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[en] • In-vessel retention (IVR) is one of the possible SA mitigation strategies ⇒ already implemented in several reactors currently operating (VVER440) or under construction (AP1000, APR1400, CPR1000+…); • In Europe, the feasibility of IVR for high power reactors still needs to be demonstrated: – Uncertainties regarding corium behaviour; – Penalizing hypotheses lead to an over conservative assessment of the thermal loads. • The IVMR project was launched to “improve the methodology by reducing the degree of conservatism in order to derive more realistic safety margins”. ⇒ extensive program covering experimental faculties and integral codes, but also the possible uses of CFD. • This presentation will focus on the work performed at EDF, partly in the frame of the IVMR project, using the NEPTUNECFD software.
[en] Highlights: • Modelling of boiling sodium flows in a multiphase flow solver. • Rod heated with a constant heat flux in a pipe liquid metal flow. • Sodium boiling flow around a rod heated with a constant heat. • Computations in progress in an assembly constituted of 19 pins equipped with a wrapped wire. - Abstract: In France, Sodium-cooled Fast Reactors (SFR) have recently received a renewed interest. In 2006, the decision was taken by the French Government to initiate research in order to build a first Generation IV prototype (called ASTRID) by 2020. The improvement in the safety of SFR is one of the key points in their conception. Accidental sequences may lead to a significant increase of reactivity. This is for instance the case when the sodium coolant is boiling within the fissile zone. As a consequence, incipient boiling superheat of sodium is an important parameter, as it can influence boiling process which may appear during some postulated accidents as the unexpected loss of flow (ULOF). The problem is that despite the reduction in core power, when boiling conditions are reached, the flow decreases progressively and vapour expands into the heating zone. A crucial investigating way is to optimize the design of the fissile assemblies of the core in order to lead to stable boiling during a ULOF accident, without voiding of the fissile zone. Moreover, in order to evaluate nuclear plant design and safety, a CFD tool has been developed at EDF in the framework of the nuclear industry. Advanced models dedicated to boiling flows have been implemented and validated against experimental data for ten years now including a wall law for boiling flows, wall transfer for nucleate boiling, turbulence and polydispersion model. This paper aims at evaluating the generalization of these models to SFR. At least two main issues are encountered. Firstly, at low Prandtl numbers such as those of liquid metal, classical approaches derived for unity or close to unity fail to accurately predict the heat transfer. In order to evaluate the wall law implemented in the CFD tool, computations have been compared with KALLA experimental results obtained in the case of a rod heated with a constant heat flux which is concentrically embedded in a pipe liquid metal flow (single-phase flow). Secondly, the incipient boiling superheat of sodium is quite different from that of conventional fluids. As a consequence, the nucleate boiling model has been improved and validated against the Charlety’s experiment where a rod heated with a constant heat flux is concentrically embedded in a pipe sodium flow. For different values of the heat flux, the pressure is measured at different locations as function of the mass flow rate. A reasonable agreement has been reached which is very encouraging for further applications. Finally, preliminary computations have been carried out in an assembly constituted of 19 pins equipped with a wrapped wire where partial experimental results are available. Computations have shown a pressure drop at the end of the heated length due to the sudden increase of the hydraulic diameter. Thus, the pressure can drop below the vapour pressure leading to liquid vaporization. This first result supports the assumption of boiling in the upper subassembly zone which could possibly lead to a sodium boiling stabilization.
[en] Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune-CFD code. • The model has been validated against 150 tests. • Neptune-CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.
