Molten Core Concrete Interaction Research Articles

Experimental study on heat transfer characteristics of steam condensation in the presence of air and CO2 outside a vertical tube

Under severe accident conditions, in addition to steam and air, there are also other non-condensable gases inside the containment, such as CO2 produced by the molten corium concrete interaction (MCCI). Due to the significantly higher molecular weight and density of CO2 compared to steam and air, CO2 will have a significantly different impact on the steam condensation process compared to air. However, few scholars have paid attention to the impact of CO2 on the condensation process, and there is no empirical correlation applicable to this condition. Therefore, this article investigates the effect of different CO2 concentrations on the steam condensation heat transfer coefficient (CHTC) in the range of system pressure from 0.2 to 1.3 MPa and subcooling from 30 to 130 °C through experimental methods. Meanwhile, based on relevant experimental data, a new correlation suitable for calculating the steam CHTC in the presence of CO2 and air was fitted. The results indicate that under low steam concentration conditions, the addition of CO2 will reduce CHTC, while under high steam concentration conditions, CO2 will increase CHTC. The increase in system pressure will strengthen the condensation promotion effect of CO2 and weaken its inhibitory effect on condensation. In addition, by comparing experimental data under different component conditions, it was found that the new empirical correlation has high computational accuracy, and the calculation results of this correlation can contain more than 90 % of the data within a 15 % error range.

A Thermal Approach for Modeling Concrete Ablation During Molten Corium–Concrete Interaction

A Thermal Approach for Modeling Concrete Ablation During Molten Corium–Concrete Interaction

Analysis of the Long-Term Interaction Between Molten Core and Dry Concrete at Fukushima Daiichi Unit 1

The latest investigations of Fukushima Daiichi Unit 1 have demonstrated that corium attack to the pedestal walls and pedestal floor has occurred in Fukushima Daiichi Unit 1 to a certain extent. The results of past analytical benchmarks, such as the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF project), have agreed with this finding. However, the latest investigation does not show evidence of unlimited molten core–concrete interaction (MCCI), which is one of the main discrepancies from the BSAF project. More recently a MCCI benchmark has been launched in the context of the OECD/NEA project ARC-F (Analysis of Information from Reactor Building and Containment Vessels of Fukushima Daiichi Nuclear Power Station). In the benchmark, common geometry, boundary, and initial conditions have been selected among all the participants. The results show an improved agreement among different codes for what concerns overall erosion, corium temperature, and hydrogen generation, confirming that to some extent, the earlier scatter found in these variables came from differences in the MCCI scenario modeled by each partner. However, common unlimited erosion, not observed by onsite visual inspections, is still predicted. Understanding the origin of this deviation might provide insights into boundary conditions, model drawbacks, or ill-posed assumptions that might need to be revisited (e.g. interfacial temperature, effective heat transfer coefficients, concrete heat transfer). In this paper, a summary of the overall results and a discussion of modeling and boundary conditions is presented to disclose the results of the activity and the future steps to be taken in the OECD/NEA project FACE (Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation).

STAR-CCM+ simulation of debris bed formation in the unheated DEFCON-S experiment

In a severe accident of light-water reactor, molten corium can be formed and discharged into the reactor cavity. If the resulting debris bed is not adequately cooled, it could threaten the containment integrity by causing a molten core-concrete interaction. This study analyzes a non-heating debris bed formation experiment of DEFCON-S using the STAR-CCM+ code. Relevant CFD models are established to reflect the interaction between particles and fluids to describe the particle behavior and flow of fluids. The shape and size of debris bed are predicted through the CFD calculation. The rolling resistance becomes an important parameter in this simulation. The prediction with the default values of STAR-CCM+ forms a spread shape of particle bed compared to the experimental result. The shape of debris bed obtained with modified rolling resistance coefficients gives consistency with the DEFCON-S experiment, having a deviation in diameter of 4.6 % and a deviation in height of 3.8 %.

Simulant molten core–concrete interaction experiments in view of understanding Fukushima Daiichi Nuclear Power Station Cs-bearing particles generation mechanism

The Fukushima Daiichi accident resulted in the release of a novel form of radioactive Cs contamination into the environment, called Cs-bearing microparticles (CsMP). CsMPs constitute a substantial portion of the radioactive pollution near the nuclear power station and traveled beyond several hundred kilometers. Extensive characterization of the CsMPs revealed an amorphous silica matrix, along with Cs and other minor or trace elements such as Fe and Zn. This study explores the unclear generation mechanism of CsMPs by conducting experimental molten core concrete interactions (MCCI) as a source of Si and analyzing the resultant aerosols. The findings demonstrate that MCCI is in capacity to produce spherical submicronic and micronic particles, primarily composed of amorphous silica and incorporating elements akin to CsMPs. A humid atmosphere is found to favour an even closer chemical composition. Examination of the internal structure of the synthesized particles unveils pores and numerous crystalline nanoinclusions possibly serving as nucleation sites for CsMP formation through the condensation of Si-rich vapors.

