99Mo Production Research Articles

Excitation functions for the cyclotron production of 99mTc and 99Mo

Excitation functions for the 100Mo(p,2n) 99mTc, 100Mo(p,pn) 99Mo and 98Mo(p,γ) 99mTc nuclear reactions were determined using 97.4 and 99.5% isotopic enrichment of 100Mo and 98Mo, respectively. The irradiations were performed on three different cyclotrons with overlapping data points from 6 to 65 MeV. The optimum energy range for the 100Mo(p,2n) 99mTc reaction is 22→12 MeV with a peak at ∼17 MeV and maximum cross section of ∼200 mb. Over this energy range the production yield of 99mTc amounts to 11.2 mCi (415 MBq)/μA h or 102.8 mCi (3804 MBq)/μA at saturation. There is no serious radionuclidic impurity problem. Production of 99Mo is not viable due to the low cross section (∼130 mb) over the proton energy range of 30 to 50 MeV. The 98Mo(p,γ) 99mTc reaction was found to have a cross section of <0.2 mb over the proton energy range studied. The earlier assignment of this reaction channel as a potential route for production of 99mTc with a medical cyclotron is refuted in this study. Countries that do not have access to fission molybdenum-99 for the 99Mo→ 99mTc generator, could consider use of a >17 MeV proton energy cyclotron for regional production of 99mTc.

97/04693 Decommissioning nuclear facilities—the legal and financial issues: Bindon, J. Energy World, May 1997, 249, 20–21

Technetium-99m, the daughter product of Molybdenum-99 is the most widely needed radionuclide for diagnostic studies in Pakistan. Molybdenum-99 Production Facility has been established at PINSTECH. Highly enriched uranium (93% 235U) U/Al alloy targets have been irradiated in Pakistan Research Reactor-1 (PARR-1) for the generation of fission Mo-99, while basic dissolution technique is used for separation of Mo-99 from target matrix activity. In line with the international objective of minimizing and eventually eliminating the use of HEU in civil commerce, national and international efforts have been underway to shift the production of medical isotopes from HEU to LEU (LEU; <20% 235U enrichment) targets. To achieve the equivalent amount of 99Mo with LEU targets, approximately 5 times uranium is needed. LEU aluminum uranium dispersion target has been developed, which may replace existing HEU aluminum/uranium alloy targets for production of 99Mo using basic dissolution technique. Neutronic and thermal hydraulic calculations were performed for safe irradiation of targets in the core of PARR-1.

Alkaline Peroxide Processing of Low-Enriched Uranium Targets for 99Mo Production—Decomposition of Hydrogen Peroxide

The recent progress on the alkaline peroxide processing of low-enriched uranium targets for the production of 99Mo, a parent nuclide of the widely used medical isotope 99mTc, is reported. Kinetic studies were undertaken to investigate the decomposition of hydrogen peroxide in alkaline solution in contact with a uranium metal surface. It was found that the decomposition of hydrogen peroxide essentially follows the kinetic trend of uranium dissolution and can be classified into two regimes, depending on the hydroxide concentration. In the low-base regime (<0.2 M), both the hydrogen peroxide and hydroxide concentrations affect the rate of peroxide decomposition. In the high-base regime (>0.2 M), the rate of peroxide decomposition is independent of alkali concentration. When the acid/base equilibrium between H2O2 and O2H− is taken into account, the overall rate of hydrogen peroxide disappearance can be described as a 0.25th order reaction with respect to hydrogen peroxide concentration over NaOH concentrations ranging from 0.01 to 5 M. Empirical kinetics models are proposed and discussed.

Dissolution Kinetics of U3Si2 Particles in Alkaline Hydrogen Peroxide

Nonproliferation concerns leading to the conversion from high- to low-enriched uranium sparked interest in U3Si2 dispersion targets as an option for 99Mo production. Dissolution of irradiated targets is an important step in recovering fission-product 99Mo. Alkaline hydrogen peroxide solutions dissolved U3Si2 particles in an open batch reactor; samples were analyzed for total peroxide and uranium concentrations as functions of time and temperature. Dissolution rates are highest at 1 to 1.5 M NaOH and change little for initial base concentrations from 0.5 to 2.5 M NaOH, indicating relatively robust process conditions. Uranium dissolution rates depend most strongly on the equilibrium concentration of the peroxyl ion (O2H−), an equilibrium product of hydrogen peroxide (H2O2) and hydroxyl ion (OH−). Temperature and equilibrium concentrations of O2H− and OH− are included in a uranium dissolution rate model.

