atw 2017-09

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atw Vol. 62 (2017) | Issue 8/9 ı August/September ENVIRONMENT AND SAFETY 532 of the project can be summarized as follows: 1) Development and validation of advanced numerical approaches for the design and safety evaluation of advanced reactors. 2) Achievement of a new or extended validation base by creation of new reference data. 3) Establishment of Best Practice Guidelines, Verification & Validation methodologies, and Uncertainty Quantification methods. The project consortium consists of 23 European partners as depicted in Figure 2. Early 2016, an international collaboration with the US Nuclear Energy Advanced Modelling and Simulation program (NEAMS) has been initiated allowing to combine thermal hydraulic forces on both sides of the Atlantic. The project ensures relevance and contact with the system designers through the establishment of a Senior Advisory Committee in which the main European liquid metal cooled reactor designers are represented. 3 Temperature fluctuations A fundamental issue for the evaluation of liquid metal reactors is the modelling of the turbulent heat transfer over the complete range, from natural to mixed and to forced convection regimes. Current engineering tools apply statistical turbulence closures and adopt the concept of the turbulent Prandtl number based on the Reynolds analogy. This analogy is valid mainly for forced convective flows with a Prandtl number of order of unity. In the particular case of liquid metal, where the Prandtl number is much smaller than 1, the turbulent Prandtl Experiment High- Fidelity Reference Simulation Flow Separation Shear / Jet number concept is not applicable, and robust engineering turbulence models are needed. Thus, a model is required which can deal with all flow regimes simultaneously in liquid metal flows. Roelofs et al. [2015] identified some promising routes for improvement which are tested on relevant available geometrically simple test cases. The main focus in Europe is now on the extension of the validation base for mixed and natural convection regimes and for geometrically complex cases. To this purpose, the validation base will be extended by including mixing jets, flow separation, mixed convection, and rod bundle test cases as summarized in Table 1. The preparation of the experiments and of the high fidelity Direct Numerical Simulations (DNS) are in progress and new reference data sets produced by these efforts are currently being produced and will become available in the second half of 2017 and 2018. Figure 3 shows a 3D view of the design of a backward facing step test section Mixed Convection | | Tab. 1. Summary of SESAME reference data for turbulence model assessment to be generated. Rod Bundle to be installed in the German KASOLA sodium facility. In addition, a pragmatic simulation model which performs well in all flow regimes simultaneously would provide much added value. Further assessment and development of the promising approaches is foreseen. Emphasis is put on the creation of missing relevant reference data, on more robust testing of the models and on testing the models in geometrically more complex cases. Some of the selected approaches will provide a more pragmatic computational alternative to the high fidelity but very expensive LES. As such LES is practically not possible for the prediction of thermal fluctuations in a wide range of flows, as is required in most cases of industrial interest. This work will therefore probably also have a positive impact also on the simulation of light water reactors. The following promising models identified in Roelofs et al. [2015] will be further developed and/or implemented and validated: | | Fig. 3. KASOLA facility at KIT (left) and it’s backward facing step test section (right). Environment and Safety The SESAME Project: State of the Art Liquid Metal Thermal Hydraulics and Beyond ı F. Roelofs, A. Shams, A. Batta, V. Moreau, I. Di Piazza, A. Gerschenfeld, P. Planquart and M. Tarantino
atw Vol. 62 (2017) | Issue 8/9 ı August/September • A dedicated Algebraic Heat Flux Model developed by NRG (AHFM- NRG) in the STAR-CCM+ [Shams et al., 2014] and OpenFOAM codes. Within the context of the European In-Vessel Melt Retention (IVMR) project, this model was recently improved [Shams et al., 2017]. • Local turbulent Prandtl number model in the OpenFOAM code • The Turbulence Model for Buoyant Flows (TMBF) described by Carteciano and Grötzbach [2003] in the TransAT code 4 Mechanical vibrations Simulation experience and validation of fluid structure interaction and specifically of flow induced vibrations (e.g. in fuel assemblies) are still limited. Some attempts on flow induced vibration modelling are reported in the open literature. However, these attempts are mostly related to rod bundles in cross flow, usually taking place in heat exchangers and in steam generators as described by Benhamadouche and