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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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Watch as solid hydrogen is extruded to feed German stellarator
In May, the Wendelstein 7-X stellarator in Greifswald, Germany, concluded an experimental campaign by sustaining a plasma with a high triple product for 43 seconds. The machine far surpassed its own previous performance with a value that the Max Planck Institute for Plasma Physics (IPP) says “exceeds previous tokamak records for long plasma durations”—in part because of a fuel pellet injection system developed by researchers at Oak Ridge National Laboratory.
Watch ORNL’s video of that fuel pellet injection system—in use since September 2024—as it extrudes a column of frozen hydrogen and then cuts individual 3.2-millimeter-long pellets. The process, which takes just half a millisecond, was captured in slow motion by ORNL engineer Steve Meitner.
Marat Margulis, Erez Gilad
Nuclear Technology | Volume 196 | Number 2 | November 2016 | Pages 377-395
Technical Paper | doi.org/10.13182/NT16-23
Articles are hosted by Taylor and Francis Online.
The application of best-estimate codes [coupled neutron kinetics (NK)/thermal hydraulics (TH)] for safety analyses of research reactors (RRs) has gained considerable momentum during the past decade. Application of these codes is largely facilitated by the high level of technological maturity and expertise that these codes allow as a safety technology in nuclear power plants, and it is largely driven by International Atomic Energy Agency activities. The present study belongs in this framework and presents the development and application of the coupled NK and TH code THERMO-T to the analysis of protected reactivity insertion accidents and loss-of-flow accidents in a typical RR with standard Materials Testing Reactor plate-type fuel elements. The coupling is realized by considering the neutronic reactivity feedbacks of the fuel and coolant temperatures and a heat generation model for the reactor power. The neutron flux in the reactor core is solved by applying point reactor kinetic equations and employing radial and axial power distributions calculated from a three-dimensional full-core model by the continuous-energy Monte Carlo reactor physics code Serpent. The evolution of temporal and spatial distributions of the fuel, cladding, and coolant temperatures is calculated for all fuel channels by using a finite volume time implicit numerical scheme for solving a three-conservation equation model. In this study, additional features, such as critical heat flux ratio prediction and decay heat model, are implemented for both highly enriched uranium and low-enriched uranium cores, and a comprehensive comparison of THERMO-T results is performed against other codes.