99Mo Production Research Articles

Efficacious selective separation of U(VI) over Mo(VI) using novel 2,9-diamide-1,10-phenanthroline ligands: Liquid-liquid extraction and coordination chemistry

Uranium and molybdenum are important strategic elements. The production of 99Mo and the hydrometallurgical process of uranium ore face difficult problems of separation of uranium and molybdenum. In this study, the four phenanthroline diamide ligands were synthesized, and extraction and stripping experiments were performed under different conditions to evaluate the potential application of these ligands for separation of U(VI) over Mo(VI). With the growth of alkyl chain, the solubility of ligands could be greatly improved, and the separation effect of U(VI) over Mo(VI) gradually increased. The SFU/Mo were around 10,000 at 4 mol/L HNO3. Three stripping agents were tested with the stripping efficiency of Na2CO3 (5%) > H2O > HNO3 (0.01 mol/L). The stripping percentages of the three stripping agents were all close to unity, indicating that the ligands had the potential to be recycled. The chemical stoichiometry of U(VI) complexes with ligands was evaluated as 1:1 using electrospray ionization mass spectrometry, ultraviolet visible spectroscopy and single-crystal X-ray diffraction. The consistency between theoretical calculation and experimental results further explains the coordination mechanism.

Neutronics analysis of a subcritical blanket system driven by a gas dynamic trap-based fusion neutron source for 99Mo production

Neutronics analysis of a subcritical blanket system driven by a gas dynamic trap-based fusion neutron source for 99Mo production

Investigating 99Mo output changes in high-sustainability uranium targets by modifying target volume and geometry

Introduction: The most common current uranium target type for 99Mo production uses plate geometry. This paper investigates the effects of geometry on 99Mo output and target sustainability.Methods: MCNP6.2 was used to model rectangular, spherical, and cylindrical targets ranging from 12.03 cm3 to 120.34 cm3 to determine the target geometry and volume effects on 99Mo output, sustainability, proliferation concerns, and heating. 4-7 days irradiation was used with a consistent target density of 2 g/cm3 UO2 for all target types.Results and Discussion: The cylindrical target was found to have the best performance due to having the second highest sustainability, the highest 99Mo output and produce the lowest amounts of 239Pu compared to the other target geometries. The heating comparison showed that there were negligible heating concerns for all target volumes and geometries.

Neutronic simulation of medical radioisotope 99Mo and 177Lu production in IPR 14 MeV neutron generator facility

An accelerator based 14 MeV neutron generator is commissioned at Institute for Plasma Research India. The generator is based on the linear accelerator concept where the deuterium ion beam impinged to the tritium target to produce neutrons. The generator is designed to produce 5 × 1012 n/s (max.). The 14 MeV neutron source facilities are an emerging tool for the lab scale experiments and research. In order to utilize the generator for the welfare of humanity, the assessment is made for the production of medical radioisotopes using the neutron facility.The usage of radioisotopes in the treatment and diagnosis of a disease is an important factor in the healthcare sector. A series of calculations are conducted to generate radioisotopes, especially 99Mo and 177Lu those are having huge applications in the medical and pharmaceutical industries. 99Mo can be also generated through neutron reactions 98Mo(n, g)99Mo and 100Mo(n, 2n)99Mo apart from fission reaction. The cross section of 98Mo(n, g)99Mo is high in the thermal energy range whereas 100Mo(n,2n)99Mo occurs at a high energy range. 177Lu can be produced using the reactions 176Lu (n, g)177Lu and 176Yb (n, g)177Yb. The cross section of both 177Lu production routes is higher at thermal energy range. The neutron flux level near the target is around 1010 cm−2s−1. In order to enhance production capabilities, the neutron energy spectrum moderators are used to thermalize the neutrons. The materials used as a moderator are beryllium, HDPE, graphite, etc. Moderators enhance the capabilities of medical isotope production in neutron generators.

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).

