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1.	Li, X., et al. (author) Flow Pattern Identification of Porous Media Based on Signal Feature Extraction and SVM 2022 In: Kung Cheng Je Wu Li Hsueh Pao/Journal of Engineering Thermophysics. - : Science Press. - 0253-231X. ; 43:11, s. 2957-2965 Journal article (peer-reviewed)abstract In this paper, the visualization experiment of gas-liquid two-phase flow in porous media is carried out. The typical flow patterns of bubbly flow, slug flow and annular flow are photographed by high-speed camera, and the corresponding differential pressure fluctuation signals are measured and recorded, Using probability density function (PDF) and power spectral density (PSD) curves, the time-domain and frequency-domain characteristics of differential pressure signals corresponding to each flow pattern are analyzed, and the quantitative characteristic parameters are introduced to construct the characteristic vector reflecting the time-frequency characteristics of differential pressure signals. A two-phase flow pattern identification method in porous media based on support vector machine (SVM) is proposed. The results show that the overall recognition rate of the three flow patterns measured by the method is 98.18%, which can provide a new technical support for the on-line recognition of gas-liquid two-phase flow patterns in porous media.
2.	Li, Xiangyu, et al. (author) Identification of two-phase flow pattern in porous media based on signal feature extraction 2022 In: Flow Measurement and Instrumentation. - : Elsevier BV. - 0955-5986 .- 1873-6998. ; 83 Journal article (peer-reviewed)abstract The statistical analysis methods based on differential pressure signals of two-phase flow are employed in the present study to identify the flow patterns in packed porous bed. The typical flow pattern images of two-phase flow in the packed porous beds are recognized and the corresponding differential pressure signals are recorded based on the visualization experiments. Then the statistical analysis methods, including probability density function (PDF), power spectral density (PSD), and wavelet energy spectrum (WES), are employed to extract the features of differential pressure signals in the time domain, frequency domain, and time-frequency domain respectively. The dimensionless parameters are proposed as the evaluation index to quantify the differences among flow patterns. The results show that the PDF, PSD, and WES methods can effectively characterize different flow patterns in the time, frequency, and time-frequency domain, respectively. The comprehensive recognition efficiency is about 88.5% using the introduced dimensionless parameters.
3.	Liu, Jiebin, et al. (author) Influence of an External Perpendicular Oscillation on Stability of a Vertical Falling Liquid Film 2020 In: Microgravity, science and technology. - : Springer Science+Business Media B.V.. - 0938-0108 .- 1875-0494. ; 32:5, s. 787-805 Journal article (peer-reviewed)abstract A vertical falling Newtonian liquid film flow is inherently unstable to surficial long-wave disturbances. Imposing external oscillation can stabilize the long-wave instability, but also triggers additional parametric instabilities. The effect of oscillation frequency on the stability is subtle. By using the “viscosity-gravity” scaling, the effect of oscillation frequency on the stability can be investigated exhaustively by separating it from other control parameters. In this paper, the effects of external perpendicular oscillation on the stability of a vertical falling liquid film are then investigated by a combination of linear stability analyses based on Floquet theory and numerical simulations with an unsteady weighted residual model (WRM). The linear analyses show that, increasing oscillation amplitude always has a stabilizing effect on the long-wave instability. On the other hand, increasing or decreasing oscillation frequency can suppress the long-wave instability, depending on whether the oscillation amplitude or the acceleration is fixed. The effect of varying oscillation frequency on the long-wave instability is opposite to that on the parametric instabilities. The long-wave and parametric instabilities compete with each other as the oscillation amplitude and frequency are varied with the Reynolds number fixed. A weakness of the long-wave instability always accompanies enhancements of the parametric instabilities, and vice versa. As a contrast, an increase of Reynolds number always results in more unstable long-wave and parametric instabilities. The numerical simulations with the WRM show that the wave amplitudes and the minimal local thickness of film are proportional to the unstable wavenumbers range rather than the growth rate of the instability. For a given oscillation frequency and Reynolds number, there exist a critical oscillation amplitude above which externally imposed oscillations perpendicular to the transversal direction of the film can also trigger a chaotic behavior in the film, just like what happens in the case where the oscillation is parallel to the stream-wise direction of the film.
