchange process, as it is used in cases were only NaOH or KOH is involved for dissolving the targets under H2 generation, is not possible due to the high concentration. Many test experiments have been performed in order to determine the optimal conditions for this step. When introducing strong basic aluminate solution into strong acid HNO3 solution in the first step, Al(OH)3 is precipitated. This hydroxide needs to dissolve immediately, otherwise it may transmute into nonsoluble configurations which may create problems in the further production steps. When the basic solution is introduced with moderate speed (50–70 mL/min) under strong mixing, one observes first a thick white precipitate that is redissolved relatively fast. Due to neutralization heat the solution is warming up. In order to bring the solution to boil additional heating is required.
As said before, we also have nitrite in the system, which is recommended to be destroyed. Simultaneously with the acidification process the nitrite is reduced with urea under nitrogen formation according to In test experiments, this gas generation looked like very fine silk. This gas generation is mixing the solution only a little, because of the microscopic fine bubbles, this effect is by far insufficient; additional strong stirring is required. The complete nitrite destruction and the re-dissolving of the primary precipitated hydroxides require refluxing under stirring for one additional hour after complete solvent transfer.
The reaction gases of this acidification process and in addition a slight carrier gas flow release also remain volatile iodine species (from iodine residue and decay of Te-parent nuclides) and radio Xe (mainly from 133I-decay). The radio iodine is retained in a gas adsorption trap filled with Ag-IONEX. This is a Zeolite exchange material that adsorbs at °C volatile inorganic and organic iodine species that is widely used in fuel reprocessing process for decontamination of acid off-gases. After passing the IONEX filter, off-gas from the acidification process that still may contain some Xe is introduced into the gas process line for further treatment.
When the nitrogen formation and iodine release is finished, the solution is cooled down to room temperature and is ready for the Al2O3 column process.
3.6. Alumina Column Process
The separation of the 99Mo from the acidified target solution is achieved via anion exchange chromatography using week acid Al2O3 as column material. The adsorption efficiency for Mo depends mainly on the salt concentration and the acidity of the solution and not so much on the absorber material itself. For the optimization of the column parameters, the control of the free acidity in the FEED solution played an important role. Due to the high salt concentration (especially that of Al3+), a direct pH-measurement is not possible. A potentiometric titration did not show the required precision (see Figure 5).
The safe and better is to dilute a sample of the solution by a factor 1 : 100 with distilled water. This solution could perfectly undergo a pH measurement with an ordinary glass electrode. The pH determined in this way was always in the region of 2.2 < pH < 2.6, which means that the free acid concentration in the original FEED solution under practical conditions was in the range of about 0.15 < [H+] < 0.7 M.
Under practical conditions, the volume of the loading solution (FEED) is around 6 L (for ~100 g target material). After the loading process, the column shall be washed with 0.5–1.0 L of 0.5 M HNO3 500 mL water and then with 1500 mL 0.01 M NH4OH. The 99Mo is then eluted with up to 2000 mL of 1 M NH4OH solution. One obtains a raw 99Mo product of already ≫99% radionuclide purity.
In order to define the optimal Al2O3 column parameters one needs to consider(i)the adsorption capacity of the exchange material Al2O3,(ii)the selectivity related to the separation from radioactive contaminations,(iii)the possible and needed loading- and elution speed which is relevant for the duration of the process.
The capacity of Al2O3 for Mo adsorption is known to be ~30 mg Mo/g Al2O3 column material. In test experiments using 50 mL of model-target solutions containing a Mo-concentration of 20–33 mg Mo/L and columns with 2 g Al2O3 column material (0.7 × 5.6 cm column dimension) using a flow rate of 7 cm/min corresponding to 2.7 mL/min the Mo could be adsorbed with an average yield of >90%. Thus, in these experiments only a small fraction (1.5–2.5%) of the capacity of the exchanger has been utilized. This corresponds to 0.5–0.8 mg Mo/g Al2O3. Based on this data one would need for processing of 3 target plates theoretically 140 g of Al2O3 corresponding to 152 mL Al2O3 for the separation process.
