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This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Koz...
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... is a four-loop pressurized water reactor with hexagonal core geometry and horizontal steam generators. The steam is supplied to a 1000 MWe turbine. The core is of open type and contains 163 hexagonal fuel assemblies. The geometry of the reactor vessel is presented in Fig. ...
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... more quantitative comparison is given in Fig. 10 which also shows the good agreement between experiment and simu- lation. The measured and calculated mean temperature of the 163 assemblies are compared, where the numbering of the assem- blies goes, in analogy to Fig. 5, line by line from the lower left to the upper right ...
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... related to the fact that RANS models are based on the hypothesis of fully developed, isotropic turbulent flow where the Reynolds stress tensor is aligned to the main strain rate tensor. This hypothesises, on which LES modelling is not based, is physically not correct for the regions of the cold leg nozzles (impinging jet), the adjustment of the Fig. 11. Comparison of loop to fuel assembly mixing coefficients which are measured for Kozloduy-6 and calculated with Trio U. downcomer (diffuser) and the narrowing gap between the ellip- tic RPV bottom and the perforated core barrel (see Fig. 2). The turbulent flow is accelerated in this gap what cannot be treated correctly with a k-ε ...
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... corresponding mixing matrix has also been determined for the Kozloduy unit 6 by means of various commissioning experiments. The loop to fuel assembly mixing coefficients which have been measured for the Kozloduy unit 6 and which have been calculated with Trio U are given in Fig. 11. The good agreement between measurement and calculation shows that CFD codes can contribute to the evaluation of mixing coeffi- ...
Citations
... Drag force Schiller and Naumann (1935) Morsi and Alexander (1972) Clift et al. (1978) Ishii and Zuber (1979) Tomiyama (1998) Universal Drag Laws (Kolev, 2007) Lift force Moraga et al. (1999) Saffman (1968,1965) Legendre and Magnaudet (1998) Tomiyama (1998), later modified by Frank et al. (2004) Hibiki and Ishii (2007) Wall lubrication Antal et al. (1991) Tomiyama Chuang and Hibiki (2017) Interfacial pressure force Vaidheeswaran and de Bertodano (2016) Interfacial shear force Antal et al. (1991) Ishii and Hibiki (2011) Virtual mass force Ishii and Mishima (1980) Basset force Ishii and Hibiki (2011) Most current research focuses on the calibration of the coefficients which appear in those models (Chuang and Hibiki, 2017). Frank et al. (2004) applied the two-fluid model with the commercial ANSYS CFX-5 package in order to predict the development of upward directed gasliquid flows in a vertical pipe. ...
... To achieve a better understanding of the stratification and mixing behaviour of the coolant flow in various conditions, boundary conditions representative of the reactor conditions are essential in CFD simulations. Early studies (Bieder et al., 2007) used a uniform velocity field at the inlet of the cold legs, which was found to be inappropriate for the prediction of buoyant asymmetric flow in the cooling loop (Boumaza et al., 2014;Farkas et al., 2016). In addition, the strong swirl induced by the main circulation pump may also have strong effects and needs to be considered as well, (Petrov and Manera, 2011). ...
Nuclear Thermal-hydraulics (NTH) is a key element of Nuclear Power Plant (NPP) design and safety. It is the study of engineering systems where energy from nuclear fuel is transferred by a coolant to a power generation turbine or to the environment, by heat transfer, phase change and flow processes.
Computational modelling of NTH is already a vital part of both the design and safety substantiation of modern NPP. State-of-the-art NTH modelling tools have the potential to enable a level of design and operational optimisation for future NPP beyond anything seen in the past, delivering both improved safety and economic benefits.
NTH shares modelling challenges and tools with many industries where fluid dynamics is important, but perhaps uniquely, has to deal with all of them simultaneously on a huge range of geometrical scales. Moreover, practical NTH computer codes cannot be completely based on first principles, so incomplete knowledge results in uncertainties that must be rigorously quantified and mitigated.
