Core Disruptive Accident Research Articles

Development of a source-term migration model for a large bubble formed in a core disruptive accident

Because sodium-cooled fast reactors are designed with high inherent safety in mind, the probability of a core disruptive accident (CDA) is extremely low. However, from a defense-in-depth perspective, the study of CDA sequences is still worthwhile to assure the safety and reliability of reactors. During a CDA, a large bubble rapidly expands inside the sodium pool, rises from the core, and covers the gas region, providing a potential migration path for source terms (radioactive materials present within the containment barriers). Source terms released initially within the cover-gas region after a few hundred milliseconds are called instantaneous source terms. We propose here an instantaneous source-term migration model that provides a simplified evaluation of the amount of source terms absorbed by coolant sodium during the ascent of the CDA bubble. In the model, the particle motion within the CDA bubble obtained from the basic momentum equation is used to calculate the amount of source terms escaping from the bubble interface. In addition, a model analogous to aerosol scavenging by precipitation is used to assess the amount of source terms absorbed by droplets present in the bubble, especially the entrained sodium droplets that form during rapid expansion of the CDA bubble. The model is further validated by a previous source term migration experiment in which a large high-pressure bubble expands and rises in a sodium pool. Good agreement with the measured retention factor of a source term demonstrates the reliability of the developed model. Given these results, some key parameters are selected for a sensitivity analysis.

Development of a numerical model for disassembly phase of hypothetical core disruptive accident and application to a medium-sized mixed oxide fuelled fast reactor

Safety assessment of fast reactors involves analyzing Hypothetical Core Disruptive Accidents (HCDA) that could result in reactor core disassembly. HCDA is generally analyzed in three distinct phases that include pre-disassembly, disassembly and post-disassembly phases. In the present work, a new coupled neutronics-hydrodynamics computer code called FAst Reactor DISassembly (FARDIS-1) is developed in 2-D cylindrical coordinate geometry to model the super-prompt critical power excursion during the disassembly phase of HCDA in a sodium-cooled fast reactor (SFR). The modelling approach followed is essentially similar to the one used in VENUS-II code but with a few improvements. The code is benchmarked against the predictions of VENUS-II for a sample case of HCDA in the FFTF reactor model. This is followed by an analysis of the disassembly phase in a medium-sized (500 MWe) mixed oxide fuelled fast reactor. Two cases are assumed at the beginning of the disassembly phase: a sodium voided core (conservative) and a partially voided core (realistic) and the influence of several core neutronics parameters on the transients is elucidated through parametric studies. The peak power, thermal energy release, peak pressure, maximum temperature and work-potential are estimated and their trends are discussed.

Development of a simplified one-dimensional CDA bubble model

Sodium-cooled fast reactors (SFRs) have high safety with an extremely low probability of a core disruptive accident (CDA). However, from a defense-in-depth perspective, the CDA sequence is still worth studying. Severe accidents in SFRs, such as unprotected loss of flow, might lead to a CDA during which fuel and some fission products could be instantly released from a large CDA bubble through potential leak paths in the top shield structure. Therefore, a reasonable prediction of dynamic behavior of a large-scale bubble within the sodium pool is vital for accurately evaluating the migration of source terms. In this study, we propose a simplified one-dimensional CDA bubble model that can handle heat and mass transfer in two-phase multicomponent materials in different computational domains during the rising of bubbles through the sodium pool toward the cover-gas region. In this model, an entrainment model based on the Rayleigh–Taylor instability and Kelvin–Helmholtz instability was used to explain the coolant entrainment through the gas/liquid boundary and jet fragmentation. The model also addresses the mitigation effect of non-condensable gases on the condensation of fuel, steel and sodium vapor at bubble interface. We evaluated the model using a past experiment on the expansion of a two-phase, large bubble with high pressure in a stagnant liquid pool conducted by Purdue University in the late 1970 s using a 1/7-scale model of Clinch River Breeder Reactor. Good agreement with the experimental data demonstrates that the developed model can reasonably represent the essential characteristics of dynamic behavior of a large, high-pressure bubble with heat and mass transfer in two-phase multicomponent materials. This is valuable for evaluating the migration of the source terms, which will be carried out in future studies.