Investigating an alternative sustainable low-enriched uranium target for the manufacture of 99Mo using MCNP6.2 modelling with CINDER90

Introduction: Current fission-based methods of 99Mo production use targets that are based on high output without taking into consideration the amount of radioactive waste produced. We examine the idea of using a low enriched target (<20%) to reduce the amount of nuclear and chemical wastes produced by the manufacture of 99Mo via fission of uranium-based targets.Methods: MCNP6.2 was used to model targets of 20%, 10%, 7%, 3%, and 1% enrichment for 99Mo output and sustainability.Results and Discussion: The 1% enriched target at the lowest density of 0.2 g/cm3 of UO2 was found to have the highest sustainability score at 6 days irradiation but it had a low 99Mo output. On the other hand, the highest output target was found to be 20% enriched with a density of 8.0 g UO2/cm3 with an irradiation time of 20 days. Target security and safeguards concerns were modelled and found to be of minimal concern.

US nonproliferation policies and Canada's medical-isotope industry: a case study in nuclear ambivalence

ABSTRACT This article analyzes how US nonproliferation policies that sought to curtail US exports of highly enriched uranium (HEU) affected Canada’s medical-isotope industry and, particularly, the Canadian MAPLE (Multipurpose Applied Physics Lattice Experiment) reactors project. US HEU-export policies established between 1978 and 2012 highlight the “nuclear ambivalence” of the Canadian isotope industry’s flagship product, molybdenum-99 (Mo-99)—a life-saving commodity widely used by US hospitals but also a material whose production process, based on HEU, came to be perceived as a potential threat to US and world security. The ambivalent status of Mo-99 production was reinforced by a state of mutual dependence between the two countries: On the one hand, Canada depended entirely on US HEU exports to maintain its dominant position in the Mo-99 world market and could not turn to other sources of HEU supply. On the other hand, US hospitals relied mainly on Mo-99 of Canadian origin, which limited the US government’s ability to enforce its policy by suspending its HEU exports to Canada. As a result, US and Canadian efforts to convert the Canadian isotope facilities to low-enriched uranium were thwarted by tensions between global security, public health, and commercial stakes, which led ultimately to ending Canada’s Mo-99 production.

Molecular dynamics simulations of Mo nanoparticles sputtering under irradiation

99Mo is an essential isotope in nuclear medicine, but the nuclear reactors used for their production reaching their end of life, problems of supply arise and new methods of production need to be considered. Here we study the possibility of using gamma and neutron irradiation of Mo nanoparticles (NPs) in suspension and use the separation of the isotopes escaping the nanoparticle by primary recoil to evaluate the efficiency of the process for 99Mo production. Molecular Dynamics simulations with empirical potential of Ackland and Thetford were used to obtain information about the 99Mo escape yield from the NPs and the resulting sputtering depending on NPs sizes and recoil energy. Results show that the best yield is obtained for 5 nm NPs irradiated with gamma particles. These results are used to guide accelerator irradiation experiments led in parallel in order to evaluate the possibility of using accelerators instead of nuclear reactors for the production of 99Mo.

Investigating Sustainability Index, 99Mo Output and 239Pu Levels in UO2 Targets by Substituting 238U with Ce

A new target material combination was modelled to replace the existing uranium-aluminium design used for 99Mo manufacture to increase the sustainability of the production process. Previous efforts to develop a more sustainable uranium target for 99Mo production, resulted in the levels of 239Pu in the target after irradiation being elevated due to the increase in 238U present. MCNP6.2 was used to model 4 different cylindrical targets based on 4–7 days irradiation to further understand this effect. To reduce the resultant 239Pu levels, ratios of 0–99% of Ce were used as a replacement for 238U. The results show that the addition of 140Ce and the removal of 238U reduced the 239Pu levels in the target significantly thus increasing the sustainability of the target and giving a slight increase to the 99Mo output of the targets.

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.

New strategies for a sustainable 99mTc supply to meet increasing medical demands: Promising solutions for current problems.

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.

