Neutronic Evaluation of Using a Thorium Sulfate Solution in an Aqueous Homogeneous Reactor

D. Pérez (1. Departamento de Energia Nuclear, Universidade Federal de Pernambuco (UFPE), Cidade Universitária, Avenida Professor Luiz Freire 1000, Recife, PE, Brasil 2. Centro Regional de Ciências Nucleares (CRCN-NE/CNEN), Cidade Universitária, Avenida Profess)
D. Milian (Departamento de Energía Renovable y Eficiencia Energética, Cubaenergía, Calle 20 No 4111 / 18a y 47, Playa, Habana, Cuba)
L. Hernández (Instituto Superior de Tecnologías y Ciencias Aplicadas (InSTEC), Universidad de La Habana, Avenida Salvador Allende y Luaces, Quinta de Los Molinos, Plaza de la Revolución, 10400, La Habana, Cuba)
A. Gámez (1. Departamento de Energia Nuclear, Universidade Federal de Pernambuco (UFPE), Cidade Universitária, Avenida Professor Luiz Freire 1000, Recife, PE, Brasil 2. Centro Regional de Ciências Nucleares (CRCN-NE/CNEN), Cidade Universitária, Avenida Profess)
D. Lorenzo (Instituto Superior de Tecnologías y Ciencias Aplicadas (InSTEC), Universidad de La Habana, Avenida Salvador Allende y Luaces, Quinta de Los Molinos, Plaza de la Revolución, 10400, La Habana, Cuba)
C. Brayner (Centro Regional de Ciências Nucleares (CRCN-NE/CNEN), Cidade Universitária, Avenida Professor Luiz Freire 200, Recife, PE, Brasil)

Atom Indonesia Journal
Vol 48, No 2 (2022): August 2022

Radioisotope 99Mo is one of the most essential radioisotopes in nuclear medicine. Its production in an Aqueous Homogeneous Reactor (AHR) could be potentially advantageous compared to the traditional technology, based on target irradiation in a heterogeneous reactor. An AHR conceptual design using low-enriched uranium for the production of 99Mo has been studied in depth. So far, the possibility of replacing uranium with a non-uranium fuel, specifically a mixture of 232Th and 233U, has not been evaluated in the conceptual design. Therefore, the studies conducted in this article aim to evaluate the neutronic behavior of the AHR conceptual design using thorium sulfate solution. Here, the 232Th-233U composition to guarantee ten years of operation without refueling, conversion ratio, medical isotopes production levels, and reactor kinetic parameters were evaluated, using the computational code MCNP6. It was obtained that 14 % 233U enrichment guarantees the reactor operation for ten years without refueling. The conversion ratio was calculated at 0.14. The calculated 99Mo production in the AHR conceptual design resulted in 24.4 % higher with uranium fuel than with thorium fuel.