[en] The NEPTUNE-CFD code, which is based on an Eulerian two-fluid model, is developed within the framework of the NEPTUNE project, financially supported by CEA (Commissariat a l'Energie Atomique), EDF, IRSN (Institut de Radioprotection et de Surete Nucleaire) and AREVA-NP. NEPTUNE-CFD is mainly focused on Nuclear Reactor Safety applications involving two-phase flows, like two-phase Pressurized Thermal Shock (PTS) and Departure from Nucleate Boiling (DNB). Since the maturity of two-phase CFD has not reached yet the same level as single phase CFD, an important work of model development and thorough validation is needed, as stated for example in NEA/CSNI Writing Group dedicated to the 'Extension of CFD Codes to Two-Phase Flow Safety Problems' (draft6c, 2009). Many of these applications involve bubbly and boiling flows, and therefore it is essential to validate the software on such configurations. In particular, this is crucial for applications to flow in PWR fuel assemblies, including studies related to DNB. This work aims at presenting the present status of NEPTUNE-CFD validation in this area, as a step in an iterative process of improvement. To this end, this paper presents NEPTUNE-CFD code validation against four test cases based on experimental results. These data have been selected to allow separate effects validation. The adequacy of the measured quantities and the corresponding basic model of the CFD code to validate is underlined in each case. The selected test cases are the following. The Liu and Bankhoff experiment (1993) is an adiabatic air-water bubbly flow inside a vertical pipe. It allows to validate forces applied to the bubbles. The Bel F'Dhila and Simonin (1992) experiment is an adiabatic bubbly air-water flow inside a sudden pipe expansion. It allows to validate the dynamic models and turbulence. The DEBORA (CEA, 2002) and the ASU (Arizona State University, Hassan 1990) facilities provide results for boiling flows inside a vertical pipe. The working fluid is refrigerant R12 for DEBORA and R113 for ASU. Both allow to validate the nucleation modeling on a heated wall, and ASU allows also the validation of the two-phase wall function (Mimouni, 2009). A key feature of this work is that all these calculations were performed with a single standard version (1.0.8) of NEPTUNE-CFD, and with a single and consistent set of models, avoiding case-dependent 'tuning' of the modeling: a RANS approach with a Reynolds Stress Model for the turbulence of the continuous phase; the drag force from Ishii (1990), the added mass from Zuber (1964), the lift force from Tomiyama (1998) and a turbulent dispersion force are chosen for the dispersed phase. The NEA/CSNI Best Practice Guidelines were followed as much as possible, especially in the mesh generation process by keeping acceptable quality for the grids, by exploring the grid convergence, and also by assessing the numerical convergence. Comparisons with experimental data show that NEPTUNE-CFD has captured experimental profiles with reasonable accuracy for dynamical quantities and void fraction. Improvement must be done for the prediction of the bubbles size distribution. The need of new experiments will also be addressed to validate other specific models, like those used for bubble condensation in subcooled convective flow, which is the goal of the new TESS program. A companion paper presenting validation computations against these very recently obtained data is also submitted to the workshop. (authors)
[en] NEPTUNE-CFD is a code based on a 3D transient Eulerian two-fluid model. It is developed within the framework of the NEPTUNE project, financially supported by CEA (Commissariat a l'energie Atomique), EDF, IRSN (Institut de Radioprotection et de Surete Nucleaire) and AREVA-NP. NEPTUNE-CFD is mainly devoted to Nuclear Reactor Safety (NRS) issues. One of the main application targets is the two-phase Pressurized Thermal Shock (PTS), which is related to PWR reactor pressure vessel lifetime safety studies, when sub-cooled water from Emergency Core Cooling (ECC) system is injected into the uncovered cold leg, and penetrates into the RPV downcomer. Following the NEA/CSNI Best Practice Guidelines (BPGs), relevant PTS-scenarios have been identified; a Phenomena Identification and Ranking Table (PIRT) process, the related state of the art of modeling and the existing data basis have been reviewed by a panel of European