Possibility for using a low-enriched target to produce 99Mo in the MAK-2 research channel of the VVR-ts reactor

Thermal-hydraulic calculations have been conducted with respect to the active part of the MAK-2 loop facility of the VVR-ts research reactor for the 99Mo production. The computational studies were undertaken both for the case of using a highly 235U enriched target and for a low-enriched target. The calculation was performed for the actual technical characteristics of the research channel. The power density for the two simulated cases was obtained in the course of a preliminary neutronic calculation and selected for the most heated channel. The problem is solved for the steady-state mode of the channel coolant flow and takes into account the dependence of the thermophysical parameters of materials on temperature. The volumetric temperature distribution in the channel was obtained in the process of the calculation. The calculation results present the maximum temperatures of the target materials for the 99Mo production. An analysis of the obtained results has shown that the maximum temperatures of the aluminum sleeve and the target filling materials do not exceed the critical values. For the analyzed calculation cases, the maximum coolant temperature is localized at a point near the sleeve wall surface and does not reach the boiling temperature for a given pressure. The study has therefore shown that it is possible to reduce the 235U enrichment of the target filling fissile material to 19.7%, provided the average density of the mixture and the amount of 235U in the target remain the same. At the same time, the amount of the medicinally important 99Mo generated will not practically change, which will lead to reduced capital costs for a highly enriched mixture of the target matrix.

Pilot Scale Validation of a Chemical Process for Uranium, Cesium, and Mercury Recovery from Cemented Radioactive Wastes

The medical isotope (Mo-99) production at Chalk River Laboratory involves the dissolution of irradiated isotope targets prior to the extraction of Mo-99. This process generates a waste that is cemented in 5-gallon containers and transferred to a waste-management facility for intermediate storage. Over the past decades, a large number of five-gallon containers of cemented radioactive waste (CRW) were produced, and Canadian Nuclear Laboratories (CNL) must develop a process to convert this material to a permanent waste form. Research has been undertaken to develop an innovative method for the recovery of U, Hg, and Cs from surrogate cemented radioactive waste (SCRW). This paper presents the pilot scale validation of the operating parameters prior to the demonstration scale testing. Leaching tests at the pilot scale were conducted with 5 kg of SCRW to validate the main operating parameters and evaluate the reuse of the leaching solution. The mean solubilization yields obtained at the pilot scale were 81.9 ± 8.3% for Cs, 99.0 ± 1% for U, and 94.9 ± 4.5% for Hg. Columns with 100 g of KNiFC-PAN and 250 g of Lewatit TP214 allow for the separation of Cs and Hg from 60 L of leaching solution without U loss. Flow rates of 12.5 BV/h and 25 BV/h were suitable to achieve 99% separation of Hg and Cs, respectively. For the Hg resin, the capacity reached 23.4 mg/g, and the capacity for the Cs resin reached 0.79 mg/g. The pilot scale U extraction results showed that the U adsorption is selective, with a breakthrough at 36 BV (capacity for U of 3.70 mg/g). Uranium elution with 1 M Na2CO3 exceeded 99%, and subsequent precipitation with NaOH achieved 99% recovery. SEM data confirmed the high purity of the U solids produced as sodium di-uranate.

GEANT4 based simulation study of converter and direct target design and optimization of target for 99Mo production using 30 MeV electron linear accelerator

A 30 MeV, 5–10 kW beam power electron linac is under development at SAMEER for radio-isotope production. An enriched 100Mo target will be irradiated to produce 99Mo which will be eluted to give 99mTc. Two approaches of 100Mo target irradiation -converter and direct are studied using GEANT4 simulations. The main aim of this study is to optimize the 100Mo target geometry for maximum yield. It is observed that direct target irradiation gives better activity as compared to converter target approach.

