Molten Core Concrete Interaction Research Articles

Thermodynamic analysis for solidification path of simulated ex-vessel corium

Molten corium is considered to form in a severe accident at the Fukushima Daiichi Nuclear Power Station (FDNPS). Investigation of the primary containment vessel (PCV) inside the FDNPS showed that a significant amount of the molten corium reached the bottom of the pedestal region. The molten corium and concrete likely caused a complex interaction called Molten Corium Concrete Interaction (MCCI). Then, ex-vessel debris was formed after the solidification of the MCCI products and accumulated in the pedestal region. The solidification hysteresis of ex-vessel debris significantly influences its properties. We performed a thermodynamic analysis using the TAF-ID database to infer the solidification path of U–Zr–Al–Ca–Si–O molten corium, which was chosen for a prototypic system of ex-vessel debris. The solidification path for the CaO-rich sim-corium showed the following sequence: (i) melting as a single liquid phase at temperatures higher than 2430 K, (ii) selective solidification of the oxide-rich corium mainly consisted of fuel materials, and (iii) solidification of the remaining materials as a silicate matrix. In contrast, the solidification path for the SiO2-rich corium indicated that (i) formation of liquid miscibility gap above 2200 K between U-rich and Zr-rich oxidic melts, (ii) individual precipitation of solid phases in each liquid phase. Identified features of these solidification paths were in good agreement with the experimental data for the solidified sim-corium.

Sensitivity Analysis of Ex-Vessel Corium Coolability Models in MAAP5 Code for the Prediction of Molten Corium–Concrete Interaction after a Severe Accident Scenario

A postulated progressing severe accident scenario has been simulated using MAAP5 code with the focus on ex-vessel cooling of molten corium in the reactor cavity. Various parameters associated with the prediction of molten corium–concrete interaction (MCCI) are identified. Accordingly, a sensitivity analysis is performed to assess the impact of these parameters on the predicted cavity floor erosion depth during this MCCI postulated accident. The sensitivity index of each variable parameter is determined using the Cotter indices method and Sobol′ indices method. At the early stage of the accident, the predicted cavity floor erosion depth is found to be highly sensitive to the downward heat transfer coefficient parameter with Cotter and Sobol′ indices of 94% and 50%, respectively. At the late phase of the accident, however, the cavity floor erosion depth becomes sensitive to melt eruption (Cotter index of 40%), water ingression (Cotter index of 13%), and particulate bed (Cotter index of 15%) parameters alongside the downward heat transfer coefficient (Cotter index of 16%) with the melt eruption parameter becoming dominant. Thus, the sensitivity of the code′s predictions can be minimized by improving the physical models associated with these parameters. Moreover, the sensitivity indices of these parameters can be used by model developers to identify unimportant parameters in a bid to reduce the dimension of the problem with the aim of improving the current predictive capabilities to conduct MCCI-related safety analyses.

Investigating the mechanical behaviour of Fukushima MCCI using synchrotron Xray tomography and digital volume correlation

A primary target towards the clean-up operation of the Fukushima disaster is the retrieval of Molten Core-Concrete Interaction (MCCI) products, presently residing on the basement of the damaged nuclear reactor Units 1–3. MCCI is a fusion of materials, composed of both nuclear fuel cladding and neighbouring structural components. Determining the currently unknown, physical and mechanical properties of MCCI is essential for successful and timely retrieval. In this paper, we aim to experimentally quantify the mechanical properties of a material fabricated to resemble MCCI. A small-scale representative specimen was mechanically tested using Hertzian indentation stepwise loading. Synchrotron X-ray computed tomography was conducted at several loading stages to reveal the sample microstructure and mechanical degradation. The acquired tomograms were analysed by digital volume correlation to measure full-field displacements and strains developed within the sample volume. Young’s modulus and Poisson ratio were determined via this combined methodology.

NPP Krško containment modelling with the ASTEC code

ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagenund Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident (SA). The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krško. The accident analysis was focused on containment behaviour; however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals, molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant containment parameters, such as compartments pressures and temperatures, is going to be discussed in the paper.

