TY - GEN
T1 - Multiphysics modeling of thorium-based (Th, u)O2 and (Th, PU)O2 fuel performance in a light water reactor
AU - Liu, Rong
AU - Cai, Jie-Jin
AU - Zhou, Wen-Zhong
AU - Wang, Ye
PY - 2018/7
Y1 - 2018/7
N2 - 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.
AB - 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.
UR - https://www.scopus.com/pages/publications/85056158878
UR - https://www.scopus.com/record/pubmetrics.uri?eid=2-s2.0-85056158878&origin=recordpage
U2 - 10.1115/ICONE26-81237
DO - 10.1115/ICONE26-81237
M3 - RGC 32 - Refereed conference paper (with host publication)
SN - 978-0-7918-5145-6
VL - 3
T3 - International Conference on Nuclear Engineering, Proceedings, ICONE
BT - ASME Proceedings - Nuclear Fuel and Material, Reactor Physics, and Transport Theory
PB - American Society of Mechanical Engineers
T2 - 2018 26th International Conference on Nuclear Engineering, ICONE 2018
Y2 - 22 July 2018 through 26 July 2018
ER -