Molybdenum-99 Production Research Articles

Radionuclide characterization and handling strategy of irradiated Fission Product Material (FPM) capsule waste from molybdenum-99 production

Abstract Irradiated FPM Capsule waste is one of the types of radioactive waste generated from Molybdenum-99 (99Mo) production. This 99Mo is a 99mTc parent radionuclide that is used for the diagnosis of human organ disorders. In its production, the capsule is used for the emplacement of uranium target for irradiation in the GA Siwabessy Reactor. Since the beginning of production until now, the irradiated FPM capsules waste are still being stored in the hot cells because no data on their characteristics and handling systems that are necessary for their treatment. However, long-term storage of the FPM capsules waste in the hot cells shall be avoided because it can disrupt the production process of 99Mo and pose radiation risk to workers. The objectives of this research are to characterize radionuclides composition in the FPM capsule waste and identify its appropriate handling system. The radionuclides characterization was carried out using ORIGEN 2.1 software whereas the selection of the handling system was based on the suitability or modification of the available facilities at Radioactive Waste Management Installation (RWMI)-BRIN. The characterization results show that up to a decay time of 50 years, an FPM capsule waste contains activation product radionuclides such as 55Fe, 59Ni, and 63Ni with a total specific activity of 7.11x106 Bq/g. Three FPM capsule waste handling systems were identified that can be implemented at the RWMI facility, namely by using a 350 -liter concrete shell and then storing in an interim storage for low and intermediate-level waste, using a 60-liter stainless-steel canister which is then stored in the interim storage pit for high-level waste, or by modifying the interim storage pit for high-level waste. Radionuclide characterization and appropriate handling systems are required for the safe management of irradiated FPM capsule waste. This FPM capsule waste management is important for human and environmental protection.

Analysis of molybdenum-99 production at the WWR-K research reactor

Analysis of molybdenum-99 production at the WWR-K research reactor

Monte Carlo simulation of Mo-99 isotope production with increased specific activity in WWR-K reactor

The activation method used for the production of molybdenum-99 in the WWR-K reactor currently allows the maximum production of the specific activity of molybdenum-99 equal to (85 ± 11) GBq/g Mo as a result of seven-day neutron irradiation of natural molybdenum (VI) oxide. It is known that the reaction 98Mo(n,γ)99Mo is induced in the thermal region, and there are also large resonances in the energy region of about 12 eV and in the range of 0.4–10 keV, and these resonances do not compete with the radiative absorption reactions of other molybdenum isotopes in this area. Thus, an increase in the proportion of epithermal and thermal neutrons will lead to an increase in the production of molybdenum-99. The purpose of this work is to develop the design of an irradiation capsule that creates specified irradiation conditions. For this purpose, various designs of irradiation capsules with different materials that moderate neutrons were considered by calculation. Based on the results of Monte Carlo simulation, the optimal design of the irradiation capsule was selected, which makes it possible to increase the production of specific activity of molybdenum-99. Irradiating trioxide molybdenum with natural composition allows to increase the specific activity by 26%, while using trioxide molybdenum with a 98.7% enrichment molybdenum-98 – by 25%. The paper provides a description of the capsule designs considered and the results of computational modeling with MCNP code for molybdenum of different enrichments.

Conceptual design of hybrid target for molybdenum-99 production based on heavywater

Molybdenum-99 (99Mo) is used for preparing Technetium-99 m (99mTc), which is the most widely used isotope in nuclear medicine. In this work, a study for 99Mo production based on a high-power electron accelerator has been performed as an alternative approach to produce 99mTc. In this study, Monte Carlo MCNPX2.6 code has been used to examine a novel idea of simultaneous hybrid production of 99Mo via both photoneutron and neutron capture reactions using an electron accelerator in heavy water tank. It is expected that this conceptual design including an arrangement of metallic plates of 100Mo and 98Mo produces total activity of 97.5 Ci at the end of 20-h continuous e-beam irradiation (30 MeV, 10 mA).

