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[en] Highlights: • New release AC2 2019 contains ATHLET/ATHLET-CD 3.2 and COCOSYS 3.0. • Validation of AC2 codes and validation matrices are discussed. • Improvements for ATHLET 3D model, AIDA lower plenum module and SAFT FP transport. • For MCCI simulation in COCOSYS, set concrete decomposition temperatures to 1800 K. • Stable coupling at multiple locations of ATHLET to CFD-codes for single-phase flows. - Abstract: The system code package AC2 by GRS for safety analyses of nuclear reactors from normal operation to severe accidents has been updated with a new release. We briefly describe the main modules of AC2 2019: ATHLET 3.2, ATHLET-CD 3.2 and COCOSYS 3.0 and selected improvements in these codes. We illustrate the improved capabilities of AC2 with selected examples. The post-test-calculation of the flooding pool of the INKA test facility demonstrated the improved ATHLET 3D model with explicit mixture level model. For ATHLET-CD, we apply the AIDA module with improvements in wall ablation and heat transfer models to a generic AP1000 lower plenum filled with molten corium. And we apply the new SAFT fission product transport module simulating three release paths through steam generator U-tubes. For COCOSYS, we present uncertainty analysis results for two MCCI experiments which justify using an elevated concrete decomposition temperature of 1800 K. Finally, we illustrate the potential of coupling ATHLET to CFD codes for safety analysis with results of a post-test calculation of the ROCOM PKLIIIT1.1 test on 3D mixing phenomena in an RPV.
[en] Highlights: • Simulation of hypothetical drop impact accident. • Still water situation and dynamic water situation. • Tensile tests of Inconel 690 material in low and high temperature cases. • Dynamic impact responses of secondary support structure. • Reliability evaluation of secondary support structure based on RCC-M Code. - Abstract: Dynamic responses of the secondary support structure are investigated to evaluate its structural reliability in a hypothetical drop impact accident of the reactor core barrel assembly in this work. Impact analyses both in still water situation (20 °C) and dynamic water situation (289 °C) are carried out by LS-DYNA. Tensile tests in both temperature cases are conducted for the secondary support structure material (Inconel 690) in order to make the simulation results more accurate with real material properties. Various aspects of structural dynamic responses during impact are analyzed, including distributions of equivalent stress, equivalent plastic strain, and maximum principal stress as well as time histories of impact force, vertical displacement, vertical velocity, and energy. In still water situation, equivalent stress of the secondary support structure exceeds the yield strength due to the higher initial impact velocity and downward base acceleration. But in dynamic water situation, there is no plastic deformation due to the lower initial impact velocity and upward base acceleration. Peak values of the maximum principal stress distributed in the secondary support structure are less than the corresponding ultimate tensile strengths respectively in both analysis cases. Results reveal that both of the secondary support structure and the reactor pressure vessel meet the design criteria of RCC-M Code with respect to stress limits, thus the structural reliability can be ensured in this drop accident.
[en] Highlights: • Assessment and calibration of different turbulent heat flux closures is performed. • Four different turbulent heat flux closures are assessed for low-Prandtl fluids. • This assessment has been performed for all three flow regimes. - Abstract: Thermal-hydraulics is recognized as a key safety challenge in the development of liquid metal cooled reactors. At nominal operating conditions, the Prandtl number of liquid metals which are used as primary coolants, such as lead and sodium, is very low: typically of the order of 0.025–0.001. Obtaining an accurate prediction of the turbulent heat transfer at such a low Prandtl number is not an easy task for the standard turbulence models and has challenged the modellers over several decades. In the framework of the EU SESAME project, an effort has been put forward to assess and/or further develop/calibrate different turbulent heat flux closures. In this regard, the present article reports an assessment of four different turbulent heat flux closures for applications involving low-Prandtl fluids. These closures include: (i) the Reynolds analogy based on a constant turbulent Prandtl number (ii) a four-equation explicit algebraic heat flux model (AHFM) (ii) a three-equation implicit AHFM called AHFM-NRG and (iv) a non-linear second-order heat flux model called Turbulence Model for Buoyant Flows (TMBF). The performance of these turbulence models has been assessed in three different test cases against high-fidelity numerical reference data been generated within the SESAME project. The three test cases are: a natural Rayleigh-Bénard convection flow, a mixed convection planar channel flow and a forced convection impinging jet flow. The shortcomings of the classical Reynolds analogy approach for low-Prandtl fluids in all flow regimes are highlighted; hence, more advanced and well-calibrated closures are recommended.