A new potential interface tension model for MPS method avoiding unphysical particle cohesion

Improvement of the capability to accurately track the interface between immiscible phases/components is of great importance to the numerical analysis with complex multi-component and multi-phase flow dynamics, which might be involved in severe accidents of light water reactors, such as Molten Core-Concrete Interaction (MCCI). In order to model the interface tension between immiscible phases/components, Zhu et, al., precedingly developed a potential interface tension model in MPS method based on Kondo et al.‘s earlier model for surface tension and succeeded to stably model the interface tension qualitatively by introducing different attractive forces for different pairs of particle types. However, it has been known that pressure is overestimated in Zhu et, al.‘s potential interface tension model. Moreover, this study has shown that the fluid flow can be numerically hindered with Zhu et al.‘s potential interface tension model. These issues of Zhu et al.‘s potential interface tension model is caused by unphysical particle cohesion due to the inter-particle attractive forces determined by the surface tension coefficients, which are required even when the free surface is not considered. Meanwhile, in the framework of SPH method, Zhou et, al., proposed to model interface tension by introducing the repulsive force between different phases. Hence, this study aimed to develop a new potential interface tension model for MPS method which can overcome the conventional drawbacks of pressure overestimation or unphysical flow hindrance, by incorporating the repulsive force between immiscible fluid phases referring to Zhou et, al.‘s idea within SPH method.The interface tension in the present method is modeled as the repulsive force derived from the “repulsive potential” and is considered only near the interface to avoid the unphysical particle cohesion due to the inter-particle attractive forces. The amplitude coefficient in the repulsive potential is determined by the methodology similar to that proposed by Kondo, et, al.With the newly developed interface tension model, the common issue of numerical flow hindrance that occurred in the previous interface tension models in MPS method has been resolved. The validation of the newly developed interface tension model shows good agreement with theoretical evaluations of Laplace pressure and the droplet deformation ratio in a shear flow. Moreover, the new interface tension model has been validated against a 3-dimensional immiscible fluid stratification test, in which the new interface tension model shows good agreement with the test results and improvement of the pressure profile evaluations compared to that by Zhu et al.‘s model. Further extension of the present model so as to incorporate surface tension and wall wetting is one of the most important issues which require future development.

Thermochemical Modeling of Metal Composition and Its Impact on the Molten Corium–Concrete Interaction: New Insights with Sensitivity Analysis

The characterization of molten corium–concrete interaction (MCCI) has increasingly become a cause of concern because, in the case of a severe nuclear accident, the core could meltdown and release radiation into the environment. The objective of this study was to determine the thermochemical impact of metal content in the corium and analyze the effect of corium metal content on ablation depth, corium temperature, its viscosity and surface heat flux, and production of hydrogen, carbon monoxide, and carbon dioxide. The governing heat transfer equations were solved while considering the various thermochemical reactions in the existing numerical code in a comprehensive way. The developed MCCI model in CORQUENCH was validated against the data available in the literature. Our findings showed that the composition of corium, especially its metal content, has a noticeable effect on mitigating or aggravating the ablation depth and nuclear reactor integrity. We observed that during molten corium–concrete interaction, zirconium plays a significant role and its presence can increase the ablation depth exponentially from 18.5 to 139 cm in the investigated case study. It was found that the presence of zirconium in the corium instigated various thermochemical reactions continuously, and thus the injected water, instead of quenching the molten corium, enhances the temperature by facilitating exothermic reactions. Additionally, due to the presence of zirconium, the production of hydrogen and carbon monoxide increases by 45 and 52 times, respectively and the generation of carbon dioxide becomes zero because the zirconium reacts with carbon dioxide continuously, converting it to carbon monoxide.

Numerical analysis on in-core ignition and subsequent flame propagation to containment in OPR1000 under loss of coolant accident

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium–concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

Thermophysical properties of molten FeO1.5, (FeO1.5)0.86–(ZrO2)0.14 and (FeO1.5)0.86–(UO2)0.14

ABSTRACT FeO1.5 is a promising candidate for a sacrificial material to lower the liquidus temperature of molten corium while keeping the viscosity as low as possible in order to mitigate the molten core–concrete interaction. Phase diagrams of such systems as FeO1.5–ZrO2 and FeO1.5–UO2 indicate that the FeO1.5 addition would significantly lower the liquidus temperature of corium, but how it would affect the viscosity has not been investigated. Thus, the viscosity of FeO1.5, (FeO1.5)0.86–(ZrO2)0.14, and (FeO1.5)0.86–(UO2)0.14 was measured using the aerodynamic-levitation technique coupled with the droplet-oscillation method. It was found that the viscosity of (FeO1.5)0.86–(ZrO2)0.14 at approximately 2100 K is almost identical to that of ZrO2 at around 3200 K, and that (FeO1.5)0.86–(UO2)0.14 maintains a low viscosity at a temperature well below the melting temperature of UO2, owing to the low viscosity of FeO1.5. These results indicate that the iron oxide addition would not only lower the liquidus temperature of the molten corium but also keep its fluidity as high as possible.

