The first HTGR developments in Germany, as well as in the US have considered High-Enriched-Uranium (HEU) and reprocessing as the reference option. Direct disposal and the use of Low-Enriched-Uranium (LEU) were initiated in the late 1970s’, due to nuclear proliferation concerns and became then widely considered since then as the sole solution for management of HTGR fuel waste management. In this paper, it is shown that, beyond evidence to substantiate this option, there are also other options, including closed cycle. For acceptability of a new generation of HTGR, expected to be deployed in the next decades for industrial process heat applications, it is important to assess the compatibility of their specific fuels with established back-end strategies and related investments in conditioning, storage, disposal, reprocessing, fuel recycling and with related regulatory requirements. This paper presents a work currently performed in the Euratom GEMINI 4.0 project based on previous European, IAEA, NEA and diverse national research programmes. It performs a compatibility assessment for different variants of TRISO fuel (enriched uranium, Pu, Th/U-233 fissile and fertile to fissile cycles and with oxide or oxicarbide fuel material options) for block type cores. It presents the technical differences and similarities with the current strategies and facilities. It identifies the remaining R&D issues. 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. Disposing of spent fuel compacts separately from the graphite blocks is expected to significantly reduce High-Level-Waste (HLW) volumes. Processes for separating the Coated-Particles (CPs) from the compacts and extracting the fuel kernels have also been evaluated. They can be considered as a head-end step for reprocessing. 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 categorisation of irradiated graphite as Intermediate-Level- Waste (ILW) in most countries. Therefore, treatment methods for reducing and/or stabilizing such isotopes for achieving a lower waste category, at much lower disposal cost, have been assessed. Reuse and refabricating options for i-graphite would be an interesting strategy towards a ‘Closed HTGR Graphite Cycle’. As the Finnish ONKALO® (Posiva Oy) repository is currently almost operable (i.e. operating license application submitted by Posiva), being therefore the first HLW disposal facility worldwide, it is taken as a reference for Waste-Acceptance-Criteria (WAC) and disposal procedures for HTGR fuel elements. Alternatively, disposing only spent fuel compacts once extracted from the fuel blocks allows a higher degree of freedom in the design of spent fuel canisters, being compatible with LWR disposal routines and WAC. Corrosion leach resistance of the fuel kernels and of the CP layers has a crucial impact on the performance assessment of CP fuel disposal. Specific leach and corrosion tests on irradiated CPs and fuel compacts are indispensable for establishing an optimised disposal concept.

HITTNER Dominique;  FUETTERER Michael;  HAVETTE Julien;  MUSZYNSKI Dominik;  OLIN Markus;  TZELEPI Athanasia;  KIEGIEL Katarzyna;  VON LENSA Werner;  TOUGAIT Olivier; 

2026-05-31

International Atomic Energy Agency

JRC144264

https://inis.iaea.org/records/s14m8-njn92,    https://publications.jrc.ec.europa.eu/repository/handle/JRC144264,