[en] Highlights: • The reference numerical database generated within the SESAME and MYRTE projects. • This includes various liquid metal flow configurations in different flow regimes. • In total, seven different flow configurations are considered. • This database will serve the further development of turbulent heat transfer model. - Abstract: Turbulent heat transfer is a complex phenomenon that has challenged turbulence modellers over various decades. In this regard, in the recent past, several attempts have been made for the assessment and further development/calibration of the available turbulent heat flux modelling approaches. One of the main hampering factors with respect to the further assessment of these modelling approaches is the lack of reference data. In the framework of the EU SESAME and MYRTE projects, an extensive effort has been put forward to generate a wide range of reference data, both experimental and numerical, to fill this gap. In that context, this article reports the numerical database that has been generated within these projects for various liquid metal flow configurations in different flow regimes. These high fidelity numerical data include seven different flow configurations: a wall-bounded mixed convection flow at low Prandtl number with varying Richardson number () values; a wall-bounded mixed and forced convection flow in a bare rod bundle configuration; a forced convection confined backward facing step (BFS) with conjugate heat transfer; a forced convection impinging jet for three different Prandtl fluids corresponding to two different Reynolds numbers of the fully developed planar turbulent jets; a mixed-convection cold-hot-cold triple jet configuration corresponding to ; an unconfined free shear layer for three different Prandtl fluids; and a forced convection infinite wire-wrapped fuel assembly. These high-fidelity numerical databases will serve the further development of turbulent heat transfer models by providing unique, new and detailed data for the thermal-hydraulic behaviour of liquid metals in various flow configurations.
[en] Highlights: • A simple steady-state model for stratified corium pool is built to estimate the IVR thermal load of SMR-IP200. • Three different configurations of molten pool (two-layer, three-layer, and water-layer) are considered. • The influences on peak heat flux, which are caused by mass components and top water, are discussed. • Some key values of calculation can provide references for the safety evaluation of SMR. - Abstract: In this paper, a simple model for stratified corium pool is built based on the concept of FIBS, to estimate the IVR thermal load of IP200 reactor (which is a small modular reactor with thermal power 220 MW). For the model validation, the IVR benchmark of AP600 reactor is calculated, and the results are compared with that of UCSB and INEEL. Then, the molten pool of IP200 with two-layer, three-layer and water-layer configurations are calculated respectively. The thermal load distributions with the effects of internal power and metal mass are analyzed. Besides, the changes of peak heat flux, which are caused by the heavy metal’s appearance and the water layer’s boiling, are discussed by the comparison of local CHF. The specific results show that, in two-layer configuration, the peak heat flux is within 279.9–667.8 kW/m2, and the thermal margin is within 0.31–0.74. In three-layer configuration, the bottom heat flux of molten pool will be enlarged. The internal heat of heavy metal can be the dominant factor for the thermal load fluctuation, when the mass composition is changed. In water-layer configuration, considering the heat transfer of film boiling can keep in a same magnitude as that of radiation, the heat dissipation of pool’s upper face will be enhanced to mitigate the peak heat flux. The present work gives out some key values of steady-state IVR under different configurations, which can provide some references for the safety evaluation of SMR.
[en] Highlights: • CFD analysis of containment gas flow inside 1F1 Primary Containment Vessel. • Wall condensation model controlled by the species concentration on boundary layer. • Implementation of wall condensation model to STAR-CCM+. • Estimation of possible melted fuel debris locations for 1F1. - Abstract: Even after seven years since the Fukushima accident, conditions inside Fukushima Daiichi Unit-1 (1F1) remain mostly unknown, making decommissioning activities challenging. Furthermore, practical data from within the primary containment vessel (PCV) of 1F1, which may provide crucial information for decommissioning activities, have not been assessed sufficiently. In this study, an alternative approach for evaluating such practical data is investigated. In this approach, the condition of debris inside the PCV is estimated through computational fluid dynamics (CFD) simulations performed using STAR-CCM+. Heat transfer inside the PCV is modeled with the developed wall condensation. The wall condensation model is firstly validated through experiment results from COPAIN tests and applied to the PCV of 1F1. Possible locations of debris and decay power that dissipated and heated the environment inside the PCV are then estimated using the simulation results. Moreover, temperature asymmetry observed from practical data on safety valves is also addressed by conducting parameter analysis for different decay powers. The results suggest that 5% of decay heat (65 kW) dissipated to heat gas inside the PCV, and temperature asymmetry on safety valves could be explained by 2% of decay heat (20 kW) as the heat source. A significant amount of debris is estimated to exist inside the pedestal and safety relief valve locations, and some may have spread to the drywell floor through the doorway opening of the pedestal. The findings of the study contribute towards improved decommissioning of 1F1.
[en] Highlights: • A finite element model is developed to simulate the mechanical behaviors of a spacer grid. • Mechanical characteristics of a spacer grid cell are implemented in the developed model. • The gap size of a spacer grid is realized by adjusting the material properties of spring and dimple. • The developed model is verified by comparing the static and dynamic tests and analysis results. - Abstract: The purpose of this study is to develop a finite element model that accurately describes the mechanical behavior of a spacer grid. The spacer grid is the most important component of a nuclear fuel assembly and supports the fuel rod with a structurally sufficient buckling strength. Therefore, the development of a reliable spacer grid model is essential to evaluate the mechanical integrity of a nuclear fuel assembly. To achieve this objective, a three-dimensional finite element model that can simulate the static and dynamic mechanical behaviors of a spacer grid was developed in this study. To simulate the exact mechanical properties of the spacer grid cell, the parameter values required for the model were determined by conducting a fuel rod drag test and spacer grid spring/dimple stiffness test. Finally, a spacer grid static compression test and dynamic impact test were performed according to the gap size of the spacer grid cell, and the model was verified by comparing the test and analysis results. The results obtained using the developed spacer grid finite element model agreed well with the mechanical test results, and it was confirmed that both the static and dynamic mechanical characteristics of the model were accurately simulated by the proposed model.
