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Melt infiltration into porous media is an intriguing phenomenon that holds immense significance across various sciences and technologies. In this work, the problem of metallic melt infiltration in particulate beds is investigated for understanding and prediction of severe accident progression associated with a molten pool penetrating through an underlying debris bed which may form in the reactor core or in the lower head of a light water reactor. The present study aims to quantify the effect of particle surface's wettability on melt infiltration kinetics. For this purpose, two categories of experiment are conceived and carried out to measure the wettability of different material surfaces by melt and to characterize melt infiltration kinetics in one-dimensional particulate beds, respectively. The melt material is tin–bismuth eutectic alloy with a melting point of 139 °C. Copper (Cu), stainless steel (SS), Tin (Sn) and tin-coated stainless steel (Sn-coated SS) are chosen as materials of substrates and particles in wettability measurement and melt infiltration study. The particulate beds, packed with 1.5 mm spheres, are preheated to 200 °C before the melt infiltration begins. The experimental data of wettability measurement shows that the contact angles of liquid Sn-Bi eutectic on the above-mentioned material surfaces range from 79° to 135°. The results of melt infiltration tests confirm the significant effect of wettability on melt penetration kinetics. The capillary force plays a significant role in the initial infiltration of particulate beds. Specifically, a wettable particulate bed enhances the initial melt infiltration, whereas non-wettable beds hinder it., QC 20240702

In the framework of the IAEA Coordinated Research Project on In-Vessel Melt Retention, a benchmark of CFD simulations, devoted to thermal-hydraulic behavior of the light metal layer involves several research organizations: KTH of Sweden, SSTC NRS of Ukraine, ÚJV Řež of Czech Republic and CEA of France. This work aims at better simulating the focusing effect phenomenon leading to a heat flux peak along the height of the light metal layer, which is formed above the oxide layer in a stratified corium pool configuration during a PWR severe accident. This is a known safety issue compromising the reactor vessel integrity. The first benchmark step provides a solid foundation to the CFD schemes (physical models, meshes) by comparing the results of CFD simulations with thermal-hydraulic experimental data obtained using water as simulating fluid in a representative and quite laminar configuration. Then a similar but highly turbulent case, of higher height, is considered for more complex validation of the numerical simulation approach. Results with different turbulent models are compared against experimental data. On the strength of this encouraging work, a simulation of the same height configuration but considering steel fluid under severe accident conditions is foreseen at the final stage of this benchmark., Part of ISBN 9784888982566QC 20231214

The severe accident scenario propagation studies of nuclear power plants (NPPs) have been one of the most critical factors in deploying nuclear power for decades. During an NPP accident, the accident scenario can change during its propagation from the initiating event to a series of accident sub-scenarios. Hence, having time-wise updated information about the current type of accident sub-scenario can help plant operators mitigate the accident propagation and underlying consequences. In this work, we demonstrate the capability of machine learning (Decision Tree) to help researchers and design engineers in finding distinctive physical insights between four different types of accident scenarios based on the pressure vessel's maximum external surface temperature at a particular time. Although the four accidents we included in this study are considered some of the most extensively studied NPPs accident scenarios for decades, our findings shows that decision tree classification could define remarkable distinct differences between them with reliable statistical confidence., QC 20230626

In a postulated severe accident, the thermo-mechanical loads from the corium debris that has relocated to the lower head of the reactor pressure vessel (RPV) can pose a credible threat to the RPV's structural integrity. In case of a vessel breach, it is vital to predict the mode and timing of the vessel failure. This affects the ex-vessel accident progression and plays a critical role in the development of mitigation strategies. We propose a methodology to assess RPV failure based on MELCOR and ANSYS Mechanical APDL simulations. A Nordic-type boiling water reactor (BWR) is considered with two severe accident scenarios: i) SBO (Station Blackout) and ii) SBO + LOCA (Loss of Coolant Accident). In addition, the approach considers the dynamic ablation of the vessel wall due to a high-temperature debris bed with the use of the element kill function in ANSYS. The results indicate that the stress failure mechanism is the major cause of the RPV failure, compared to the strain failure mechanism. Moreover, the axial normal stress and circumferential normal stress make the dominant contributions to the equivalent stress sigma at the lower head of RPVs. As expected, the region with high ablation is most likely the failure location in both SBO and SBO + LOCA. In addition, comparisons of the failure mode and timing between SBO and SBO + LOCA are described in detail. A short discussion on RPV failure between ANSYS and MELCOR is also presented., QC 20220322