Parametric study of molten core concrete interaction and assessment of corium coolability in the light of MCCI experiments

Molten Corium Concrete Interaction (MCCI) is an important phenomenon observed in the late phase of reactor severe accidents. A realistic analysis of MCCI can save cost and efforts to mitigate severe accidents. The heat transfer mechanisms in MCCI, such as melt eruption and water ingression, are the primary focus of current research owing to complications in their quantification and replication in severe accident analysis codes. This study investigated MCCI using top flooding strategy with MELCOR 2.2 code for two concrete types, and three experiments that were conducted under (OECD/MCCI-1 program). The results demonstrated that the ablation process terminated in the case of Limestone Common Sand (LCS) concrete. In the simulation of siliceous (SIL) concrete, the maximum and average melt front locations confirmed the asymmetrical ablation progression. The LCS concrete showed a smaller axial ablation front in this study; however, the comparison with the concrete ablation volume (0.21 m3) showed an exact match, and the difference between the experimental and predicted ablated mass of concrete was only 0.61%. The quantitative comparison between the prediction of the present study, CORQUENCH code, and experimental results is provided. The parametric values of MCCI phenomenon are also predicated and model improvement for MELCOR 2.2 code are discussed as well.

Analysis of High-Pressure extended SBO accident in VVER-1000: A sensitivity study on lower plenum structures modeling and core region nodalization in MELCOR

The lower plenum structures of VVER-1000 reactors exhibit a unique shape. This paper aims to investigate the sensitivity of various important parameters related to safety using different models in the MELCOR 1.8.6 code's COR package. The study focuses on modeling the lower plenum structures and the core region using different CVH package nodalizations during a High Pressure Extended Station Black-Out accident. At first, MELCOR calculated parameters are investigated during the course of a severe accident for a basic model of the lower plenum structures and core region nodalization. Different important parameters from a safety point of view are also examined. These include the total amount of in-vessel hydrogen production, the lower head breach time and elevation, the synchronicity between lower head breach and debris ejection, the total amount of hydrogen produced during molten corium concrete interaction, the total amount of ejected debris to the cavity, and the total amount of radioactive material release. Then, by considering sixteen different simulations corresponding to four lower plenum structures modeling and four core region nodalizations, the sensitivity of the mentioned parameters is investigated. The obtained results show a sensitivity of about 12.7% for the timing of the accident (equivalent to), 103% for the hydrogen produced inside the vessel (equivalent to 417.9 to 847.8 kg), 143% for the hydrogen produced outside the vessel (equivalent to 445.7 to 1081.4 kg), 47% for the total hydrogen produced (equivalent to 1207.5 to 1780.7 kg), and 76% for total radioactive material release (equivalent to 625.311 to 1101.892 kg) up to 24 h after the accident.

Development and validation of diffusion based CFD model for modelling of hydrogen and carbon monoxide recombination in passive autocatalytic recombiner

In water-cooled power reactor, hydrogen is generated in case of steam zirconium reaction during severe accident condition and later on in addition to hydrogen; CO is also generated during molten corium concrete interaction after reactor pressure vessel failure. Passive Autocatalytic Recombiners (PARs) are provided in the containment for hydrogen management. The performance of the PARs in presence of hydrogen and carbon monoxide along with air has been evaluated. Depending on the conditions, CO may either react with oxygen to form carbon dioxide (CO2) or act as catalyst poison, reducing the catalyst activity and hence the hydrogen conversion efficiency. CFD analysis has been carried out to determine the effect of CO on catalyst plate temperature for 2 & 4% v/v H2 and 1–4% v/v CO with air at the recombiner inlet for a reported experiment. The results of CFD simulations have been compared with the reported experimental data for the model validation. The reaction at the recombiner plate is modelled based on diffusion theory. The developed CFD model has been used to predict the maximum catalyst temperature and outlet species concentration for different inlet velocity and temperatures of the mixture gas. The obtained results were used to fit a correlation for obtaining removal rate of carbon monoxide inside PAR as a function of inlet velocity and concentrations.