From debris bed formation to self-leveling behaviors in core disruptive accident of sodium-cooled fast reactor

Understanding the Core Disruptive Accident (CDA) is crucial for the safety analyses of Sodium-cooled Fast Reactors (SFR), despite its exceedingly low possibility. During a CDA, the potential release and subsequent relocation of molten materials from the core could result in the formation of a debris bed inside the reactor vessel due to their quenching and fragmentation from the interactions with the subcooled sodium coolant. Debris beds with varying shapes are expected to form due to the different flow-regime characteristics during their formation. Further, the heat decay from this high-temperature debris might trigger sodium coolant boiling, instigating the debris bed self-leveling behavior, which flattens the debris beds. Considering the importance of the debris bed shapes to the debris transfer, heat removal and neutronic criticality, lots of knowledge about the mechanisms and characteristics of debris bed formation and self-leveling behaviors has been gathered in recent years through extensive experimental, modeling and numerical outcomes. This study aims to systematically and critically summarize these past explorations into the debris bed formation and self-leveling behaviors, offering a beneficial direction for future research on the improved safety evaluations in SFR severe accidents.

Analysis method for aerosol deposition behavior under sodium spray fire

Sodium spray fire is a significant containment-related safety concern in sodium-cooled fast reactors (SFR). In a highly-energetic accident such as a core disruptive accident (CDA), the sodium spray fire not only poses a threat to containment integrity but also causes the generation and release of sodium aerosol source terms with strong radioactivity and chemical hazards. In our method, the sodium aerosol generation model, log-normal size distribution model, agglomeration models, and deposition models were applied. In particular, different turbulent energy dissipation rates were considered in the agglomeration models according to the different thermal-hydraulic states of containment and confinement, and the simplified Crump model and Beresnev models have been selected to describe the gravity-diffusion and the thermophoresis deposition behavior respectively, due to the specific conditions of sodium spray fires. Simulation studies have been divided into two categories based on a potential design concept: HTCA, AB5, and AB6 experiments were introduced to describe the drastic thermodynamic state variations caused by spray fires in the confinement, while ATF and AB7 experiments characterized a relatively stable environment similar to the containment. From the simulation results, the analysis method can be considered reasonable with respect to experimental values. Moreover, calculations have been performed for a hypothetical sodium spray fire caused by a rotator plug failure case in a paper-based European SFR.

Numerical analysis of sodium-argon two-phase flow in narrow channels under severe accident conditions in pool type SFR

A two-phase flow model for sodium-argon mixture flow through the reactor vessel head penetrations under the energetic core disruptive accident in a sodium fast reactor (SFR) is presented. The model evaluates the total sodium mass in sodium-argon mixture released to the reactor containment building (RCB) through penetrations. The driving pressure at the penetration entrance is given as input. The two-phase frictional and local pressure drops, inertial and piezo-metric heads are accounted for in the model. Depending on the argon volume fraction, either a homogenous or separated flow approach is adopted to evaluate the mixture flow. Reactor-scale analysis show that the presence of argon in vessel head penetrations can significantly mitigate the total sodium release mass during the slug impact phase. Also, the two-phase frictional pressure drop plays a mitigating role and needs to be included in the model formulation to avoid overestimation of the sodium release mass. Fluid-structure interaction effects such as vessel head bending and hold-down bolt elongation have negligible influence on sodium release mass as long as the deformations are within the elastic limit. The model is conceived to be part of an integrated code system to evaluate the sodium fire and the maximum RCB pressure under severe accident conditions in a pool-type SFR.

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

ABSTRACT Boron carbide (B4C)–stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using B4C pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid B4C with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid B4C–liquid SS reaction based on the reduced thickness of B4C pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid B4C–solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid B4C–liquid SS reaction, it was found that similar temperature dependency was identified between solid B4C–liquid SS and solid B4C–solid SS. Besides, the reaction rate constants of solid B4C–liquid SS were smaller than those of solid B4C–solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for B4C side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid–solid contact for B4C side/SS side.