experts, mainly within the ECORA and NURESIM projects. Consistently, the following five experiments were selected for the NEPTUNECFD validation presented in this paper. The first four are useful for separate effects validation. The Fabre et al., 1987, experiment is a co-current smooth and wavy Air Water Stratified (AWST) flow in a rectangular channel with detailed measurements of turbulence and velocities. It allows to validate the dynamic models (turbulence and interfacial friction). The Lim et al., 1984, experiment is a co-current smooth and wavy Steam Water Stratified (SWST) flow in a rectangular channel with measurements of the steam flow rates at six axial positions along the channel. It allows to validate the condensation models. The Bonetto and Lahey, 1993, and the Iguchi et al., 1998, experiments deal with a water jet impingement on a water pool free surface in air environment. In the first one, the void fraction and the mean velocities are measured whereas in the second one, mean and rms velocities are measured. Both allow to validate the dynamic models in the situation of a jet impinging a pool free surface - a challenging case for two-phase CFD - the first one mainly versus gas entrainment phenomena and the second one mainly versus turbulence. Finally, the COSI experiment represents a cold leg scaled 1/100 for volume and power from a 900 MW PWR under LOCA conditions, and therefore can be used for global validation. The measurements include condensation rates and temperature profiles at eight axial positions in the pipe, at various ECC flow rates, inlet steam flow rates and water level in the cold leg. It allows to validate all the models involved in a PTS. The five experiments were calculated with NEPTUNE-CFD 1.0.8 with the same set of models. It includes the Large Interface Method (LIM) and a RANS approach with (k-ε) transport equations in each phase. The available measurement uncertainties are generally smaller than typical calculation / measurement discrepancies. Unfortunately there are often lacks in the available experimental data which stress the need for new ones such as the on-going TOPFLOW-PTS. Following the BPGs, the mesh sensitivity is investigated. The five experiments all deal of course with free surfaces. In this case, the BPGs concede that it is not possible to obtain completely grid-independent results and this is actually what we found. However, some calculations show that the LIM transfer models at the free surface, which are written under the format of wall-functions, allow to better master some mesh size effects, confirming the adequacy of this modeling approach for the industrial application. (authors)
[en] Highlights: ► The two-phase Pressurized Thermal Shock (PTS) is a key thermohydraulics issue for PWR safety. ► The dynamic and condensation models are firstly validated separately. ► Then the global validation is done with the COSI experiment. ► All the calculations performed with the same set of models both in the Large Interface Method and in the k–ε approach for turbulence substantiate the application of the tool to PTS. - Abstract: NEPTUNECFD is a code based on a 3D transient Eulerian two-fluid model. One of the main application targets is the two-phase Pressurized Thermal Shock (PTS), which is related to PWR Reactor Pressure Vessel (RPV) lifetime safety studies, when sub-cooled water from Emergency Core Cooling (ECC) system is injected into the possibly uncovered cold leg and penetrates into the RPV downcomer. Five experiments were selected for the validation, a selection reviewed by a panel of European experts. The dynamic models are validated with a co-current smooth and wavy air–water stratified flow in a rectangular channel with detailed measurements of turbulence and velocities. The condensation models are validated with a co-current smooth and wavy steam-water stratified flow in a rectangular channel with measurements of the steam flow rates. The dynamic models are validated in the situation of a jet impinging a pool free surface with two experiments dealing with a water jet impingement on a water pool free surface in air environment. Finally, all the models involved in the reactor conditions are validated with the COSI experiment. The calculations are done with the same set of Large Interface Method models and a RANS (k–ε) approach for turbulence. They substantiate the application of the tool to PTS studies.
