Scaling in nuclear reactor system thermal-hydraulics

D'Auria, Francesco Saverio; Galassi, G. M.

doi:10.1016/j.nucengdes.2010.06.010

Cited by 79 publications

(18 citation statements)

References 53 publications

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“…ref. [3] (see also section 3 below). The reference SYS TH codes are APROS**, ATHLET*, CATHARE*, KORSAR*, MARS**, RELAP*, SPACE**, TRAC*, TRACE** (where: * = precursor code; ** = lately developed code).…”

Section: -1980mentioning

confidence: 99%

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Perspectives in System Thermal-Hydraulics

D'Auria¹

2012

Nuclear Engineering and Technology

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Section: -1980mentioning

confidence: 99%

“…The scaling issue in licensing is considered in ref. [3] and a roadmap for addressing the issue is proposed. Even though there is no common acceptance for the proposed roadmap, the reader can refer to this paper for details.…”

Section: ) the Scaling Issue And Experimentsmentioning

confidence: 99%

Perspectives in System Thermal-Hydraulics

D'Auria¹

2012

Nuclear Engineering and Technology

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“…In a nuclear reactor, the thermal‐hydraulic phenomena take place at diverse spatial scales whose characteristic lengths vary from meters down to nanometers. Hence, the simulations are typically classified into three main scales: System scale: the phenomenon which is tightly related to the dynamics of the overall nuclear power plant or facilities is concerned at this scale, eg, the large break loss of coolant accident, where the primary cooling system of the reactor fails and loss the coolant Component scale: the phenomenon occurring in special reactor components such as the reactor core and the heat exchanger is concerned at this scale, eg, the departure from nucleate boiling which is very important concerning the reactor safety. Mesoscale: the fine description of the reactor thermal‐hydraulics is concerned at this scale, eg, the unsymmetrical coolant mixing in the downcomer and lower plenum (coolant of different temperatures entering the reactor vessel from different inlet nozzles will mix with each other in the vessel) in the reactor pressure vessel (RPV) …”

Section: Introductionmentioning

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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“…More detailed discussion and support of implementing system evaluation models for scaling of experiments is available in Ref. [12]. This approach allows an iterative process of improvement of both: scaled test facility during design phase and applied system evaluation models.…”

Section: Introductionmentioning

confidence: 99%

“…d ¼ small change j ¼ isothermal compressibility, m 2 /N j s ¼ isentropic compressibility, m 2 /N q ¼ density, m 3 /kg R ¼ summation U ¼ agent of change w ¼ state variable W a ¼ multiplier in Eqs (12). and(13), see Eq (14).…”

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confidence: 99%

Application of Fractional Scaling Analysis for Development and Design of Integral Effects Test Facility

Dzodzo

Oriolo

Ambrosini

et al. 2019

Journal of Nuclear Engineering and Radiation Science

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The aim of this paper is to present a fractional scaling analysis (FSA) application for a system with interacting components where multiple figures of merits need to be respected during complex transient accident scenario with several consecutive time sequences. This paper presents FSA application to the International Reactor Innovative and Secure (IRIS) reactor and Simulatore Pressurizzato per Esperienze di Sicurezza 3 (SPES3) integral effects test (IET) facility. The FSA was applied for the small break loss of coolant accident (SBLOCA) on the direct vessel injection (DVI) line as the most challenging transient scenario. The FSA methodologies were applied for two figures of merits: (1) reactor and containment vessels pressure responses, and (2) reactor vessel water collapsed level response. The space decomposition was performed first. The reactor vessel and containment vessel were divided in components so that important phenomena and their consequences can be evaluated in each of them. After that, the time decomposition in consecutive time sequences was performed for the considered transient (DVI SBLOCA) based on the starts, or ends, of the defining events. The configuration of the system in each time sequence might be different and dependent on the control system actions connecting, or disconnecting, various components of the system due to the valves openings, or closings. This way, the important phenomena and their consequences can be evaluated for each component and time sequence. Also, this paper presents and discusses options for deriving nondimensional groups and calculation of distortions between prototype and model responses for complex transients containing multiple consecutive time sequences. The input data for scaling analysis are based on the results of RELAP/GOTHIC analysis performed for IRIS and RELAP analysis performed for SPES3. The scaling analysis was applied iteratively several times for different IRIS and SPES3 configurations. Based on the intermediate results, some components in the IRIS and SPES3 were redesigned so that the distortions between IRIS and SPES3 responses are decreased.

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Scaling in nuclear reactor system thermal-hydraulics

Cited by 79 publications

References 53 publications

Perspectives in System Thermal-Hydraulics

Perspectives in System Thermal-Hydraulics

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

Application of Fractional Scaling Analysis for Development and Design of Integral Effects Test Facility

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