Balance equations

Bestion, D.; Morel, Christophe

doi:10.1016/b978-0-08-100662-7.00005-1

Cited by 3 publications

(4 citation statements)

References 19 publications

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“…1-D, 3-D, two fields, three fields, etc.) and constitute the current state of art in nuclear system thermal-hydraulic modeling, Bestion and Morel (2017). • Computer codes are needed for the analysis of complex transient scenarios in nuclear reactors: those codes can be considered as the baskets of expertise and database in nuclear thermal hydraulics, Bestion (2017a).…”

Section: Ormentioning

confidence: 99%

“…Entropy evaluation and the 2nd principle of thermodynamics are connected with energy changes and are considered in the formulations of current PDE, see e.g. Bestion and Morel (2017). Otherwise, various irreversible flow evolutions may lead to entropy increase during an accident scenario: at the level of overall nuclear reactor system, one may look at blow-down and, at a local level, one may look at pressure drops at geometric discontinuities, or even at situations connected with flow regimes changes and non-fully developed flow conditions.…”

Section: The Challengesmentioning

confidence: 99%

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Challenges and concerns for development of nuclear thermal-hydraulics

D'Auria

Hassan

2021

Nuclear Engineering and Design

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Section: Ormentioning

confidence: 99%

Section: The Challengesmentioning

confidence: 99%

Challenges and concerns for development of nuclear thermal-hydraulics

D'Auria

Hassan

2021

Nuclear Engineering and Design

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“…Nuclear power reactors use the common steam cycles to convert heat into electricity. Predicting the temperature and velocity distributions in the reactor's various sections is one of the main goals of reactor thermalhydraulics (TH), [10], [11], [12], [13], [14], [15], [16], [17], [18], [19], [20], [21], [22]. The reactor core, which generates heat and has the greatest temperatures, is the most significant component of the nuclear reactor.…”

Section: Introductionmentioning

confidence: 99%

“…∂𝑢𝑖/∂𝑡+∂/∂𝑥(𝑢𝑖𝑢𝑗)=𝑓 𝑖 −(1/𝜌)(∂𝑝/∂𝑥 𝑖 )+(𝜇/𝜌)(∂ 2 𝑢𝑖/∂𝑥𝑗∂ 𝑥𝑗)+(1/3)(𝜇/𝜌)(∂/∂𝑥𝑖)(∂𝑢𝑘/∂𝑥𝑘)(11)The momentum equation may be further reduced for incompressible fluids, where the continuity equation necessitates ∂𝑢 𝑘 /∂𝑥 𝑘 =0. 𝑢𝑖/∂𝑡+∂/∂(𝑢𝑖𝑢𝑗)=𝑓𝑖−(1/𝜌)(∂𝑝/∂𝑥𝑖)+(𝜇/𝜌)(∂ 2 𝑢𝑖/∂𝑥𝑗∂𝑥𝑗)(12)…”

mentioning

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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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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An investigation of numerical performance enhancement of RELAP5: Numerical stability, higher resolution and an alternative constitutive relation

Shirvan

et al. 2018

Nuclear Engineering and Design

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Balance equations

Cited by 3 publications

References 19 publications

Challenges and concerns for development of nuclear thermal-hydraulics

Challenges and concerns for development of nuclear thermal-hydraulics

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

An investigation of numerical performance enhancement of RELAP5: Numerical stability, higher resolution and an alternative constitutive relation

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