Laurence [2003] and Simoneau et al. [2011]. In contrast, the flow in a typical nuclear fuel assembly is oriented in axial direction parallel to the rod bundle. Nevertheless, Benhamadouche and Le-Maître [2009] report that even for water cooled reactors these flows may induce vibrations. Because of the high density of liquid metals, such effects are expected to be more pronounced and more important in liquid metal fast reactors. Attempts to analyze such situations are reported in De Pauw et al. [2015], De Ridder et al. [2013] and De Santis and Shams [2017]. Experimental data for validation are however missing. Therefore, a fundamental experiment in a seven pin rod bundle without spacers is foreseen, as depicted in Figure 4. This experiment uses water as simulant fluid allowing to obtain missing validation data both | | Fig. 4. Matched Index of Refraction vibration experiment (TUD). | | Fig. 5. Computational set-up for the high fidelity infinite wire wrap simulation (NRG). on the structure and on the fluid motion, thus supporting the development and validation of numerical approaches. The preparations of this experiment are in progress. In the meantime, model development is ongoing using different multi-physics simulation platforms. These simulations are also used to support the design of the experiment. 5 Core thermal hydraulics Most liquid metal cooled reactors employ wire wraps as a spacer in the fuel assemblies. Although liquid metal experiments are being carried out in the European context on wirewrapped fuel assemblies and to a lesser extent on fuel assemblies with grid spacers, the data to be retrieved from those experiments will be limited to the thermal field and will not include detailed information on the flow field. To derive reference data for the flow field in wire wrapped fuel assemblies, ideally a combination of experimental data based on a Matched Index of Refraction (MIR) technique and reference high fidelity numerical simulations is required. This is foreseen in the project. Firstly, a reference experiment is foreseen in a 7-pin rod bundle using water as a simulant fluid to obtain validation data for the flow field. In addition to this, numerical simulation data will be generated for a 61-pins rod bundle [Obabko et al., 2016] and an infinite number of pins rod bundle as shown in Figure 5 [Shams et al., 2015]. These data will allow closing the gap in the validation process of wire wrapped fuel assemblies. This high fidelity simulation for an infinite wire wrap ran on about 400 processors in parallel for about 9 months consuming 2.6 million core hours. Due to the design specific nature, validation of fuel assemblies using grid spacers is limited. Many publications can be found on grid spacers for water cooled reactors. However, specific geometric details on the grid spacer design are revealed seldom due to Intellectual Property Rights (IPR) issues. Data on the performance of grid spacers for liquid metal cooled reactors are largely missing, although some work has been performed and reported by Pacio et al. [2014] studying the coolant behavior in a lead cooled fuel assembly employing grid spacers. The grid spacer employed in this experiment is not representative for a grid spacer to be employed in a real reactor. Therefore, the SESAME project attempts to provide these missing data for the ALFRED reactor design by performing experiments in a liquid metal rod bundle. Such experiments will be performed for grid spacers without blockages and with blockages and will be accompanied by CFD simulations [Marinari et al., 2016]. Evaluation on the influence of the inter-wrapper flow is an important topic as a step towards a complete core simulation. Some experimental work has been performed in the past by Kamide et al. [1998] on this topic in Japan. However, as experience, both experimentally and computationally, is limited within Europe, it is foreseen to conduct a new experiment in which three reduced scale 7-pin fuel assemblies including their inter-wrapper space are employed. This experiment under design as depicted in Figure 6, will provide unique data for model development and validation. As multiple fuel assemblies need to be modelled to evaluate this, reduced or low resolution (coarse grid) CFD approaches as described by Roelofs et al. [2012] are foreseen from a computational point of view. Keeping in mind that the end goal will be to increase the modelling capabilities of a complete core, also multi-scale approaches will be tested by e.g., combining sub-channel codes with CFD codes. As a final step and by lack of experimental data, a code to code comparison will be performed employing ENVIRONMENT AND SAFETY 533 Environment and Safety The SESAME Project: State of the Art Liquid Metal Thermal Hydraulics and Beyond ı F. Roelofs, A. Shams, A. Batta, V. Moreau, I. Di Piazza, A. Gerschenfeld, P. Planquart and M. Tarantino
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