In a postulated severe accident, a molten pool with decay heat can form in the lower head of a reactor pressure vessel, threatening the vessel’s structural integrity. Natural convection in molten pools with extremely high Rayleigh (Ra) number is not yet fully understood as accurate simulation of the intense turbulence remains an outstanding challenge. Various models have been implemented in many studies, such as RANS (Reynolds-averaged Navier–Stokes), LES (large-eddy simulation), and DNS (direct numerical simulation). DNS can provide the most accurate results but at the expense of large computational resources. As the significant development of the HPC (high-performance computing) technology emerges, DNS becomes a more feasible method in molten pool simulations. Nek5000 is an open-source code for the simulation of incompressible flows, which is based on a high-order SEM (spectral element method) discretization strategy. Nek5000 has been performed on many supercomputing clusters, and the parallel performance of benchmarks can be useful for the estimation of computation budgets. In this work, we conducted scalability tests of Nek5000 on four different HPC clusters, namely, JUWELS (Atos Bullsquana X1000), Hawk (HPE Apollo 9000), ARCHER2 (HPE Cray EX), and Beskow (Cray XC40). The reference case is a DNS of molten pool convection in a hemispherical configuration with Ra = 1011, where the computational domain consisted of 391 million grid points. The objectives are (i) to determine if there is strong scalability of Nek5000 for the specific problem on the currently available systems and (ii) to explore the feasibility of obtaining DNS data for much higher Ra. We found super-linear speed-up up to 65536 MPI-rank on Hawk and ARCHER2 systems and around 8000 MPI-rank on JUWELS and Beskow systems. We achieved the best performance with the Hawk system with reasonably good results up to 131072 MPI-rank, which is attributed to the hypercube technique on its interconnection. Given t, QC 20230213

Fluid infiltration, solidification, and remelting in a particle bed are complex phenomena that can occur in the lower head of a reactor pressure vessel (an in-vessel phenomenon) or in the reactor cavity below the vessel (an ex-vessel phenomenon) during a severe accident in a nuclear power plant. When the non-homogeneous corium, consisting of metal and oxide components, reheats, the lower-melting metals will melt first and move downward to the bottom of the reactor pressure vessel. This will change the global debris bed configuration and its physical and chemical properties, and thereby actively influence the accident progression, specifically the mode and timing of possible vessel failure and the melt characteristics upon release. Similar ex-vessel debris can form on the cavity floor below the vessel, which can threaten containment integrity if stable cooling is not established. In this paper, we present an experimental program employing recently constructed MRSPOD (multicomponent remelting, relocation, and solidification in porous debris) facility that mainly investigates melt infiltration, solidification, remelting, and relocation in a particulate debris bed. The facility uses a 12 x 130 cm(2) (OD x Length) quartz tube in a cylindrical furnace and allows a debris bed to be configured, heated, and/or pressurized prior to fluid infiltration through the bed. The MRSPOD experiments were instrumented with thermocouples (TCs), fiber Bragg grating (FBG) sensors, laser sensor, video, and infrared cameras, which are essential in describing the overall melt infiltration and solidification behavior. Here, a eutectic Sn-Bi melt with superheat temperature between 50 and 70 & DEG;C is poured into a preheated particle bed consisting of 1.5-mm spherical particles made of either copper (Cu), Sn-coated Cu, stainless steel (SS), Sn-coated SS, and/or glass beads to study the effect of thermal properties and wettability on the melt infiltration. Moreover, melt infiltration into a, QC 20230125

QC 20220620

For Light Water Reactor (LWR) safety, spray cooling during severe accidents is one of the promising approaches to achieve In-Vessel Retention of corium by External Reactor Vessel Cooling (IVR-ERVC). To study the efficiency of multi-nozzle spray cooling (nozzles of 2×3 matrix) on a downward-facing FeCrAl heated surface, a lab-scale experimental facility was built. It should be emphasized, however, that a direct measurement of Heat Transfer Coefficient (HTC) on the sprayed side is challenging due to the strong interference of water flow and intrusiveness of standard instrumentation methods. In this paper, a 3D numerical model has been established with the same geometric and material parameters as the foil sample in a multi-nozzle upward spray cooling. Given the experimental temperature profiles on the sample's dry side measured by an IR camera, the complementary numerical simulations have revealed the HTCs and corresponding temperature profiles on the sprayed side, which enabled the prediction of the maximum heat fluxes (MHFs). The maximum heat fluxes for the given spray cooling conditions can reach up to 3.25 MWm2, which is more than adequate for what is required for a successful IVR-ERVC for high-power reactors. At the same time, the maximum temperature on the dry side at the highest input power is still much lower than the expected failure temperature of the sample material., QC 20230622