Chemical characterisation of degraded nuclear fuel analogues simulating the Fukushima Daiichi nuclear accident

The Fukushima Daiichi accident generated degraded nuclear fuel material, mixed with other reactor components, known as molten core-concrete interaction (MCCI) material. Simulant MCCI material was synthesised, excluding highly radioactive fission products, containing depleted U, and incorporating Ce as a surrogate for Pu. Multi-modal µ-focus X-ray analysis revealed the presence of the expected suite of U-Zr-O containing minerals, in addition to crystalline silicate phases CaSiO3, SiO2-cristobalite and Ce-bearing percleveite, (Ce,Nd)2Si2O7. The formation of perclevite resulted from reaction between the U-Zr-O-depleted Ce-Nd-O melt and the silicate (SiO2) melt. It was determined that the majority of U was present as U4+, whereas Ce was observed to be present as Ce3+, consistent with the highly reducing synthesis conditions. A range of Fe-containing phases characterised by different average oxidation states were identified, and it is hypothesised that their formation induced heterogeneity in the local oxygen potential, influencing the oxidation state of Ce.

Influence of corium temperature, concrete composition and water injection time on concrete ablation during MCCI: New insights

Molten corium, a mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural elements, can originate in a nuclear plant accident after a reactor core meltdown. This un-cooled corium could penetrate through the reactor pressure vessel and cause concrete ablation via basement melt-through, a process known as Molten Corium Concrete Interaction (MCCI). The MCCI analysis because of its complex nature is still uncertain and needs thorough investigation of various parameters. In this study the use of CORQUENCH simulator is presented to model the molten corium, composition of concrete and heat transfer along with related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH is successfully utilized enabling the modeling of interaction between molten corium and concrete. The developed model is validated against experimental data at PWR and BWR conditions. The results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete was found to be the main controlling factor to mitigate ablation. An alternative to concrete is to utilize igneous rock (pyrolite) and this approach could lead to comparatively very low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a cooling agent) into the reactor cavity should also be optimized to enhance corium quenching to avoid ablation via basement melt-through. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus and ablation temperatures, respectively.

Mechanical properties and thermal conductivity of (U,Zr)SiO4

During the core meltdown at the Fukushima Daiichi Nuclear Power Plant (1F), molten core materials penetrated the reactor pressure vessel (RPV), and they may have reacted with the concrete at the bottom of the primary containment vessel (PCV). Through this molten core concrete interaction (MCCI), uranium could potentially dissolve into ZrSiO4 and form a (U,Zr)SiO4 solid solution. For fuel debris retrieval to proceed smoothly, detailed information on the physical properties and critically characteristic of all possible fuel debris, including MCCI products, is necessary. However, no studies have reported on the physical properties of uranium in ZrSiO4 because it is difficult to synthesize. In this study, we fabricated dense (U,Zr)SiO4 samples through the spark plasma sintering (SPS) of SiO2 and (U,Zr)O2 and analyzed the solubility of uranium in ZrSiO4. The solubility limit of uranium in ZrSiO4 was found to be within 1.6 – 2.6 at%. The mechanical properties of the (U,Zr)SiO4 solid solution were calculated using the samples’ sound velocities. To understand the effect zirconium substitution by uranium has on heat transport, the thermal conductivity was calculated using the thermal diffusivity data measured with laser flash. This is the first study on the thermophysical properties of (U,Zr)SiO4, and it provides us with a better understanding of the properties of MCCI products.