Effect of coolant boiling induced by accumulated debris on flow-regime characteristics of debris bed formation behavior for sodium-cooled fast reactor: Experimental and modeling study with gas injection

Understanding the Debris Bed Formation (DBF) behavior is important to the optimized integral evaluation for Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactor (SFR). Aimed at clarifying the characteristics of flow regimes during DBF, extensive experimental works were carried out by releasing solid particles into a water pool. According to the preliminary studies using bottom heating and nitrogen gas injection, it was found that the effect of coolant boiling should be crucial to the flow-regime transition. However, the gas flow rate considered in these studies (0–50 L/min) was not adequate wide to cover the range of vapor flow rate from the possible drastic boiling in actual CDA of SFR. Therefore, in current study, to elucidate the influence of coolant boiling on DBF flow-regime characteristics more comprehensively and provide more extensive data for supporting the predictive model development, a greater number of experiments are conducted with a much-enlarged range of gas flow rates (up to 100 L/min) under varying parametric conditions (including particle size, particle type, water depth and release pipe diameter). Through analyzing the flow-regime transition, an extended functional form for the effect of coolant boiling is successfully proposed. By incorporating this form into our base model, the model predictability can be extended for the DBF at the conditions of coolant boiling induced by accumulated debris in accordance with the verification of the regime map. Present study is believed to be valuable for further developing and verifying the models and codes for SFR safety analysis in China.

Numerical study of debris bed formation behavior for severe accident in sodium-cooled fast reactor by using least square MPS-DEM method

During a Core Disruptive Accident (CDA) in a Sodium-cooled Fast Reactor (SFR), debris beds may form in the lower plenum of the reactor vessel as a result of the accumulation of discharged debris, which is formed from the solidification and fragmentation of core molten materials during their release due to interactions with the sodium coolant. Since the configuration of debris beds plays a vital role in their long-term coolability and neutronic subcriticality for achieving in-vessel retention, it is essential to understand and assess the Debris Bed Formation Behavior (DBFB). In this study, an advanced Moving Particle Semi-implicit (MPS) method, called least square MPS (LSMPS), is utilized and coupled with the Discrete Element Method (DEM) for the numerical investigations of DBFB characteristics in a two-dimensional liquid pool. The applicability and reliability of the proposed LSMPS-DEM method are validated by comparing simulation results with experiments using simulant materials. The LSMPS-DEM method is then employed to study DBFB in sodium. It is observed that debris properties significantly impact the sedimentation and accumulation processes of DBFB, leading to variations in the characteristics of the formed debris beds. The present study is expected to be useful for further investigations of DBFB under more realistic conditions and for the design improvement of safety devices to mitigate the consequences of CDA in SFR.

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

ABSTRACT The typical initiating events of a core disruptive accident (CDA) in small sodium-cooled fast reactors (SFR) are evaluated with the computational code, SAS4A. CDA is one of the hypothetical events in the safety assessment of the SFR, as the SFR has the potential to experience power excursion due to the sodium voiding and the core degradation. To improve the safety of future SFRs, the development of SFRs with low void reactivity has been promoted. Small SFRs can have a negative void coefficient of reactivity. The analysis of the CDA event sequence in small SFRs is valuable for the investigation of the reactor characteristics for the future R&D of SFRs. The event progression during ULOF and UTOP, which are the typical initiating events of CDA, is investigated through the numerical analysis. The event progression of these accidents in the low void reactivity reactor is found to be slow due to the effective insertion of the negative reactivity feedback and the absence of significant positive reactivity insertion. No power excursion occurs in the initiating phase. In ULOF, the cladding melt and motion behavior becomes more important for the evaluation of the event progression due to its positive reactivity.

