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September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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.
Ki-Yong Choi, Seok Cho, Hyoung-Kyu Cho, Chul-Hwa Song
Nuclear Technology | Volume 170 | Number 1 | April 2010 | Pages 54-67
Technical Paper | Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Thermal Hydraulics | doi.org/10.13182/NT10-A9445
Articles are hosted by Taylor and Francis Online.
The 6 × 6 reflood test facility for Advanced Thermal Hydraulic Evaluation of Reflood phenomena (ATHER) has been operated by Korea Atomic Energy Research Institute to investigate the reflooding phenomena in a rod bundle. A series of bottom reflood tests was carried out by varying several parameters affecting the reflooding process such as the flooding velocity, inlet coolant subcooling, system pressure, initial maximum rod wall temperature, and rod power. Subsequently, counterpart reflood tests of rod bundle heat transfer data from The Pennsylvania State University were conducted for comparison, focusing especially on the effects of the heat flux on the peak cladding temperature (PCT) and the quenching behavior. The best-estimate thermal-hydraulic system analysis code MARS3.1 was assessed with the obtained data to investigate the parametric effects on its prediction accuracy. It was found that the prediction accuracy of the PCT is reasonable on the whole but that the MARS code predicts delayed quenching behavior compared with the data, especially for high heat flux conditions. In particular, the prediction becomes deteriorated downstream, far from the inlet of the test section.