NDLI: Multiphysics Modeling of Thorium-Based (Th, U)O2 and (Th, Pu)O2 Fuel Performance in a Light Water Reactor

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Liu, Rong Cai, Jie-Jin Zhou, Wen-Zhong Wang, Ye
Copyright Year	2018
Abstract	ThO2 has been considered as a possible replacement for UO2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO2 fuel. In this study, the thermal, mechanical properties of Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel were reviewed with updated properties and compared with UO2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th0.923U0.077O2 fuel were found to decrease the fuel centerline temperature, while Th0.923Pu0.077O2 fuel was found to have a bit higher fuel centerline temperature than UO2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO2 fuel at high fuel burnup. The Th0.923U0.077O2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th0.923Pu0.077O2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO2 fuel. So the fuel performance could be expected to be improved by applying Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel.
Sponsorship	Nuclear Engineering Division
File Format	PDF
ISBN	9780791851456
DOI	10.1115/ICONE26-81237
Volume Number	Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory
Conference Proceedings	2018 26th International Conference on Nuclear Engineering
Language	English
Publisher Date	2018-07-22
Publisher Place	London, England
Access Restriction	Subscribed
Subject Keyword	Temperature Light water reactors Mechanical properties Nuclear reactors Fuels Modeling Pressure Nuclear fission
Content Type	Text
Resource Type	Article

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