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 techn...
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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.& COPY;2023 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://***/licenses/by-nc-nd/4.0/).
This paper discusses design selection with variation of the height-to-diameter (H/D) ratios and core-blanket configuration for a long-life modular gas-cooled fast reactor (GFR). The modular GFR is a nuclear reactor co...
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This paper discusses design selection with variation of the height-to-diameter (H/D) ratios and core-blanket configuration for a long-life modular gas-cooled fast reactor (GFR). The modular GFR is a nuclear reactor concept that proposes a longer operation time than a traditional nuclear reactor. The potential advantage of GFRs is the expected technological application in electricity, heat processing, nuclear burning, and breeding capability. However, the GFR design still needs an investigation to find the prospective core configuration design. The main objective is to understand the feasible design of long-life modular GFR using the openmc code. The openmc is an open-source Monte Carlo code that offers the exact solution to solve the neutron transport equation in a high-fidelity model and detailed three-dimensional geometry using Evaluated Nuclear Data File (ENDF/B-VII.1/V2) nuclear data and continuous energy. The H/D ratios give various core-type, that is, pancake, balance, and tall core, while core-blanket configuration explains different core layouts, that is, homogeneous, radial heterogeneous, and axial heterogeneous. The neutronics parameters characterized are the value of keff, fission reaction rate, neutron flux, fissile and fertile material, and conversion ratio distribution to know the optimum core design. The study implemented the traditional mesh tallies for the fission reaction rate distribution and the functional expansion tallies (FET) for the flux distribution. The FET feature used the Legendre polynomial for the axial calculation, whereas the Zernike polynomial for the radial calculation. Finally, the balance core-type with homogeneous configuration is the prospective design for advanced research.
In this work, openmc code has been used for reactor physics calculations of the IAEA 10 MW MTR benchmark. To perform the burnup dependent studies i.e. Beginning of Life (BOL) and End of Life (EOL) calculations are car...
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In this work, openmc code has been used for reactor physics calculations of the IAEA 10 MW MTR benchmark. To perform the burnup dependent studies i.e. Beginning of Life (BOL) and End of Life (EOL) calculations are carried out using openmc. The isotopic number densities have been generated using WIMS code and comparison of global core parameters has been performed. Along with the prevalent 9-isotope vector methodology, a set of 16-isotopes, based on their relative importance, have been employed in the whole core calculations and the corresponding computed values of integral parameters been compared. Comparison of effective core multiplication factor is made with studies performed by various organizations, employing different codes based on diffusion theory and Monte Carlo methodologies. The openmc predicted values of the multiplication factor and cell averaged thermal fluxes in central flux trap for HEU and LEU cores are found in good agreement with the corresponding published data. (C) 2015 Elsevier Ltd. All rights reserved.
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