4.	Liu, J., et al. (author) Linear stability of a fluid mud–water interface under surface linear long travelling wave based on the Floquet theory 2021 In: European journal of mechanics. B, Fluids. - : Elsevier BV. - 0997-7546 .- 1873-7390. ; 86, s. 37-48 Journal article (peer-reviewed)abstract The Floquet theory is combined with the unsteady Orr–Sommerfeld equations for the first time to model the linear stability of a fluid mud–water interface under the influence of a linear long travelling wave (or linear shallow water wave). The modelling results reveal three instability modes that could appear on the fluid mud–water surface: the Kelvin–Helmholtz (K–H) and finite-wavelength (F-W) instabilities, which are also present in steady two-layer systems, and parametric instability, which is only seen in periodic problems. The growth rate of the parametric instability is generally small, but it affects the growth rate curves of the other two instabilities. The K–H and F-W instabilities are found to be dominant, and each plays an important role in determining the evolution of the fluid mud–water interface. Both the K–H and F-W instabilities grow with increasing water depth and decreasing wave period as well as with decreasing thickness and density of the mud layer. However, they exhibit distinct dependencies on the fluid mud-to-water viscosity ratio and compete near the critical conditions. For unstable flow near the critical conditions, the K–H instability dominates over the F-W instability at a low viscosity ratio and vice versa at a high viscosity ratio, while for unstable flow far beyond the critical conditions, the K–H instability is dominant regardless of the viscosity ratio. These results are practically instructive for waterway and harbour construction and protection since they provide valuable insights into the early dynamics of the instability mechanisms of the fluid mud–water interface.
5.	Lu, Junjing, et al. (author) An improved sectional model to simulate multi-component aerosol dynamics in a containment of pressurized water reactor 2021 In: Journal of Aerosol Science. - : Elsevier BV. - 0021-8502 .- 1879-1964. ; 157 Journal article (peer-reviewed)abstract Simulating an evolving aerosol population in a reactor containment is essential for estimating the radioactivity that is possible to leak to the environment. In this study, a sectional model is developed to simulate multi-component aerosol dynamics in the containment during severe ac-cidents of a pressurized water reactor by improving the widely used MAEROS (Multicomponent AEROSol) model. An important advantage of the improved model is its simplified calculation method by introducing a series of correction factors to the equation coefficients when the thermal boundary conditions and the aerosol particle density in the containment change continuously. In addition, the restriction of the maximum section number in the MAEROS model is removed. The reliability of the model is validated against four analytical solutions and three sets of test data. Moreover, the improvements in the model are also proven to be necessary to effectively capture the influences of thermal boundary conditions and aerosol particle density on aerosol dynamics.
6.	Yu, Peng, et al. (author) A numerical study of heat transfer in bottom-heated and side/top-cooled liquid metal layers with different aspect ratios 2022 In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 177, s. 109328- Journal article (peer-reviewed)abstract A liquid metal layer heated from bottom and cooled from both side and top can be encountered in indus-trial applications. A special interest is from safety design of advanced pressurized water reactors that adopt the so-called in-vessel melt retention (IVR) to mitigate severe accident risk. Quantification of heat transfer in a stratified melt pool in the lower head of a reactor pressure vessel (RPV) is of great impor-tance to the qualification of the IVR strategy. The upper liquid metal layer of the stratified melt pool is heated by the lower molten oxide layer (with decay heat) underneath, and cooled by water outside the reactor vessel and by radiation or flooded water at the top. This is essentially a problem of natural convection and heat transfer in a liquid metal layer heated from bottom and cooled from both side and top. The present study is conducted to numerically investigate the heat transfer characteristics of such layer with an emphasis on the influence of the aspect ratio (ratio of radius to height; R/H) of the liq-uid metal layer. Based on the numerical outcomes, three correlations of heat transfer coefficients (for downward, upward and sideward flows) are also developed to account for the impact the aspect ratio. The numerical simulation results show that, under the same Rayleigh number, the bulk temperature and the upward and sideward heat fluxes all increase with R/H, but the downward heat flux decreases with R/H. The Nusselt numbers in all directions decrease with increasing R/H, as a reduced cooling -heating area ratio due to increasing R/H shall suppress the cooling efficiency and the convection. When R/H is larger than a threshold (-8), the heat transfer characteristics are no longer sensitive to R/H. Each correlation of heat transfer coefficient is developed as the product of two terms: a base correlation of heat transfer coefficient that is Ra dependent only, and an aspect ratio factor that considers the effect of aspect ratio R/H. The developed correlations are compared with the numerical simulation results of cases with different aspect ratios and Rayleigh numbers, and good agreements achieved.