Structural integrity evaluation of irradiated LEU targets for the production of molybdenum-99 using thermo-mechanical behavior simulation coupled with pressure of fission gas release calculation

Structural integrity evaluation of irradiated LEU targets for the production of molybdenum-99 using thermo-mechanical behavior simulation coupled with pressure of fission gas release calculation

Future development of global molybdenum-99 production and saving of atmospheric radioxenon emissions by using nuclear fusion-based approaches

Technetium-99m, the decay product of molybdenum-99, is the most used medical isotope in diagnostic imaging. The future disruptions of molybdenum-99 supply, due to the final shut down of some old producing reactors, has led some current global supplies to plan the expansion of their production capacity. While other countries are developing own production facilities to supply their domestic demand. The global increase of molybdenum-99 production in the coming years could increase by about five times the current demand, with about the 50 percent of additional production in North America. Xenon radionuclides are an inevitable by-product of the nuclear plants production, and their periodically release into the atmosphere, contribute to the background that is also revealed by the IMS stations of the CBTO treaty. In this framework, the development of new technologies, posing no risk in relation to nuclear proliferation and do not result in emissions of radioxenon, could mitigate the issues related to the forecast increase of molybdenum-99 production worldwide. In Italy, an alternative 99Mo production project, the project ENEA Sorgentina, based on the irradiation of molybdenum by neutrons produced by a deuterium-tritium nuclear fusion process, is under development. This facility will not release radioxenon into the atmosphere, so it will not affect the background value in the atmosphere in Southern Europe.

PHWR Reactivity Device Incremental Macroscopic Cross Sections and Reactivities for a Molybdenum-Producing Bundle and a Standard Bundle

A new 37-element PHWR fuel bundle, designed for molybdenum-99 production, has been proposed previously. The new bundle has been shown to have lattice properties and reactivity feedback effects equivalent to the standard PHWR bundle. This study looks at the effect the use of molybdenum-99-producing bundles has on the reactivity worth of reactivity devices, through the prism of reactivity-device macroscopic-cross-section increments. The study utilizes three-dimensional supercell configurations and the neutron transport code DRAGON to calculate and compare the incremental macroscopic cross sections and supercell reactivity for adjuster absorbers, shutoff absorber rods and liquid zone controllers when surrounded by molybdenum-99-producing bundles and by regular bundles. Two geometrical representations of fuel bundles are used: a detailed, cluster, representation, whereby all fuel pins are modeled separately, and an annularized representation, whereby each ring of fuel pins and corresponding coolant is represented as a homogeneous annulus. The latter model is the one customarily used in production calculations for finding cross-section increments of reactivity devices. The study finds that reactivity-device cross-section and supercell reactivity increments are very similar (< 2% difference in reactivity increments) for the case of the molybdenum-producing bundle and the regular bundle. The study also finds that the use of a detailed, cluster, geometrical representation of the fuel bundle produces slightly different cross-section increments and supercell reactivity increments than the use of an annularized geometrical representation. The supercell reactivity-increment difference between the two representations is found to be ~8.0% for adjuster absorbers and ~11.0% for shutoff absorber rods.

Pyrochlore glass-ceramics for the immobilization of molybdenum-99 production wastes: Demonstrating scalability and flexibility to waste stream variance

Pyrochlore glass ceramics have been fabricated via in-situ crystallization under reducing conditions by both sintering and hot isostatic pressing (HIPing) as candidate wasteforms for the acidic waste biproduct of Mo-99 radiopharmaceutical production. The tailored wasteform demonstrates flexibility in the wasteform design to receive the required waste variability, it also suitably passes high-level waste performance requirement, and successfully scales to 1 kg scale with 45 wt.% waste loading. U-rich pyrochlore as the major phase was confirmed by X-ray diffraction, scanning electron microscopy and energy dispersive X-ray spectroscopy, with residual glass and minor secondary phases. The presence of both U4+ and U5+ valences in the wasteforms was revealed by diffuse reflectance spectroscopy. Addition of glass content had little influence on the pyrochlore composition but facilitated minor perovskite formation. The up-scaled dense, HIPed sample showed elemental releases of < 2 g/L for all elements in durability experiments.

