Mo Yield Research Articles

Refueling Considerations for 99Mo Production in a CANDU Reactor

Operating CANDU reactors have the potential to produce significant quantities of molybdenum-99 (99Mo) because of their ability to be refueled online, high thermal neutron flux, and fuel design flexibility. A new molybdenum-producing fuel bundle (MPB), previously designed for CANDU reactors, has as its principal attribute that it is neutronically and thermal hydraulically equivalent to the standard 37-element fuel bundle typically used in CANDU reactors. Given that the typical irradiation time for MPBs is 20 days while the typical refueling period for a channel is on average 6 months, the refueling strategy needs to be adjusted to accommodate the shorter irradiation time of MPBs. This study evaluates a new refueling strategy suitable for employing the new MPBs in the core. A full-core, three-dimensional model is constructed in the diffusion code DONJON, and a fueling strategy for achieving the desired weekly yield of 99Mo is developed. The adequacy of the proposed refueling scheme is evaluated using a series of time-average calculations, which show that a small increase in the core reactivity (<0.4 mk) can be expected when irradiating a set of four MPBs in three different fuel channels in the inner region of the core. The small increase in the core reactivity can be managed by slightly increasing the discharge burnup in the non-MPB-bearing fuel channels, thus also improving slightly the fuel utilization in the reactor.

Optimization of target system for the production of 99Mo via 100Mo(γ,n)99Mo reaction

With an increase of stopping operation of nuclear reactors worldwide, the supply of medical 99Mo becomes difficult and thus many efforts have been made to find an alternative. A process based on an electron linear accelerator (linac) system and a100Mo target via the 100Mo (γ,n)99Mo reaction receives a lot of attention due to the relatively low level of co-produced impurities. This process has been recently developed at the Institute of Modern Physics (IMP) and the Monte Carlo simulation was used to optimize the target system before operating pilot irradiation experiments. First, tungsten and tantalum, as mostly used converter materials, were tested. The yield of 99Mo was evaluated with respect to the converter thickness and the electron beam energy by means of Geant4 simulations. Besides, the specific activity of 99Mo produced from one-stage approach (100Mo target without a converter) and two-stage approach (100Mo target with a converter) was compared when varying the testing conditions. The two-stage approach was selected for the experiment due to the higher specific activity of produced 99Mo at all tested conditions. A target consisting of a 10 mm thickness of the 100Mo tablets and a 2.4 mm thick Ta converter was irradiated for 40 h (50 MeV with 0.2 μA). The Geant4-calculated specific activity of generated 99Mo at the end of bombardment agreed well with the experimental value, which proved high level of accuracy of the Geant4 simulation. In future studies, the Geant4 simulation will be used to optimize the production process when using high power linac system.

PHOTONUCLEAR PRODUCTION OF 99Mo/ 99mТс-GENERATOR WITH USING EFFECT SZILARD-CHALMERS AND ELECTROLYSIS METHOD

The possibility of photonuclear production of 99Mo medical radioisotope using recoil nuclei of molybdenite nanoparticles from reaction 100Mo(γ,n)99Mo was investigated. The enrichment of radioactive isotopes is carried out by the effect of Szilard-Chalmers. Molybdenite nanoparticles were irradiated by bremsstrahlung with Emax = 39 MeV. The recoil nuclei of 99Mo are separated by electrolysis. The yield of 99Mo from the extractable phase amounted to ~3%. This technology will allow producing 0.8 GBq/g of 99mTc with high specific activity per day. Key words: linear electron accelerator, photonuclear production, electrolysis, specific activity, recoil nuclei, molybdenite nanoparticles.

Neutronics study of a subcritical system driven by external neutron source for 99Mo production

A subcritical system with the Low Enriched Uranium (LEU) solution was proposed to produce 99Mo, which is driven by an external D–T neutron source. The neutronics model of the subcritical system is approximately a concentric cylindrical structure with the maximum radius and height of the system as limited to 100 cm and 150 cm, respectively. The materials and dimensions of the zones were designed based on neutronics analysis including the sub-critical multiplication factor ks and distribution of neutron flux to achieve high production of 99Mo, as well as to keep the system in a safe and deep sub-critical state (i.e. 0.90<ks<0.95). With intensity of neutron source as 5 × 1013 n/s, the yield of 99Mo in the system can reach to 167 Ci/day, which could be used for around ten thousand nuclear medical procedures.