7.	Yu, Peng, et al. (author) An assessment of the lumped parameter model for the two-layer melt pool heat transfer 2023 In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 180 Journal article (peer-reviewed)abstract Lumped parameter codes for the two-layer melt pool heat transfer in IVR analysis are usually only verified against the UCSB case and lack of validation. We assessed the performance of the lumped parameter model against the recently conducted two-layer LIVE2D experiment. Both test series with/without top water cooling and with different pool heights were simulated. Influences of the heat transfer correlations in both layers were also investigated. Results showed that the lumped parameter model with existing correlations tends to over -predict sideward heat flux, regardless of the selection of heat transfer correlations in the bottom and top layers. The deficiency could be related to the simplified treatment by modelling the top layer heat transfer as two independent mechanisms: correlations obtained independently from Rayleigh-Be ' nard convection and sideward cooled convection are directly applied to calculate the corresponding upward/downward and sideward heat transfer coefficients, respectively. Evidences from calculations of top layer experiments also support this observation. To solve this issue, additional consideration of the interaction between the two convection mech-anisms may be needed and then perhaps proper model corrections be introduced, or new correlations be developed for this specific convection.
8.	Zhang, Zhengzheng, et al. (author) Experimental and numerical studies on the two-dimensional flow characteristics in the radially stratified porous bed 2022 In: International Communications in Heat and Mass Transfer. - : Elsevier BV. - 0735-1933 .- 1879-0178. ; 133, s. 105940- Journal article (peer-reviewed)abstract The experiment and numerical simulations are both conducted in the present study to better understand the twodimensional flow characteristics in the radially stratified porous bed. Spherical particles of two different sizes are packed in the left part and right part of a cylindrical test section separately to form a two-layer bed with the configuration of radial stratification. The variations of pressure drops in each part of the stratified bed are measured when water flows up through the packed bed. Meanwhile, the numerical simulation is also carried out to investigate the flow field in the stratified bed, especially the flow characteristics around the interface of two parts. The results indicated that the pressure drops in the two layers of radial stratified bed are almost equal. When the fluids flow up through the radially stratified porous layers, the lateral flowing from the low permeability layer to the high permeability layer leads to a decrease of pressure drop in the low permeability layer and an increase of pressure drop in the high permeability layer. Most of the lateral flow occurs in the initial part of the test section. Besides, the lateral and vertical pressure gradient can be well predicted by Ergun equation respectively.
9.	Zhang, Zhengzheng, et al. (author) Investigation on Flow Characteristics in Radial Stratified Debris Bed 2022 In: Yuanzineng Kexue Jishu/Atomic Energy Science and Technology. - : Atomic Energy Press. - 1000-6931. ; 56:10, s. 2032-2040 Journal article (peer-reviewed)abstract During the severe accident of light water reactors (LWRs), the particulate debris bed with porous structure may be formed at different places in the reactor after molten corium fuel coolant interaction (FCI). The coolability of the debris bed therefore plays an important role in corium risk quantification, which is crucial to the stabilization and termination of a severe accident in LWRs. Many experimental and analytical studies have been conducted towards quantitative understanding of debris bed coolability. However, most of previous studies were conducted based on the homogeneous debris beds packed with single size particles, and only a few investigations were performed with the heterogeneous debris beds like stratified debris bed. In fact, scoping studies on debris bed formation and configuration based on FCI experiments indicate that the stratified debris bed would be most expected. In order to study the flow characteristics in heterogeneous debris beds, the packed porous beds with radial stratification were constructed in the present study using two different sizes glass spheres with the diameter of 2 mm and 8 mm respectively. Besides, the homogeneous packed beds packed with single size particles and uniform mixture by the above two size particles were also constructed for comparison. The particles were packed in a cylindrical test section with the inner diameter of 120 mm and the height of 600 mm. Single-phase flow tests were performed on the homogeneous beds and heterogeneous bed firstly to investigate the flow resistance characteristics in the packed beds with different configurations. Then numerical simulation was also conducted to reveal the flow redistribution of stratified bed, especially on the flow field at the stratified interface. The experimental results show that the pressure drops of single-phase flow in the homogeneous beds can be well predicted by Ergun equation. For the radial stratified packed bed with different permeability layers, the pressure drops in each layer of the stratified bed are almost equal and increase with the liquid inlet flowrate. Comparing with those in the homogenous beds packed with the same size particles as those in different layer of stratified bed, the pressure drops in the stratified bed are much lower than those of homogenous bed packed with smaller size particles, while slightly higher than those with larger size particles. The numerical simulation results state that there is a two-dimensional flow phenomenon in the radial stratified bed. In addition to dominate upward flow in the stratified bed, a lateral flow flows from low permeability layer to high permeability layer. The two-dimensional flow in stratified bed decreases the flowrate and pressure drops in low permeability layer and increases the pressure drops in high permeability layer. With the increase of liquid flowrate, the average lateral flowrate at the stratified interface increases, but the ratio of lateral volume flowrate to total fluid volume flowrate decreases.
