Diffusion of Fission Products in Nuclear Graphite: A Review

Abstract Nuclear graphite in high-temperature gas-cooled reactors (HTGRs) acts as a moderator, a structural material, as well as an accommodation for fission products (FPs). Given the release of FPs under normal and accident conditions, the diffusion of FPs in nuclear graphite should be elucidated for the estimation of source term and reactor safety. However, the existing experimental data of diffusion are enriched in data of the nuclear graphite types (i.e., H451 and A3), that may not apply to next-generation nuclear graphite. The proposed diffusion theory and experimental systems are conducive to developinging a novel experimental method for the more accurate collection of the diffusion data and an insight into FPs diffusion in graphite. For this reason, this paper reviews the information regarding the diffusion of metallic FPs (e.g., Cs, Sr and Ag) in nuclear graphite. The diffusion theory adopted to calculate the diffusion coefficient is summarized. Emphasis is placed on the measurements for determination the diffusion coefficient and the discussion of detection limits, accuracy and experimental methods. Given the diffusion coefficient data, the variables of diffusion are addressed. The mentioned results lay a basis for gaining insights into the diffusion theory and can be exploited by employing with a valid measurement method to determine the diffusion coefficients of FPs in various nuclear graphite materials under a particular range of conditions.