The Cupid Code Development and Assessment Strategy

Jeong, Jae Jun; Yoon, Han Young; Park, I.K.; Cho, Hyoung Kyu

doi:10.5516/net.2010.42.6.636

Cited by 72 publications

(10 citation statements)

References 37 publications

(46 reference statements)

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“…The CUPID code adopts the two-fluid model for two-phase flows [2]. In the two-fluid model, the mass, energy, and momentum equations for liquid and vapor phases are established separately, and then, they are linked by the interfacial mass, momentum, and energy transfer models.…”

Section: Two-phase Flow Modelsmentioning

confidence: 99%

“…They have been verified against standard conceptual problems of single-and two-phase flows and validated for thermal-hydraulic experiments in our previous studies [3][4][5]. The assessment strategy for the future verification and validation was outlined in Jeong et al [2]. Recently, the CUPID code was coupled with a system analysis code, MARS [6] and a threedimensional kinetics code, MASTER [7], which is expected to provide the advanced multi-scale and multi-physics calculations for many safety-related issues of the light water reactors [8].…”

Section: Introductionmentioning

confidence: 99%

“…These features require more detailed analysis with enhanced accuracy because of the complicated multi-dimensional effects. Motivated by these issues, the development of a numerical solver for a multi-dimensional thermal hydraulic analysis code, named CUPID [2], is in progress at Korea Atomic Energy Research Institute (KAERI). The objective of the development is to support a resolution for the thermal hydraulic issues regarding the transient multidimensional two-phase flow which can arise in an advanced light water reactor.…”

Section: Introductionmentioning

confidence: 99%

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Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Cho

Yun

Yoon

et al. 2012

Journal of Nuclear Science and Technology

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For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the downcomer boiling experiment (DOBO) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change from a bubbly flow to churn and mist flows and a circulation of liquid accelerated by bubbles. The twophase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling.

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Section: Two-phase Flow Modelsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Cho

Yun

Yoon

et al. 2012

Journal of Nuclear Science and Technology

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“…Numerical stability and accuracy of the CUPID code has been verified using conceptual problems [9]. Each of the problems was designed so that a specific term in the governing equations or a specific algorithm could be verified.…”

Section: Validation Of the Cupidmentioning

confidence: 99%

“…A two-phase CFD code, CUPID [6][7][8][9], has been developed at KAERI for the analysis of transient two-phase flows in nuclear reactor components. The CUPID code employs a two-fluid three-field model.…”

Section: Introductionmentioning

confidence: 99%

Multi-Scale Thermal-Hydraulic Analysis of PWRS Using the Cupid Code

Yoon¹,

Cho²,

Lee³

et al. 2012

Nuclear Engineering and Technology

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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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The Cupid Code Development and Assessment Strategy

Cited by 72 publications

References 37 publications

Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Assessment of the two-phase flow models in the CUPID code using the downcomer boiling experiment

Multi-Scale Thermal-Hydraulic Analysis of PWRS Using the Cupid Code

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

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