Comparison of Characteristics of Solution and Conventional Reactors for 99Mo Production

Nearly all the 99mTc administered to patients is obtained from eluting a radionuclide generator. The generators manufactured by the U.S. radiopharmaceutical companies use only the high-specific activity molybdenum produced by the fission of uranium. The dominant production methods are those used by Cintichem, Inc. and Nordion International. There are, however, competing methods of the production of fission-based 99Mo. One of the most promising proposed alternatives is the use of solution reactors (or homogeneous reactors). The operational characteristics of conventional reactors (i.e., Cintichem process) and those of solution reactors to produce 99Mo for use in manufacturing 99Mo/99mTc generators are examined. The use of conventional reactors has the disadvantage of generating large amounts of radioactive waste. The use of solution reactors can significantly reduce this problem. Both methods require rigorous processing to meet the purity requirements due to the presence of fission product contamination.

Kinetics of Dissolution of Uranium Metal Foil by Alkaline Hydrogen Peroxide

To develop a new process for the production of 99Mo using low-enriched uranium targets, uranium dissolution in alkaline hydrogen peroxide was studied. Molybdenum-99 is a parent of the widely used m...

96/01586 Validation of CATHENA againts the CS28-1 high temperature experiment under an impaired cooling condition

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.

Recovery treatment of nonfissioned uranium in the production of99Mo

An effective modification of the chemical processes has been obtained to dissolve uranium(IV) and extract it as uranium triperoxocomplex, which facilitates its manipulation and final conversion into uranium nitrate as a high purity concentrate.

Influence of the postreaction time in the concentration of141Ce anionic species in uranium tricarbonated solutions

An effective method to minimize the presence of141Ce in the final hexahydrated uranyl nitrate recovery product has been obtained. This condition is considered as one prepurification stage in the recovery process of residual nonfissioned uranium in the production of99Mo of fission.

Radionuclide production with > 70-MeV proton accelerators: current and future prospects

The role of proton accelerators (i.e. cyclotrons) in the production of medical radionuclides is increasing as a result of new trends in medical research and due to clinical needs associated with the expanding field of positron emission tomography (PET). Accelerators also produced radionuclides needed in single photon emission computerized tomography (SPECT). Modern high-intensity H − cyclotrons are now commercially available, allowing the use of simultaneous beams, as well as a highly reliable operation. Furthermore, new accelerator-based methods for the production of several useful positron emitters and for the large-scale production of 99mTc and 99Mo, are being investigated. Based on new data, an ample supply for both PET and SPECT radiopharmaceuticals is feasible and may be obtained from centralized regional facilities operating a modern accelerator. For these facilities, an accelerator capable of delivering > 70-MeV protons is, by far, the best overall choice. This rationale is explained here and supported with a summary of radionuclide production data. An increasing role for modern proton accelerators in nuclear medicine and other applications is believed to be scientifically proven and forthcoming although further research and development of the technical and economical basis are required.

Cyclotron production of NCA 99mTc and 99Mo. An alternative non-reactor supply source of instant 99mTc and 99Mo→ 99mTc generators

The direct production of no-carrier-added (NCA) 6.02 h 99mTc and of 66 h 99mMo using proton beams on natural Mo targets was investigated. The major objective of this work was to evaluate the potential of utilizing high-intensity proton accelerators as a supply source of 99mTc and 99Mo for use in diagnostic nuclear medicine. The excitation functions for the production of the directly-made 99mTc via the 100Mo(p, 2n) 99mTc ( Q = −7.85 MeV) reaction, and of its parent 99Mo via the 100Mo(p, pn) 99Mo ( Q = −8.30 MeV) and 100 Mo(p, 2 p) 99 m Nb(15 s) → 99 Mo (Q = −11.14 MeV) reactions, were measured the 68-8 MeV energy range. Single and cumulative yields for 99mTc and 99Mo, and for other Tc, Mo, Zr, Nb and Y radiocontaminants were also determined. The prospects of integrating the use of enriched 100Mo targets with high-intensity, dual beam, H − accelerators was analyzed. The potential of this combined method to replace or complement the current reactor-based supply sources of 99Mo→ 99mTc generators, is also discussed. Finally, a brief analysis is made on the potential use of this combined technology to support the anticipated expansion of nuclear medicine in developing nations.