10.	Bandini, G., et al. (author) Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors 2015 In: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 281, s. 22-38 Journal article (peer-reviewed)abstract 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.
11.	Bechta, Sevostian, et al. (author) On the EU-Japan roadmap for experimental research on corium behavior 2019 In: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 124, s. 541-547 Journal article (peer-reviewed)abstract A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.
12.	Bian, Boshen (author) CFD Study of Molten Pool Convection in a Reactor Vessel during a Severe Accident 2023 Doctoral thesis (other academic/artistic)abstract During severe accidents in nuclear reactors, the core and internal structures can melt down and relocate into the reactor pressure vessel (RPV) lower head (LH) forming there a stratified molten corium pool. The pool generally consists of superheated oxidic and metallic liquid layers imposing thermo-mechanical loads on the RPV wall. The in-vessel retention (IVR) strategy employs external cooling with water to maintain RPV integrity. Investigating the thermo-fluid behaviour of corium and predicting heat flux distribution on the vessel wall are crucial. The molten pool exhibits natural convection, which can typically consist of two stratified layers. There exists internally heated (IH) natural convection in the oxidic layer and Rayleigh-Bénard (RB) convection in the surface metallic layer.This study starts by illustrating the mathematical models that involve the numerical study of natural convection flow in molten corium. A verification work of the model has been done using a previous direct numerical simulation (DNS) study, and the results show good agreement. In addition, a scaling theory of the natural convection flow is demonstrated to facilitate the pre-estimation based on the Rayleigh number (Ra) and Prandtl number (Pr). After that, the numerical approaches involved in the numerical simulation of the corium are illustrated, especially focusing on the DNS method. A DNS mesh strategy is proposed in the form of a pipeline from the pre-estimation to the post-check. A scalability study of Nek5000 is performed on four different HPC clusters based on a DNS case of the IH molten convection in a hemispherical geometry with Ra=1.6×1011. The results show a super-liner speedup property of Nek5000 on each cluster within a certain range.Then, three numerical studies focusing on turbulent natural convection flow within both the oxidic and metallic layers of corium are demonstrated and discussed. Through these simulations, the thermos-fluid behaviour of the system is examined in detail, including flow configuration, temperature distribution, heat flux profiles on cooling boundaries, and turbulent quantities.1. A DNS investigation is performed on the IH molten pool convection within a hemispherical domain, employing a Rayleigh number of 1.6×1011 and a Prandtl number of 0.5. The results show a turbulent flow characterized by three distinct regions, consistent with the observation from the BALI experiments. Detailed information regarding turbulence, including turbulent kinetic energy (TKE), turbulent heat flux (THF), and temperature variance, is presented. Furthermore, the study offers comprehensive 3D heat flux distributions along the boundaries, showing heat flux fluctuations along the top boundary due to nearby turbulent eddies and a nonlinear increase in heat flux along the curved boundary from bottom to top.2. A numerical study investigates the effect of Prandtl number on the natural convection of an IH molten pool in a 3D semi-circular test section. Prandtl numbers of 3.11, 1.0, and 0.5 are considered, with a Ra= 6.54×1011. Smaller Prandtl numbers result in more vigorous turbulent motion and a thicker layer of intense turbulent mixing in the upper region. The descending flow extends further down the bottom, creating a stronger circulation at the bottom with smaller Pr. Additionally, smaller Pr leads to more thermal stripping structures and less stable stratification layers. Comparing heat fluxes on the top and curved walls reveals higher fluctuation frequency with smaller Pr for heat fluxes to the top boundary. However, the maximum heat fluxes to the side walls are lower with smaller Pr.3. A numerical study investigates the turbulent natural convection in a 3D fluid layer based on the BALI-Metal 8U experiment. Different methods, including DNS and three Reynolds-averaged Navier-Stokes (RANS) models, are employed. The results are compared with experimental data, and the performance of the RANS models is evaluated using DNS as a reference. DNS reproduces a two-distinct region flow structure observed in experiments, while the k-ω SST model exhibits similar flow patterns and TKE profiles. However, all simulations overpredict temperature compared to experimental data, with DNS providing the closest results. The DNS results also achieve better agreement with experimental data in terms of heat flux distribution and energy balance, specifically capturing the transient maximum heat flux on the lateral cooling wall. This transient behaviour plays a crucial role in accurately estimating the ‘focusing effect’.