For defining the column dimensions one needs to find a compromise between the needed amount of the Al2O3 material and reasonable high applicable elution speed. Furthermore one has to consider losses due to irreversible bound Mo with increasing Al2O3 quantities. Assuming the following practical conditions:(i)the total volume of liquids that has to pass the column is ~11 L composed from 5.8 L acidified target solution, 3.3 L wash solutions, and 2.0 L elution volume,(ii)the linear filtration speed is 7 cm/min (50 mL/min for loading and eluting and 80 mL/min washing), (iii)the diameter of the column shall be 5 cm,
one obtains a volume flow speed of 137 mL/min under practical conditions.
Considering the previous determined 140 g or 152 mL Al2O3 absorber material, one would obtain an absorber bed height of 7.8 cm. If one increases the dimensions by at least a factor 2 for compensating not optimal conditions, the length of the Al2O3 column becomes 15.6 cm filled with 304 mL Al2O3 absorber material.
During the commissioning, it has been demonstrated that a 250 g alumina column bed is acceptable which corresponds to a column bed volume of about 275 mL. Table 2 summarizes the Al2O3 column process parameters.
Using the parameters shown in Table 2, the profile for eluting the 99Mo from the Al2O3 column has been performed. As seen in Figure 6, the 99Mo is eluted in a relatively sharp peak and 1000 mL of 1.0 M NH3 solution is sufficient. The 99Mo retention at the column using model solutions was practically 100%, and the 99Mo recovery was measured to be 91.2%.
The Al used in the target material contains some quantities of Si. It is well known that Si forms very unpleasant nonsoluble Mo Si-species which may cause dramatic losses in the 99Mo yield. Certain limited quantities of Mo-carrier can help solving this problem. The other way around would be to elute the Al2O3 column with higher-concentrated NH3 (2 M instead of 1 M) or with NaOH.
3.7. DOWEX-1 Column Process
Molybdenum in its anionic form is adsorbed directly from the ammonia solution eluted from the Al2O3 column at strong basic anion exchange resins as DOWEX-1 (configuration ). The distribution coefficients has been determined to be for adsorption from 1 M NH4OH and for the desorption with 1 M (NH4)2CO3 solution.
The dimensions of a suitable DOWEX-1 column and its operation parameters are determined in a similar way as demonstrated for the Al2O3 column. For a column of about 26 × 120 mm, a linear flow speed of 13 cm/min is the maximum. If the volume of the Mo solution is 2000 mL one would need theoretically 54.6 g of the ion exchange resin. DOWEX-1 in the dry form. Considering the density of 0.65 g/mL resin, this would give an 84 mL volume of the resin. For rinsing the column, 4 bed volumes are required which correspond to 340 mL. Table 3 summarizes the parameters for the DOWEX column process. The corresponding elution profile is shown in Figure 7.
3.8. Evaporation Step
The purification step at the DOWEX column delivers 200 mL of the 99Mo molybdate in 1 M (NH4)2CO3 solution. In the following step this eluted solution is evaporated to the dryness in a special evaporator, with condenser. During evaporation, the (NH4)2CO3 is being decomposed; thus no additional salts are introduced into the final configured [99Mo] molybdate solution.
The residue is redissolved in the desired volume of diluted NaOH solution forming the final product solution [99Mo]Na2MoO4. This final product solution is then transferred into a corresponding plastic vial and transferred into the hot cell 3 for further processing, precise measurement and distribution.
A sublimation step (at 1000°C) is foreseen as an additional reserve for improving purity, if required. In this case the residue after evaporation shall be redissolved in diluted HNO3 or NH4OH.
3.9. Radioactivity Balance during the Process
Careful studies have been performed to obtain a full picture on the behavior of the 99Mo, of the most important impurities in 99Mo preparations, for other fission products, and for the target element itself. On one hand, tracer activities of 99Mo and 131I have been used, and after having optimized the separation conditions the same full protocol has been applied to study the separations technology with weak irradiated original target material (activity level ~4 GBq). Figure 8 summarizes gamma-spectroscopic measurements that illustrate how powerful the individual separation steps are. Segments of original measured gamma spectra are shown in one graph. Signals of the most critical radionuclides are clearly identified. In order to have a good overview, the original data of the different spectra have been expanded using a factor shown in the graph.
