Verification of the OpenMC homogenized MYRRHA-1.6 core model

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Verification of the OpenMC homogenized MYRRHA-1.6 core model. / Hernandez Solis, Augusto; Malambu Mbala, Edouard; Stankovskiy, Alexey; Van den Eynde, Gert.

PHYSOR 2020: Transition to a Scalable nuclear future. Unknown, 2020. p. 1-1.

Research output: Contribution to report/book/conference proceedingsIn-proceedings paper

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@inproceedings{422d637f2685479dbb3e217de2d7ad5d,
title = "Verification of the OpenMC homogenized MYRRHA-1.6 core model",
abstract = "The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a k_eff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15{\%} and 1{\%}, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.",
keywords = "MYRRHA, OpenMC, PHISICS, Homogenization, Multi-group Monte Carlo",
author = "{Hernandez Solis}, Augusto and {Malambu Mbala}, Edouard and Alexey Stankovskiy and {Van den Eynde}, Gert",
note = "Score=3",
year = "2020",
month = "3",
day = "29",
language = "English",
pages = "1--1",
booktitle = "PHYSOR 2020: Transition to a Scalable nuclear future",
publisher = "Unknown",

}

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TY - GEN

T1 - Verification of the OpenMC homogenized MYRRHA-1.6 core model

AU - Hernandez Solis, Augusto

AU - Malambu Mbala, Edouard

AU - Stankovskiy, Alexey

AU - Van den Eynde, Gert

N1 - Score=3

PY - 2020/3/29

Y1 - 2020/3/29

N2 - The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a k_eff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15% and 1%, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.

AB - The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a k_eff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15% and 1%, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.

KW - MYRRHA

KW - OpenMC

KW - PHISICS

KW - Homogenization

KW - Multi-group Monte Carlo

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/37234110

M3 - In-proceedings paper

SP - 1

EP - 1

BT - PHYSOR 2020: Transition to a Scalable nuclear future

PB - Unknown

ER -

ID: 6052919