13.	Bian, Qingzhen, 1988-, et al. (author) Vibronic coherence contributes to photocurrent generation in organic semiconductor heterojunction diodes 2020 In: Nature Communications. - : NATURE PUBLISHING GROUP. - 2041-1723. ; 11:1 Journal article (peer-reviewed)abstract Charge separation dynamics after the absorption of a photon is a fundamental process relevant both for photosynthetic reaction centers and artificial solar conversion devices. It has been proposed that quantum coherence plays a role in the formation of charge carriers in organic photovoltaics, but experimental proofs have been lacking. Here we report experimental evidence of coherence in the charge separation process in organic donor/acceptor heterojunctions, in the form of low frequency oscillatory signature in the kinetics of the transient absorption and nonlinear two-dimensional photocurrent spectroscopy. The coherence plays a decisive role in the initial 200 femtoseconds as we observe distinct experimental signatures of coherent photocurrent generation. This coherent process breaks the energy barrier limitation for charge formation, thus competing with excitation energy transfer. The physics may inspire the design of new photovoltaic materials with high device performance, which explore the quantum effects in the next-generation optoelectronic applications.
14.	Buck, M., et al. (author) The LIVE program : Results of test L1 and joint analyses on transient molten pool thermal hydraulics 2010 In: Progress in nuclear energy (New series). - : Elsevier BV. - 0149-1970 .- 1878-4224. ; 52:1, s. 46-60 Journal article (peer-reviewed)abstract The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e.g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO3-NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results of the LIVE activities also provide data for a better understanding of in-core corium pool behaviour. The experimental results are being used for the development and validation of mechanistic models for the description of molten pool behaviour, In the present paper, a range of different models is used for post-test calculations and comparative analyses. This includes simplified, but fast running models implemented in the severe accident codes ASTEC and ATHLET-CD. Further, a computational tool developed at KTH (PECM model implemented in Fluent) is applied. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE (Nuclear Safety Institute of Russian Academy) within the RASPLAV project and was further improved within the ISTC 2936 Project.
15.	Buerger, M., et al. (author) Coolability of particulate beds in severe accidents : Status and remaining uncertainties 2010 In: Progress in nuclear energy (New series). - : Elsevier BV. - 0149-1970 .- 1878-4224. ; 52:1, s. 61-75 Journal article (peer-reviewed)abstract Particulate debris beds may form during different stages of a severe core melt accident; e.g. in the degrading hot core, due to thermal stresses during reflooding, in the lower plenum, by melt flow from the core into water in the lower head, and in the cavity by melt flow out of a failing RPV into a wet cavity. Deep water pools in the cavity are used in Nordic BWRs as an accident management measure aiming at particulate debris formation and coolability. It has been elaborated in the joint work of the European Severe Accident Research Network (SARNET) in Work Package (WP) 11.1 that coolability of particulate debris, reflooding of hot debris as well as boil-off under decay heat (long-term coolability), is strongly favoured by 2D/3D effects in beds with non-homogeneous structure and shape. Especially, water inflow from the sides and via bottom regions strongly improves coolability as compared to 1D situations with top flooding, the latter being in the past the basis of analyses on coolability. Data from experiments included in the SARNET network (DEBRIS at IKE and STYX at VTT) and earlier ones (e.g. POMECO at KTH) have been used to validate key constitutive laws in 2D codes as WABE (IKE) and ICARE/CATHARE (IRSN), especially concerning flow friction and heat transfer. Major questions concern the need of the explicit use of interfacial friction to adequately treat the various flow situations in a unified approach, as well as the adequate characterization of realistic debris composed of irregularly shaped particles of different sizes. joint work has been supported by transfer of the WABE code to KTH and VTT. Concerning realistic debris, the formation from breakup of melt jets in water is investigated in the DEFOR experiments at KTH. Present results indicate that porosities in the debris might be much higher than previously assumed, which would strongly support attainment of coolability. Calculations have been performed with IKEJET/IKEMIX describing jet breakup, mixing and settling of resulting particles. Models about debris bed formation and porosity are developed at KTH. The codes have been applied to reactor conditions for analysing the potential for coolability in the different phases of a severe accident. Calculations have been performed with WABE (MEWA) implemented in ATHLET-CD and with ICARE/ICATHARE for degraded cores and debris beds in the lower plenum, under reflooding and boil-off. Ex-vessel situations have also been analysed. Strong effects of lateral water inflow and cooling by steam in hot areas have been demonstrated. In support, some typical basic configurations have been analysed, e.g. configurations with downcomers considered as possible AM measures. Melt pool formation or coolability of particulate debris is a major issue concerning melt retention in the core and the lower head. Present conclusions from those analyses for adequate modelling in ASTEC are outlined as well as remaining uncertainties. Experimental and analysis efforts and respective continued joint actions are discussed, which are needed to reach resolution of the coolability issue.
