Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

Anderson, Nolan; Hassan, Yassin A.; Schultz, Richard R.

doi:10.1016/j.nucengdes.2007.06.008

Cited by 34 publications

(7 citation statements)

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“…In recent years, several additional efforts have been focused on similar approaches. Anderson et al (Anderson, Hassan and Schultz 2008) analyzed the Very High Temperature Reactor (VHTR). Bertolotto et al (Bertolotto, et al 2009) performed single phase mixing studies.…”

Section: Motivationmentioning

confidence: 99%

A novel multi-scale domain overlapping CFD/STH coupling methodology for multi-dimensional flows relevant to nuclear applications

Grunloh

Manera

2017

Nuclear Engineering and Design

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A novel multi-scale domain overlapping coupling methodology designed to couple a computational fluid dynamics (CFD) code with a system thermal hydraulic (STH) code was developed and its performance was investigated. The methodology has been implemented in the coupling infrastructure code Janus, developed at the University of Michigan, providing methods for the on-the-fly data transfer through memory between the commercial CFD code STAR-CCM+ and the US NRC best-estimate thermal hydraulic system code TRACE. Coupling between these two software packages is motivated by the desire to extend the range of applicability of TRACE to scenarios in which local momentum and energy transfer are important, such as three-dimensional mixing of localized slugs of deborated or cold water in the downcomer and lower plenum of a reactor pressure vessel. The intra-fluid shear forces necessary to correctly capture these effects are neglected in the TRACE equations of motion, but are readily calculated from CFD solutions. CFD/STH coupling implementations therefore have applications in reactor transients such as boron dilution scenarios, Anticipated Transient Without Scram (ATWS) and Main Steam Line Break (MSLB). The proposed method is based on aliasing all spatial sources and sinks of momentum in the CFD domain as frictional losses in the system code domain. The internal velocity fields and, consequently, the inertial component of the pressure field are maintained consistent between the CFD and STH domains through a complementary velocity-matching interface. In this paper, coupled simulations are performed on Cartesian and cylindrical geometry with emphasis on consistency, convergence, and stability during transient scenarios. Results show that the presented domain overlapping coupling method is capable of adjusting pressure and velocity profiles of multi-dimensional system code solutions to match CFD solutions accurately. Important characteristics of transient simulations were found to include the background flow rate, specifically the stabilizing effect of viscous forces, as well as the time derivative of the flow rate. Under certain adverse conditions, the basic coupling method is found to produce unstable behavior. A stabilization method for adjusting CFD data is laid out and found to significantly improve the method's performance under the most challenging conditions. Recommendations are laid out for further improving the coupling via advanced time stepping methods.

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Section: Motivationmentioning

confidence: 99%

A novel multi-scale domain overlapping CFD/STH coupling methodology for multi-dimensional flows relevant to nuclear applications

Grunloh

Manera

2017

Nuclear Engineering and Design

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“…To date, tremendous efforts have been dedicated to enabling the coupling between CFD and system codes. Anderson et al (2008) analysed a very high-temperature reactor (VHTR) using a coupled RELAP/CFD system where the 3-D flow in the outlet plenum was modelled using CFD. Papukchiev et al (2011) coupled ATHLET and ANSYS CFX, and then validations were done based on a pressure thermal shock related experiment for pressurized water-cooled reactor (PWR).…”