Experiments of melt jet-breakup for agglomerated debris formation using a metallic melt

In severe accidents in a light water reactor, the relocated molten core (so-called corium or melt) can form a debris bed. The debris bed coolability is a critical issue for prevention and mitigation of the molten core–concrete interactions. Agglomeration has a serious impact on assessment of debris bed coolability if agglomeration forms massive debris (so-called agglomerated debris) by merging of melt particles with others when the melt particles accumulate on a floor. This paper presents the results of melt jet-breakup experiments for agglomerated debris formation using a simulant metallic melt. The experiments injected a melt jet of a low-melting point metal through a circular nozzle into a test section filled with coolant water. The particles were generated due to the melt jet-breakup accumulated on to a catcher, which is a flat plate made of stainless steel, installed in the test section. A high-speed video camera imaged particle formation and accumulation on the catcher plate. Agglomerated debris was confirmed by morphological investigation of the recovered debris. The experimental results revealed the effects of the melt jet injection conditions (melt temperature, coolant temperature, and coolant depth) on the mass fraction of agglomerated debris. On the basis of the experimental results, we proposed a simple correlation to estimate the mass fraction. The simple correlation successfully reproduced the mass fraction of agglomerated debris obtained in the DEFOR-A test [Kudinov et al., Nucl. Eng. Des., 301 (2013), 284–295]. The experimental data base presented in this paper makes further contributions to the modeling and validation of mechanistic models or simulation tools for agglomerated debris formation.

Experimental study on molten corium-concrete interaction with simulant metal and oxide

Under hypothetical Reactor Pressure Vessel (RPV) failure accidents in Light Water Reactor (LWR), Molten Corium-Concrete Interaction (MCCI) will cause erosion of cavity concrete, possibly resulting in containment failure due to basemat penetration and overpressure. The CINA (Corium-Concrete Interaction Apparatus) experiment was conducted to investigate the MCCI of metallic and oxidic corium with siliceous concrete. Simulant melt material metallic iron and oxidic alumina generated by exothermic thermite chemical reaction was used to investigate the two-dimensional erosion of a cylindrical crucible in CINA experiment. The crucible was fabricated from siliceous concrete with an inner diameter of 300 mm and a height of 500 mm containing reinforcement (rebars). Decay heat in the melt was simulated by subsequent sustained addition of thermite. During the experiment, the melt temperature was monitored by two thermocouples and a two-color optical endurance pyrometer. Besides, the phenomenon that appeared on the melt surface was recorded by the video camera placed above the crucible and the formation process of the crust anchoring was observed. After the experiment, the crucible was split in half by a wire saw to accurately measure the ablation depth. Axial-radial ablation depths were 25 mm and 13 mm, respectively. Current findings contribute to the further understanding of MCCI mechanisms and the optimization of MCCI mitigation strategies.

Experimental and numerical investigation on characteristics of MCCI with exothermic thermite

Investigation was conducted experimentally and numerically for ex-vessel melt behavior and concrete interaction during Molten Corium-Concrete Interaction (MCCI) in current research. The CINA (Corium-Concrete Interaction Apparatus) experiment was carried out with a total of 94 kg melt generated by exothermic thermite chemical reaction to react in a two-dimensional (2-D) cylindrical siliceous crucible with an inner diameter of 300 mm and a height of 500 mm. Meanwhile, the variation of melt temperature was monitored during the experiment. Besides, the accurate ablation depth and thickness of stratified layers were measured after the experiment. Numerical simulation was used to study the mechanism of high-temperature melt heat transfer and concrete ablation during the MCCI process. The slag film model to calculate the melt/concrete interface heat transfer coefficient and the stratified melt model were both used to simulate the above experiment. The results showed that the numerical results were in good agreement with experiment measurements by the slag film model under separated layers condition.

The influence of the water ingression and melt eruption model on the MELCOR code prediction of molten corium-concrete interaction in the APR-1400 reactor cavity

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ∼38% and ∼32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

Analyses of hydrogen risk in containment filtered venting system using MELCOR

Hydrogen risk in the containment filtered venting system (CFVS) vessel was analyzed, considering operation pressure and modes with the effect of PAR and accident scenarios. The CFVS is to depressurize the containment by venting the containment atmosphere through the filtering system. The CFVS could be subject to hydrogen risk due to the change of atmospheric conditions while the containment atmosphere passes through the CFVS. It was found that hydrogen risk increased as the CFVS opening pressure was set higher because more combustible gases generated by Molten Core Concrete Interaction flowed into the CFVS. Hydrogen risk was independent of operation modes and found only at the early phase of venting both for continuous and cyclic operation modes. With PAR, hydrogen risk appeared only at the 0.9 MPa opening pressure for Station Black-Out accidents. Without PAR, however, hydrogen risk appeared even with the CFVS opening set-point of 0.5 MPa. In a slow accident like SBO, hydrogen risk was more threatening than a fast accident like Large Break Loss-of-Coolant Accident. Through this study, it is recommended to set the CFVS opening pressure lower than 0.9 MPa and to operate it in the cyclic mode to keep the CFVS available as long as possible.

