Development and testing of a coupled SuperMC and SUBCHANFLOW code for LWR simulations

Yu, Louhanrong; Wang, Lei; Liu, Chao; Wu, Bin; Song, Jing; Wu, Yican

doi:10.1016/j.anucene.2020.107465

Cited by 3 publications

(2 citation statements)

References 8 publications

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“…As it is a new type of power plant (VVER-1200), more research is needed to understand its thermal hydraulics behavior and for upgrading its operation capacity. In conclusion, despite significant developments in CFD over the past couple of decades that have made it possible to perform thermal hydraulics analysis along with neutronics, further study is needed to develop suitable coupling codes such as SuperMC and SUBCHANFLOW, [28], [65], [70], [71], [72], [73]. [74], [78].…”

Section: Discussionmentioning

confidence: 99%

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Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Mollah

2023

1790-5044

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A nuclear power reactor's primary use is to generate thermal energy, which in turn produces electricity. The primary heat source is a nuclear fission event occurring inside the fuel rod. The convection heat is transmitted through the coolant by the heat energy generated at the fuel rod wall boundary. Better heat transfer is produced in the flow area by turbulence and irregularity. As a result, turbulent flow heat transfer may present a significant challenge when predicting and assessing the thermal performance of nuclear power reactors. Computational techniques in convective heat transfer have become indispensable for solving challenging issues in the fields of science and engineering thanks to the development of current sophisticated numerical methods and high-performance computer hardware. The development of novel computational techniques and models for complicated transport and multi-physical phenomena is constantly in demand throughout applicable disciplines. This chapter's objective is to provide some recent developments in computational techniques for convective heat transfer, taking into account research interests in the community of mass and heat transfer, and to showcase relevant applications in nuclear power plant engineering domains including future directions. This study describes the most recent advancements in nuclear reactor convective heat transfer research utilizing the computational fluid dynamics (CFD) method, particularly at Ansys Fluent. This work examines the convective heat transfer and fluid dynamics fluid dynamics for turbulent flows across three rod bundle sub-channels that are typical of those employed in the PWR-based VVER type reactor. In this paper, CFD analysis is carried out using the software tool Ansys Fluent. Temperature distribution profile, velocity profile, pressure drop, and turbulence properties were investigated in this study. Boundary conditions i.e. temperature, velocity, pressure, heat flux, and heat generation rate were applied in the sub-channel domain. The main obstacles and bright spots for the CFD methods in nuclear reactor engineering are discussed, which helps to further its further uses. We intend to research a full-length fuel bundle model for VVER-1200 in the future to gather specific fluid characteristic data and use the findings to analyze safety and operate nuclear power facilities in Bangladesh. This paper presents a thorough analysis of the sub-channel thermal hydraulic codes used in nuclear reactor core analysis. This review discusses several facets of previous experimental, analytical, and computational work on rod bundles and identifies potential future directions based on those earlier studies.

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Section: Discussionmentioning

confidence: 99%

“…The output of certain CFD simulations can then be combined with SYS-TH codes to get more precise approximations. Flow 3D [62], PorFlow [63], Flica-4 [64], Sub-chanflow [65]. Ansys Fluent/CFX [66], [67], OpenFoam [68], Comsol [69], and STAR-CCM+ [70] are a few examples of CFD programs.…”

Section: Computer Codes For Thmentioning

confidence: 99%

Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Mollah

2023

1790-5044

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Steady-state analyses of an smDFR with coupled Serpent-OpenFOAM calculation

Liu

Zhang

Wang

2022

Nuclear Engineering and Design

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Core design and neutronics analysis of lead liquid metal-based heat pipe cooled reactor for undersea purpose

Yinusa,

Mei,

Chen

et al. 2024

Kerntechnik

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Various heat pipe-cooled reactors have been conceptualized and designed in recent years, ranging from kilowatts to megawatts, suitable for remote and extreme environments. However, existing designs employ monolith cores, which are saddled with heat transfer deteriorations due to thermal expansion and subsequent deformations are some of the key challenges facing monolith heat pipe reactors. Addressing this, a novel 300 kW heat pipe has been designed by introducing lead liquid instead of the monolith. A comprehensive Monte Carlo neutronics analysis of the designed core has been conducted including its thermal analysis. The core design comes with the following parameters: fresh core excess reactivity of 2.9914 $ that can support a 10-year life cycle, and a negative temperature coefficient of −0.1648 pcm/K. The axial and radial peak power factors at critical states were calculated to be 1.67 and 1.35 respectively. Preliminary thermal hydraulics analysis shows maximum temperatures of 1044 K under normal operation and worst-case accident scenarios (3 heat-pipe failures) of 1,149.9 K. These temperatures are lower than the melting point of uranium nitride fuel and lower than the possible heat pipe boiling failure limit at 1,300 K. Thus, the possibility of fuel melting/heat pipe boiling accidents is avoided. A hypothetical criticality accident was simulated, assuming the core was flooded with water, leading to an increased fission rate. However, the implementation of three emergency rods in the core ensures the subcritical of the reactor.

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Development and testing of a coupled SuperMC and SUBCHANFLOW code for LWR simulations

Cited by 3 publications

References 8 publications

Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

Steady-state analyses of an smDFR with coupled Serpent-OpenFOAM calculation

Core design and neutronics analysis of lead liquid metal-based heat pipe cooled reactor for undersea purpose

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