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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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DOE issues new NEPA rule and procedures—and accelerates DOME reactor testing
Meeting a deadline set in President Trump’s May 23 executive order “Reforming Nuclear Reactor Testing at the Department of Energy,” the DOE on June 30 updated information on its National Environmental Policy Act (NEPA) rulemaking and implementation procedures and published on its website an interim final rule that rescinds existing regulations alongside new implementing procedures.
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 | 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.