16.	Chen, Yangli, et al. (author) A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor 2019 In: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 343, s. 22-37 Journal article (peer-reviewed)abstract This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.
17.	Chen, Yangli, et al. (author) Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds 2022 In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 165 Journal article (peer-reviewed)abstract The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-termcoolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bedformed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of theMELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debrisbed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integralanalysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debrisbed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on thebed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns.Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity poolsaturation and containment venting.
18.	Chen, Yangli, et al. (author) Development and application of a surrogate model for quick estimation of ex-vessel debris bed coolability 2020 In: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 370 Journal article (peer-reviewed)abstract During a hypothetical severe accident of a Nordic boiling water reactor (BWR), an ex-vessel particulate debris bed is expected to form in the flooded lower drywell due to melt-coolant interactions after vessel failure. The key parameter to evaluate debris bed coolability is the dryout heat flux (DHF) or dryout power density, representing the limit of heat removal capacity by the coolant. Several numerical codes such as COCOMO have been developed to simulate thermal hydraulics in multi-dimensional debris beds and predict the cooling limit, but they are computationally expensive and not suitable for probabilistic risk analysis. This paper aims to develop a surrogate model which can serve as a quick-estimate tool for the dryout power density of a heap-like debris bed in a saturated water pool. The dryout power density predicted from the COCOMO code is treated as the full model. A characteristic factor is introduced as the dryout power density ratio between the multi-dimensional debris bed (predicted by COCOMO code) and the corresponding one-dimensional debris bed (predicted by Lipinski 0-D model). The characteristic factor is correlated by the Kriging method with six parameters: bed porosity, particle diameter, debris mass, bed slope, cavity radius and containment pressure. After the surrogate model is trained and validated, it is employed to analyze the coolability of prototypical debris beds of a reference Nordic BWR, given the bed mass and containment pressure from MELCOR simulation. Coolability maps are produced as quick look-up diagrams for identification of coolable domain with the variation of porosity, particle diameter and slope angle. A preliminary uncertainty analysis is performed to demonstrate the effect of uncertain input parameters on non-coolable domain.
19.	Chen, Yangli, et al. (author) Development of surrogate model for debris bed coolability analysis 2019 In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019. - : American Nuclear Society. ; , s. 6770-6779 Conference paper (peer-reviewed)abstract The cornerstone of severe accident management (SAM) strategy of a Nordic boiling water reactor (BWR) is to flood the reactor cavity with water from the pressure suppression pool before failure of the reactor pressure vessel (RPV). The idea is to form a deep water pool which can accommodate the corium ejected from the RPV breach and cool the debris bed in the reactor cavity. Hence, assessment of debris bed coolant in the deep water pool is of paramount importance to the qualification of this SAM strategy. For the coolability analysis of a debris bed, one needs to estimate the dryout heat flux/power density of the particle bed, which is considered as the limit for heat removal capacity of coolant. For a multi-dimensional debris bed, the dryout power density can be assessed only by numerical simulation of two-phase flow and heat transfer in porous media. Since the numerical simulation is computationally expensive, it is neither suitable for massive calculations, nor feasible to be implemented into a system code (e.g. MELCOR). There is a clear need to develop a fast-running tool to estimate the dryout power density of a prototypical debris bed. The present study is concerned with development of a surrogate model which is sufficient for PSA study or capable of coupling with the MELCOR code without significant sacrifice of computational efficiency. The surrogate model is conceived from the coolability database predicted by COCOMO which is a mechanistic code for simulating thermal-hydraulic response of debris bed and has been extensively validated and applied in our previous studies [1][2]. The comparative results show that the surrogate model is not only able to predict the coolability limit of a debris bed, but also employed in the sensitivity study of bed’s characteristics (e.g., particle diameter, bed geometry and porosity) and the uncertainty and risk analysis.
