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

Yoon, Han Young; Cho, Hyoung Kyu; Lee, Jae Ryong; Park, Ik Kyu; Jeong, Jae Jun

doi:10.5516/net.02.2012.716

Cited by 32 publications

(6 citation statements)

References 21 publications

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“…An overestimation of the flow rate appears to be associated with underestimating the pressure drop along the cooling channel, especially in the upper vertical part of the cooling channel. For more detailed analysis, Case 4-1 was analyzed again using the CUPID code [13], which is a 3D two-phase flow analysis code [14,15]. The results showed a recirculation flow in the vertical part of the cooling channel at a quasi-steady state.…”

Section: Discussion On the Problems Of Calculation Resultsmentioning

confidence: 99%

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Lee¹,

Choi²,

Ha³

et al. 2017

Science and Technology of Nuclear Installations

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A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i) it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii) it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii) the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH) code and has a multidimensional TH component. The facility was modeled by using both one-and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

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Section: Discussion On the Problems Of Calculation Resultsmentioning

confidence: 99%

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Lee¹,

Choi²,

Ha³

et al. 2017

Science and Technology of Nuclear Installations

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“…108,109 At KAERI and Seoul National University (SNU), MARS was coupled to the CFD code CUPID 110 . 111 These codes were applied to the simulation of tests such as PASCAL, VAPER, 112 and FRIGG and to assess the PWR steam generator (SG). 113 Some other code pairs, eg, SAS4A-SASSYS-1/STAR-CD, 114 CATHERA/STAR-CCM +, 115 and FLOW1D/NPHASE.…”

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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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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“…Rather, a simple relation between heat transfer coefficient ℎ and Nusselt number was used to estimate a first approximation of the heat transfer coefficient. This relation is given as = ℎ (5) The Nusselt number is estimated from the Dittus-Boelter correlation for turbulent heat transfer:…”

Section: Selection Of Artificial Convective Heat Transfer Coefficientmentioning

confidence: 99%

“…System-level computer modeling of complex nuclear systems is increasingly becoming a trend due to the availability of advanced multi-physics computer programs and the increasing use of multiprocessor-based parallel computing hardware and software. Recently, many works have been published on thermal-hydraulics simulations of fluid flow and heat transfer in a single reactor component or in a complex large-scale assembly [1][2][3][4][5][6][7]. This type of system-level thermal-hydraulics model helps to better understand and to accurately predict the fluid flow and heat transfer not only in individual components but also the overall system and the interaction with each other.…”

Section: Introductionmentioning

confidence: 99%

Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

Mohanty

Soppet

Majumdar

et al. 2015

Nuclear Engineering and Design

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This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal-mechanical stress analysis were performed for typical thermalmechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.

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

Cited by 32 publications

References 21 publications

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

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

Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

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