The recovery of uranium from the waste solution of fission 99Mo production

A procedure is described for the solvent extraction of enriched uranium from the waste solution of fission 99Mo production. An extraction process has been developed for recovering residual uranium from waste solutions of the fission 99Mo production. The procedure and the apparatus meet the very low uranium concentrations (∼10 g U/L) by using special air lift extractors which work satisfactorily at low organic/aqueous phase ratios. The design of the plant is briefly described.

Production of99Mo and its application in nuclear medicine

Molybdenum-99 is produced in large quantities as the parent radioisotopes of99mTc, which has been used recently in nuclear medicine. The neutron capture reaction on molybdenum and the nuclear fission of uranium are used for the large scale production of99Mo. The products by these methods are used properly according to the objects of diagnosis. In this paper, the production of99Mo is reviewed and the development of the production in JAERI is outlined.

Epithermal neutron self-shielding in the reactor (n, γ) production of 99Mo

Activation data from routine irradiations of natural MoO 3 targets having different 98Mo densities were used to calculate an effective epithermal absorption cross section. This allowed the estimation of the self-shielding effects for enriched 98Mo targets, for the same irradiation conditions, without detailed knowledge of the flux spectrum and without experimental ad-hoc irradiations.

The determination of molybdenum in geological samples by Neutron Activation Analysis

A metal-silicate extraction technique combined with neutron activation analysis has been developed to determine molybdenum in geological samples. The samples are equilibrated with Femetal powder at high temperatures. Molybdenum is completely extracted into the metal phase because of very reducing conditions in the furnace. The metal spherule is separated from the silicates, irradiated and dissolved in an acid solution. The molybdenum is precipitated as a sulfide and the precipitate is dissolved in aqua regia and counted on a Ge/Li/detector. The radiochemical yield is obtained by irradiation of the solution. The method avoids production of99Mo from induced fission of235U by performing the metal-silicate separation before irradiation. The precipitation step may be necessary to remove the high background from the decay of59Fe. Mo concentrations down to 15 ng/g have been obtained using this method.

Large scale production of fission 99Mo by using fuel elements of a research reactor as starting material

A procedure is described for the production of fission 99Mo. Starting from short period irradiated fuel elements of a research reactor the following steps are included: dissolution of the aluminium cladded fuel element in 6 N HNO 3, separation of 99Mo from uranium and the bulk of other fission products by adsorption/desorption on alumina, and purification of the molybdenum fraction by means of high temperature volatitization. The procedure is particularly suitable for those who do not dispose of highly enriched uranium from high flux reactors. The total yield proved to be about 1200 Ci 99Mo from one fuel element containing 36 g 235U after irradiation at 5·10 13 neutrons/cm 2s for 200 h. The final product is well suited for manufacturing high-quality 99 m Tc generators.

Dissolution and solidification of aluminum capsule in production of Mo-99 by sublimation from neutron-irradiated UO2.

(1980). Dissolution and Solidification of Aluminum Capsule in Production of Mo-99 by Sublimation from Neutron-Irradiated UO2. Journal of Nuclear Science and Technology: Vol. 17, No. 1, pp. 83-85.

Dissolution and Solidification of Aluminum Capsule in Production of Mo-99 by Sublimation from Neutron-Irradiated UO2

Dissolution and Solidification of Aluminum Capsule in Production of Mo-99 by Sublimation from Neutron-Irradiated UO₂

The production of carrier-free 99Mo

The production of carrier-free 99Mo

Excitation functions and yields of the ( d, p) reactions on natural molybdenum for deuteron energies less than 13 MeV

The excitation functions of the reactions 98Mo( d, p) 99Mo and 100Mo( d, p) 101Mo have been determined by irradiation of stacked foils with deuterons of energies less than 13 MeV and non-destructive determination of the absolute activity of the Mo radioisotopes by semiconductor γ-ray spectrometry. From the excitation functions, the thick-target yields and the saturation production rates of 99Mo and 101Mo for deuteron energies of 13.0 MeV and 11.7 MeV have been calculated. Implications for the production of 99Mo for generators of 99 m Tc are discussed.