Accident Management Strategy Research Articles

Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

Threat of Combustible Mixture Formation in Small Rooms of an NPP’s Unit Containment during a Severe Accident

Protection of the containment integrity is one of the key objectives pursued in managing a severe accident at an NPP. Possible hydrogen inflammation in the containment can be regarded as the most dangerous threat for the containment integrity. Inert state of a steam–air–hydrogen mixture can be secured by maintaining a high steam concentration. With a steam concentration above 55%, the mixture is in an non-flammable state whatever the concentrations of its other components. The containment reinforced concrete walls lined with metal sheets behave as heat sinks with a very high thermal inertia. Heat removal through walls may have a significant influence on the mixture composition in the containment or in its individual part (especially if the rooms have comparatively small volumes) if part of steam from the mixture undergoes long-term condensation, as a result of which its concentration in the mixture decreases. The article presents an investigation, by means of detailed modeling, of the heat transfer processes through the walls of containment rooms with a view to reveal factors that are of importance in considering hydrogen challenge. It is found that the wall surface area to the room volume ratio is a factor having an effect on the formation of dangerous hydrogen-containing mixture compositions. It is shown that the wall heat absorption capacity decreases only slightly in the course of an accident. In rooms with a small volume, there may be a real risk that the medium in them may quite quickly lose its inertness as a result of heat removal through the walls. The possibility of a detonation mixture to form is not even excluded. In the case of a severe accident, there will be a risk of dangerous hydrogen-containing mixture compositions formation in small rooms within the containment irrespectively of the strategies implemented for the containment’s main rooms unless special measures and means are taken, e.g., application of accident management strategies or installation of hydrogen recombiners.

Prandtl number effect on thermal behavior in volumetrically heated pool in the high Rayleigh number region

The in-vessel retention by external reactor vessel cooling (IVR-ERVC) is a severe accident management strategy to hold or mitigate the molten radioactive material in the reactor pressure vessel (RPV). Highly turbulent natural convection is one of the most important phenomenon determining the thermal behavior of the molten pool. In the safety evaluation of IVR-ERVC, thermal behavior of the corium pool has been determined using experimental correlations. Property differences between the experimental simulant and the corium have been considered negligible. The effect of property differences represented by Prandtl number (Pr) on the thermal behavior of oxide pool is numerically investigated. The algebraic heat flux model (AFM) is used to simulate the turbulent natural convection and is validated with the BALI experiment. The Pr effect on the flow characteristics and the thermal behavior is investigated in the high Rayleigh number region. The present results show that the upward heat transfer tends to increase at the same Ra′ in the case of the molten pool which is lower Pr fluid than the simulant material.

Analysis of the effect of MELCOR modelling parameters on in-vessel accident progression in Nordic BWR

Nordic Boiling Water Reactors (BWRs) rely on the flooding of the lower drywell (LDW) as a severe accident management (SAM) strategy. The termination of a SA is achieved by fragmenting and quenching of the melt released from the vessel. Success of SAM strategy depends on melt release and water pool conditions. The characteristics of the melt release are the major source of uncertainty in quantification of the risk of SAM failure. Vessel failure and melt release modes are subject to aleatory and epistemic uncertainties at the in-vessel accident progression stage. In this work we focus on predicting the properties of debris relocated to the lower plenum using MELCOR code. We address the effect of epistemic uncertainty in modeling parameters and models in the MELCOR code in different severe accident scenarios on main characteristics of the in-vessel accident progression in Nordic BWRs. Sensitivity analysis is performed to rank the importance of MELCOR modelling parameters and the effect of different MELCOR models is addressed by using different versions of the code. The results provide valuable insights regarding the effect of MELCOR models, modelling parameters and sensitivity coefficients on code predictions.

Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

An analysis on the consequences of a severe accident initiated steam generator tube rupture

An analysis on the thermal hydraulic behavior, core damage progression and radiological releases of a steam generator tube rupture event initiated by a station black out event for APR1400 is provided in this paper. The analysis was performed by using MELCOR, where counter-current natural circulation, mixing in the steam generator inlet plenum, creep ruptures of reactor coolant system components and U-tube, and mitigation actions by the operator are modeled. A station black out accident with an opening of atmospheric steam dump valves (ADVs) at the time of severe accident management strategy entry condition with a failure of a supply of auxiliary feed water to the steam generators selected as a base case resulted in a creep rupture of U-tubes with a direct release of radioactive materials into the atmosphere. It was shown that a mitigation action by the operator for an aggressive cooldown of the reactor coolant system using ADVs and supplying the water to the steam generator even after the creep rupture of U-tubes was effective. It arrested core melt progression and significantly reduced the amount of radiological releases to the atmosphere.

Accident risk management strategy at un-signalized intersection

The increasing of traffic sign and/or light violations is a commonly views in many countries. Even it was occurred intentionally which indicates that drivers accept its possible consequences. The constant fatality index of motorcyclist strongly indicates that determinant variables behind their risky behaviour and/or accident involvement should be further investigated and managed systematically. Accordingly, this paper focuses on motorcyclist accident risk management, particularly at un-signalized segment, by combining the aggregated-individual and expert expectancy approaches. Therefore a questionnaire and braking manoeuver test were undertaken at closed circuit course. The result shows that speeding behaviour was caused by trip purpose and triggered by perception about their braking and hazard detection abilities. In addition, only 24 % of 56 % of riders who believed that their braking capability was above average could apply high braking capability so that most riders might involve in crash due to the average critical crossing gap choice at the monitored intersection. This explains that their speed choice should be deal with their braking capability. However, the overlay policy at the monitored intersection indicates that the unbalanced between mobility and safety still to be a latent issue, which virtually could be bridged by using a standardized braking capability.