Direct determination of aluminum in low-enriched UAlx targets (UAlx-Al) by Inductively Coupled Plasma Optical Emission Spectrometry

The production of molybdenum-99 (99Mo) using low-enriched uranium targets (&lt; 20% 235U) dispersed in aluminum (UAlx) is a very promising strategy towards the independence in 99Mo local production. A thorough control must be performed to ensure that these targets meet the regulatory requirements to achieve the expected efficiency in the reactor. The determination of the targets’ composition is of high interest, because the distribution of Al in different phases may have an impact on the U concentration. Among the techniques used for this purpose, inductively coupled plasma optical emission spectrometry (ICP OES) stands out because of its high sensitivity and precision, allowing for simultaneous determination of several elements in a variety of samples and matrices. However, because U exhibits a complex emission spectrum, spectral interferences are prone to affect the analysis of Al, calling for time consuming preparation steps to remove the U from the matrix. This study proposes a method of direct determination of Al in UAlx targets through the selection of specific emission lines enabled by the evaluation of the associated interferences on the recovery values.

Simulating the Production of Medical Radioisotopes in a Fast Thorium-Cycle ADS with SERPENT

Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced continuously and transported as quickly as possible to the medical units where they are used. Neutron-rich medical radioisotopes are generally produced in research reactors, like technetium-99m, lutetium-177, holmium-166 and iodine-131. On the other hand, proton-rich radioisotopes are produced via reactions with charged particles from accelerators like fluorine-18, gallium-67, iodine-123 and thallium-201. Beside this, innovative nuclear reactors are advocated as solutions to the issues of nuclear waste production and proliferation threats. Fast neutron, thorium-cycle and accelerator-driven subcritical (ADS) reactors are some of the most promising of them, proposed as safer fuel breeders and “waste burners”. This article examines the use of a fast thorium-cycle ADS with liquid lead-bismuth eutectic coolant for the production of molybdenum-99/technetium-99m and lutetium-177. Burnup simulation has been made with the Monte-Carlo (MC) code SERPENT. It is demonstrated that MC codes can advantageously be used to determine the optimal irradiation time for a given radioisotope in a realistic reactor core. It is also shown that fast thorium-cycle ADS is an economical option for the production of medical radioisotopes.

Neutronic analysis of DECY-13 cyclotron target system as a neutron source for SAMOP

The Centre for Accelerator Science and Technology (PSTA) has developed a subcritical reactor for the molybdenum-99 (99Mo) production (SAMOP). This device requires an external neutron source. For experimental purposes, SAMOP is still driven by an external neutron source from a critical reactor, i.e. from the radial beam-port of the Kartini reactor. PSTA is also developing a 13 MeV cyclotron (DECY-13) for the production of fluorine-18 (18F) which can generate neutron as by-products. This cyclotron has an opportunity to be used as an external neutron source for the SAMOP. In this work, particle transport simulations have been carried out to determine the characteristics of the neutron produced by the DECY-13. The simulation results show that DECY-13 produces an average neutron flux of 2.6347 × 109 particles/cm2s and around 97% of them are the fast neutron. Based on the analysis, it concluded that the DECY 13 could be used as an external neutron source for the SAMOP by the addition of a neutron collimator.

