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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
2021 Student Conference
April 8–10, 2021
North Carolina State University|Raleigh Marriott City Center
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Nuclear Science and Engineering
Fusion Science and Technology
Baranwal reviews virtual STEM lessons for U.S. tribal communities
In a blog post to the Department of Energy’s website on November 23, Rita Baranwal, assistant secretary for the Office of Nuclear Energy, commended recent virtual lesson projects from the Office of Nuclear Energy and the Nuclear Energy Tribal Working Group to increase STEM opportunities for Native American tribes.
The spotlighted lesson discussed in the article focused on a 3D-printed clip that turns a smartphone or tablet into a microscope with the ability to magnify items by 100 times. The Office of Nuclear Energy shipped nearly 1,000 of these microscope clips to students across the country, many of them going to U.S. tribal communities.
H. Y. Yoon, I. K. Park, J. R. Lee, S. J. Lee, Y. J. Cho, S. J. Do, H. K. Cho, J. J. Jeong
Nuclear Science and Engineering | Volume 194 | Number 8 | August-September 2020 | Pages 633-649
Technical Paper | dx.doi.org/10.1080/00295639.2020.1727698
Articles are hosted by Taylor and Francis Online.
A high-fidelity safety analysis method for pressurized water reactors (PWRs) is presented using a multiscale and multiphysics coupled code. Computational resolution of the conventional safety analysis can be greatly improved using this method in which the whole reactor vessel is modeled at a subchannel scale with around 5 million calculation meshes. Three-dimensional thermal hydraulics inside the reactor vessel is simulated using CUPID-RV with subchannel-scale thermal-hydraulic models for the reactor core. The subchannel models were validated using the legacy rod bundle experiments including single- and two-phase flow tests that were used in the validation of other subchannel analysis codes. The three-dimensional mesh was generated for the reactor vessel. Structured meshes were used in the core region for the subchannel model, and body-fitted unstructured meshes were applied for the downcomer, lower and upper plenums, and hot and cold legs. The number of meshes was optimized for a practical calculation. A three-dimensional core kinetics code (MASTER) and a one-dimensional system analysis code (MARS) were coupled with CUPID-RV for an accident analysis of PWRs. Subchannel-scale full-core steam line break accident analysis of the OPR1000 PWR was realized using the coupled code (MASTER/CUPID-RV/MARS) with a reasonable computation time, and thus, the present method can be used as a practical tool for three-dimensional safety analysis of PWRs.