Thermochemical Assessment of Oxygen Gettering by SiC or ZrC in PuO2-x TRISO Fuel
- ORNL
Particulate nuclear fuel in a modular helium reactor is being considered for the consumption of excess plutonium and related transuranics. In particular, efforts to largely consume transuranics in a single-pass will require the fuel to undergo very high burnup. This deep burn concept will thus make the proposed plutonia TRISO fuel particularly likely to suffer kernel migration where carbon in the buffer layer and inner pyrolytic carbon layer is transported from the high temperature side of the particle to the low temperature side. This phenomenon is oberved to cause particle failure and therefore must be mitigated. The addition of SiC or ZrC in the oxide kernel or in a layer in communication with the kernel will lower the oxygen potential and therefore prevent kernel migration, and this has been demonstrated with SiC. In this work a thermochemical analysis was performed to predict oxygen potential behavior in the plutonia TRISO fuel to burnups of 50% FIMA with and without the presence of oxygen gettering SiC and ZrC. Kernel migration is believed to be controlled by CO gas transporting carbon from the hot side to the cool side, and CO pressure is governed by the oxygen potential in the presence of carbon. The gettering phases significantly reduce the oxygen potential and thus CO pressure in an otherwise PuO2-x kernel, and prevent kernel migration by limiting CO gas diffusion through the buffer layer. The reduction in CO pressure can also reduce the peak pressure within the particles by ~50%, thus reducing the likelihood of pressure-induced particle failure. A model for kernel migration was used to semi-quantitatively assess the effect of controlling oxygen potential with SiC or ZrC and did demonstrated the dramatic effect of the addition of these phases on carbon transport.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 979208
- Journal Information:
- Journal of Nuclear Materials, Vol. 397, Issue 1-3; ISSN 0022-3115
- Country of Publication:
- United States
- Language:
- English
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