Calculation of ablation instabilities during MCCI with the TIM model

The Transient Interface Model (TIM) (Seiler and Combeau, 2014; Seiler and Jamet, 2019) dedicated to Molten Core-Concrete Interaction simulation is used to investigate ablation instabilities. The ablation instability leads to concrete ablation in preferential directions, potentially jeopardizing the mitigation criterion based on residual concrete thickness. The ablation instability occurs when all the corium/concrete interfaces do not undergo the same ablation regime. These regimes are specific of the interface temperature; in the homogeneous Low Interface Temperature (LIT) regime, the interface temperature is close to the liquidus temperature of concrete matrix, whereas in the homogeneous High Interface Temperature (HIT) regime, the interface as well as the pool temperatures are close to the pool liquidus temperature. When these different ablation regimes are present simultaneously (non-homogeneous situation), only the surfaces under LIT regime are ablated. It is shown that ablation instabilities are favored at low power dissipation and at small test scale. Calculations suggest that the limit between HIT and LIT dominant regimes can be represented in the pseudo-binary phase diagram. The surfaces initially under HIT regime (not ablated) may return to the LIT regime (with ablation) after a time delay which depends on the extension of the surface initially under HIT regime, on the value of the mass transfer coefficient and, to some extent, on the melt to interface heat transfer. Finally, the calculation at reactor scale with siliceous concrete indicates that an ablation instability may occur in the initial stage of the MCCI transient. Such ablation instability does not induce significant preferential ablation in the reactor configuration if it is limited to the horizontal surface of the reactor pit. If it affects the lateral walls, and due to the small thickness of the corium layer, a significant ablation depth could be reached (∼1 m in a reactor pit of 6 m inner diameter).

A simplified core catcher model for the containment code AC2/COCOSYS

In Generation 3+ NPPs such as EPR or WWER-1200 the ex-vessel core catcher presents a key component of the severe accident mitigation concept. A simplified core catcher model in the containment code COCOSYS (part of AC2 code package) is developed to consider core melt retention and cooling phenomena in core catchers for containment simulations. The model enables the characterization of the corium in the core catcher including processes in the corium conditioning phase (melting of sacrificial material in the core catcher, oxidation of metallic components, possible layer flip) and the long-term cooling phase (evolution of long-term corium pool bulk temperature and changes in physical state). The core catcher model is based on the existing molten corium / concrete interaction (MCCI) model in COCOSYS with additions for melt retention within water-cooled structures and input data can be adapted to simulate various core catcher design features of e. g. NPPs with EPR or WWER.The current state of the model, including basic assumptions, calculation methods, simplifications, and two application examples are outlined. The recalculation of the water-cooled basemat test OECD WCB-1 shows that the experimental data regarding e. g. the ablation of the sacrificial concrete on top of the water-cooled basemat, corium pool temperature and heat fluxes at water-cooled interfaces can be adequately reproduced by the COCOSYS simulation, when using empirical data for the heat transfer, including the MCCI phase. These empirical heat transfer data agree well with findings obtained from recent analytical work on MCCI tests. However, the model does not predict high local bottom plate temperatures as observed in the WCB-1 test due to a restriction of the model to quasi-steady state processes. The second calculation example demonstrates the current modelling features when applied to an axisymmetric geometry of a reactor-scale core catcher. With view to reactor-scale, the model is ready for a further verification, like e.g. comparison with other analytical core catcher approaches.

Thermodynamics of aerosols during a molten core-concrete interaction at Fukushima Daiichi Unit 2 estimated conditions

Several radioactive releases from the containment took place during all the events leading up to Fukushima Daiichi endstate. Radionuclides released into the environment may differ in their composition or chemical form depending on the unit involved in the discharge and depending on the course of the accident. These differences may result from the path of the radionuclides from the core to the environment (e.g. through suppression pool water) and from the chemical and physical properties within the reactor core and the containment. On March 14th and 15th 2011, spherical glassy Cs-bearing microparticles (type-A CsMP) were collected for the first time. These microparticles might result from molten core-concrete interaction (MCCI) inside Unit 2, where the Zircaloy from the fuel cladding interacted with the SiO2 of the concrete pedestal at high temperature. The result of this interaction is a significant release of silica-rich gases or aerosols from the melt pool, which condense in colder locations. Calculations using ThermoCalc with OECD/NEA’s database TAF-ID version 11 were performed on the thermodynamics of vaporization during an MCCI to find correlations between the chemical and physical properties of the reactor atmosphere and the chemical composition of type-A CsMP. Data from ORIGEN2 and MELCOR calculations were used to estimate the core melt and atmosphere composition at the assumed time of the reactor pressure vessel failure. The vapour temperature at which type-A CsMP may have condensed in the containment has been evaluated. Using Fe/Si, U/Si and Cs/Si ratios in the observed CsMP as an indicator for vapour composition, this suggests that type-A CsMP originate from condensation of vapours around 2000–2200 °C from a low oxidised composition of corium, rich in zirconium and poor in stainless steel.