[en] NEPTUNE CFD is a Computational Multi-(Fluid) Dynamics code dedicated to the simulation of multiphase flows, primarily targeting nuclear thermo-hydraulics applications, such as the departure from nuclear boiling (DNB) or the two-phase Pressurized Thermal Shock (PTS). It is co-developed within the joint research/development project NEPTUNE (AREVA, CEA, EDF, IRSN) since 2001. Over the years, to address the aforementioned applications, dedicated physical models and numerical methods have been developed and implemented in the code, including specific sets of models for turbulent boiling flows and two-phase non-adiabatic stratified flows. This paper aims at summarizing the current main modeling capabilities of the code, and gives an overview of the associated validation database. A brief summary of emerging applications of the code, such as containment simulation during a potential severe accident or in-vessel retention, is also provided. (authors)
[en] In this paper, we present numerical results obtained with the multifield CFD code NEPTUNE_CFD in the framework of the OECD/NRC PWR Subchannel and Bundle Tests (PSBT) international benchmark, focusing on the simulation of five selected runs of the steady-state subchannel exercise. The propagation of the estimated experimental uncertainties on the simulations results is investigated, as well as the mesh sensitivity of the axial evolution of the mean void-fraction by using three grid levels. Last, calculation results using a devoted model for the bubble-size distribution are presented. (author)
[en] Nuclear Power Plants are subjected to a variety of ageing mechanisms and, at the same time, exposed to potential Pressurized Thermal Shock (PTS) – characterized by a rapid cooling of the Reactor Pressure Vessel (RPV) wall. In this context, NEPTUNE-CFD is developed and used to model two-phase PTS in an industrial configuration, providing temperature and pressure fields required to assess the integrity of the RPV. Furthermore, when using CFD for nuclear safety demonstration purposes, EDF applies a methodology based on physical analysis, verification, validation and application to industrial scale (V&V), to demonstrate the quality of, and the confidence in results obtained. By following this methodology, each step must be proved to be consistent with the others, and with the final goal of the calculations. To this effect, a chart demonstrating how far the validation step of NEPTUNE-CFD is covering the PTS application will be drawn. A selection of the code verification and validation cases against different experiments will be described. For results consistency, a single and mature set of models – resulting from the knowledge acquired during the code development over the last decade – has been used. From these development and validation feedbacks, a methodology has been set up to perform industrial computations. Finally, the guidelines of this methodology based on NEPTUNE-CFD and SYRTHES coupling – to take into account the conjugate heat transfer between liquid and solid – will be presented. A short overview of the engineering approach will be given – starting from the meshing process, up to the results post-treatment and analysis.
[en] Highlights: • The physical closures of two CFD codes are compared on a series of experiments. • Both sets of closures deliver acceptable predictions for the experimental database. • The differences in the closures are discussed. • A systematic study of the propagation of the uncertainties was also performed. • The void fraction evidences the highest sensitivity to the closure parameters. - Abstract: The nuclear industry is interested in better understanding the behavior of turbulent boiling flows and in using modern computational tools for the design and analysis of advanced fuels and reactors and for simulation and study of mitigation strategies in accident scenarios. Such interests serve as drivers for the advancement of the 3-dimensional multiphase Computational Fluid Dynamics approach. A pair of parallel efforts have been underway in Europe and in the United States, the NEPTUNE and CASL programs respectively, that aim at delivering advanced simulation tools that will enable improved safety and economy of operations of the reactor fleet. Results from a collaboration between these two efforts, aimed at advancing the understanding of multiphase closures for pressurized water reactor (PWR) application, are presented. Particular attention is paid to assessment and analysis of the different physical models implemented in the CFD tools respectively used in the NEPTUNE and the CASL programs, for application to turbulent two-phase bubbly flows. The experiments conducted by Liu and Bankoff (Liu, 1989; Liu and Bankoff, 1993a,b) are selected for benchmarking, and predictions from NEPTUNECFD and STAR-CCM+ codes are presented for a broad range of flow conditions and with void fractions varying between 0 and 50%. Comparison of the CFD simulations and experimental measurements reveals that a similar level of accuracy is achieved in the two codes. The differences in both sets of closure models are analyzed, and their capability to capture the main features of the flow over a wide range of experimental conditions are discussed. Finally, a parametric sensitivity study for the set of closures used in STAR-CCM+ is included to serve as a preview of how uncertainty quantification methods can provide insights into interactions between closures of different phenomena. In conclusion, it is seen that, the multi-CFD-code approach and uncertainty analysis of a set of closures in a particular CFD code, are both of great value in assessing the limitations and the level of maturity of multiphase hydrodynamic closures, and can serve as aids in further improving them.