The Effect of Fuel Aerosol Particle Size on Aerosol Release Potential in Energetic Core Disruptive Accident Studies

In nuclear reactor accident safety studies, the radiological source term is a metric that quantifies the release of radiological material from the reactor to the environment. The present work evaluates heat transfer between high-temperature vapor bubbles and the surrounding coolant and the effect these interactions have on the source term for postulated core disruptive accident scenarios associated with an oxide-fueled, liquid metal–cooled fast reactor class. It is shown that aerosol particle size can influence heat transfer, and it is suggested that the extent of the influence depends on the fineness of the particles in the aerosol. The results are consistent with legacy experiments conducted in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory and offer a more comprehensive assessment of vapor condensation by treating the bubble constituents, in the context of radiation heat transfer, as participating media. The model, which couples classical scattering theory to the equation of radiative transfer and the energy equation, provides a means for estimating size-affected radiative cooling times. Solutions are obtained via the P-1 method of spherical harmonics with improved, higher-order boundary conditions. Outcomes include the development of an “extinction-time ratio” criterion for assessing whether ejection of aerosol from the bubble to the cover region is likely. Aerosol release from the coolant pool is evaluated using this criterion with the potential to extend this work to reactor-scale accidents. A baseline evaluation is provided that shows that omission of participatory effects could lead, in a relative sense, to cooling time offsets in excess of 14%. In addition to enhancing previous evaluations of FAST results, these modeling outcomes contribute to knowledge management efforts aimed at developing a more mechanistic assessment of the source term while suggesting potential enhancements to severe accident safety analysis through the use of more comprehensive radiative heat transfer models.

Development of new treatment of fuel isotope vector in the core degradation accident analysis of fast breading reactor

ABSTRACT In a design of future Sodium-cooled Fast Reactor (SFR), there is a demand for evaluation of event sequences and consequences of core disruptive accidents. Future SFRs include a unique core design with axially or horizontally heterogeneous core arrangement having complex fuel isotope distribution. A new model to flexibly represent fuel isotope distribution, called as the Pu-vector model, has been developed in this study for inclusion in the SIMMER-III and SIMMER-IV codes (simply called as SIMMER). The model calculates movement of individual fuel isotopes, assuming they always accompany the convecting fuel in the fluid-dynamics model. The accuracy of the Pu-vector model was confirmed by comparing with the standard Monte Carlo static neutronics calculation. The new model can improve some of the limitations in the current SIMMER code, in which the fuel isotopes are represented only by two groups, fertile and fissile fuels. Assignment of a number of fuel isotopes to the two groups requires a detailed examination of different combinations of fuel isotopes to determine an optimized combination. The Pu-vector model can eliminate this complicated procedure to be performed prior to a SIMMER analysis, and more importantly provides accurate spatial distribution of nuclide isotopes and thus will improve the applicability of SIMMER to the analyses of future large heterogeneous reactors.

Spatio-temporal evolution of sodium aerosol and temperature distribution during sodium pool fire in a confined environment

In case of a Core Disruptive Accident in sodium-cooled fast reactor, sodium will be ejected from the reactor vessel into the Reactor Containment Building (RCB) and results in sodium fire. Liquid sodium combustion generates aerosol, which gets dispersed and remains suspended in RCB based on geometry of the building and degree of turbulence. The Environment Source Term (EST) primarily depends on the dynamics of sodium aerosols inside the containment. Hence, it is essential to carry out a safety analysis of SFR using computational tools and models shall be validated with experimental data, particularly aerosol dynamics inside the containment along with thermal hydraulics. Besides, all codes assume uniform aerosol concentration in large containment buildings and the evolution of suspended concentration was predicted. However, in actual case a non-uniform concentration profile exists, which leads to stratification of aerosol concentration inside the containment. Towards assessment of aerosol dynamics in realistic scenario, spatial–temporal characteristics of sodium aerosol and temperature distribution were studied by conducting sodium pool fire experiments in a large containment.The sodium fire experiment was conducted by using ∼2.0 kg of sodium in a pool area of ∼0.25 m2. The measured maximum pool and gas temperatures are 685 °C and 126 °C respectively, and the temperature gradient at the wall boundary layer was 7.5 °C/cm. The estimated average sodium burning rate and aerosol release rate are 17.3 kg/m2hr and 0.24 g/s, respectively. The spatial variation of aerosol characteristics was measured for both horizontal and vertical planes. The measured maximum aerosol concentration in the horizontal plane ranges from 2.76 to 5.32, 2.16–3.30 and 2.09–4.21 g/m3 for the bottom, middle and top elevations respectively. For the vertical plane, aerosol concentration varied from 2.16 to 3.65 g/m3 for central positions and 2.53–5.32 g/m3 for off-center positions. The trend of average suspended aerosol concentration at each elevation showed that the concentration was higher at the top elevation than at the middle and bottom for up to 15 min i.e., during aerosol generation period and after that concentration was found to be higher at the bottom elevation for the next 25 min. Based on the measurements, it is observed that the aerosol distribution is not well mixed within the experimental period. A significant difference in aerosol mass concentration was observed between each elevation, between estimation and observations based on a zero-dimensional model. The evolutions of aerosol median size by theoretical simulation capture the overall trend reasonably.