Section: Introductionmentioning

confidence: 99%

Sub-channel CFD for nuclear fuel bundles

Liu

Moulinec

et al. 2019

Nuclear Engineering and Design

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This paper presents a novel Computational Fluid Dynamics (CFD)-based sub-channel framework for nuclear power plants, which combines the advantageous features of modern CFD and traditional 1-D subchannel codes. The new method is capable of producing CFD-level 3-D results with locally desirable refinement when coupled with embedded resolved models, but because a very coarse mesh is used over most part of the domain, the computing cost is typically significantly smaller than that of a conventional CFD simulation. In this new Sub-Channel CFD (SubChCFD), a dual-mesh system is used, comprising, (i) a filtering mesh which aligns with the mesh used in typical sub-channel codes, enabling the use of existing engineering correlations to account for integral wall friction and heat transfer effects, and (ii) a computing mesh, which provides a platform for the solution of the governing equations with turbulence modelled using a mixinglength-type model. The method has been implemented in an open-source finite volume CFD code Code_Saturne and validated initially using a 5×5 bare bundle case based on the OECD/NEA MATiS-H benchmarking experiment. It has been found that SubChCFD is able of satisfactorily predicting both the velocity and temperature fields. To further investigate the performance for complex flow conditions, SubChCFD was applied to two full 3-D cases. The first is a 5×5 rod bundle case with local blockage in one of the sub-channels, creating significant localised cross flows. The second is a two-parallel-assembly channel with different input mass flow rates at the inlet of each assembly, allowing strong inter-assembly mixing. For both cases, SubChCFD has produced results which agree well with experiments and simulations using resolved CFD. It has also been demonstrated that SubChCFD exhibits excellent flexibility in comparison with traditional sub-channel codes and that it has the potential to serve as a substitution to sub-channel codes.

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“…In particular, promising is the coupling of CFD codes to best estimate system codes, adopted for safety analysis and transient thermal hydraulic calculations, as RELAP5 or TRACE (Aumiller et al, 2002;Bertolotto et al, 2009). The coupling between the codes limits the application of the more computationally expensive CFD to those areas where three dimensional flow effects and mixing phenomena are important, avoiding at the same time the modeling of the entire geometry (Anderson et al, 2008).…”

Section: Introductionmentioning

confidence: 99%

CFD study of an air–water flow inside helically coiled pipes

Colombo

Cammi

Guédon

et al. 2015

Progress in Nuclear Energy

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a b s t r a c tCFD is used to study an airewater mixture flowing inside helically coiled pipes, being at the moment considered for the Steam Generators (SGs) of different nuclear reactor projects of Generation IIIþ and Generation IV. The two-phase mixture is described through the EulerianeEulerian model and the adiabatic flow is simulated through the ANSYS FLUENT code. A twofold objective is pursued. On the one hand, obtaining an accurate estimation of physical quantities such as the frictional pressure drop and the void fraction. In this regard, CFD simulations can provide accurate predictions without being limited to a particular range of system parameters, which often constricts the application of empirical correlations. On the other hand, a better understanding of the role of the centrifugal force field and its effect on the two-phase flow field and the phase distributions is pursued.The effect of the centrifugal force field introduced by the geometry is characterized. Water is pushed by the centrifugal force towards the outer pipe wall, whereas air accumulates in the inner region of the pipe. The maximum of the mixture velocity is therefore shifted towards the inner pipe wall, as the air flows much faster than the water, having a considerably lower density. The flow field, as for the singlephase flow, is characterized by flow recirculation and vortices. Quantitatively, the simulation results are validated against the experimental data of Akagawa et al. (1971) for the void fraction and the frictional pressure drop. The relatively simple model of momentum interfacial transfer allows obtaining a very good agreement for the average void fraction and a satisfactory estimation of the frictional pressure drop and, at the same time, limits the computational cost of the simulations. Effects of changes in the diameter of the dispersed phase are described, as its value strongly affects the degree of interaction between the phases. In addition, a more precise treatment of the near wall region other than wall function results in a better definition of the liquid film at the wall, although an overestimation of the frictional pressure drop is obtained.

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Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

Cited by 34 publications

References 1 publication

A novel multi-scale domain overlapping CFD/STH coupling methodology for multi-dimensional flows relevant to nuclear applications

A novel multi-scale domain overlapping CFD/STH coupling methodology for multi-dimensional flows relevant to nuclear applications

Sub-channel CFD for nuclear fuel bundles

CFD study of an air–water flow inside helically coiled pipes

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