Evaluation of cementation of intermediate level liquid waste produced from fission 99Mo production process and disposal feasibility of cement waste form

The Korea Atomic Energy Research Institute (KAERI) is planning the construction of the KIJANG Research Reactor (KJRR) for stable supply of 99Mo. The Fission 99Mo Production Process (FMPP) of KJRR produces solid waste such as spent uranium cake and alumina cake, and liquid waste in the form of intermediate level liquid waste (ILLW) and low level liquid waste (LLLW). This study thus established the operating range and optimum operating conditions for the cementation of ILLW from FMPP. It also evaluated whether cement waste form samples produced under optimum operational conditions satisfy the waste acceptance criteria (WAC) of a disposal facility in Korea (Korea radioactive waste agency, KORAD). Considering economic feasibility and safety, optimum operational conditions were achieved at a w/c ratio of 0.55, and the corresponding salt content was 5.71 wt%. The cement waste form samples prepared under optimum operational conditions were found to satisfy KORAD's WAC when tested for structural stability and leachability. The results indicate that the proposed cementation conditions for the disposal of ILLW from FMMP can be effectively applied to KJRR's disposal facility.

Increasing the production of the Mo-99 isotope by modernizing the design of targets irradiated in the experimental channels of the VVR-c reactor

There are various ways to obtain Mo-99. Some of them are widely used in industrial production, others are in the research stage with the aim of increasing the product yield. The main industrial method for obtaining Mo-99 using a nuclear reactor is the fragmentation method. This method provides for the presence of a uranium target and a nuclear reactor. The target is placed in the channel of the reactor core and irradiated with neutrons for the required time. After that, the target is removed from the channel to the “hot” chamber for the chemical separation of Mo-99. This is how Mo-99 is obtained practically all over the world. The paper considers the fragmentation method for producing Mo-99, which is implemented on the basis of the engineering and technological complex of the VVR-c research nuclear reactor. In order to increase the yield of Mo-99, a modernized model of the “tube-in-tube” target is proposed. The assessment of the production of Mo-99 and the cooling efficiency of the modernized target was carried out. The calculations were performed using the VisualBurnOut and Ansys CFX software packages. Computational studies have shown an increase in the energy release and the amount of the produced Mo-99 isotope in the target of the modernized design. In the most stressed zones, the target wall temperature exceeds the water saturation temperature. Surface boiling occurs in these zones. As a result, turbulization and mixing of the near-wall boundary water layer increases. This improves heat dissipation.

Radioactive Fission Waste from Molybdenum-99 Production and Proliferation Risks

Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

Neutronics analysis of a stacked structure for a subcritical system with LEU solution driven by a D-T neutron source for 99Mo production

Neutronics analysis of a stacked structure for a subcritical system with LEU solution driven by a D-T neutron source for 99Mo production

Porous activated carbon monoliths as a novel target material for the production of 99Mo by fission

Porous activated carbon monoliths as a novel target material for the production of 99Mo by fission

Theoretical insights into the substitution effect of phenanthroline derivative ligands on the extraction of Mo (VI)

With the rapid development of nuclear medical imaging, the production of 99Mo has attracted much attention, because the important medical isotope 99mTc can be obtained from 99Mo. N,N′-diethyl-N,N′-ditolyl-2,9-diamide1,10-phenanthroline (Et-Tol-DAPhen) has been proven to be an excellent ligand for the extraction of U(VI), while there are still few reports on the extraction of Mo(VI) using this ligand. To investigate the coordination structures of MoO22+ with Et-Tol-DAPhen, we carried out theoretical calculations using scalar relativistic density functional theory (DFT). The analyses of Wiberg indices (WBIs), quantum theory of atoms in molecules (QTAIM) and natural orbitals of chemical valence (NOCV) indicate that the MoO22+ have stronger complexation ability with O atoms compared to N atoms. The bonds between MoO22+ ions and ligands are mainly ionic interactions. The analyses of EDA and the most negative binding energy indicates that complex [MoO2L(NO3)]+ is energetically favorable and Et-Tol-DAPhen has also good extraction ability for MoO22+. Moreover, although both the electron-donating and electron-withdrawing groups have slightly effect on the electronic structures of the [MoO2L(NO3)]+ complexes, Et-Tol-DAPhen modified by the electron-donating group can enhance the extraction ability of MoO22+. This study is helpful to understand the complexation behavior of Et-Tol-DAPhen with MoO22+ and provides useful structural and thermodynamic information for the MoO22+ extraction with phenanthroline derivative ligands.