Burn‐up calculation of the neutronic and safety parameters of thorium‐uranium mixed oxide fuel cycle in a Westinghouse small modular reactor

Abstract

Summary Thorium fuel is presently a globally known future nuclear fuel alternative, having good neutronic, physical and chemical properties in addition to its spent nuclear fuel characteristic proliferation resistance. This research focused on the neutronic and safety parameters of thorium‐uranium mixed oxide fuel cycle, utilising three fissile enrichment zones, a departure from the conventional single enrichment. The aim was to determine the range of three fissile zones adequate for thorium‐uranium fuel cycle; investigating the performance efficiency of the fuel neutronic and inherent safety parameters in response to temperature differentials, which determines the viability of the fuel and core composition. Use was made of the MCNPX 2.7 code integrated with the CINDER90 fuel depletion code for steady‐state and burn‐up calculations. The keff, moderator temperature coefficient (MTC) and fuel temperature coefficient (FTC) of reactivity are affected by the range of fissile enrichment and fuel temperature which decreased with their respective increases. The MTC for all the moderator temperatures was within 0 to −40 pcm/K design value for UO2 fuel. Similarly, the FTC was within −3.5 to −1 pcm/K design value for all the fuel temperatures except after 2000 days, where a positive reactivity feedback was introduced. At ~86 MWd/kgHM single discharge burn‐up, the result shows that ~90% of the initial fissile load was utilised for energy production at the normal reactor operating temperature (600 K) with a slight reduction at higher fuel temperature. The total fissile inventory ratio (FIR), 233U/kg‐232Th and 239Pu/kg‐238U inventory ratios were significantly large and increased with burn‐up. It is remarkable that the FIR and the 233U/kg‐232Th inventory ratio did not reach conversion equilibrium until exit burn‐up. The large percentage fuel utilisation supports the advantage of fissile enrichment zoning in a thermal nuclear reactor core, making the chosen novel three fissile enrichment zones for thorium‐uranium fuel cycle reliable.

Neutronic and burnup characteristics of potential dual‐cooled annular ( Th‐ 233 U‐ 235 U ) O 2 fuel for the advanced pressurized water reactors: An assembly‐level analysis

“…The FIR is obtained by dividing the assembly fissile inventory at a given point in time by the initial assembly fissile inventory. In this study, we consider Pa‐233 and Np‐239 to be fissile materials to calculate FIR for all cases because of their relatively fast conversion into U‐233 and Pu‐239 via beta‐minus decay, 27 and 2.4 days, respectively 19,38 . Figure 6A depicts the FIR variation, which decreased with burnup.…”

“…In this study, we consider Pa-233 and Np-239 to be fissile materials to calculate FIR for all cases because of their relatively fast conversion into U-233 and Pu-239 via beta-minus decay, 27 and 2.4 days, respectively. 19,38 Figure 6A depicts the FIR variation, which decreased with burnup. However, when Th-232 is the fertile nucleus, the FIR values are greater than when U-238 is the fertile nucleus.…”

“…As a direct consequence, most studies used high assay low enriched uranium (HALEU). 18,19 According to relevant studies, (Th-233 U)O 2 fuel is a practical combination capable of self-sustaining breeding in the thermal spectrum. 3,20,21 The fissile U-233 can be created in a reactor or accelerator by bombarding Th-232 and then recovered from irradiated thorium via chemical separation in a nuclear reprocessing facility.…”

“…Because of Th‐232's higher thermal neutron capture cross‐section, thorium‐uranium fuel requires a higher initial enrichment to achieve acceptable burnups. As a direct consequence, most studies used high assay low enriched uranium (HALEU) 18,19 …”