The upper spectrum has been taken from a small fraction of the filter cake (precipitate) followed by the spectrum from a sample from the filtrate. It is clearly seen that only few gamma lines are left in the filtrate, which correspond to 99Mo, its daughter 99mTc, the radioiodine’s, and some fractions of Ru. The strongest signals in the precipitate (239Np, and 140La) are not seen in the filtrate solution. As said before the filtrate is passed through a silver-coated Al2O3 column prior to the acidification process. Thus comparing the spectra 2 and 3, one clearly sees that iodine is missing in the FEED solution (see also Figure 4). When interpreting spectrum 3, one needs to consider that the strongest gamma signal from 99Mo is 739 keV with the branching of 12.13% (not shown in this graph). The two gamma lines here at 181 keV and 366 keV have a branching of only 5.99 and 1.19%, respectively. Finally the last spectrum below is taken from a sample of the final 99Mo preparation. All measurements have been performed using a Pb absorber to suppress the strong gamma signal from 99mTc (and other low-energetic radiation).
3.10. Radioactivity Distribution between Precipitate and Filtrate
In total the precipitate collects more than 60% of the radioactivity formed in the nuclear process in the chemical form of hydroxides, oxides, or carbonates of the fission products. This corresponds to the fission products of Ba and Sr, the rear earth elements and actinides, Zr/Nb. Te and Sb are nearly quantitatively collected in the precipitate. The results of a corresponding tracer experiment, are summarized in Table 4. In this experiments target plates of the original composition were used.
The most important impurities have been followed up quantitatively throughout the process as good as gamma-spectroscopy could do under practical conditions with limited measuring time. The results are summarized in Table 5. The FEED solution (filtrate after passing the silver column) contains already relatively clean 99Mo, however in presence of high salt concentration (Al, Na including 24Na and fission-Cs). Up to 80% of the Ru is found in the filtrate, at the silver column already about 22% are retained. The remaining Ru is passing the Alumina column during the loading procedure. Careful washing avoids the transfer of Ru to the next purification steps. Ru shows a nonstandard behavior, sometimes we observed significant higher Ru-retention in the filter cake.
The 132Te is nearly quantitatively coprecipitated. The highest 132Te-content in the filtrate was 3.7% of the original quantity. About half of this fraction is retained at the silver column, the other half fraction at the Alumina column. In the filtrate and wash solutions from the Alumina column the 132Te could not be detected any more (with the applied spectrometric parameters).
The iodine is nearly quantitatively retained at the silver column. The small fraction that is passing the silver column is then distributed throughout the system, mainly in the waste solutions through the washing procedures of the columns. Summarizing, the separation and purification process is efficient and easy.
3.11. 99Mo Production Facility at PINSTECH
The 99Mo Production Facility (MPF) is installed at PINSTECH Phase-1 building near reactor hall of PARR-1. The technical realization of the ROMOL-99 process in a semiautomated separation facility has been carried out by ITD Dresden GmbH (former Hans Wälischmiller (HWM) GmbH, Branch office Dresden). The main working areas of this facility are the Hot Cell complex (3 Hot Cells), interim liquid storage tanks, charcoal filter beds for iodine retention, xenon delay tanks, and the operator and service areas interconnected with it. Additionally, there are the so-called lock rooms through which the activated targets, the final product, and solid and liquid wastes are moved. Still, there are areas for personnel, preparation of reagents, storage, dosimetry, measurement, and decontamination. Further equipment in other rooms or buildings, which participate in the 99Mo production, is the existing equipment of the main exhaust system with filter chamber, Secomak blowers and the main exhaust blower. The spent target material (loaded filter plates enclosed in screw shut cans) is stored in Spent Fuel bay of PARR-1. The solid low-radioactive wastes (spent ion-exchange columns, tubes, interconnections, and other one-way materials) are stored, while decayed radioactive liquid waste is cementized in the radioactive waste management Group building.
