Use of coupled code technique for Best Estimate safety analysis of nuclear power plants

Peng

Science and Technology of Nuclear Installations

et al. 2021

This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT. The implementations and parallel performances of block Gauss–Seidel-type and block Jacobi-type Picard iterative algorithms have been investigated. In addition, this paper introduces two adaptive load-balancing algorithms into the neutronics and thermal-hydraulics coupled simulation to reduce the time cost of computation. Considering that the different scalability of OpenMC and FLUENT limits the performance of block Gauss–Seidel algorithm, an adaptive load-balancing algorithm that can increase the number of nodes dynamically is proposed to improve its efficiency. Moreover, with the natural parallelism of block Jacobi algorithm, another adaptive load-balancing algorithm is proposed to improve its performance. A 3 x 3 PWR fuel pin model and a 1000 MWt ABR metallic benchmark core were used to compare the performances of the two algorithms and verify the effectiveness of the two adaptive load-balancing algorithms. The results show that the adaptive load-balancing algorithms proposed in this paper can greatly improve the computing efficiency of block Jacobi algorithm and improve the performance of block Gauss–Seidel algorithm when the number of nodes is large. In addition, the adaptive load-balancing algorithms are especially effective when a case demands different computational power of OpenMC and FLUENT.

Section: Introductionmentioning

confidence: 99%

An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing

Peng

Science and Technology of Nuclear Installations

et al. 2021

“…In the frame of different European projects, eg, NURESIM, NURISP, and NURESAFE, multiscale coupling approaches were systematically elaborated and various safety‐relevant phenomena were investigated in detail from a multiscale perspective. Different publications provide guidelines for potential coupling approaches.…”

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

2020

Int J Energy Res

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.

Progress in Nuclear Energy

“…In the past, the lack of a full understanding has been solved with the adoption of conservative safety limits that guarantee safety operation despite limiting the optimal exploitation of the plant. Nowadays, overcome those limitations is made possible by the more and more powerful computational resources available for the optimization of the design and the improvement of the safety evaluations (Bousbia-Salah and D'Auria, 2007).…”

Section: Introductionmentioning

confidence: 99%

CFD study of an air–water flow inside helically coiled pipes

Colombo

Cammi

Guédon

et al. 2015

a b s t r a c tCFD is used to study an airewater mixture flowing inside helically coiled pipes, being at the moment considered for the Steam Generators (SGs) of different nuclear reactor projects of Generation IIIþ and Generation IV. The two-phase mixture is described through the EulerianeEulerian model and the adiabatic flow is simulated through the ANSYS FLUENT code. A twofold objective is pursued. On the one hand, obtaining an accurate estimation of physical quantities such as the frictional pressure drop and the void fraction. In this regard, CFD simulations can provide accurate predictions without being limited to a particular range of system parameters, which often constricts the application of empirical correlations. On the other hand, a better understanding of the role of the centrifugal force field and its effect on the two-phase flow field and the phase distributions is pursued.The effect of the centrifugal force field introduced by the geometry is characterized. Water is pushed by the centrifugal force towards the outer pipe wall, whereas air accumulates in the inner region of the pipe. The maximum of the mixture velocity is therefore shifted towards the inner pipe wall, as the air flows much faster than the water, having a considerably lower density. The flow field, as for the singlephase flow, is characterized by flow recirculation and vortices. Quantitatively, the simulation results are validated against the experimental data of Akagawa et al. (1971) for the void fraction and the frictional pressure drop. The relatively simple model of momentum interfacial transfer allows obtaining a very good agreement for the average void fraction and a satisfactory estimation of the frictional pressure drop and, at the same time, limits the computational cost of the simulations. Effects of changes in the diameter of the dispersed phase are described, as its value strongly affects the degree of interaction between the phases. In addition, a more precise treatment of the near wall region other than wall function results in a better definition of the liquid film at the wall, although an overestimation of the frictional pressure drop is obtained.