Generation IV reactors are designed to provide sustainable energy generation, minimize waste production and excel in safety. Due to lack of operational experience, ever evolving design and stringent safety requirements, these novel reactors have to rely heavily on simulations.

Best estimate one-dimensional (1D) system thermal-hydraulics (STH) codes, originally intended for simulating water-cooled reactor systems with high coolant mass flow rates, are unable to capture complex three-dimensional (3D) phenomena in liquid metal cooled pool-type reactors. Computational fluid dynamics (CFD) codes are capable of resolving the 3D effects, however applying these methods with high resolution for the whole primary system results in prohibiting computational cost.

At the same time, there are system components where flow can, with reasonable accuracy, be approximated with 1D models (e.g. core channels, some heat exchangers, etc.). One of the proposed solutions in order to achieve adequate accuracy and affordable computational efficiency in modelling of a Generation IV reactor is to divide the primary system into 1D and 3D regions and apply coupled STH and CFD codes on the respective sub-domains.

Successful validation is a prerequisite for application of both, standalone and coupled STH and CFD codes in design and safety analysis of Generation IV systems. In this work we develop and apply different aspects of code validation methodology with an emphasis on (i) STH code analysis in support of validation experiment design (facility and test conditions), (ii) calibration of uncertain code input parameters and validation of standalone STH code, (iii) development of an approach to couple STH and CFD codes.

A considerable part of the thesis work is related to the development of a loop-type, 3 leg, liquid metal experimental facility TALL-3D for code validation. Particular focus was on identification of test conditions featuring complex feedbacks between 1D-3D phenomena, which can be challenging for the codes. Standalone STH code (RELAP5) was validated against experimental data. The domain of natural circulation instabilities in TALL-3D operation parameters was discovered using a validated STH code and global optimum search algorithms. Then existence of growing natural circulation oscillations was experimentally confirmed. An international benchmark was initiated in the framework of EU SESAME project based on the obtained experimental data.

Simulations were performed to define dimensions and location of a new test section for coolant solidification experiments that would also enhance possibilities for studying natural circulation instabilities in the future tests.

An approach to automated input calibration and code validation is developed in order to minimize possible “user effect” in case of multiple uncertain input parameters (UIPs) and system response quantities (SRQs). These methods were applied extensively in the development of RELAP5 input models and identification of the natural circulation instability regions.

Domain overlapping approach to coupling of RELAP5 and Star-CCM+ codes was proposed and resulted in considerable improvement of the predictive capabilities in comparison to standalone RELAP5.

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. 

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.

Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called "user effects". The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the SRQ of primary interest. The ranges of the input parameter were defined based on the experimental data and results of the calibration process. Then GA was used in order to identify combinations of the uncertain input parameters that provide maximum deviation of code prediction results from the experimental data. Such approach provides a conservative estimate of the possible discrepancy between the code result and the experimental data.

Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel-coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input can result in large changes in the final mass of debris. Typical ranges of the debris properties in different scenarios are discussed in detail.

The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

The system thermal-hydraulic (STH) code ATHLET was coupled with the commercial 3D computational fluid dynamics (CFD) software package ANSYS CFX to improve ATHLET simulation capabilities for flows with pronounced 3D phenomena such as flow mixing and thermal stratification. Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), validation activities for coupled thermal-hydraulic codes are being carried out. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. GRS carried out pre-test simulations with ATHLET-ANSYS CFX for the TALL-3D experiment T01, while KTH scientists perform these analyses with the coupled code RELAP5/STAR CCM+. In the experiment T01 the main circulation pump is stopped, which leads to interesting thermal-hydraulic transient with local 3D phenomena. In this paper, the TALL-3D behavior during T01 is analyzed and the results of the coupled pre-test calculations, performed by GRS (ATHLET-ANSYS CFX) and KTH (RELAP5/STAR CCM+) are directly compared.

APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with ID,'3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBt) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and lETs). The design is aimed at reducing experimental uncertainties and allowing fall separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D lest matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) arc defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments arc determined in order to minimize the code input uncertainties. The applied methodology and the results arc discussed in detail.

Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed configurations in the lower plenum.