Selecting Burnup Algorithms in OpenMC Using the Calculated Benchmark of LEU Assembly and MOX Fuel

Selecting Burnup Algorithms in OpenMC Using the Calculated Benchmark of LEU Assembly and MOX Fuel

OpenMC is a state-of-the-art Monte Carlo neutron transport simulation code that uses the Python programming language as an API. OpenMC supports eight burnout simulation algorithms. This study presents the results of choosing an integration method for modeling the burnup of fuel assemblies with burna...

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Bibliographic Details
Main Authors: Hamza A Tanash, Denis A Solovyov, Vyacheslav G Zimin, Alexey L Lobarev, Denis A Plotnikov, Nikolay V Schukin
Format: Article
Language:English
Published: Национальный исследовательский ядерный университет МИФИ 2023-04-01
Series:Глобальная ядерная безопасность
Subjects:
openmc
выгорание
методы интегрирования
бенчмарк oecd
реакторов ввэр-1000
метод переноса нейтронов
предиктор интегратор (pi)
ce/li интегратор
endf/b-vii.1
коэффициент размножений нейтронов
Online Access:https://glonucsec.elpub.ru/jour/article/view/178

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