Characteristics and Mechanisms of Debris Bed Formation Behavior in Severe Accidents of Sodium-Cooled Fast Reactors: Experimental and Modeling Studies

A Sodium-cooled Fast Reactor (SFR) is one of the optimized candidates in Generation IV nuclear reactor systems, but safety is an essential issue for SFR development and application. Most knowledge was accumulated through SFR safety investigations, especially for Core Disruptive Accidents (CDAs). During the CDA of SFRs, the molten materials in the core region are likely to discharge into subcooled sodium and form a debris bed on the lower region of the reactor vessel. Noticing that elaboration on the characteristics and mechanisms of Debris Bed Formation (DBF) behavior should be essential for the subsequent analysis of debris bed coolability and accident progression through various experimental and modeling studies, much knowledge was obtained during the past decades. Motivated to promote future investigations on CDAs of SFRs, the previous experiments and modeling studies on DBF behavior are systematically reviewed and discussed in this paper. The experimental results showed that the flow-regime and accumulated-bed characteristics during DBF were influenced by varying parameters and realistic conditions. Through the modeling studies, several empirical models were proposed for predicting the flow regime and accumulated-bed characteristics in DBF. In addition, to promote further development of research, the future prospects concerning DBF behavior are also described.

Experimental study on design optimization of the core catcher for the CEFR

ABSTRACT In severe accidents, sodium-cooled fast reactors (SFRs) are more concerned with core disruptive accidents (CDAs), and in-vessel retention (IVR) is an important mitigation measure to ensure the integrity of the reactor vessel. The distribution and accumulation uniformity of the melt in the lower chamber seriously affects the residual heat discharge capability. The installation of a core catcher is an effective mitigation method, and its structure and arrangement play a key role in mitigating severe accidents. In this study, the experimental facility is designed based on the core catcher of the China Experimental Fast Reactor (CEFR). An experimental platform is established to simulate the falling and relocation of melts in the core catcher under the condition of CDAs. At the same time, a quantitative evaluation index is established to analyze the influence of the number of chimneys, the inclination angle of the chimney canopy, the central chimney height, and the spacing of chimneys on relocation of melts in the core catcher. Finally, some suggestions for correlation optimization design are given.

X線イメージングを用いた球充填層内のボイド率分布計測

In the safety study on a sodium-cooled fast reactor, it is assumed that the high-temperature fuel debris are formed because of a core disruptive accident. During the cooling of the fuel debris, gas-liquid two-phase flow can be generated in the debris due to the coolant boiling. To improve the prediction accuracy of the flow characteristics, detailed measurement of the gas-liquid two-phase flow in the debris bed is required. In this study, gas-liquid two-phase flow in a quasi-two-dimensional sphere-packed bed, which simulates the debris bed, is visualized by using X-ray imaging. The experimental results show that the local void fraction increases near the splitting section in the packed bed and decreases near the coalescence section.

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications

Abstract A new subassembly type passive reactor shutdown device is proposed to expand the diversity and robustness of core disruptive accident prevention measures for sodium-cooled fast reactors (SFRs). The device contains pins with a fuel material that is in a solid state during normal operation but melts and fluidizes during an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident. By rapidly transferring the liquefied device fuel into the lower plenum region of the pins via gravitation alone, the device passively provides high negative reactivity to the core. This study evaluated the nuclear and thermal properties of the device subassembly with metallic fuel to determine the device specifications for proper device operation during ULOF and UTOP accidents. The results of the transient analysis of the ULOF initiating phase in a 750-MWel-class mixed-oxide-fueled SFR core confirmed that a conventional homogeneous core maintains stable cooling of the core before coolant boiling in the driver fuel subassemblies. In contrast, the negative reactivity required to terminate the event by device operation was slightly higher in the low sodium void reactivity core than in the conventional homogeneous core.