This review highlights a selection of underlying phenomena most important to NPP, including: turbulence; heat transfer; bubble and droplet thermodynamics and transport; flow induced vibration; surface effects and ageing of structures, together with the multiscale aspects linking of all the above. Current practice and procedures by designers, regulators and utilities are outlined, including consideration of uncertainty management and the challenges of scaling from experiments. Classical ‘whole system’ tools are briefly summarised followed by more detailed discussion on finer grained modelling by 3D Computational Fluid Dynamics (CFD). This review also includes sections summarising emergent technology and international benchmarks and projects.
... This mixing pattern is the consequence of a complex three-dimensional fluid flow. Progress in computer hardware and numerical techniques has made it viable to predict these mixing patterns using CFD codes, see for instance the contributions by Moretti [13], Bieder [14], and Höhne ([16], [6]). However, as CFD codes contain more or less empirical models (for instance turbulence models) it is necessary to validate the predicted results using experimental data. ...
... Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of such thermal asymmetric loop operation. Such asymmetric flow distribution was reproduced and analyzed with Trio_U code using CAD data of the fabricated reactor pressure vessel [3]. However, the reactor part above the core inlet was not modelled and the perforated core barrel and support columns were modelled by flow resistance. ...
... The fluid is assumed to be incomepressible and the temperature dependency of the density is considered by the Boussinesq approximation. The physical properties of the fluid are those of pure water at 270℃ and 16 MPa [3]. ...
... The Temperature distribution of the vessel wall is shown in Fig. 10. The flow does not rotate significantly in counter clockwise direction due to the non-uniform and asymmetric azimuthal distribution of the cold leg nozzle as detected experimentally and in the calculation of Bieder et al. [3]. ...
Computational fluid dynamics (CFD) research for nuclear reactor safety dedicates to real scale reactor circuits under realistic thermal hydraulic conditions. In the framework of an OECD/NEA benchmark, CEA has attempted 10 years ago with the code TrioCFD to study the temperature distribution at the core inlet in a main steam line break (MSLB) accident scenario in a Bulgarian VVER1000 reactor. This work is resumed here by completing the geometry of the reactor pressure vessel (RPV) and by capitalizing both code development and high performance computing (HPC) resources. Before modelling the full scale RPV thermal-hydraulics, a PIRT (Phenomena Identification and Ranking Table) was performed to classify the existing physical phenomena in a ranking table. Three single effect validation test cases were defined in a test matrix. The CFD approach was validated single effect by single effect by reproducing the defined well suited test cases. The core outlet temperature distribution was measured during a commissioning steam generator separation test at Kozloduy nuclear power plant. This temperature distribution is compared to the CFD calculations and helps to validate integrally the full scale reactor calculation. Tetrahedral meshes of 50 to 400 million velocity control volumes were generated for the complete RPV; self-evidently the mesh refinement reflects the restrictions of the former defined test matrix. In the OECD benchmark, the core inlet temperature was calculated from the measured core outlet temperature by simple energy conservation. With the integral calculation we were able to review this process with the calculated core inlet and outlet temperature.
... So far, some researchers have already carried out 3D CFD numerical simulations of the complex flow and heat transfer phenomena in different parts of a PWR. Bieder et al. (2007) simulated the asymmetric flow distribution in the RPV of VVER-1000 V320 reactor with the Trio U code, a CFD code developed by the CEA Grenoble. For certain flow patterns, a Large Eddy Simulation (LES) turbulence model which can treat correctly hydraulic instabilities was better than RANS models to show the swirl formation. ...
... Transporting the colder water into the core region can lead to power excursion and core damage. LACYDON experiments of CEA (French 900 MWe PWR) and a commissioning experiment of the VVER1000 reactor at Kuzloduy [29] were used for validation; the temperature field in a French 900 MWe reactor has been analyzed without coupling to neutronics. ...
