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Center for Used Fuel Research: Building confidence in storage and transport
Used nuclear fuel storage and transportation have reached a critical juncture.
Dozens of utilities need reliable data on how used nuclear fuel performs in dry storage casks and canisters to extend regulatory licenses at sites across the United States. Likewise, the Department of Energy expects to take ownership of the used nuclear fuel—termed “spent nuclear fuel” in the laws and regulations governing its stewardship—and transfer it to one or more federal staging facilities for management and disposition.
Meanwhile, dozens of reactor companies are testing prototypes of advanced reactors and advanced reactor fuels. Eventually, regulators and industry must also verify the safety and security of storage methods for these advanced fuel types.
To help address these challenges, the DOE established the Center for Used Fuel Research (CUFR) in January 2026 for work related to the long-term storage and transport of used nuclear fuel.
Mariusz P. Dąbrowski, Hirofumi Ohashi
Nuclear Technology | Volume 212 | Number 7 | July 2026 | Pages 1663-1672
Research Article | doi.org/10.1080/00295450.2025.2562640
Articles are hosted by Taylor and Francis Online.
A design of the research-demo High-Temperature Gas-cooled Reactor (HTGR) based on Japanese High Temperature Engineering Test Reactor (HTTR) technology that is updated and adjusted to Polish market requirements is presented. It is a collaborative design project of the National Centre for Nuclear Research (NCBJ) in Poland and of the Japan Atomic Energy Agency in Japan and is going to be built at the site of NCBJ near to the existing research light water reactor MARIA. The reactor thermal power is 30 MW, and it carries the name High-Temperature Gas-cooled Reactor—POLish Atomic (HTGR-POLA). The main objective of the design is focused on applications of high-temperature steam acting continuously in polygeneration. Steam temperature of 540°C, electricity, and district heat as well as perspectival hydrogen production are considered. The reactor’s core is prismatic (hexagonal graphite blocks), helium cooled, and TRISO fueled and possesses many inherent safety features. The current status of the project is basic design and large pieces of the Preliminary Safety Analyses Report finalized, ready to start detailed design and licensing.
