http://orcid.org/0000-0001-7410-6551
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Abstract
A combination of the neutronics features of gas-cooled high-temperature reactors by using the fuel in the form of ceramic-coated particles, called tristructural-isotropic, and the heat removal feature of molten salt reactors by using molten salt as a coolant is an attractive option in designing a reactor with a high-power density operation without compromising the safety aspects. Neutronics feasibility of such a combination of the molten salt (LiF-BeF2) as a coolant and thorium-based fuel, in particular (Th-233U)O2, in a graphite-moderated system is investigated. This technical note presents the influence of the heavy metal (HM) loading on neutronics features of a pebble lattice cell, that is, infinite multiplication factor (K-inf), temperature coefficients of reactivity (TCR), the void reactivity coefficient, etc. In addition, enriched uranium fuel has also been studied just to make a comparison with thorium-based fuel. Furthermore, the minimum HM loading of fuel per pebble that is needed to achieve negative coolant-temperature reactivity coefficients and void reactivity coefficients has been estimated for molten salt coolant.
The analyses show that Th2/U3 fuel gives a less negative fuel temperature reactivity coefficient as compared with that of uranium-based fuel. This study also shows that all the TCR of both fuel types improve, becoming less positive or more negative, by increasing HM loading per pebble. Further, the burnup dependence of K-inf and the reactivity coefficients are studied for limiting HM loadings, e.g., 30 g per pebble. The change in the spectrum and the four-factor formula are used to explain the behavior of the reactivity coefficients as a function of HM loading and burnup.
Nomenclature
| 1/ | = | inverse of average neutron speed |
| f = | = | thermal utilization factor |
| p = | = | resonance escape probability |
| T = | = | temperature (K) |
| VF = | = | fuel volume |
| VM = | = | moderator volume |
| ϵ = | = | fast fission factor |
| ς = | = | thermal disadvantage factor (ratio of flux in the moderator to flux in the fuel) |
| η = | = | number of neutrons released per thermal absorption in the fuel |
| Σ = | = | macroscopic cross section |
| = | moderation ratio |
Acknowledgments
The authors sincerely thank Alain Hebert and R. Karthikeyan for the fruitful discussion held during this study regarding the lattice code DRAGON.