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Meeting 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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The U.S. Million Person Study of Low-Dose-Rate Health Effects
There is a critical knowledge gap regarding the health consequences of exposure to radiation received gradually over time. While there is a plethora of studies on the risks of adverse outcomes from both acute and high-dose exposures, including the landmark study of atomic bomb survivors, these are not characteristic of the chronic exposure to low-dose radiation encountered in occupational and public settings. In addition, smaller cohorts have limited numbers leading to reduced statistical power.
K. Lisa Reed, Farzad Rahnema, Dingkang Zhang, Dan Ilas
Nuclear Technology | Volume 206 | Number 11 | November 2020 | Pages 1686-1697
Technical Paper | doi.org/10.1080/00295450.2020.1757962
Articles are hosted by Taylor and Francis Online.
In this paper, a set of stylized numerical benchmark problems is developed. These problems are based on the Oak Ridge National Laboratory preconceptual design of a fluoride-salt-cooled small modular advanced high-temperature reactor, or SmAHTR, that uses prismatic fuel assemblies with cylindrical pins/rods containing tri-isotropic fuel particles. A detailed description of the benchmark problems is achieved by closing several outstanding design gaps and modifying the coolant channel shape to reduce bypass flow for improved coolant and fuel temperature distributions. The benchmark problems, while stylized, retain the important thermal-hydraulic and reactor physics features (e.g., fuel particles) necessary for benchmarking tools for reactor core analysis.
In addition to the full description, detailed reference results such as the eigenvalue (keff) and fuel pin and assembly-averaged fission density distributions are provided for five benchmark problems: full-length fuel assemblies with control rods fully withdrawn and inserted, and full core with all control rods withdrawn, all control rods fully inserted, and some control rods fully inserted (near-critical core). The provided results are calculated using the continuous-energy Monte Carlo code MCNP.