Development and verification of LOOP: A Linkage of ORIGEN2.2 and OpenMC

Gul, Anas; Chaudri, Khurrum Saleem; Khan, R.; Azeem, Muhammad Waqar

doi:10.1016/j.anucene.2016.09.016

Cited by 11 publications

(5 citation statements)

References 13 publications

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“…The flux unit in OpenMC code is normalized per source neutron with the default unit is neutron‐cm/source. The flux tally formula is shown in Equation ) 19 . We need the formula of flux normalization factor (FNF) to get the unit as neutrons/cm 2 ‐s.…”

Section: Resultsmentioning

confidence: 99%

“…We need the formula of flux normalization factor (FNF) to get the unit as neutrons/cm 2 ‐s. The FNF formula is represented by Equation ), in which the unit is source/s 19,20 . The default unit of OpenMC for the fission reaction rate is the reaction/source, so we also need the normalization factor by considered the thermal power, total of fuel rods, and fission energy to get the unit as reactions/s.…”

Section: Resultsmentioning

confidence: 99%

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Reflector materials selection for core design of modular gas‐cooled fast reactor using OpenMC code

et al. 2020

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Summary This paper discusses the selection of reflector materials for the core design of modular gas‐cooled fast reactor (GFR). Ideally, the reflector materials can be recycled, compatible with helium as the primary coolant, and resistant to high working temperature in the modular GFR. The modular GFR is a reactor design concept that maintains the criticality to achieve a longer operation time. However, maintaining the neutron population in the fast reactor core is still a challenge due to the high neutron leakage. In this research, the main objective is selecting potential reflectors materials for a modular GFR using open‐source Monte Carlo, OpenMC code. The OpenMC code applied the Monte Carlo method that provides high fidelity three‐dimensional (3D) geometry modeling, implementing continuous energy cross‐sections, and enabling to use of the nuclear data of Evaluated Nuclear Data File (ENDF/B‐VII.b5). The various candidate reflector materials, such as pure nickel, pure magnesium, pure lead, PbO, Ba2Pb, BeO, SiC, and Zr3Si2, are calculated from the neutron physics phenomenon to determine a good neutron reflector. The considered parameters of neutron physics included the multiplication factor profile, neutron energy distribution, flux and fission rate distribution, neutron leakage, core lifetime, mass evolution of fissile and fertile material, and thickness of reflector. The most crucial parameter of GFR core design is determined based on the conditions at the beginning of life (BOL) and the end of life (EOL). Finally, we concluded that the BeO material and silicon‐based materials are good reflector candidates for the core design of modular GFR.

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Section: Resultsmentioning

confidence: 99%

Section: Resultsmentioning

confidence: 99%

Reflector materials selection for core design of modular gas‐cooled fast reactor using OpenMC code

et al. 2020

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“…The UO2 fuel cell pin model High burnup on the Westinghouse standard PWR 17x17 reactor type fuel assembly was chosen in benchmarking this program. This model has been analyzed using the CASMO program using the collision probability method [7], MCODE [8], and LOOP [1] with the Monte Carlo transport theory method. The comparative program specifications in this study are for UO2 cell pin cases using the program can be seen in table 1 below.…”

Section: Benchmark Op-ormentioning

confidence: 99%

“…The study about neutronics and depletion analysis will be helpful when designing the reactor core. Research on calculations to find the accurate composition of fuel when the reactor is operating has been done long ago, using the CITATION programs with deterministic methods, but with the limited ability to burn-up and the number of isotopes is insufficient [1,2].…”

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confidence: 99%

“…The ORIGEN2.2 program was chosen as an additional program in performing the burnup calculation process as a tandem for the OpenMC program. This selection is based on research obtained in previous studies [1], where the coupling between OpenMC and ORIGEN has been proven to be accurate. In this program, static parameters of the system will be attained using the OpenMC program, while the production and loss of radioactive isotopes is determined using the ORIGEN program.…”

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confidence: 99%

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Development of Linkage Program Code OpenMC and ORIGEN2.2 for Neutronic Analysis and Burnup Nuclear Reactor Program

Ilham¹

2022

J. Phys.: Conf. Ser.

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This paper discusses the development of program code that coupling the ORIGEN2.2 for burn-up analysis with the Monte Carlo program OpenMC for neutron analysis program called OpenMC-ORIGEN (Op-OR). The results of the program have been compared with benchmark results from previously published results and MCNP6 program. The acquired results show a good agreement with benchmark results. This linkage program codes performed well for designing a new advanced reactor and analysing the neutronics parameter and burnup/depletion calculation.

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