More than 50 commercial batches of fission based 99Mo using ROMOL-99 process have been successfully completed. After the evaporation step, the residue is dissolved in the desired volume of diluted NaOH solution forming the final product solution [99Mo]Na2MoO4. This final product solution is then transferred to the PAKGEN 99mTc generator production site at PINSTECH. These generators are then distributed to the 35 nuclear medical centers in Pakistan. The performance of these generators is comparable to that of generators produced from imported fission 99Mo. The quality of the 99Mo preparations produced at PINSTECH corresponds to the required international standard (Table 6). Details about the preparation of PAKGEN 99mTc generators and their quality control have already been reported . The next steps at PINSTECH related to the routine 99Mo-production are upscaling the production capacity and transmutation to LEU (low enriched uranium) as target fuel.
The ROMOL-99 process allows dissolving UAlx/Al clad dispersion targets under reduced pressure conditions without generation of hydrogen at temperatures between 70 and 80°C. The technology implements the separation of NH3 and radio-iodine prior to the 99Mo separation. Generated nitrite is safely destroyed during the acidification process by urea to N2. The technical realization of the ROMOL-99 process in a semiautomated separation facility has been carried out by ITD Dresden GmbH (former Hans Wälischmiller (HWM) GmbH, Branch Office Dresden). More than 50 commercial batches of fission-based 99Mo using the ROMOL-99 process have been successfully completed at PINSTECH. PAKGEN 99mTc generators were prepared by using this locally produced high purity fission 99Mo and distributed to 35 nuclear medical centers in Pakistan. The performance of these generators is comparable to that of generators produced from imported fission 99Mo.
Conflict of Interests
The authors declare that they have no conflict of interests.
- L. G. Stang, BNL 864 (T-347), 1964.
- D. Novotny and G. Wagner, “Procedure of small scale production of Mo-99 on the basis of irradiated natural uranium targets,” in Proceedings of the IAEA Consultancy Meeting on Small Scale Production of Fission Mo-99 for Use in Tc-99m Generators, Vienna, Austria, July 2003.
- R. Muenze, O. Hladik, G. Bernhard, W. Boessert, and R. Schwarzbach, “Large scale production of fission 99Mo by using fuel elements of a research reactor as starting material,” International Journal of Applied Radiation and Isotopes, vol. 35, no. 8, pp. 749–754, 1984.
- O. Hladik, G. Bernhardt, W. Boessert, and R. Muenze, “Production of fission 99Mo by processing irradiated natural uranium targets,” Fission Molybdenum for Medical Use, IAEA-TECDOC-515, IAEA, Vienna, Austria, 1989.
- G. J. Beyer, R. Muenze, D. Novotny, A. Mushtaq, and M. Jehangir, “ROMOL-99—a new innovative small-scale LEU-based Mo-99 production process,” in Proceedings of the 6th International Conference on Isotopes, Seoul, Republic of Korea, May 2008.
- G. Beyer, R. Muenze, D. Novotny, M. Ahmad, and M. Jehangir, “S8-4 ROMOL-99: a new innovative small-scale LEU-based Mo-99 production process,” in Proceedings of the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR '08), Washington, DC, USA, October 2008.
- G. Beyer, R. Muenze, D. Novotny, M. Ahmad, and M. Jehangir, “S13-P6 ROMOL-99: a new innovative small-scale LEU-based Mo-99 production process,” in Proceedings of the 31th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR '09), Beijing, China, November 2009.
- A. Mushtaq, “Specifications and qualification of uranium/aluminum alloy plate target for the production of fission molybdenum-99,” Nuclear Engineering and Design, vol. 241, no. 1, pp. 163–167, 2011.
- NuDat 2.5, http://www.nndc.bnl.gov/nudat2/.
- A. A. Sameh and H. J. Ache, “Production techniques for fission molybdenum -99,” Radiochimika Acta, vol. 41, pp. 65–72, 1987.
- M. V. Wilkinson, A. V. Mondino, and A. C. Manzini, “Separation of iodine produced from fission using silver-coated alumina,” Journal of Radioanalytical and Nuclear Chemistry, vol. 256, no. 3, pp. 413–415, 2003.
- A. Mushtaq, S. Pervez, S. Hussain et al., “Evaluation of Pakgen c generators loaded with indigenous fission 99Mo,” Radiochim Acta, vol. 100, pp. 793–800, 2012.