On the contribution of the electronuclear reaction to the photonuclear production of Mo-99 and other radioisotopes

We used Monte Carlo simulations to estimate contribution of the electronuclear (e−,e-'n) reaction to the one-stage approach photonuclear production of 99Mo. Two types of targets were considered, and it was shown that in case of a thick (2 cm) target, electronuclear yield is two orders of magnitude lower than the photonuclear one. However, in case of a thin target electronuclear yield becomes comparable to the photonuclear or even exceeds it. For a 1 mm thick 100Mo target contribution of the 99Mo yield is about 20%, while for a 0.1 mm thick target this contribution exceeds 70%. We have shown that photonuclear yields of the radioisotopes produced in thin targets could be greatly underestimated if electronuclear reactions are not taken into consideration.

Evaluation of 99Mo production in a small modular thorium based molten salt reactor

Producing medical radionuclide 99Mo in molten salt reactors is a very attractive choice to solve its global shortage. In this study, we evaluated the yield of 99Mo in the small modular thorium based molten salt reactor (SM-MSR). Firstly, the fuel burn-up analysis of SM-MSR was carried out by an in-house developed code (MOBAT), which takes the unique characteristics of molten salt reactor into account, and the variation of fission yield of 99Mo with burn-up time was obtained. The minimum value of the fission yield of 99Mo is 1.13 × 10−3 (6-day TBq/MW/s) at approximately 600 equivalent full power day. Then, based on the behavior of noble metals in the fuel salt and the experimental results of MSRE gas sample composition measurements, 99Mo migration probability from the primary loop to the off-gas module with the burn-up time was calculated, and the equilibrium value of the migration probability is obtained as 18.4%. When the load factor of SM-MSR is 0.75 (300 MWth), the annual 99Mo amount of migration to the off-gas module would be 1.96 × 106 (6-day TBq) under the most conservative calculation. Finally, a filter system was added in the off gas module, and 99Mo would be extracted from the off gas. As long as the utilization percentage of 99Mo in the off gas module can reach 0.94%, the global demand under current data could be met, which implies a huge additional economic value for the SM-MSR.

USE OF MOLYBDENITE NANOPARTICLES FOR PHOTONUCLEAR PRODUCTION OF TECHNETIUM-99m

The possibility production of 99Mo radioisotope with using recoil nuclei of molybdenite nanoparticles from reaction 100Mo(γ,n)99Mo was investigated. The use of thermally stable molybdenite allows solving the problem of thermal loads of high-power electron beams. The enrichment of radioactive isotopes carried out by the effect of SzilardChalmers. Molybdenite nanoparticles were irradiated by bremsstrahlung with Emax=39 MeV. The recoil nuclei of 99Mo separated by electrolysis at low concentration. The yield of 99Mo from extractable phase amounted 3% for size nanoparticles 350 nm.

Cyclotron Production of 99Mo Radionuclide by Bombarding Zirconium Targets with Alpha Particle Beams

A possibility of manufacturing 99Mo via a nuclear reaction with 96Zr (α, n) in the R-7 cyclotron of the Tomsk polytechnic university is investigated. The beam energy of α-particles is 27 MeV for the beam current values 13, 15 and 22 μA. The targets are manufactured from natural zirconium shaped as disks measuring 30 mm in diameter and 1.5 mm in thickness. As a result of the experiments the designed yield of 99Mo is found to be on the order of (1.62 ± 0.15) (MBq/(μA·h) for 96Zr targets with 100% enrichment, which closely agrees with the yields reported by other authors. It is shown that a sufficient amount of 99Mo can be output for technetium-99m to be manufactured at a regional level.