Safety Analysis of Gas Pressure in a Subcritical Assembly for Molybdenum-99 Production System

The Subcritical Assembly for Molybdenum-99 Production system is a subcritical system fueled by uranium nitrate, which utilizes the Kartini reactor’s beam port as the neutron source. One of the problems in using uranium nitrate fuel involves the radiolysis reactions and gaseous fission products that form in the cavity above the Subcritical Assembly for Molybdenum-99 Production fuel tube, resulting in a buildup of pressure. To address this issue, this study examined the total accumulated gas pressure in each Subcritical Assembly for Molybdenum-99 Production tube contributed by gaseous fission products and water radiolysis by neutron and gamma radiation during 7 days of operation. Examinations were performed by combining the Subcritical Assembly for Molybdenum-99 Production and Kartini reactor geometry to obtain the burnup power using a tally within the Monte Carlo N-Particle eXtended code. Subcritical Assembly for Molybdenum-99 Production system was then simulated for 7 days with the obtained burnup power with the same code. Outputs from the code were then calculated and analyzed to determine the total accumulated pressure on each fuel tube from each of the pressure contributors. This research showed that the maximum accumulated pressures were 0.45 atm and 0.5 atm for Kartini reactor’s power of 100 kW and 110 kW, respectively. These pressures are lower than the atmospheric pressure; hence, the current Subcritical Assembly for Molybdenum-99 Production system can be operated safely for 7 days.

Molybdenum-99 production via lead and bismuth moderators and milli-structure-98Mo samples by indirect production technique using Monte Carlo method

Molybdenum-99 production via lead and bismuth moderators and milli-structure-⁹⁸Mo samples by indirect production technique using Monte Carlo method

Molybdenum-99 production via lead and bismuth moderators and milli-structure- 98Mo samples by the indirect production technique using the Monte Carlo method

We describe the design of a neutron activator that contains a beryllium target, a lead/bismuth moderator, graphite/PbF2 reflectors, and a boron absorber. Production of the 99Mo isotope using millimeter-sized 98Mo samples located in various regions of the activator was simulated using the MCNPX code. If the 98Mo samples are placed in the reflector region, the yield increases due to a larger flux of epithermal neutrons.

Synthesis and Characterization of Molybdenum Phthalocyanine as Target Material for High Specific Activity Molybdenum-99 Production

High specific activity is a necessity in the fabrication of 99Mo/99mTc radioisotope generators. Recoil reaction, or the Szilard-Chalmers effect, is a method that could be used as an alternative method for increasing specific activity in radioisotope production in light of tightening regulation of highly enriched uranium (HEU) irradiation. Phthalocyanine compounds are usually used as the target material in recoil reactions for the production of high specific radioisotope activity via the (n,γ) reaction. Molybdenum phthalocyanine (Mo-Pc) could be a promising target material in recoil reactions for producing high specific activity of 99Mo. Mo-Pc was synthesized via solid-state reaction between ammonium heptamolybdate and phthalonitrile in a reflux system at 300 °C for 3 h. This optimum condition was identified after performing several variations of temperature and time of reaction, considering FTIR spectra, the yield of product and melting point of the product. XRD measurement showed that Mo-Pc synthesized at optimum condition was free from MoO2, phthalimide and unreacted molybdenum. Mo-Pc has UV-vis properties of Q-band absorption between 600 and 750 nm when dissolved in tetrahydrofuran, dimethylformamide and trifluoroacetic acid. Splitting at absorption peak in tetrahydrofuran and dimethylformamide solution indicated that protonation had occurred. This split peak did not appear in a trifluoroacetic acid solution. In the preliminary study of irradiation of 1 g Mo-Pc at 3.5x1012 n cm–2 s–1 neutron flux, followed by dissolution in tetrahydrofuran and extraction of Mo-99 into NaOH, we obtained Mo-99 solution with a specific activity of 682.35 mCi/g Mo, this being 254.61 times higher than in the regular MoO3 target.

