Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors

Toti, A; Vierendeels, Jan; Belloni, Francesco

doi:10.1016/j.anucene.2017.01.002

Cited by 38 publications

(11 citation statements)

References 12 publications

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“…It was applied for the simulation of natural and forced circulation tests and a heavy liquid metal system—TALL/3D (with a new semi‐implicit method) . At SCK‐CEN, RELAP5‐3D is coupled with FLUENT and tested using the data from the lead‐bismuth eutectic (LBE)—cooled test loop and facility . It was also applied to the analysis of a protected loss flow transient in an LBE‐cooled experimental reactor …”

Section: A Review Of the Multiscale Thermal‐hydraulic Coupled Codesmentioning

confidence: 99%

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The multiscale thermal‐hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

Zhang

2020

Int J Energy Res

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Summary Simulation is becoming increasingly important in the safety analysis of nuclear reactors nowadays. The physical phenomena in a nuclear power plant happen on three classified scales: system scale (phenomenon over the whole plant is concerned), component scale (phenomenon in specific component is concerned), and mesoscale (phenomenon in a small part of a component is concerned). Owing to the particular emphases, various codes are developed to simulate particular problems. System codes intend to predict the behavior of the whole power plant during normal or accidental phases (system scale). Subchannel codes are for core behavior predictions (component scale). CFD codes can simulate the thermal‐hydraulic in a fixed part of the plant (mesoscale). Those codes are coupled together to better predict the conditions in a nuclear reactor in last the two decades, which is the multiscale thermal‐hydraulic simulation approach for nuclear power systems. Diverse coupling approaches are developed and various coupling codes are implemented. This paper first proposes a classification of those approaches. It tells that a multiscale coupling is composed of five items: coupling architecture, operation mode, domain coupling, field mapping, and temporal coupling. Numbers of options are available for each item. For coupling architecture, it can be internal coupling, via‐IO coupling, server‐client, or serverless coupling. For operation mode, it can be either parallel or serial. For domain coupling, it can be either domain‐decomposition or domain‐overlapping coupling. For field mapping, it can be manual‐definition, processed by user‐developing toolkit, or handled by third‐party libraries. For temporal coupling, it can be explicit coupling, semi‐implicit coupling, or implicit coupling. An evaluation of the approaches is performed based on new‐proposed criterion. A general review of the multiscale thermal‐hydraulic coupled codes is made based on the classification. Especially, a review of the domain‐overlapping approach is present considering it is the most promising but challenging method for multiscale thermal‐hydraulic simulation.

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Section: A Review Of the Multiscale Thermal‐hydraulic Coupled Codesmentioning

confidence: 99%

“…82 At SCK-CEN, RELAP5-3D is coupled with FLUENT and tested using the data from the leadbismuth eutectic (LBE)-cooled test loop and facility. 83,84 T A B L E 7 Review of the multiscale thermal-hydraulic coupling of system and sub-channel codes…”

Section: Coupling Of System Code and Cfd Codementioning

confidence: 99%

The multiscale thermal‐hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

Zhang

2020

Int J Energy Res

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“…In the available literature, examples of the application of coupled STH/CFD codes to liquid metal thermal hydraulics exists, proving the capabilities of the adopted method in predicting the considered experimental conditions. Particularly, several coupled simulations were performed in works by Toti [4][5][6], aiming at reproducing the behavior of the MYRRHA reactor in postulated conditions, highlighting the advantages with respect to the stand-alone STH and CFD calculations.…”

Section: Introductionmentioning

confidence: 99%

STH/CFD Coupled Simulation of the Protected Loss of Flow Accident in the CIRCE-HERO Facility

et al. 2020

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The paper presents the application of a coupling methodology between Computational Fluid Dynamics (CFD) and System Thermal Hydraulic (STH) codes developed at the University of Pisa. The methodology was applied to the CIRCE-HERO facility in order to reproduce the recently performed experimental conditions simulating a Protected Loss Of Flow Accident (PLOFA). The facility consists of an internal loop, equipped with a fuel pin simulator and a steam generator, and an external pool. In this coupling application, the System code RELAP5 is adopted for the simulation of the internal loop while the CFD code ANSYS Fluent is used for the sake of simulating the pool. The connection between the two addressed domains is provided at the inlet and outlet section of the internal loop; a thermal coupling is also performed in order to reproduce the observed thermal stratification phenomenon. The obtained results are promising and a good agreement was obtained for both the mass flow rates and temperature measurements. Capabilities and limitations of the adopted coupling technique are discussed in the present paper also providing suggestions for improvements and developments to be achieved in the frame of future applications.

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“…Bavière et al (2014) simulated a sodium-cooled nuclear system with a coupled simulation of CATHARE2 and Trio_U, which allows energy and momentum feedback from CFD to the system code. Toti et al (2017) implemented an implicit domain decomposition algorithm to enable the coupling between the system code RELAP5-3D and ANSYS FLUENT for high fidelity safety analyses of pool-type reactors. The method showed good agreement with the experimental data of a loss of flow transient induced by local 3-D phenomena.…”

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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Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors

Cited by 38 publications

References 12 publications

The multiscale thermal‐hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

The multiscale thermal‐hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

STH/CFD Coupled Simulation of the Protected Loss of Flow Accident in the CIRCE-HERO Facility

Sub-channel CFD for nuclear fuel bundles

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