This paper investigates the dynamics of melt infiltration through a preheated porous debris bed which is of importance to severe accident modeling in nuclear power plants. Proper understanding of the flow physics and affecting parameters is needed to define flow regime(s) according to combination of the driving forces, i.e. capillary and gravity. A model development and validation therefore should consider various effects and competing mechanisms. After a careful study of the governing equations and scaling rules, a known analytical model is validated against existing experimental data from REMCOD experiment. The predictions of this model are in good agreement with the experimental data. Furthermore, a global sensitivity analysis identifies the most influential parameters and reveals the need for further experiments with different range of affecting parameters. The results underline the importance of permeability as the most influential parameter., QC 20210726

The re-melting of multi-component debris is important for both in-vessel and ex-vessel phases of severe accident progression in nuclear power plants. However, current knowledge is limited with respect to understanding the associated complex phenomena and their interactions. In this paper, the phenomenon of melt infiltration through a porous debris bed with and without solidification is examined by synthesizing the data obtained from ongoing experimental research (REMCOD facility). In this regard, results obtained from 12 experiments are analyzed. Eight tests were conducted for melt infiltration through debris at temperatures above solidification. At this condition, two flow regimes are identified for the melt flow inside the hot porous debris, which is initially dominated by capillary forces and hydrostatic head and then later by the gravity forces. In addition, 4 tests were performed for melt penetration into cold debris where melt infiltration is limited by solidification. It was found that the depth of penetration is correlated with the difference between "sensible heat of melt" and "the amount of heat required to heat the bed up to the melting point of specific melt composition.", QC 20210714

The reactor pressure vessel (RPV) of a nuclear reactor is one of the key safety barriers preventing radioactive environmental releases during a severe accident. One of the promising strategies of severe accident management (SAM) is to retain the molten core having continuous decay heat inside the RPV by natural water cooling of the external vessel surface. The feasibility of such a strategy relies on complex safety analyses including accurateprediction of vessel thermo-mechanical behavior which can be assessed by mechanical stresses and strains. In this paper, we present the stress–strain response of an ablated RPV of a Nordic boiling water reactor (BWR) to dynamic thermomechanical loads set by expanding volumetrically heated molten pool inside the RPV cooled by water at the external surface. MELCOR 2.2.9541 severe accident code is used to simulate the in-vessel behavior and provides the input conditions for dedicated structural analysis of the RPV using ANSYS® Mechanical APDL 19.2. A creep model of the SA533B1 vessel steel is validated against uniaxial creep tests carried out by INEL (Idaho National Engineering Laboratory) and creep tests performed at CEA (French AlternativeEnergies and Atomic Energy Commission) as part of the OLHF (OECD Lower Head Failure) Project. Two generic severe accident scenarios are considered: (i) Station Blackout (SBO) and (ii) Station Black-out and Loss-of-coolant Accident (SBO + LOCA). In both scenarios, we found that the RPV has maintained structural integrity considering two failure criteria: stress-based and strain-based., QC 20210521

Spray cooling is a versatile technology for various cooling applications involving high surface heat fluxes. Experimental facility was built to study heat transfer performance of an upward multi-nozzle array of water sprays impacting a surface of heated plate made of reactor vessel grade steel. The effect of inclination angles of the steel surface on the cooling performance was investigated to assess heat transfer in complex semispherical/ semielliptical geometry of large reactor lower head and to address possible application of spray cooling in severe accident management (SAM) of light water reactors (LWRs) based on In-vessel melt retention with external reactor vessel cooling (IVR-ERVC). Joule heating of SA302B steel foil of 0.15 mm thickness and surface area of 96 cm2 enabled prototypic heat fluxes to be evacuated during reactor accident. A 2×3 array of full jet narrow-coned pressure-swirl spray nozzles was used to reproduce multi-nozzle cooling. The tests were conducted as a series of consequent steady states realized at stepwise increasing power and surface heat fluxes up to the maximum values of 29 kW and 2.97 MW/m2 limited in the specific facility design. Seven surface inclinations, between 0o and 90o were tested and no significant variations in spray cooling performance with the inclination of the heated surface was found. The results indicated a promising prospect of using a multi-nozzle array system for cooling of large surface area of reactor lower head. Much higher heat fluxes can be safely extracted by spray cooling in comparison with the critical heat fluxes that appeared at RPV water pool cooling at natural convection. The maximum value of heat flux at direct spray impact zones and its drop-off slightly from the center to the periphery of the spray cone was detected in the tests. The water flow rate and liquid subcooling significantly influenced maximum steel surface temperature but had no noticeable effects on surface temperature uniformity. The spra, QC 20210218

