Production of Molybdenum-99 using Neutron Capture Methods

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Abstract

Pacific Northwest National Laboratory (PNNL), operated by Battelle, has identified a reference process for the production of molybdenum-99 (99Mo) for use in a chromatographic generator to separate the daughter product, technetium-99m (99mTc). The reference process uses the neutron capture reaction of natural or enriched molybdenum oxide via the reaction 98Mo(n,γ)99Mo. The irradiated molybdenum is dissolved in an alkaline solution, whereby the molybdenum, dissolved as the molybdate anion, is loaded on a proprietary ion exchange material in the chromatographic generator. The approach of this investigation is to provide a systematic collection of technologies to make the neutron capture method for Mo-99 production economically viable. This approach would result in the development of a technetium Tc99m generator and a new type of target. The target is comprised of molybdenum, either natural or enriched, and is tailored to the design of currently operating U.S. research reactors. The systematic collection of technologies requires evaluation of new metallurgical methods to produce the target, evaluation of target geometries tailored to research reactors, and chemical methods to dissolve the irradiated target materials for use in a chromatographic generator. A Technical specification for testing the target and neutron capture method in a research reactor is also required. This report includes identification of research and demonstration activities needed to enable deployment of neutron capture production method, including irradiations of prototypic targets, chemical processing of irradiated targets, and loading and extraction tests of Mo99 and Tc99m on the sorbent material in a prototypic generator design. The prototypical generator design is based on the proprietary method and systems for isotope product generation. The proprietary methods and systems described in this report are clearly delineated with footnotes. Ultimately, the Tc-99m generator solution provided by the system has exactly the same chemical and radiochemical characteristics as the Tc-99m currently produced by standard generator systems. Analysis results indicate: • The production of Mo99 is a function of the neutron flux in the thermal and epithermal region, the target volume, and the target geometry. Calculations show that neutron self-absorption is not very important such that large (2 cm OD or more) cylinders of molybdenum can be irradiated without significant losses. • Efficient use of target volume design is function of simultaneously optimizing the amount of molybdenum that can be inserted into each irradiation capsule and the amount of interconnected porosity within the specimen body to enhance the rate of post-irradiation dissolution. • Neutron capture of natural molybdenum may effectively achieve up to 1 Ci/g Mo99 in a 144 hour irradiation period, when using the fuel annulus plus a beryllium reflector configuration.

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