Influence of GA Siwabessy Reactor Irradiation Period on The Molybdenum-99 (99Mo) Production by Neutron Activation of Natural Molybdenum to Produce Technetium-99m (99mTc)

Production of 99Mo by neutron activation of natural Mo in multipurpose reactor GA Siwabessy is an alternative solution to overcome the 99Mo shortage, particularly in Indonesia. The aim of this study is to evaluate the influence of the irradiation period of the reactor on the quality of the produced 99Mo and 99mTc. A natural molybdenum was packed in quartz ampule and aluminium capsule, irradiated in the research reactor for around 100 hours. The 99Mo - 99mTc separation was conducted in Center for Radioisotope and Radiopharmaceutical Technology using zirconium-based material (ZBM). The observed parameters are 99Mo activity, 99mTc yield percentage, adsorption capacity ZBM and 99mTc quality. Both the obtained 99Mo activity and 99mTc yield percentage were influenced by the irradiation period. On the other hand, neither the adsorption capacity of ZBM nor quality parameters of 99mTc were influenced by the irradiation period.

Numerical analysis of isotope production in molten salt reactors: A case study for Molybdenum-99 production

With the aging of current isotope-producing reactors and the rising demand for medical isotopes, future shortages are predicted to greatly increase the cost of Mo-99. Molten salt reactors have been proposed for the production of Mo-99, and have the potential for safe, efficient operation at a variety of scales. However, while Mo-99 is a major fission product, it is unclear how accessible it will be for direct extraction from a fuel salt given its relatively short half-life. In this study, a computationally efficient mathematical model was used to analyze fission product accumulation in an MSR, with the specific goal of assessing Mo-99 production by continuous or near-continuous extraction from the reactor. The model includes generation and removal of isotopes in an MSR from fission, decay, activation and extraction. The extraction process was assumed to be mass-transfer limited, owing to the noble behavior and relatively low concentration of molybdenum in the reactor. It was found that the concentration at which Mo was present in the reactor had a profound effect on the specific activity of the Mo-99 that could be successfully extracted. If molybdenum is allowed to buildup in the reactor, the majority of Mo-99 is lost by decay or activation, leaving only a small fraction of Mo-99 in the extracted molybdenum. Therefore, separation processes that require a high relative molybdenum concentration for extraction are not desirable. On the other hand, separation processes become more difficult at low concentrations. Thus, a trade-off exists between the difficulty of the separation and effectiveness of the extraction as measured by the recovery fraction and product quality. It was found that increasing the power density enabled improved Mo-99 recovery for separation processes at increased concentrations. These results demonstrate the use of a model to explore the viability of Mo-99 extraction from an MSR, the influence of operating parameters on that extraction, and the conditions under which Mo-99 extraction can be optimized.

Criticality analysis of Kartini Reactor connected to test facility of subcritical assembly Molybdenum-99 production (SAMOP)

Criticality Analysis of the Kartini Reactor as an external neutron source for the test facility of the Subcritical Assembly Molybdenum-99 Production (SAMOP) connected via its radial beamport is presented using The Monte Carlo N-Particle Six Version (MCNP6) software. The reactor was simulated at 100 kW operating power and the fuel of the SAMOP was a solution of uranyl nitrate with a concentration of 300 gU/L. The criticality values obtained from this research was 1.00749 ± 0.00045 and 0.99 ± 0,00041 for the Kartini and SAMOP respectively . When the both reactors were connected, criticality value was 1.01536 ± 0.00009. It could be concluded that the Kartini reactor remains critical although it was used as the neutron source for the SAMOP facility.

Isolation of molybdenum(vi) from simulated leachates of irradiated uranium-aluminum targets using diluted and undiluted sulfate ionic liquids

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.

Photoneutronic aspects of the molybdenum-99 production by means of electron linear accelerators

Monte Carlo simulations were used to study 99Mo photonuclear production. Simulation showed that direct production of 99Mo in the cylindrical 100Mo target provides higher values of 99Mo specific activity compared to production using separate bremsstrahlung converter. For 50 MeV 100 kW electron beam we obtained the maximum value of 61 Ci/g specific activity of produced 99Mo at saturation. Calculations for the targets with optimized shape showed that even higher levels of specific activity are achievable: for optimized target the calculated value of 99Mo specific activity at saturation exceeds values of 145.89 Ci/g for setup with bremsstrahlung converter and 170.07 Ci/g without the converter.