VERIFICATION OF THE OpenMC HOMOGENIZED MYRRHA-1.6 CORE MODEL

Augusto Hernandez-Solis,Edoaurd Malambu,Alexey Stankovskiy,Yohannes Molla,Gert Van Den Eynde,M Margulis,P Blaise

doi:10.1051/epjconf/202124704002

Augusto Hernandez-Solis, Edoaurd Malambu + Show 5 more

Open Access

https://doi.org/10.1051/epjconf/202124704002

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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 keff 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.

Highlights

MYRRHA (Multi-purpose hYbrid Research Reactor) is a multi-functional irradiation facility for innovative applications, and is intended to be ready by 2035 [1]
The use of other Monte Carlo codes that are very much reactor-physics oriented and more dedicated towards neutron transport have emerged in the academic community, as it is the case of OpenMC [6]
For the energy-integrated flux, a maximum of 1% relative difference from the OpenMC multi-group Monte Carlo (MGMC) solution is observed for the fast irradiation assembly located at the center of the core

Summary

Introduction

MYRRHA (Multi-purpose hYbrid Research Reactor) is a multi-functional irradiation facility for innovative applications, and is intended to be ready by 2035 [1]. This core design brings interesting neutronic characteristics (such as strong flux-spectra gradients) among its different components such as fuel assembly batches, irradiation test sections and control rod banks. Studies have already been conducted about employing OpenMC for the creation of so-called nodal databases of homogenized and few-group cross-sections (for the most recent studies see references [7,8]), with the aim of being utilized as input parameters in deterministic transport codes that can be employed at low scales (i.e. from the pin cell up to a fuel assembly), or even larger scales (i.e. a group of assemblies configuration or a part of a core)

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