TY - JOUR
T1 - A Stylized 3-D Benchmark Problem Set Based on the Pin-Fueled SmAHTR
AU - Reed, K. Lisa
AU - Rahnema, Farzad
AU - Zhang, Dingkang
AU - Ilas, Dan
N1 - Publisher Copyright:
© 2020, © 2020 American Nuclear Society.
PY - 2020
Y1 - 2020
N2 - 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density distributions are provided for five benchmark problems: full-length fuel assemblies with control rods fully withdrawn and inserted, and full core with all control rods withdrawn, all control rods fully inserted, and some control rods fully inserted (near-critical core).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.
AB - 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.
KW - Benchmark problem description
KW - fluoride-salt-cooled high-temperature reactor
KW - pin-type fuel
KW - small modular advanced high-temperature reactor
UR - http://www.scopus.com/inward/record.url?scp=85087478312&partnerID=8YFLogxK
U2 - 10.1080/00295450.2020.1757962
DO - 10.1080/00295450.2020.1757962
M3 - Article
AN - SCOPUS:85087478312
SN - 0029-5450
SP - 1686
EP - 1697
JO - Nuclear Technology
JF - Nuclear Technology
ER -