The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.

Computational Fluid Dynamics (CFD) provides means for high-fidelity 3D thermal-hydraulics analysis of Generation IV pool-type nuclear reactors. However, to be used in the decision making process a proof of code adequacy for intended application is required. This paper describes the Verification, Validation and Uncertainty Quantification (VVUQ) of a commercial CFD code Star-CCM + for forced, natural and mixed convection regimes in lead-bismuth eutectic (LBE) coolant pool flows. Code qualification is carried out according to an iterative VVUQ process aiming to reduce user effects. Validation data is produced in TALL-3D experimental facility - a 7 m high LBE loop featuring a 3D pool-type test section. Accurate prediction of mutual interaction between thermal stratification and mixing in the pool and the loop dynamics requires 3D analysis, especially during natural circulation. Solution verification is used to reduce the numerical uncertainty during code validation activities. Sensitivity Analysis (SA) is used to identify the effect of the most influential uncertain input parameters (UIPs) on numerical results. Two new visualization methods are proposed to enhance interpretation of the SA results. Dedicated experiments are performed according to the SA results to reduce the uncertainties in the most important UIPs. Automated calibration method for large CFD models is tested and demonstrated in combination with manual calibration using detailed temperature profile measurements in the pool. Calibration reveals the deficiencies in the modeling of heat losses owing to the presence of thermal bridges and other local effects in thermal insulation that are not explicitly modeled. It is demonstrated that Star-CCM + is able to predict thermal stratification and mixing phenomena in the pool type geometries. The results are supported by an Uncertainty Analysis (UA).

Coolant solidification is a phenomenon of potential safety importance for Liquid Metal Cooled Fast Reactors (LMFRs). Coolant solidification can affect local flow, heat transfer and lead to partial or complete blockage of the coolant flow paths jeopardizing decay heat removal function. It is also possible that reduced flow circulation may increase coolant temperature, counteract solidification and prevent complete blockage of the flow. Complex interactions between local physical phenomena of solidification and system scale natural circulation make modelling of solidification uncertain. Development and validation of adequate models requires validation grade experimental data. In this work we discuss results of analysis carried out in support of experiment development, specifically, design of a solidification test section and a test matrix for TALL-3D experimental facility (lead-bismuth eutectic (LBE) thermal-hydraulic loop). The aim of the analysis is experimental design that satisfies requirements stemming from the process of model qualification. We focus on two aspects: (i) design of solidification test section (STS) for investigation of solidification phenomena in lead-bismuth eutectic (LBE), and (ii) effect of the STS pool on the system scale behavior of the TALL-3D facility. Selection of the STS characteristics and experimental test matrix is supported using computational fluid dynamic (CFD) and system thermal-hydraulic (STH) codes. 

Pool-type primary system can improve the economy of lead-cooled fast reactors. However, partially filled pool of heavy liquid metal poses safety concerns related to seismic loads. Violent sloshing during earthquake-initiated fluid-structure interaction can lead to structural failures, gas entrapment and potential core voiding. Seismic isolation systems can be used to reduce the structural stresses, but its effect on sloshing is not straightforward. This paper presents a numerical study of seismic sloshing in ELSY reactor. The purpose is to evaluate the effects of seismic isolation system on sloshing at different levels of earthquake. Sloshing is modeled using computational fluid dynamics with a volume of fluid free surface capturing model. Earthquake is simulated using synthetic seismic data produced in SILER project as a boundary condition. Simultaneous verification and validation of the numerical model using a dam break experiment is presented. The adverse resonance effect of seismic isolation system is demonstrated in terms of sloshing-induced hydrodynamic loads and gas entrapment. Effectiveness of seismic isolation system is discussed separately for design and beyond design seismic levels. Partitioning baffles are proposed as a potential mitigation measure in the design and their effect is analyzed.

Steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues for pool type liquid metal cooled fast reactors. During SGTL/R, water is injected from high-pressure secondary side to low-pressure primary side. The possible consequences of such an event include void transport to the core that has adverse effects on the reactor performance including heat transfer deterioration and reactivity insertion. This paper addresses the potential transport of steam bubbles to the core and subsequent void accumulation in the primary system in ELSY conceptual reactor. A CFD model of the primary coolant system for nominal operation is developed and verified. Bubble motion is simulated using Lagrangian tracking of steam bubbles in Eulerian flow field. The effects of uncertainties in the bubble size distribution and bubble drag are addressed. A probabilistic methodology to estimate the core and primary system voiding rates is proposed and demonstrated. A family of drag correlations by Tomiyama et al. (1998) provide the best agreement with the available experimental data. Primary system and core voiding analysis demonstrate that the smallest (sub-millimeter) bubbles have the highest probability to be entrained and remain in the coolant flow. It is found that leaks at the bottom region of the SG result in larger rates of void accumulation.