The main objective of the present paper is to give an overview of the actual state of the TrioCFD project. TrioCFD is a software developed for about 20 years in the Nuclear Energy Division of CEA. The code is designed to treat efficiently various physical problems, such as turbulent flows, fluid/solid coupling, multiphase flows or flows in porous media. The domains of application are mainly related to the nuclear industry, in particular to the interaction of turbulent flow with the solid structures of nuclear reactors, that TrioCFD is able to handle successfully thanks to its massive parallelism. The TrioCFD project integrates a major procedure of Verification and Validation (V&V). Numerous verification tests are performed to avoid unintentional modifications and assure a proper implementation of the numerical methods, models and options. Non-regression tests are automatically launched and provide an evaluation of the differences between the development version and the latest released version. The validation is the other part of procedure. Its purpose is to carry out reference calculations on a large range of physical problems and compare their results to analytical, experimental or literature results. It leads to the definition of compromises between accuracy, robustness and restitution time and guarantees that the requirement of a correct quality of results is met. Some typical examples of recent qualification studies in nuclear domain are presented here. Most of them correspond to international experimental or industrial facilities and combine various complex physical phenomena.
... Earlier the azimuthal rotation of loops flows in the reactor vessel was also shown by the results of the other coolant mixing start-up tests,f or example, during the commissioning of the Kozloduy NPP Unit 6. The V1000CT-2 thermalhydraulics benchmark is based on the measurements at these tests and many works have been performed in the framework of this benchmark [14]. ...
Nowadays the demands for accuracy of determining the weighted mean reactor power significantly increases due to implementing programs aimed at increasing installed power at nuclear power plants with WWER-1000 reactors. This accuracy strongly depends on the coolant temperature stratification in the primary circuit pipes, especially in the hot legs. Thus, the acute problem is to investigate the phenomenon on the basis of the full-scale tests experimental data and the calculated data of computational fluid dynamics simulations. The paper presents the full-scale experimental data on the coolant temperature distribution in the hot legs obtained via I&C systems, primarily via in-core monitoring system. These data were obtained at the pilot operation stage of Unit 4 Kalinin nuclear power plant in a stationary mode at nominal power level. To present the results of calculating simulation for the same mode CFD codes are used.
... Then the assembly outlet temperature (in related outlet section) increases to 605 K, which is 4 K higher than nominal condition. This behavior can be seen in Fig. 5. Similar results, which is about the coolant mixing in the core region, are reported in (Böttcher, 2008;Böttcher and Krüßmann, 2010;Bieder et al., 2007). ...
In this study, the Reactor Pressure Vessel (RPV) of the VVER-1200 (AES-2006) was modeled and thermal-hydraulic behavior of coolant in the RPV was investigated. The model set for the study includes 3 dimensional geometry description of the RPV. While the geometry is defined, the console elements (which help keeping RPV in its original position) are ignored since they are small and their effects on thermal or hydraulic behavior on the flow is negligible. The outer sections of the RPV were modeled in detail. For the modeling of inner sections, many of the core supporting structures locating at the lower plenum are modeled almost as exactly as they are. For the parts and components that are not modeled in detail and are simplified, porous media approach was adopted in order to perform realistic simulations. For the modeling, a detailed description of the VVER-1000 (V-392) RPV was used as the base model. The design difference in fuel assembly and RPV height for the VVER-1200 RPV is taken into consideration. As the result of using the described 3 dimensional modeling for the simulations of the thermal-hydraulic behavior of the VVER-1200 reactor core, the temperature distribution and pressure drop are calculated. The calculated values are in good agreement with the values reported in the literature.
... The growth of easily accessible computational power allows the CFD modeling of large components. Several studies reported in literature focused on the three-dimensional flow behavior in the reactor pressure vessel of different reactor types (Tinoco, 2002;Kwon et al., 2003;Bieder et al., 2007;Popov et al., 2007;Jeong and Han, 2008;Tallman et al., 2008;Yan and Mallner, 2009;Matsumoto et al., 2010;Tinoco et al., 2010), as this has a direct influence on the flow characteristics at the inlet of the reactor core and is therefore relevant for safety analyses (e.g. distributions of boron concentration and fluid temperature during main-steam line breaks of boron dilution scenarios). ...