The development of a long-term predictive model of the changes in debris properties during interaction with aqueous solutions. Thermodynamic–kinetic modeling

This article presents a long-term predictive model of the properties’ changes of the molten core–concrete interaction (MCCI) debris material formed during the Fukushima Daiichi nuclear power plant (NPP) accident based on the available experimental data. For the development of the model, samples of MCCI debris were synthesized to represent the composition of the actual Fukushima debris. The samples' composition simulated different stages of ablation of the concrete basement beneath the NPP unit, i.e., various “concrete–corium” ratios. MCCI simulated debris samples (monoliths and powders) were exposed to aqueous solutions under the following conditions:1) continuous contact with water, i.e., no renewal of aqueous solutions (batch leaching experiments);2) periodic renewal of aqueous solutions by fresh ones (conditions simulating an open system with a constant inflow of fresh portions of aqueous solution).Experiments were performed under various conditions (temperature, pH) in two types of aqueous solutions: deaerated deionized and aerated deionized water. During the experiments, aliquots were taken to measure the concentrations of dissolved components (Ca, Si, Al, U, Zr, Fe, Ni, and Cr) in the aqueous solution.MCCI debris “aging” experiments were interpreted using kinetic curves and their relationship with physicochemical conditions (temperature, pH, type of aqueous solution). In this interpretation, the chemical processes of the MCCI debris components in water phase (dissolution and precipitation) were considered.The model was extended to 50 years and used to interpret debris aging experiments: (1) prediction of components’ releases into the aqueous phase caused by dissolution of solid phases, and (2) prediction of dusty secondary phase formation at the “debris–aqueous solution” interface.

Analyses of Critical Hydrogen Enrichment in PWR Containment Compartments

During a loss-of-coolant accident in a pressurized water reactor (PWR), steam of varying quality is released from the primary circuit into the equipment compartments of the containment, followed by the release of a hydrogen-steam mixture during the core degradation phase. In the case of long-lasting accidents, findings of detailed code analyses indicate an enrichment of hydrogen in lower peripheral containment compartments in the reference PWR plant under investigation. During the late accident phase with ex-vessel molten core–concrete interaction, even in the case of an operating passive autocatalytic recombiner system, this poses a threat for local hydrogen combustion later on. Such hydrogen phenomena are not expected and have not been widely studied up to now. Therefore, corresponding experiments have been performed at the THAI test facility operated by Becker Technologies. One of these tests had been precalculated with the COntainment COde SYStem (COCOSYS) as part of the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) code system AC2 and has been used to validate the code. The 60-m3 THAI test vessel has been divided into an inner compartment that has been connected to the surrounding vessel, simulating the upper and peripheral containment part, by very small flow openings at the bottom representing the clearance between door frames and door leaves and one opening at the top representing typical openings by burst disks. The paper discusses both the experimental findings of a test series on the potential enrichment of hydrogen in lower containment compartments and the COCOSYS calculations demonstrating the applicability of the code under complex flow conditions including stratification phenomena.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part II: Analysis of containment thermal-hydraulics and ex-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs).This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF).The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability.The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA).For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated.The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS).In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated.The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results.

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Assessment of corium concrete interaction and impact of water ingression mechanism using top flooding strategy for a small PWR with MELCOR 2.2

Molten Corium Concrete Interaction (MCCI) process is a critical phenomenon in the management of the late phase of a severe accident. Determination of water ingression (WI) and melt eruption (ME) impact on the ablation process is necessary for the assessment of cavity erosion. A model is developed to study the MCCI for Siliceous (SIL) and Limestone Common Sand (LCS) concretes in dry and water-filled cavities using MELCOR 2.2 code for a small PWR. It is found in this study that the ablated volume is greater in the SIL as compared to the LCS concrete. WI and ME were able to successfully halt the ablation process for LCS concrete under water flooding. Even with WI, SIL concrete ablation cannot be completely stopped in the axial and radial directions. In terms of concrete ablation, the cavity constructed with LCS is found to be safer than the cavity constructed with SIL concrete.