Recoil Atom Yield in 100Mo(p, x)99Mo During 28 MeV Proton Irradiation of Nanosize Molybdenum Layers

According to the Szilard–Chalmers effect, 99Mo recoil atoms can be obtained in nuclear reactions and recorded in a collector. Knowledge of the dependence of the yield of 99Mo atoms on the thickness of the molybdenum layer is necessary for their efficient collection. The yield of 99Mo recoil atoms from molybdenum nanolayers in the nuclear reaction 100Mo(p, x)99Mo was measured as a function of the thickness of the nanolayer. Nanolayers of metallic molybdenum were fabricated by magnetron sputtering on sapphire plates. The measurements were performed after the nanolayers were irradiated by 28 MeV protons in the U-150 cyclotron. The yield of 99Mo recoil atoms for 38–205 nm thick nanolayers was 65–8%. It was found that the maximum 99Mo yield obtains with molybdenum layer thickness 80 ± 5 nm. It was found that the free path of 99Mo recoil atoms in native metallic molybdenum is equal to 34 ± 9 nm.

Production cross sections of Mo, Nb and Zr radioisotopes from α-induced reaction on natZr

Cross sections of α-induced reactions on natural zirconium were measured up to 50 MeV using the stacked-foil technique, activation method and high resolution γ-ray spectrometry. The production cross sections of 93m,99Mo, 90g,92m,95g,95m,96Nb and 88,89g,95Zr were determined and compared with other experimental data measured earlier and result of theoretical calculations. The integral thick target yield of 99Mo was deduced from the measured cross section data.

Measurement of 99Mo Yield in 100Mo(p, x) with 30 MeV Proton Irradiation of Multicomponent Submicron Particles

The radionuclide 99Mo was produced and separated in the process of irradiating powdered mixtures of molybdenum compounds and buffer particles by 30 MeV protons. The separation is based on the Szilard–Chalmers effect wherein 99Mo recoil particles are fixed in inert buffer particles. Two types of targets were investigated by using mixtures: soluble molybdenum compound–insoluble buffer and insoluble molybdenum compound–soluble buffer. 99Mo yield equal to 20% with enrichment coefficient 18.3 was obtained by using a target consisting of a mixture of submicron 100MoO3 and Al2O3 particles; the 99Mo yield was equal to 20% and the enrichment coefficient 18.3. For a target with the composition 100MoS2 + KCl, making it possible to separate 99Mo recoil atoms directly into solution, the 99Mo yield was equal to 8.7% and the enrichment coefficient 30.7.

Molybdenum transmutation via 98Mo samples using bismuth/lead neutron moderators

99mTc is the usual medical isotope for imaging in nuclear medicine. The use of accelerators to produce 99Mo is an alternative to nuclear reactors. Here, the neutron production from a new neutron source is investigated to achieve a proper production yield. The induced neutron was simulated from 30 MeV protons on a beryllium target. The Adiabatic Resonance Crossing (ARC) method was evaluated using lead and bismuth moderators, and lead(II) fluoride reflector around a moderator region. The flux and energy of the aggregated neutrons within different regions of the proposed activator were estimated by the MCNPX code. The epithermal flux in the reflector region was greater than that in the moderator region, since the maximum amount was estimated to be when the bismuth moderator was employed. The outcomes demonstrated that the production yield of 99Mo improved using the bismuth material, whereas the 98Mo sample at a distance from the target showed a greater amount by 5.03E + 5 MBq through a 0.1 cm sample thickness. When the sample was positioned in the reflector region, the obtained yield increased taking advantage by the more accumulated flux in the epithermal range. The ARC method can replace conventional reactor-based prototypes of the 99Mo-99mTc generator, or complement them, because it is safe and more accessible.

99Mo Yield Using Large Sample Mass of MoO3 for Sustainable Production of 99Mo

A neutron source from the C(d,n) reaction has the unique capability of producing medical radioisotopes such as 99Mo with a minimum level of radioactive waste. Precise data on the neutron flux are crucial to determine the best conditions for obtaining the maximum yield of 99Mo. The measured yield of 99Mo produced by the 100Mo(n,2n)99Mo reaction from a large sample mass of MoO3 agrees well with the numerical result estimated with the latest neutron data, which are a factor of two larger than the other existing data. This result establishes an important finding for the domestic production of 99Mo: approximately 50% of the demand for 99Mo in Japan could be met using a 100 g 100MoO3 sample mass with a single accelerator of 40 MeV, 2 mA deuteron beams.

