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Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on â¹â¹Mo recovery and purification by its precipitation with α-benzoin oxime. Parameters that were studied include concentrat...
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Context 1
... of these tests are shown in Table 1. As can be seen, molybdenum recoveries are above 90% for all solutions except for test 5, which has the lowest molybdenum and α-benzoin oxime concentrations. ...
Citations
... During the production steps, about 50 kg of weapons-grade HEU are usually handled, and only a very minute amount (~3%) is utilized in the entire process (NAP, 2009;NEA, 2010;IAEA, 2015). In order to eliminate these nuclear proliferation issues, HEU targets have to be fully substituted by LEU or natural uranium targets (Wu et al., 1995;Cols et al., 2000;Conner et al., 2000). In addition, the aging fleet of the currently used irradiation facilities is also of concern. ...
The continuing rapid expansion of 99mTc diagnostic agents always calls for scaling up 99mTc production to cover increasing clinical demand. Nevertheless, 99mTc availability depends mainly on the fission-produced ⁹⁹Mo supply. This supply is seriously influenced during renewed emergency periods, such as the past ⁹⁹Mo production crisis or the current COVID-19 pandemic. Consequently, these interruptions have promoted the need for 99mTc production through alternative strategies capable of providing clinical-grade 99mTc with high purity. In the light of this context, this review illustrates diverse production routes that either have commercially been used or new strategies that offer potential solutions to promote a rapid production growth of 99mTc. These techniques have been selected, highlighted, and evaluated to imply their impact on developing 99mTc production. Furthermore, their advantages and limitations, current situation, and long-term perspective were also discussed. It appears that, on the one hand, careful attention needs to be devoted to enhancing the ⁹⁹Mo economy. It can be achieved by utilizing ⁹⁸Mo neutron activation in commercial nuclear power reactors and using accelerator-based ⁹⁹Mo production, especially the photonuclear transmutation strategy. On the other hand, more research efforts should be devoted to widening the utility of ⁹⁹Mo/99mTc generators, which incorporate nanomaterial-based sorbents and promote their development, validation, and full automization in the near future. These strategies are expected to play a vital role in providing sufficient clinical-grade 99mTc, resulting in a reasonable cost per patient dose.
... Some authors have reported modifications of the CINTICHEM process. 10,[12][13][14][15] Other authors have explored various column materials to treat acidic UAl x leachates. [16][17][18][19] Another strategy that has received some attention is the exploitation of differences in the volatility of oxides and chlorides of the fission products. ...
A method is presented to separate molybdenum from other elements commonly present in oxidative alkaline leachates of irradiated uranium-aluminum targets for the production of molybdenum-99. The separation was accomplished by selective extraction of molybdate anions using triazolium and phosphonium sulfate ionic liquid extractants, either diluted in 1-octanol or undiluted. Molybdenum was then stripped from the organic phase using a sodium hydrogen carbonate solution. The extractant was regenerated by contacting the organic phase with an alkaline sulfate solution. The extraction mechanism and the influence of the diluent on the extractant performance were investigated. The reported method provides a promising alternative to state-of-the-art chromatographic processes, showing potential for limiting the production of radioactive waste.
... A collaboration is continuing between BATAN and Argonne National Laboratory (ANL) under the aegis of the RERTR (Reduced Enrichment for Research and Test Reactors) program to develop means for 99 Mo production using LEU-metal foil targets. Earlier work has been reported at previous RERTR meetings12345678910. This paper provides a progress report on the process-demonstration phase of our collaboration. ...
In March and September 1999, demonstrations of the irradiation, disassembly, and processing of LEU metal foil targets were performed in the Indonesian BATAN PUSPIPTEK Facilities. These demonstrations showed that (1) irradiation and disassembly can be performed so that the uranium foil can be easily removed from the target body, and (2) with only minor changes to the current process, the LEU foil can produce yield and purity of the {sup 99}Mo product at least as great as that obtained with the HEU target. Further, because of these modifications, two hours are cut from the processing time, and the liquid waste volume is reduced. Results of these demonstrations will be presented along with conclusions and plans for future work.
... A collaboration is underway between BATAN and Argonne National Laboratory (ANL) under the aegis of the RERTR (Reduced Enrichment for Research and Test Reactors) program to carry out R&D and demonstration on the production of 99 Mo using LEU-metal foil targets. This paper gives the results of a continuation of earlier work reported at previous RERTR meetings123456789. EXPERIMENTAL Six new targets were fabricated at ANL for irradiation in Indonesia; two other targets were already at PUSPIPTEK from an earlier shipment. ...
... Early R&D [30,31] was based on our knowledge of the Cintichem process found in patents [13,14]. Later R&D [32][33][34][35] was based on a firm knowledge of the process due to our cooperative project with BAT AN and SNL. Details of the work presented below can be found in these publications. ...
... Argonne National Laboratory and the University of Illinois at Urban/Champaign are collaborating with the National Atomic Energy Agency (BATAN) of Indonesia to develop and demonstrate the use of LEU targets in the Cintichem process. This work is a follow-up to work on this project reported last year [3][4][5][6]. In the next few months, we plan to perform the demonstration of processing a fully irradiated LEU metal foil at the PUSPIPTEK Radioisotope Production Center in Serpong, Indonesia. ...