Carbon-free nuclear power production can help to relieve the increasing energy demand in the world. New generation reactors with improved safety, sustainability, reliability, economy and security are being developed. Lead-cooled fast reactor (LFR) is one of them.

Lead-cooled reactors have many advantages related to the use of liquid metal coolants and pool-type primary system designs. Still, several technological challenges are to be overcome before their successful commercial deployment. The decisions on new designs and licensing of LFRs require use of code simulations, due to lack of operational experience and the fact that full scale experimental investigations are impractical. Code development and validation is necessary in order to reduce uncertainty in predictions for risk assessment and decision support. In order to make the decisions robust, i.e. not sensitive to remaining uncertainty, the process of collecting necessary evidences should iteratively converge. Extensive sensitivity and uncertainty analysis is necessary in order to make the process user-independent.

The pool-type design introduces 3D phenomena that require application of advanced modeling tools such as Computational Fluid Dynamics (CFD). The focus of this thesis is on the development and application of CFD modeling and general code validation methodology to the following issues of safety importance for LFR: (i) pool thermal stratification and mixing, (ii) steam generator tube leakage, (iii) seismic sloshing, and (iv) solidification.

Thermal mixing and stratification in a pool-type reactor can affect natural circulation stability and thus decay heat removal capability and pose thermal stresses on the reactor structures. Standalone System Thermal-hydraulics (STH) and CFD codes are either inadequate or too computationally expensive for analysis of such phenomena in prototypic conditions. In this work an approach to coupling of STH and CFD is proposed and implemented. For validation of standalone and coupled codes TALL-3D facility (lead-bismuth eutectic (LBE) loop) and test matrix was designed featuring significant two-way thermal-hydraulic feedbacks between the pool-type test section and the rest of the loop.

The possibility of core voiding due to bubble transport in case of steam generator tube leakage/rupture (SGTL/R) is a serious safety concern that can be a showstopper for LFR licensing. Voiding of the core increases risks for reactivity insertion accident (RIA) and/or local fuel damage due to deteriorated heat transfer. Bubble transport analysis for the ELSY (European Lead-cooled SYstem) reactor design demonstrated the effect of uncertainty in drag correlation, bubble size distributions and leak location on the core and primary system voiding. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.

The effect of seismic isolation (SI) system on the sloshing of the heavy metal coolant and respective mechanical loads on the vessel structures is studied for design basis and beyond design basis earthquakes. It was found that SI shifts the frequencies closer to resonance conditions for sloshing, thus increasing the loads due to slamming of coolant waves onto the vessel internal structures including reactor lid. Possible mitigation measures, such as partitioning baffles have been proposed and their effectiveness was analyzed. 

Coolant solidification is another serious safety issue in liquid metal cooled reactors that can cause partial or even full blockage of the coolant flow path and thermo-mechanical loads on the primary system structures. Validation of melting/solidification models implemented in CFD is a necessary pre-requisite for its application to risk analysis. In this work, the parametric analysis in support of the design of a solidification test section (STS) in TALL3D facility is carried out. The goals and requirements for the validation experiment and how they are satisfied in the design are discussed in the thesis.

In this chapter, an overview of the verification, validation, and uncertainty quantification process is offered. First, the context of the dialog with the safety authorities is explained, and the need for a thorough code validation procedure able to meet the regulatory safety requirements is highlighted. Then, the concept of code verification is introduced, and the main steps are described. The validation process is depicted next. Emphasis is made upon the identification of the physical phenomena of interest and upon the choice of adequate computational tools to capture them. The targeted validity domain of these computational tools and its dependence on available and accurate experimental data are detailed with respect to the issue of scaling. Finally, an overview of selected techniques for uncertainty and sensitivity analysis is provided. 

TALL-3D facility is a lead-bismuth eutectic (LBE) thermal-hydraulic loop designed to provide experimental data on thermal-hydraulics phenomena for validation of stand-alone and coupled System Thermal Hydraulics (STH) and Computational Fluid Dynamics (CFD) codes. Pre-test analysis is crucial for proper choice of experimental conditions at which the experimental data would be most useful for code validation and benchmarking. The goal of this work is to identify these conditions at which the experiment is challenging for the STH codes yet minimizes the 3D-effects from the test section on the loop dynamics. The analysis is focused on the identification of limit cycle flow oscillations in the TALL-3D facility main heater leg using a global optimum search tool GA-NPO to find a general region in the parameter space where oscillatory behavior is expected. As a second step a grid study is conducted outlining the boundaries between different stability modes. Phenomena, simulation results and methodology for selection of the test parameters are discussed in detail and recommendations for experiments are provided.