Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5

Aragonés, José María Mateu; Ahnert, C.; Cabellos, Ó.; García-Herranz, Nuria; Aragonés-Ahnert, Vanessa

doi:10.13182/nt04-a3484

Cited by 12 publications

(5 citation statements)

References 5 publications

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“…There are also diverse varieties for the semi‐implicit and implicit coupling methods. Moreover, some other acceleration techniques are also massively developed and implemented to the codes temporal coupling . They aimed at the optimization of the intercodes synchronization as well .…”

Section: The Multiscale Thermal‐hydraulic Coupling Approachesmentioning

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: The Multiscale Thermal‐hydraulic Coupling Approachesmentioning

confidence: 99%

“…Moreover, some other acceleration techniques are also massively developed and implemented to the codes temporal coupling. [50][51][52][53] They aimed at the optimization of the intercodes synchronization as well. 54 gives a review of the temporal coupling methods in detail.…”

Section: A Classification Of the Approachesmentioning

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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“…An example of the staggered grid with the value of 9 = = = 1.5 was analyzed by the Universidad Politécnica de Madrid, coupling SIMTRAN and RELAP-5 and employed to analyze a MSLB (Aragonés et al 2004) and with COBAYA3 and COBRA-TF (G. ).…”

Section: Staggered Time Gridsmentioning

confidence: 99%

Study and lmplementation of Neutronic-Thermohydraulic Coupling Methods for Detailed Analysis of Transients in Nuclear Reactors for Safety Margins Optimization

Alcaráz¹

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Understanding the performance of a nuclear reactor is a complex task that covers different physical fields as the neutron behavior or fluid dynamics. These fields are not isolated, and there is a feedback among them. Along nuclear engineering history, different codes have been developed to analyze each of the physics in which the problem can be split in an isolated way. An example of this are the neutronic or the thermal-hydraulic codes. To study the performance of the reactor in a realistic way, it is necessary to take into account the interaction among the physics, or in other words, the codes have to be coupled. xiii ha facilitado en todo lo posible que pudiera seguir con mis estudios. Además, mis compañeros, Jorge, Paloma, Alfonso, Rosario, Carmen, Lustolde, David y Ruben se han asegurado de que acabara y me han hecho pasar buenos ratos, con unos buenos partidos y con frases como: "le vamos a decir a Edith que te quite el wifi" o "mucho ves tú Juego de Tronos".

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“…Within the nodes, a linear change of the transversal leakage is assumed to approximate the 6 averaged valuesL k of the transversal leakage at the sides of the hexagon from the average of the considered node L 0 and the average values L 0,k of the 6 neighbouring nodes (k = 1,2,..., 6).…”

Section: Transverse Leakage Termsmentioning

confidence: 99%

“…Similar work is done at ENEA for RELAP5/PARCS to simulate lead cooled fast reactors [43]. Previous work was done at UPM coupling SIMTRAN and RELAP5 [6].…”

Section: Contextmentioning

confidence: 99%

Core physics and safety analysis of Generation-IV Sodium Fast Reactors using existing and newly developed computational tools

Valero¹

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Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5

Cited by 12 publications

References 5 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

Study and lmplementation of Neutronic-Thermohydraulic Coupling Methods for Detailed Analysis of Transients in Nuclear Reactors for Safety Margins Optimization

Core physics and safety analysis of Generation-IV Sodium Fast Reactors using existing and newly developed computational tools

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