[en] Highlights: • Nuclear power plant stress tests for emergency service water system failure. • Two-dimensional simulation of the flooding process of the services galleries. • Enhancing Iber modelling tool for the simulation of flows in occupied galleries. • Modelling of mixt open-channel and pressurized flow, punctual discharges and flow through culverts. • Significant influence of occupation on the speed of the flood propagation. - Abstract: In the context of the stress tests that have to be applied to nuclear power plants, this work presents the study of the flooding processes of the service galleries of a nuclear plant caused by a hypothetical failure of some of the pipes of the Essential Services Water System (ESWS). To assess the flood propagation along the galleries, two-dimensional hydraulic modelling tools, based on the solution of the shallow water equations with the finite volume method, were used. Due to the complexity and special features of the geometry and hydraulic processes, when compared with more standard urban flood assessment works, several specific modules were developed. A relevant one is a new module to consider the effect of the occupation of the galleries on the advance of the waterfront. This module was developed and verified prior to being applied to a case study. The results show the suitability of the proposed methodology to be used as part of the stress tests to ensure high security standards of nuclear power plants.
[en] Highlights: • Current state-of-the-art experimental measurements in complex geometries of advanced nuclear reactors. • Turbulent flows in wirewrapped fuel bundle for liquid metal reactors. • Randomly packed beds for gas-cooled and molten-salt reactors. • High-fidelity experimental measurements for CFD code validations. • Time-resolved particle image velocimetry (TR-PIV) and matching-index-of-refraction approaches. - Abstract: Modeling and simulation performed with advanced tools are important for thorough understanding of existing power plant response to accidents; for life extension decisions of existing plants; and to support licensing activities for new power plants. The use of computational fluid dynamics (CFD) tool in nuclear R&D has gained significantly due to its capabilities to predict complex flow phenomena at very fine resolutions. High-fidelity numerical simulations including direct numerical simulation (DNS) and large-eddy simulation (LES) have been considered as reliable CFD tools for the development and validation of turbulence models along with experiments. Compared to other CFD techniques, DNS is the most computationally expensive approach, and limited to flow studies at low to moderate Reynolds numbers. LES subgrid-scale (SGS) models require the specification of model coefficients that cannot be generally used to simulate a wide spectrum of flows. Performances of LES with modified SGS model coefficients need to be verified and validated versus high-resolution experimental database or DNS results. In this paper, we present the current state-of-the-art experimental measurements in complex geometries of advanced nuclear reactors, such as turbulent flows in wirewrapped fuel bundle for liquid metal reactors and randomly packed beds for gas-cooled and molten-salt reactors. It is important to achieve an in-depth understanding of flow phenomena and complex flow characteristics within these reactor cores because they are related to the safety and design scenarios. High-fidelity experimental measurements of velocity fields are acquired featuring a combination of time-resolved particle image velocimetry (TR-PIV) and matching-index-of-refraction approaches. Experimental results are obtained at high spatial and temporal resolutions of velocity fields and the first- and second-order flow statistics are suitable for the verification and validation of CFD codes currently used in nuclear engineering applications.
[en] Highlights: • Approximately 11,540 µCi/m3 of tritium generated from molten FLiBe salt. • Measured tritium permeation through bare and alumina-coated SS316 tubes. • Tritium permeability SS316 was significantly reduced by alumina coating. • Alumina coating via plasma thermal spray has excellent thermal stability. - Abstract: This study experimentally investigates the reduction efficiency of tritium permeation through 316 stainless steel tubing coated with alumina as a tritium permeation barrier (TPB) in support of the development of molten salt nuclear reactors, particularly for fluoride salt-cooled high-temperature nuclear reactors (FHRs). The TPB coatings composed ofan intermediate bond layer of NiCr, a transition layer of NiCr + alumina, and a pure alumina layer were successively added onto the outer surface of commercial 316 stainless steel tubing via plasma thermal spray. In order to generate a continuous gaseous tritium source, 35 g of purified natural-lithium FLiBe salt was irradiated by thermal neutron flux at 620 °C in the Massachusetts Institute of Technology Research Reactor (MITR). The preliminary results suggest that the TPB coatings on tube surfaces significantly reduced the tritium permeation rate at 700 °C. To get a better understanding of the TPB, the microstructure of the coated tubes was characterized with various techniques.