Measurement and Estimation of the 99Mo Production Yield by 100Mo(n,2n)99Mo

We, for the first time, measured the yield of 99Mo, the mother nuclide of 99mTc used in nuclear medicine diagnostic procedures, produced by the 100Mo(n,2n)99Mo reaction with accelerator neutrons. The neutrons with a continuous energy spectrum from the thermal energy up to about 40 MeV were provided by the C(d,n) reaction with 40 MeV deuteron beams. It was proved that the 99Mo yield agrees with that estimated by using the latest data on neutrons from the C(d,n) reaction and the evaluated cross section of the 100Mo(n,2n)99Mo reaction given in the Japanese Evaluated Nuclear Data Library. On the basis of the agreement, a systematic calculation was carried out to search for an optimum condition that enables us to produce as much 99Mo as possible with a good 99Mo/100Mo value from an economical point of view. The calculated 99Mo yield from a 150 g 100MoO3 sample indicated that about 30% of the demand for 99Mo in Japan can be met with a single accelerator capable of 40 MeV, 2 mA deuteron beams. Here, by referring...

Investigation of 99Mo potential production via UO2SO4 liquid target irradiation in a 5 MW nuclear research reactor

Abstract The activation method for 99Mo production in comparison to fissionable target irradiation in research reactors is less preferable. Therefore, 99Mo yield using UO2SO4 samples was theoretically investigated. Computational results revealed admirable potential of the liquid samples for 99Mo production. Low-concentrated uranyl sulphate samples could easily be handled by the irradiation box. The sample geometry optimization improves thermal hydraulic conditions and production yield. The optimized geometry including only 0.12 g 235U produced 57Ci99Mo at end-of-irradiation (EOI) with a temperature peak of 72°C during the irradiation.

Computational simulation of natUO2, 232ThO2 and U3O8-Al pills to estimate (p,fission) 99Mo yield in the modeled targets irradiated by CYCLONE30 accelerator

99Mo is important for both therapy and imaging purposes. Accelerator and reactor-based procedures are applied to produce it. Newly proton-fission method has been taken in attention by some research centers. In the present work, computationally investigation of the 99Mo yield in different fissionable targets irradiated by proton was aimed. The results showed UO2 pill target could be efficiently used to produce 11.12Ci/g-U saturation yield of 99Mo using 25MeV proton irradiation of the optimized-dimension target with 70µA current.

Fission-Produced 99Mo Without a Nuclear Reactor.

99Mo, the parent of the widely used medical isotope 99mTc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of 99Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of 99Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The 99Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq 99Mo production run are presented.

Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

99mTc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of 99Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of 99Mo is investigated. A modified 37-element fuel bundle, suitable for the production of 99Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of 99Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium.The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of 99Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated 99Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of 99Mo.

Determination of the Radiochemical Yield of Fission-Fragment 99Mo In VVR-Ts

The parameters of the neutron flux in the experimental channels of VVR-Ts are determined. It is shown that to evaluate the 99Mo yield it is necessary to take account of, aside from the neutron flux, the dependence of the fission cross section of 235U on the temperature of the neutron gas. The blocking coefficient is calculated for an experimental uranium target. The computed radiochemical yield of 99Mo is compared with the experimentally measured value. It is shown that the average yield equals 0.75. A correlation is found between the neutron flux density, which is directly proportional to the power of the VVR-Ts nuclear reactor, at the target location and the radiochemical yield of 99Mo.

Development of Industrial-Scale Fission 99Mo Production Process Using Low Enriched Uranium Target

Abstract Molybdenum-99 (99Mo) is the most important isotope because its daughter isotope, technetium-99m (99mTc), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of 99Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of 99Mo technology developments. Most of the industrial-scale 99Mo processes have been based on the fission of 235U. Recently, important issues have been raised for the conversion of fission 99Mo targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of 99Mo yield, caused by a significant reduction of 235U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission 99Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the 99Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.