Development and verification of LOOP: A Linkage of ORIGEN2.2 and OpenMC
Author:
Funder
Higher Education Commission of Pakistan
Publisher
Elsevier BV
Subject
Nuclear Energy and Engineering
Reference22 articles.
1. Burnup dependent Monte Carlo neutron physics calculations of IAEA MTR benchmark;Chaudri;Prog. Nucl. Energy,2015
2. Croff, A.G., 1980. A User ’s Manual for the ORIGEN2 Computer Code. ORNL/TM-7175, Tennessee.
3. Initial MCNP6 release overview;Goorley;Nucl. Technol.,2012
4. VVER-1000 cross-section library generation for ORIGEN-II based on MCNP calculations;Hadad;Int. J. Hydrogen Energy,2015
5. Analysis of MNSR core composition changes using the codes WIMSD-4 and CITATION;Haj Hassan;Appl. Radiat. Isot.,2008
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1. Heterogeneous model based burnup and its impact on integral parameters of IAEA MTR benchmark;Annals of Nuclear Energy;2023-02
2. Development of Linkage Program Code OpenMC and ORIGEN2.2 for Neutronic Analysis and Burnup Nuclear Reactor Program;Journal of Physics: Conference Series;2022-08-01
3. Validation of deterministic code DRAGON5 for the fuel depletion analysis of a PWR pin-cell benchmark;Radiation Physics and Chemistry;2021-09
4. Depletion capabilities in the OpenMC Monte Carlo particle transport code;Annals of Nuclear Energy;2021-03
5. Reflector materials selection for core design of modular gas‐cooled fast reactor using OpenMC code;International Journal of Energy Research;2020-10-14
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