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Duy Long Ta

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Featured publications

Validation of UNIST Monte Carlo code MCS for criticality safety calculations with burnup credit through MOX criticality benchmark problems6 citations
Nuclear Engineering and Technology · 2020 · DOI
A spent nuclear fuel source term calculation code BESNA with a new modified predictor-corrector scheme3 citations
Nuclear Engineering and Technology · 2022 · DOI
Evaluation of Source Terms Using the Besna Code for Reference Spent Fuels for Future Underground Disposal System of South Korea1 citations
SSRN Electronic Journal · 2023 · DOI
Evaluation of source terms using the BESNA code for reference spent fuels for future underground disposal system of South Korea1 citations
Nuclear Engineering and Technology · 2024 · DOI
Coulped neutronics/thermal-hydraulics calculation of VVER-1000 fuel assembly
Nuclear Science and Technology · 2016 · DOI
Neutronics benchmark calculations of the NuScale-like SMR core using SCALE/KENO-VI
Nuclear Engineering and Technology · 2025 · DOI
A Study on minor actinides transmutation as burnable poison in VVER-1000 LEU and MOX fuel assemblies
· 2017

About the field of Physics

Common research topics in this field:

Graphene research and applicationsQuantum Dots Synthesis And Properties2D Materials and ApplicationsPhotonic and Optical DevicesOptical Network TechnologiesChalcogenide Semiconductor Thin Films

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Duy Long TaInstitute for Nuclear Science and Technology
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