20.	Chen, Yangli (author) MELCOR Capability Development for Simulation of Debris Bed Coolability 2021 Doctoral thesis (other academic/artistic)abstract The severe accident management (SAM) strategy for a Nordic boiling water reactor (BWR) employs cavity flooding prior to vessel failure, so that the core melt (corium) discharged from the vessel could fragment and form a particulate debris bed. The key to the success of this SAM strategy is the coolability of ex-vessel debris beds.The safety analysis involves knowledge about the reactor response to severe accidents under this SAM strategy, which requires the integral simulation of a system code such as MELCOR. Since currently the MELCOR code lacks the modeling of ex-vessel particulate debris beds, the present study aims to develop the capability of MELCOR for the simulation of debris bed coolability through the coupling of MELCOR with other codes, which are dedicated to this phenomenon.The study is started from the qualification of a MELCOR model for severe accident analysis of a reference Nordic BWR, with the aim to help identify a proper core nodalization. For this purpose, three different core meshes (coarse, medium, and fine) are employed to obtain their impacts on corium release conditions. It is found the coarse mesh is sufficient in the present study, since it is not only computationally efficient, but also predicting earlier vessel failure and faster corium release, providing a more conservative condition for debris bed coolability analysis.Two couplings are then adopted: (i) coupling of MELCOR with the COCOMO code, which is a mechanistic code for simulation of thermal hydraulics in debris beds; and (ii) coupling of MELCOR with a surrogate model developed in the present study. The first method can simulate the transient behavior of a debris bed during quench process. The second method can efficiently predict the coolability limit (dryout power) required in safety analysis. The surrogate model is developed based on the COCOMO prediction of two-dimensional debris beds.The developed simulation tools, including the coupled codes and the surrogate model, are applied to the safety analysis of the reference Nordic BWR. The coupled MELCOR/COCOMO simulation is used to investigate the debris bed properties. The effective particle diameter is found as approximately 10% larger than the surface mean diameter of a debris bed with distributed sizes, quantified by the quench rate. For the effect of debris bed shape, it shows a faster quench process with a lower bed slope angle. The quench front propagation as well as the responses of local temperature and containment pressure are obtained.The coupled MELCOR/surrogate model simulation is performed to estimate the coolability of ex-vessel vessel debris beds. The results show that debris beds are coolable under prototypical conditions with probable bed properties. The surrogate model is used to generate coolability maps, which show the debris bed coolability with the variation of bed properties. The sensitivity analysis indicates that the porosity and the geometry are the most influential to coolability limit. An uncertainty analysis methodology is proposed to obtain the probability of non-coolable debris beds.
21.	Chen, Yaodong, et al. (author) Numerical investigation of Fukushima Daiichi-2 SBO scenario 2014 In: International Congress on Advances in Nuclear Power Plants, ICAPP 2014. - 9781632668264 ; , s. 995-1003 Conference paper (peer-reviewed)abstract Simulations of the severe accident progression for Fukushima Daiichi NPP Unit 2 (1F2) are performed using the MELCOR code. Detailed modeling of the plant is developed to represent the whole reactor system and its safety systems. The predicted results are compared with the plant data measured during the accident. By applying the main actions taken during the accident and the assumptions into the full plant MELCOR modeling, the major physical phenomena from core uncovery and degradation till reflooding of reactor core by fire pump injection are reproduced in the simulations. The trend of simulation results agree in general with the limited data (e.g., pressures) measured by the plant. The closed RCIC cycle, which involved steam flow and working process, and its interacting with reactor cooling status was modeled by user defined control function in the simulation. The simulations reveal that: The operations of RCIC kept the reactor core flooded to the top for more than 70 hours after the earthquake until the suppression pool water got saturated. Sea water might have flooded into the TORUS room to more extent than as assumed, which kept cooling of suppression pool, and delayed the failure of RCIC. Around 2 hours before the cooling water by fire pump was able to inject water into the reactor, the core damage started at around 76.5hr and got oxidized severely within 2 hours. While no further degradation occurred, the core geometry was maintained, and capable of being cooled by sea water injection.. A leakage has possibly occurred somewhere in RCS steam phase region, to account for pressurization of containment dry well before suppression pool got saturated.
