Quantitative analysis of relationships between irradiation parameters and the reproducibility of cyclotron-produced99mTc yields

Estimating 225Ac yield in thorium metal targets

In this work we estimate the yield of the radioisotope [Formula: see text] in a thorium metal target geometry similar to that described by Roberston et al.. We do so in three different yet complimentary studies. In the first study, we pose a three-way coupled time-dependent model describing beam position, temperature field, and local growth of the activity of the radioisotope and solve this numerically. In the second study, we present an analytical solution of the model equations for a generalized solid target in the "beam-thin" limit, i.e. where only a small fraction of the incoming energy of the proton beam is deposited into the thorium material. In the third study, we use the insight gained from the analytical solution and describe an operating strategy to maximize yield by modulating the beam flux temporally.

Analysis of radioactive waste generated during the cyclotron production of 99mTc

Past and prospective shortages of medical radioisotopes have driven recent developments in the direct production of ^99mTc via the ¹⁰⁰Mo(p,2n)^99mTc reaction. The cyclotron-based production method has been shown to successfully produce ^99mTc, however trace impurities present in the enriched molybdenum target can also lead to the unintended creation of other radioisotopes which constitute waste. The isotopic composition of the waste has to be investigated in order to determine how it can be handled, transported and safely stored. In this article, we report which waste radioisotopes are created alongside ^99mTc during target irradiation. Results are based on the gamma spectroscopy of waste produced. Significant complexities in the emission spectra made automated identification of radioisotopes inaccurate; complexities were resolved using a manual radioisotope identification procedure. The impact of target composition, integrated beam current and duration of target irradiation on the waste produced was studied. Results indicate that an average of 0.059  ±  0.003 GBq of waste is generated per 1 GBq of ^99mTc produced. Two-thirds of the total waste activity produced was attributed to ⁹⁹Mo (T _1/2  =  66 h) alone, while a total of fifty radioisotopes were found in the waste. Long-lived isotopes (T _1/2  >  2 months) constituted only 1% of the total waste activity at end of beam (EOB). In conclusion, it was determined that the waste generated during cyclotron-based ^99mTc production was acceptably low for routine clinical production.

Cyclotron production of 99mTc: Comparison of known separation technologies for isolation of 99mTc from molybdenum targets

Intensive efforts were undertaken during the last few decades for the separation of cyclotron-produced ^99mTc from ⁹⁹Mo and new papers have been published on this topic since the last review [1]. In the future the cyclotron-based methods can replace reactor-based technology in producing this medical radioisotope and the nuclear reaction ¹⁰⁰Mo(p,2n)^99mTc appears to be the most worthwhile approach. New ways of producing of ^99mTc require efficient separation methods. Several strategies for separation of ^99mTc from ⁹⁹Mo have been already developed. The advantages, disadvantages and technical challenges toward application potential of investigated methods to separate ^99mTc from irradiated ¹⁰⁰Mo target are discussed. These methods include column chromatography, solvent extraction, chemical precipitation and thermochromatography.

Recent achievements in Tc-99m radiopharmaceutical direct production by medical cyclotrons

^99mTc is the most commonly used radionuclide in the field of diagnostic imaging, a noninvasive method intended to diagnose a disease, assess the disease state and monitor the effects of treatments. Annually, the use of ^99mTc, covers about 85% of nuclear medicine applications. This isotope releases gamma rays at about the same wavelength as conventional X-ray diagnostic equipment, and owing to its short half-life (t_½ = 6 h) is ideal for diagnostic nuclear imaging. A patient can be injected with a small amount of ^99mTc and within 24 h almost 94% of the injected radionuclide would have decayed and left the body, limiting the patient's radiation exposure. ^99mTc is usually supplied to hospitals through a ⁹⁹Mo/^99mTc radionuclide generator system where it is produced from the β decay of the parent nuclide ⁹⁹Mo (t_½ = 66 h), which is produced in nuclear reactors via neutron fission. Recently, the interruption of the global supply chain of reactor-produced ⁹⁹Mo, has forced the scientific community to investigate alternative production routes for ^99mTc. One solution was to consider cyclotron-based methods as potential replacement of reactor-based technology and the nuclear reaction ¹⁰⁰Mo(p,2n)^99mTc emerged as the most worthwhile approach. This review reports some achievements about ^99mTc produced by medical cyclotrons. In particular, the available technologies for target design, the most efficient extraction and separation procedure developed for the purification of ^99mTc from the irradiated targets, the preparation of high purity ^99mTc radiopharmaceuticals and the first clinical studies carried out with cyclotron produced ^99mTc are described.