Modeling and simulation of VERA core physics benchmark using OpenMC code

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficie...

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Bibliographic Details
Published in:	Nuclear engineering and technology Vol. 55; no. 9; pp. 3388 - 3400
Main Authors:	Albugami, Abdullah O., Alomari, Abdullah S., Almarshad, Abdullah I.
Format:	Journal Article
Language:	English
Published:	Elsevier B.V 01-09-2023 Elsevier 한국원자력학회
Subjects:	Core design Core simulation Monte Carlo method Neutronics OpenMC code Reactor physics Validation VERA benchmark 원자력공학 Reactor physics VERA benchmark Validation Monte Carlo method Core simulation OpenMC code Neutronics Core design
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Summary:	Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.
ISSN:	1738-5733 2234-358X
DOI:	10.1016/j.net.2023.05.036