Cooling by water spray is a well-known technology that can reach significantly higher Critical Heat Flux (CHF) compared to other cooling methods. For the light water reactor safety, the in-vessel retention (IVR) by external reactor vessel cooling (ERVC) is a comprehensive severe accident management strategy to arrest and confine the corium in the lower head of the reactor pressure vessel. Heat fluxes up to 1.5 MW/m2 have already been assumed attainable in low-power nuclear reactors while cooling required in high-power reactors is expected to reach 2.5 MW/m2. Instead of reactor lower head flooding and relying on cooling due to natural convection, a viable and more efficient alternative is to spray the external surface of the vessel. Given all the advantages of spray cooling reported in the literature, a lab-scale experimental facility was built to validate the efficiency of multi-nozzle spray cooling of a downward-facing heated surface inclined at different angles up to 90o. The facility employed a 2×3 matrix of spray nozzles to cool the FeCrAl alloy foil with an effectively heated surface area of 96 cm2 using water as the coolant. Heat loads and surface inclinations were varied parameters in the test matrix. The results show that no significant variations in spray cooling performance concerning the inclination of the heated surface. A surface heat flux of 2.5 MW/m2 was achieved at every inclination of the downward-facing surface. The results also indicate that more uniform liquid film distribution could be obtained for some inclinations, which in turn leads to maintaining low surface temperature. The obtained surface heat flux margin by spray cooling indicates that it is feasible to adopt IVR-ERVC strategy for a large power reactor., QC 20210105

Vessel penetrations are important features of both pressurized water reactors and boiling water reactors. The thermal and structural behaviour of instrumentation guide tubes (IGTs) and control rod guide tubes (CRGTs) during a severe accident is vital in the assessment of the structure integrity of the reactor pressure vessel. Penetrations may fail due to welding failure, nozzle rupture, melt-through, etc. It is thus important to assess the failure mechanisms of penetrations with sufficient details. The objective of this paper is to assess the timing and failure modes of IGTs at the lower head during a severe accident in a Nordic boiling water reactor. In this study, a three-dimensional local finite element model was established using Ansys Mechanical that includes the vessel wall, the nozzle, and the weld joint. The thermo-mechanical loads of the finite element model were based on MELCOR results of a station blackout accident (SBO) combined with a large-break loss-of-coolant accident (LBLOCA) including an external vessel cooling by water as a severe accident management strategy. Given the temperature, creep strain, elastic strain, plastic strain, stress and displacement from the ANSYS simulations, the results showed the timing and failure modes of IGTs. Failure of the IGT penetration by nozzle creep is found to be the dominant failure mode of the vessel. However, it was also found that the IGT is clamped by the flow limiter before the nozzle creep, which means that IGT ejection is unlikely., Part of ISBN 978-488898256-6QC 20230921

In order to investigate the mechanisms of steam explosion which may occur during a severe accident of light water reactors (LWRs), the MISTEE facility was developed at Royal Institute Technology (KTH) to visualize the micro interactions of steam explosion when a single molten droplet was falling into a water pool. For preparation of a molten droplet, an aerodynamic levitation system was proposed to prevent the droplet from falling out of the crucible during heating in an induction furnace by injecting argon gas through a purging line connected to the bottom nozzle of the crucible. To support the design of such levitation system, a numerical simulation of the aerodynamic levitation system was performed using the CFD code ANASYS FLUENT v16.2. The problem was simplified as adiabatic two-phase flow dynamics in a 2-D axisymmetric geometry. The VOF method is employed to track the interface of two phases (liquid metal and argon gas), and the SST k-omega model was chosen for turbulence. Various characteristics of droplet dynamics in incorporated with argon gas flowrates through the crucible were examined in the numerical simulation. The simulation results suggested there exists an optimal range of argon gas flowrate for levitating a coherent shape of droplet in the crucible. The wall adhesion had a considerable effect on initiating the levitation of the droplet, which means the properties of the inside surface of the crucible may play an important role in the levitation and discharge of the droplet. Proof-of-concept tests were carried out on the prototype of the design, and it was confirmed that the levitation system was able to fulfill its function, i.e., to keep the droplet in the crucible during its melting process, although the actual argon gas flowrates for levitation was higher than the predicted ones, probably due to the leakage of flow path and heat transfer which were not considered in the simulation. Generally speaking, the numerical simulation did not only help u, QC 20200918