A computational fluid dynamic (CFD) model for the pressure vessel of the evolutionary pressurized reactor (EPR™) was developed and validated. The aim of this model is the simulation of transients where three-dimensional effects play a strong role, such as boron dilution and main steam line break (MSLB) scenarios. First, a full solid (CAD) model has been built, that includes all details of the reactor pressure vessel (RPV) and the internals which are important for fluid dynamic analyses. The solid model has then been used as basis for the generation of the computational mesh necessary to carry out CFD simulations. Both a hexahedral and a polyhedral mesh have been created. The CFD model has been validated against experimental results of the JULIETTE facility, a 1:5 scaled mock-up of the EPR™ reactor RPV built by AREVA and equipped with advanced instrumentation.The performances of the hexahedral and the polyhedral meshes are investigated in relation to the agreement with experimental data, convergence and CPU requirements. In addition, the effect of the cold-leg swirls on the velocity field inside the RPV is investigated. These swirls mimic the effects of the main coolant recirculation pumps on the flow field at the entrance of the RPV. It is shown that the CFD model is able to capture the shift of the maximum velocity in the downcomer annulus observed in the experimental results. Good qualitative as well as quantitative agreement with the experimental data is achieved.
... The method used in Trio_U relies on a finite volume/finite element (FV/FE) numerical scheme and offer possibilities to switch between pure FE and mixed FE/FV formulations. The variables are computed as a combination of base functions that are assigned to each element, depending on the nodal values: the discretization is based on a staggered positioning: P1-non-conforming for the velocity and P0/P1 for the pressure (Bieder et al. (2007)). This discretization method yields a strong velocity/pressure coupling, while ensuring global conservation of the transported quantities (momentum and scalar in our case). ...
Since a long time, the Large Eddy Simulation (LES) concept is considered as a very promising candidate for advanced thermal hydraulic modeling in Nuclear Reactor Safety. This paper shows how LES is successfully applied in an industrial framework to free shear flows at high Reynolds numbers and to the associated transport of scalars (e.g. boron). Extensive verification and validation attempts towards this objective have already been performed for the Trio_U code (Höhne et al., 2006; Bieder et al., 2007; Bieder and Graffard, 2007). In the first part, this paper presents a short overview of what has been done for predicting the boron concentration at the core inlet under accident conditions. These calculations are then related to the demands of Best Practice Guidelines (BPG), which have been discussed by Mahaffy et al. (2007). It is shown, that high quality LES simulations for free sheer flows can be performed on tetrahedral meshes, what significantly simplifies the mesh generation procedure in topologically complex geometries. Guidelines specifically devoted to the LES framework are proposed to analyse the capability of numerical schemes to treat correctly the scalar transport.
... FZK presented full vessel CFD simulation results. In addition to these results some participants have published additional studies (Bieder, 2005(Bieder, , 2007Boettcher, 2008aBoettcher, , 2008bHoehne, 2007a;Yamaji, 2006) and comparisons of different turbulence models in accordance with the BPG (Mahaffy, 2008;Menter, 2002). ...
... This file is based on the actual geometry of the vessel V1000CT-2: SUMMARY RESULTS OF EXERCISE 1 ON VESSEL MIXING SIMULATION -© OECD/NEA 2010 and not the conceptual design geometry. It was tested in TRIO_U LES validation runs and in a sensitivity study (Bieder, 2004(Bieder, , 2005(Bieder, , 2007 which indicated the importance of using the actual vessel geometry for reliable results. ...
... A sensitivity study was performed with TRIO_U LES (Bieder, 2004(Bieder, , 2005(Bieder, , 2007 to analyse the significance of different modelling hypotheses. The results indicate the importance of the use of the actual vessel geometry. ...
