HTGR TRISO Fuel and Graphite Waste Management Strategies

This paper presents a review of possible high-temperature gas-cooled reactor (HTGR) fuel and graphite cycle back-end routes performed in the Euratom GEMINI 4.0 project based on previous European, International Atomicc Energy Agency, Nuclear Energy Agency, and diverse national research programs. It considers different variants of tristructural-isotropic (TRISO) fuel (enriched uranium, Pu, Th/²³³U fissile and fertile to fissile cycles, with oxide or oxicarbide fuel material options) for block-type cores. It identifies the remaining research and development issues.

As the Finnish ONKALO® (Posiva Oy) repository is currently almost operable (i.e. operating license application submitted by Posiva), being therefore the first high-level waste disposal facility worldwide, its waste acceptance criteria are taken as a reference for the direct disposal of HTGR fuel elements. Spent fuel management for HTGRs is strongly impacted by the large graphite and coating volumes compared to the small portion of the fissile kernel. Therefore, a separation of the fuel compacts from the graphite block has been investigated. Disposal of spent fuel compacts separately from the graphite blocks is indeed expected to significantly reduce the HLW volumes.

Processes for separating the TRISO particles from the compacts are also considered, as well as the packages for disposal of compacts or TRISO particles. For the direct disposal or disposal of separated compacts or particles, the corrosion leach resistance of the fuel kernels and of the TRISO coating layers has a crucial impact on the performance assessment of TRISO fuel disposal. Specific leach and corrosion tests on irradiated TRISO particles and fuel compacts are indispensable for establishing an optimized disposal concept.

An ultimate step of separation can be to extract the fuel kernels from the TRISO coatings. It can be considered as a head-end step for reprocessing, the feasibility of which is examined. On the other hand, irradiated graphite management is a specific challenge for all graphite-moderated reactors due to the large associated volume, the specific contamination, and the degradation caused by neutron irradiation. Long-lived activation products, such as ¹⁴C and ³⁶Cl, lead to the categorization of irradiated graphite as an intermediate-level waste in most countries. Therefore, treatment methods for reducing and/or stabilizing such isotopes for achieving a lower waste category, at a much lower disposal cost, are assessed. Reuse and refabricating options for irradiated graphite would be an interesting strategy toward a closed HTGR graphite cycle.