CFD and LP code benchmark evaluating the onset of par operation in case of extremely low oxygen concentration

In many reactor containments Passive Autocatalytic Recombiners (PAR) are installed to mitigate the hydrogen threat in case of a severe nuclear reactor accident. PARs oxidize hydrogen catalytically into water vapor. The heat of catalytic reaction drives flow through the PAR unit by natural convection, which continuously brings “fresh” hydrogen-air mixtures to the PAR unit, thereby developing a self-sustained process until either hydrogen or oxygen falls below a minimum concentration.The test series (HR-33 to HR-35) of the OECD/NEA THAI-2 program investigated the influence of temperature, pressure and steam on the onset of hydrogen recombination in case of low or extremely low oxygen concentrations. These conditions with oxygen concentrations below 1 vol–% can occur in accident scenarios (e.g. late accident phase with molten core concrete interaction). A better understanding of the PAR performance in oxygen-limited conditions is important for hydrogen management in severe accidents.The test data of HR-35 was used for a blind benchmark exercise within the OECD/NEA joint project. Lumped parameter codes (MELCOR and COCOSYS) have been used as well as 3D/CFD codes (GASFLOW, GOTHIC and CFX). A variety of PAR models were used, including models based on engineering correlations, models with a detailed nodalization of the PAR unit, and a model with detailed heat and mass transfer process (REKO-DIREKT). Consequently, the benchmark provided a good opportunity to assess the PAR models that are deployed in the state of the art safety analysis tools, especially under conditions representative for the late phase of a severe accident.

PAR model development exercise in the framework of SAMHYCO-NET

The simulation of passive auto-catalytic recombiner (PAR) operation is a pre-requisite for the assessment of the hydrogen combustion risk and hydrogen mitigation measures in light water reactors (LWRs). However, the ex-vessel boundary conditions of a severe accident pose a challenge for numerical models due to lack of suitable experimental data for validation. Especially the PAR interaction with carbon monoxide (CO) released in significant amounts during molten corium-concrete interaction has not been systematically investigated yet.The international network SAMHYCO-NET has performed several numerical benchmark exercises related to the behavior of combustible gases (hydrogen and carbon monoxide) inside LWR containments during severe accidents. In the field of PAR simulation, an international group of 8 institutions, supported by Framatome and Becker Technologies, have been collaborating towards a realistic and reliable assessment of PAR operation under challenging ex-vessel conditions, including the effect of oxygen starvation and the presence of CO. Focusing on the Areva PAR design, the partners applied various approaches to model the recombination rate, such as the engineering correlation provided by the vendor, a mechanistic model (REKO-DIREKT), and a full chemistry model (SPARK). Based on the experimental data obtained in the THAI and REKO-3 facilities, a step-wise approach towards PAR model assessment has been implemented. While the comparison of model predictions with experimental data shows in general good agreement, significant discrepancies still exist especially for oxygen-lean conditions.

Passive Autocatalytic Recombiner Perfomance Assessment Using COCOSYS

The progression of a severe accident in nuclear reactors is composed by a diversity of phenomena that areBeyond Design Basis, such as Direct Containment Heating (DCH), Molten Corium Concrete Interaction(MCCI), hydrogen detonation, and others. Currently, there are several devices and systems that allowmitigating the progression of this events, avoiding the failure of the physical barriers between the nuclearpower plant and the environmental. In this context, the present work aims to reproduce the HR-14experiment carried out at the Thermal-hydraulic, Hydrogen, Aerosols and Iodine (THAI) test facility throughthe Passive Autocatalytic Recombiners (PAR) performance assessment with the COCOSYS code. The analysisof the convergence of the results was performed using the Fast Fourier Transform Based Method (FFTBM)and showed that the results had sufficient accuracy with the experimental data.

Synthesis of Fukushima Daiichi Cs-bearing microparticles through molten core-concrete interaction in nitrogen atmosphere

Cs-bearing microparticles, released during the accident of Fukushima Daiichi, are a significant concern for health and environment due to their abundance over wide areas, associated with a low solubility. Since their discovery, efforts have been made to characterise these particles and use them as direct witnesses of the various events inside the reactors. This study reports the experimental synthesis of micrometric particles through molten core-concrete interaction. Primary chemical analyses on Al/Si, K/Si and Na/Si atomic ratios seem to confirm concrete as source of Si. Cs has also been identified in low concentrations.