22.	Chen, Yangli, et al. (author) Sensitivity and uncertainty analysis of trace simulation against FIx-II experiments 2016 In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM). Conference paper (peer-reviewed)abstract In a previous study [1], the US NRC code TRACE was employed to simulate the FIX-II tests which were carried out to investigate the loss of coolant accident (LOCA) of a boiling water reactor (BWR). Results exhibited that the TRACE simulation was sensitive to modelling parameters. In order to further qualify the TRACE code for BWR safety analysis, and to increase our confidence in the simulation results, sensitivity and uncertainty analysis is performed in this paper for the possible uncertain parameters, so as to identify the most influential ones. 12 parameters related to the simulated physical phenomena are selected by resorting to phenomena identification and ranking tables (PIRTs) in relative references. The sensitivity analysis method chosen is based on Finite Mixture Models (FMM) together with Hellinger distance and Kullback-Leibler divergence. Kolmogorov-Smirnov test is first introduced to combine FMM, and it has better performance in screening. Sensitivity analysis results of FMM method show that decay power, choked flow multipliers and break area have the most important influence on calculating peak cladding temperature (PCT). Although previous study failed to predict PCT, uncertainty analysis provides a certain range that successfully covers experiment result.
23.	Chen, Yangli, et al. (author) Uncertainty quantification for TRACE simulation of FIX-II No. 5052 test 2020 In: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 143 Journal article (peer-reviewed)abstract The Best Estimate Plus Uncertainty approach requires the knowledge of input uncertainties for the uncertainty propagation with best-estimate codes. Inaccurate judgement of some model parameter uncertainties related to the dominant physical phenomena may result in misestimation of the safety margin. This paper presents a framework of inverse uncertainty quantification (UQ) to quantify model parameter uncertainties in order to address this issue. It is applied to TRACE simulation of a large break loss of coolant accident conducted on the FIX-II facility, and peak cladding temperature (PCT) is the simulation output. Sensitivity analysis identifies the parameters of the critical flow model as the most influential to the PCT. The inverse UQ is performed based on Bayesian framework, which adopts Markov Chain Monte Carlo sampling and surrogate modelling algorithms. The quantified uncertainties of the model parameters are the desired results from the inverse UQ process, which are useful in BEPU studies.
24.	Cheng, X., et al. (author) European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems 2015 In: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 290, s. 2-12 Journal article (peer-reviewed)abstract Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.
25.	Chikhi, N., et al. (author) Evaluation of an effective diameter to study quenching and dry-out of complex debris bed 2014 In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 74, s. 24-41 Journal article (peer-reviewed)abstract Many of the current research works performed in the SARNET-2 WP5 deal with the study of the coolability of debris beds in case of severe nuclear power plant accidents. One of the difficulties for modeling and transposition of experimental results to the real scale and geometry of a debris bed in a reactor is the difficulty to perform experiments with debris beds that are representative for reactor situations. Therefore, many experimental programs have been performed using beds made of multi-diameter spheres or non-spherical particles to study the physical phenomena involved in debris bed coolability and to evaluate an effective diameter. This paper first establishes the ranges of porosity and particle size distribution that might be expected for in-core debris beds and ex-vessel debris beds. Then, the results of pressure drop and dry-out heat flux (DHF) measurements obtained in various experimental setups, POMECO, DEBRIS, COOLOCE/STYX and CALIDE/PRELUDE, are presented. The issues of particle size distribution and non-sphericity are also investigated. It is shown that the experimental data obtained in "simple" debris beds are relevant to describe the behavior of more complex beds. Indeed, for several configurations, it is possible to define an "effective" diameter suitable for evaluating (with the porosity) some model parameters as well as correlations for the pressure drop across the bed, the steam flow rate during quenching and the DHF.

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