... Molybdenum precipitation is quantitative, and the precipitate contains very little impurities. In our previous tests, we found that molybdenum can be also precipitated quantitatively with α-benzoin oxime from a nitric acid solution [3,4]. However, to prove the feasibility of using the nitric acid alone as a dissolver solution, we had to verify that radionuclide decontamination of the 99 Mo product is not degraded by this modification. ...
... As can be seen from the comparison in Table 1, there is no significant effect of eliminating sulfuric acid for either molybdenum recovery or radioisotope decontamination. As reported in the past [3][4][5], most of the decontamination is done in the precipitation of molybdenum with α-BO, and the following purifications are polishing steps. We must note that our tracer tests only indicate chemical behavior; verification of this behavior will require full-scale demonstrations using fully irradiated uranium-foil targets. ...
Presented here are recent experimental results on tests of a modified Cintichem process for producing â¹â¹Mo from low enriched uranium (LEU). Studies were focused in three areas: (1) testing the effects on â¹â¹Mo recovery and purity of dissolving LEU foil in nitric acid alone, rather than in the sulfuric/nitric acid mixture currently used, (2) measuring decontamination factors for radionuclide impurities in each purification step, and (3) testing the effects on processing of adding barrier materials to the LEU metal-foil target. The experimental results show that switching from dissolving the target in the sulfuric/nitric mixture to using nitric acid alone should cause no significant difference in â¹â¹Mo product yield or purity. Further, the results show that overall decontamination factors for gamma emitters in the LEU-target processing are high enough to meet the purity requirements for the â¹â¹Mo product. The results also show that the selected barrier materials, Cu, Fe, and Ni, do not interfere with â¹â¹Mo recovery and can be removed during chemical processing of the LEU target.
... The samples were analyzed by gamma-ray spectrometry for fission product nuclides by using a high-purity germanium detector, and by a gas-flow proportional counter (Eberline, Model SAC-4) for alpha-emitting nuclides. These data are comparable to results obtained at the University of Illinois [6]. ...
Two demonstrations of the use of the Cintichem process on simulated low enriched uranium (LEU, < 20% {sup 235}U) targets were run by personnel in the BATAN Isotope Production Facilities at PUSPIPTEK (Serpong, Indonesia). These demonstrations were done using a solution of either natural or depleted uranium spiked with irradiated high enriched uranium (HEU). The activity levels were low enough to perform the process in a fume hood. The volumes, equipment, and procedures used were the same as those used in the actual processing of irradiated HEU targets in a shielded cell. These results, when combined with data obtained at the University of Illinois and Argonne National Laboratory, show that substitution of LEU for HEU is possible for the Cintichem process, perhaps, with no modification.
... Experiments are planned to determine the parameters that control this process, including the effect of carbonates on interfering with the uranium precipitation. Since coprecipitation loss of 99 Mo was experienced with bulky uranyl hydroxide [15], coprecipitation with sodium diuranate will be assessed. Processes for recycling the recovered uranium and their cost effectiveness will be investigated. ...
Low-enriched uranium silicide targets designed to recover fission product â¹â¹Mo were dissolved in alkaline hydrogen peroxide (HâOâ plus NaOH) at about 90C. Sintering of matrix aluminum powder during irradiation and heat treatment retarded aluminum dissolution and prevented silicide particle dispersion. Gas evolved during dissolution is suspected to adhere to particles and block hydroxide ion contact with aluminum. Reduction of base concentrations from 5M to O.lM NaOH yielded similar silicide dissolution and peroxide destruction rates, simplifying later processing. Future work in particle dispersion enhancement, â¹â¹Mo separation, and waste disposal is also discussed.
... [16] [17] [18] [19] [20] [21] [22] [23] [24] ...
Presented here are recent experimental results of the continuing development activities associated with converting current processes for producing fission-product â¹â¹Mo from targets using high-enriched uranium (HEU) to low-enriched uranium (LEU). Studies were focused in four areas: (1) measuring the chemical behavior of iodine, rhodium, and silver in the LEU-modified Cintichem process, (2) performing experiments and calculations to assess the suitability of zinc fission barriers for LEU metal foil targets, (3) developing an actinide separations method for measuring alpha contamination of the purified â¹â¹Mo product, and (4) developing a cooperation with Sandia National Laboratories and Los Alamos National Laboratory that will lead to approval by the US Federal Drug Administration for production of â¹â¹Mo from LEU targets. Experimental results continue to show the technical feasibility of converting current HEU processes to LEU.
A new method is presented here for digesting irradiated low-enriched uranium foil targets in alkaline carbonate media to recover 99Mo. This method consists of the electrolytic dissolution of uranium foil in a sodium bicarbonate solution, followed by the precipitation of carbonate, base-insoluble fission products, activation products, and actinides with calcium oxide; most of the molybdenum, technetium, and iodine remain in solution. An electrochemical dissolver and mixing vessel were designed, fabricated, and tested for the processing of a full-sized irradiated foil under ambient pressure and elevated temperature. Over 92% of the fission-induced 99Mo was recovered in a product solution that was compatible with an anion-exchange column for retaining molybdenum and iodine.
