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September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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Latest News
DOE on track to deliver high-burnup SNF to Idaho by 2027
The Department of Energy said it anticipated delivering a research cask of high-burnup spent nuclear fuel from Dominion Energy’s North Anna nuclear power plant in Virginia to Idaho National Laboratory by fall 2027. The planned shipment is part of the High Burnup Dry Storage Research Project being conducted by the DOE with the Electric Power Research Institute.
As preparations continue, the DOE said it is working closely with federal agencies as well as tribal and state governments along potential transportation routes to ensure safety, transparency, and readiness every step of the way.
Watch the DOE’s latest video outlining the project here.
T. Kakuta, S. Hirata, S. Mori, S. Konishi, Y. Kawamura, M. Nishi, Y. Ohara
Fusion Science and Technology | Volume 41 | Number 3 | May 2002 | Pages 1069-1073
Blanket Material and Process | Proceedings of the Sixth International Conference on Tritium Science and Technology Tsukuba, Japan November 12-16, 2001 | doi.org/10.13182/FST02-A22748
Articles are hosted by Taylor and Francis Online.
In this design work, a combination of the hydrogen pumps in charge of the function of hydrogen isotopes recovery and the oxygen pump was adopted to the blanket tritium recovery system for the prototype fusion reactor designed by Japan Atomic Energy Research Institute (JAERI). The main functions of this system are described below. 1) Transport of tritium with helium purge gas: tritium released from ceramic breeding material in the blanket is transported to the tritium recovery system by the helium purge gas which contains a small amount of hydrogen gas. 2) Steam electrolysis and removal of oxygen gas: the oxygen pump with electrolyte of oxygen ionic conductors electrolyzes the steam (H2O and HTO) contained in the puige gas into hydrogen isotopes and oxygen, and simultaneously removes impurity of oxygen by electrical membrane permeation. 3) Recovery of hydrogen isotopes: the hydrogen pump with electrolyte of protonic conductors electrically recovers the pure hydrogen isotopes (HT and H2) from the purge gas. Based on the experimental data obtained by feasibility study and the present design effort, it was revealed that the simple and continuous tritium recovery system for gaseous stream is possible and attractive for fusion power reactors.