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Preliminary Study

AbstractFollowing the Fukushima Nuclear Power Plant accident in 2011, it has become increasingly important for reactor safety designs to consider measures that can prevent the occurrence of severe accidents. This report proposes a novel subassembly-type passive reactor shutdown device that expands the diversity and robustness of core disruptive accident (CDA) prevention strategies for sodium-cooled fast reactors. The developed device contains pins with a fuel material that is in the solid state during normal operation but melts into a liquid when the temperature exceeds a certain value (i.e., during a potential accident). When an unprotected loss of flow (ULOF) or unprotected transient overpower (UTOP) accident occurs, the device can passively provide significant negative reactivity by rapidly transferring liquefied device fuel into the lower plenum region of the pins via gravitation alone. The reactors containing some of the proposed devices in place of original fuel subassemblies become subcritical before the driver fuels are damaged, even if ULOF or UTOP transient events occur. The present study evaluates candidate materials for device fuels (e.g., metallic alloy, chloride), optimal device pin structures for liquefied fuel relocation, and nuclear and thermal-hydraulic characteristics of the device-loaded core under accident conditions to demonstrate the engineering applicability of the proposed device. This report discusses preliminary results regarding the nuclear requirements for inducing negative reactivity to achieve reactor shutdown under the expected device conditions during an accident.

An improved multiphase SPH algorithm with kernel gradient correction for modelling fuel–coolant interaction

Fuel–coolant interaction (FCI) has a pivotal role in the development of core disruptive accident in a sodium-cooled fast reactor. The drastic deformation of multiphase interface in the FCI is hard to deal with in the traditional grid method. In this paper, an improved multiphase smoothed particle hydrodynamics (SPH) algorithm corrected with the kernel gradient correction (KGC) technique is presented for multiphase flow with a large density ratio and complex interfacial behaviors. The density discontinuity across the multiphase interface is described with the use of special volume, which only depends on the position information of adjacent particles. This multiphase SPH algorithm, which is troubled by an unstable and mixed-interface problem under a large density ratio, is significantly improved with the KGC technique. The accuracy and robustness of the improved method are demonstrated in the numerical simulations of the deformation of a square droplet, Rayleigh–Taylor instability, and bubble rising in water. The verified corrected multiphase SPH is applied to simulate the hydrodynamic behaviors of the general FCI and the interaction between fuels coated with stainless steel film and coolant. Boundary layer stripping, Rayleigh–Taylor instability, and Kelvin–Helmholtz instability are observed as important mechanisms of hydraulic fracturing. The fragments’ distribution and the influence of stainless steel film are analyzed. The existence of a stainless steel film has been shown to inhibit breakage.

THEFIS Test Simulation to Validate a Freezing Model of ASTERIA-SFR Core Disruptive Accident Analysis Code

The mechanical consequences of core disruptive accidents (CDAs) are a major safety concern in sodium-cooled fast reactors. Once core disruption occurs, liquefied core materials rapidly disperse vertically and radially. The dispersed materials penetrate the pin bundles and control rod guide tubes (CRGTs) before freezing at the edge of the penetration zone as heat is transferred to surrounding structures. Such freezing phenomena can suppress the negative reactivity feedback of fuel dispersion. The discharge of core materials can be impeded, resulting in a molten core pool formation when tight blockages occur inside CRGTs due to frozen material. Accordingly, freezing phenomena of core materials play a key role in governing the mechanical consequences of a CDA. To validate a freezing model implemented in our CDA analysis code, ASTERIA-SFR, a preliminary simulation of the THEFIS RUN#1 test, was performed. The calculation results show that freezing on the structural wall and crust formation were key phenomena affecting the penetration behavior, and the structural heat transfer is an important parameter. A remarkable reduction of the heat transfer coefficient was required to reproduce the penetration length observed in the experiment. This suggests that the momentum exchange and flow regime at the leading edge as well as heat transfer should be well modeled to predict the freezing phenomena in rapidly evolving CDAs.