Numerical simulation on jet breakup in the fuel-coolant interaction using smoothed particle hydrodynamics

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.

Algorithms of three-dimensional concrete ablation front tracking (CAFT) and crust growth

A 3D MCCI code, CORCAAB, has been developed to account for uncertainty factors in the long-term molten corium–concrete interaction including changes in the three-dimensional cavity contour, which affect the convective heat transfer of the molten debris and late ablation behaviors. In this work, a 3D concrete ablation front tracking algorithm was formulated using the piecewise linear interface calculation method. The ablation and crust growth rates are governed by the local heat transfer across the 1D boundary layer that consists of molten debris, crust, slag, and concrete. This combination of the 3D and 1D methods led to the energy mismatch that was shown to be small enough even under a significantly large change of the cavity contour shape; thus, the proposed algorithm is compatible with the energy conservation law of the entire MCCI code. In order to test the efficacy and restriction of the proposed ablation algorithm, a simple rectangular cavity connected with a double sump on the cavity floor was adopted. Ablation surface formations including saddle points with multiple curvatures were confirmed to be adequate. A moderate dependence on the cell size was observed in predicting the disappearance of the thin walls separating the two partitions.

ESTIMATED CRITICALITY LOWER-LIMIT MULTIPLICATION FACTOR OF LOW-ENRICHED URANIUM DIOXIDE-CONCRETE SYSTEM USING THE BOOTSTRAP METHOD

The present paper aims to evaluate the estimated criticality lower-limit multiplication factor of fuel debris in a form of uranium dioxide-concrete mixture for a study of criticality control on the fuel debris generated through the molten core concrete interaction in a severe accident of a light water reactor. The estimated criticality lower-limit multiplication factor is evaluated using the bootstrap method where the assumption of the normal distribution is not necessary. In addition, it is calculated taking into account correlation coefficients that represent the degrees of neutronic similarity between the target system and benchmark critical experiment systems, experimental uncertainties of benchmark data, and statistical uncertainties of calculated effective multiplication factor by a continuous energy Monte Carlo code. This paper shows that the estimated criticality lower-limit multiplication factor using the bootstrap method can be comparable with a baseline upper-subcritical-limit which is evaluated by Whisper-1.1 without margins of subcriticality for uncertainties from nuclear covariance data and undetected errors in software.

Occurrence of highly radioactive microparticles in the seafloor sediment from the pacific coast 35 km northeast of the Fukushima Daiichi nuclear power plant

To understand the properties and significance of highly radioactive particles in the marine environment, we have examined seafloor sediment with a radioactivity of ∼1200 Bq/kg (dry weight; after decay correction to March 2011) collected 35 km northeast of the Fukushima Daiichi Nuclear Power Plant (FDNPP). Among the 697 highly radioactive particles separated from the sediment, two particles, D1-MAX and D1-MID, had a total Cs radioactivity of ∼56 and 0.67 Bq (after decay correction to March 2011), respectively. These particles were characterized with a variety of electron microscopic techniques, including transmission electron microscopy. The 134Cs/137Cs radioactivity ratio of D1-MAX, 1.04, was comparable to that calculated for Unit 2 or 3. D1-MAX consisted mainly of a Cs-rich microparticle (CsMP) with a silica glass matrix. The data clearly suggested that D1-MAX resulted from a molten core–concrete interaction during meltdowns. In contrast, D1-MID was an aggregate of plagioclase, quartz, anatase, and Fe-oxide nanoparticles as well as clay minerals, which had adsorbed soluble Cs. D1-MID was likely a terrestrial particle that had been transported by wind and/or ocean currents to a site 35 km from the FDNPP. The radioactive fractions of D1-MAX and D1-MID were 15% and 0.36%, respectively, of the total radioactivity in the bulk sediment. These highly radioactive particles have a great impact on the movement of radioactive Cs in the marine environment by carrying condensed Cs radioactivity with various colloidal and desorption properties depending on the host phase.