Feasibility study on the medical isotopes production with solution target using OSTR: (99)Mo and related isotopes.

摘要

Molybdenum-99 (99Mo) is the parent nuclide of Technetium-99m (99mTc), a radioisotope which is widely used in nuclear medicine. 99Mo is produced from the fission of 235U or the irradiation of 98Mo. This study shows the feasibility of the using an 'aqueous homogeneous uranium solution target' for the production of a medical isotope, 99Mo. Some of the advantages that the solution target has over a solid target include the inherent reactor safety features offered by large negative temperature and power reactivity coefficients, the fabrication convenience, the straightforward extraction process, and a low volume of waste generated.;To evaluate the core configuration and the production rate of 99Mo, a three-dimensional model of the Oregon State University TRIGA Reactor (OSTR) core was developed for use with the Monte Carlo N-Particle Transport Code (MCNP) and then verified by comparing with the measured values. Two values are in good agreement within one percent in the keffective values calculated.;Two types of solution targets are analyzed for the OSTR. The first one has the same outer-dimensions as an OSTR fuel element but is filled with a uranium solution. The other is the continuous flow target system (CFTS) like solution fuel reactors. Uranyl nitrate and uranyl sulfate solutions enriched to 20% or 93% are investigated as a target material without raising any safety concern to the OSTR operation. A seven-day irradiation of ten tube-type-93% enriched uranyl nitrate solution targets would produce 43% of the 99Mo required in the US for one week. The CFTS would generate 31% of the required 99Mo in a 7-day cycle. The conceptual chemical extraction processes for irradiated solution targets are developed. This work also includes an analysis of nuclear safety issues such as the radiolytic gas, thermal hydraulics, the waste, and the radiological impacts of an accident.;The production of 99Mo in the OSTR with the uranium solution is technically feasible as demonstrated in this work. The use of the uranium solution would increase the production efficiency by good neutron economy, reduction of the processing period, through the reuse of uranium, and by minimizing the waste generation.