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A comparative analysis of the neutronic performance of thorium mixed with uranium or plutonium in a high‐temperature pebble‐bed reactor
Author(s) -
Kabach Ouadie,
Chetaine Abdelouahed,
Benchrif Abdelfettah,
Amsil Hamid,
El Banni Fadi
Publication year - 2021
Publication title -
international journal of energy research
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.808
H-Index - 95
eISSN - 1099-114X
pISSN - 0363-907X
DOI - 10.1002/er.6935
Subject(s) - nuclear engineering , fissile material , thorium fuel cycle , mox fuel , burnup , thorium , spent nuclear fuel , fission products , plutonium , waste management , environmental science , materials science , natural uranium , radiochemistry , uranium , neutron , nuclear physics , chemistry , engineering , physics , metallurgy
Summary High‐temperature gas‐cooled reactors (HTGRs) present an interesting concept for the next generation of nuclear reactors due to them having many advantages. In particular, their fuel is based on coated particles known as tri‐structural‐isotropic (TRISO) particles, and these provide a strong container capacity for fission products. Solid graphite acts as a moderator by delaying the transient temperature in the event of an accident, while neutronically inert helium acts as a coolant. What is more, several TRISO particle models are in various stages of development. Nowadays, numerous studies have sought to investigate the integration of thorium as a fertile absorber with a driver fuel (ie, U‐233, U‐235, or Pu‐293) in nuclear reactors due to its natural abundance and its potential for producing fissile material (U‐233). In addition, there are numerous perspectives for recycling and reusing the fissile isotopes released during thorium irradiation and the spent fuels from current reactors, with the main goal being to dispose of the world's growing stocks of radioactive waste and surplus weapons material. The primary objective of this study is not just to discuss the advantages of using thorium in a (pebble‐bed type) HTGR but also to investigate how thorium can be incorporated into the fuel with different drivers and how this affects the neutronic parameters in comparison to the nominal HALEU (UO 2 17 wt%) fuel. These parameters include the burnup behavior, spent fuel compositions, effective delayed neutron fractions, neutron spectra, and temperature coefficients. The investigated fuel mixtures include thoria (ThO 2 ) fuel mixed with high‐assay, low enriched uranium (HALEU 20 wt%), reactor‐grade plutonium (RGPuO 2 ), weapons‐grade plutonium (WGPuO 2 ), and U‐233. Neutronic analyses show that cores fueled by (Th + 233 U)O 2 and ThO 2 + WGPuO 2 achieved higher excess reactivity at the beginning of cycle (BOC), compared with the UO 2 core which achieves higher discharge burnups. Cores fueled by ThO 2 + RGPuO 2 , in contrast, can achieve higher discharge burnup despite having the lowest excess reactivity at BOC. Further comparisons of the examined fuel mixtures reveal that the (Th + 233 U)O 2 and ThO 2 + PuO 2 mixtures have less favorable effective delayed neutron fractions and temperature coefficients than the UO 2 . However, the moderator temperature coefficients of the (Th + 233 U)O 2 mixtures were sufficiently negative for all time steps.