Стратегии управления авариями при полной длительной потере электроснабжения на ядерных энергоустановках

Стратегии управления авариями при полной длительной потере электроснабжения на ядерных энергоустановках

The Influence of Crust Layer on Reactor Pressure Vessel Failure Under Pressurized Core Meltdown Accident

The so-called in-vessel retention (IVR) was considered as a severe accident management strategy and had been certified by Nuclear Regulatory Commission (NRC) in U.S. as a standard measure for severe accident management since 1996. In the core meltdown accident, the reactor pressure vessel (RPV) integrity should be ensured during the prescribed time of 72 h. However, in traditional concept of IVR, several factors that affect the RPV failure were not considered in the structural safety assessment, including the effect of corium crust on the RPV failure. Actually, the crust strength is of significant importance in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the lower head (LH) of the RPV. Consequently, the RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multilayered crust formation conditions with consideration of detailed thermal characteristics, such as high-temperature gradient across the wall thickness. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative RPV to figure out the possibility of high temperature induced failures with the effect of crust layer.

CHF correlation development under ERVC conditions by using the local liquid velocity from PIV measurements

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management strategy that removes the decay heat from the molten corium in the reactor vessel by flooding the reactor cavity. In ERVC conditions, the critical heat flux (CHF) on the outer wall of the vessel is a crucial factor that indicates a thermal margin of the system. Previous studies used averaged mass flux to predict CHF (Haramura and Katto, 1983; Lee and Mudawwar, 1988, Katto, 1990; Jeong et al., 2005; Park et al., 2013). The averaged value cannot indicate the degree of liquid supply which is the most critical factor in the CHF model. Local liquid velocity near the heated surface is one of the most crucial factors that should be quantified to develop a CHF model. In this study, the local liquid velocity near the surface has been measured. A curved rectangular flow channel was devised to simulate the gap between the external reactor vessel wall and the insulation. The boiling heat transfer and CHF were simulated with air injection at the inner wall of the test section. In the experiment, the flow field under the simulated ERVC conditions was measured using the particle image velocimetry (PIV) technique. The PIV measurement was validated by comparing the measured velocity data and the analytic solution under circular pipe flow. The liquid velocity profile at the middle plane of the curved rectangular test section was obtained in the multiple mass flux and inclination angle conditions. Based on the velocity data, a local velocity correlation was developed. The local velocity correlation was applied to improve the CHF prediction model.

Penentuan Prioritas Strategi Pengelolaan K3 Proyek Pemasangan Pipa Gas Menggunakan Pendekatan Metode AHP

The pipeline is an immobile transport medium that functions to distribute liquids and gases. The pipeline installation project is one of the Engineering Procurement and Construction (EPC) projects that has the potential for accident and occupational safety risks. The purpose of this research is to determine the priority of work accident management strategies gas pipeline by applying the AHP method. Brainstorming and pair-wise comparison approaches were conducted by experts. A total of three experts are used as key informants in assessing paired comparisons. The results show that there are four main factors that could potentially result in the risk of working accidents such as humans, machines, methods, and materials. Furthermore, based on the analysis result shows that human factor is the highest priority and environmental factor as the lowest priority. The AHP method application in this study indicates that this method can be used to determine the priority of each criterion and sub-criteria on the health and safety management strategy of the gas pipeline installation project. The implication of this in the application of AHP in other issues required more in-depth review related to the criteria and sub-criteria that are used to obtain more comprehensive results.

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

Analysis of Steam Explosion under Conditions of Partially Flooded Cavity and Submerged Reactor Vessel

A steam explosion in a reactor cavity makes a mechanical load of the pressure pulse, which can result in a failure of the containment isolation. To prove the integrity of the containment during the ex-vessel steam explosion, the effects of water conditions on a steam explosion have to be identified, and the impulse of a steam explosion has to be exactly assessed. In this study, the analyses for steam explosions were performed for the conditions of a partially flooded cavity and a submerged-vessel in a pressurized water reactor. The entry velocity of a corium jet for the scale of the test facility was varied to simulate the two plant conditions. The TEXAS-V code was used for simulating the phases of premixing and explosion, and the load of a steam explosion was estimated based on the pressure variation. The impulse of a steam explosion under the condition of a corium jet falling into water without a free-fall height is bigger than that under a free-fall height. The fragmented mass of corium in an explosion phase and the distribution of steam fraction are the main parameters for the total load of the steam explosion. This study is expected to contribute to analyses of a steam explosion for a severe accident management strategy.

Experimental and numerical investigation of molten corium behavior in lower head under external subcooling and boiling conditions

The in-vessel melt retention by flooding the reactor vessel externally is regarded as an effective severe accident management (SAM) strategy. According to this strategy, the corium will be stabilized within the lower head, by transferring the decay heat through the wall into the containment via external cooling. One key question of this strategy is how the melt pool heat transfer reacts to different external cooling conditions. In this paper, the melt’s thermal–hydraulic behavior under different external cooling conditions is studied experimentally in two LIVE tests performed in the frame of the LIVE program investigating late in-vessel melt pool behavior and calculated with the lower head module AIDA of ATHLET-CD. One LIVE test was performed under nucleate boiling condition, the other under sub-cooling condition. Melt temperature, heat flux along the curved vessel wall and the crust behavior are described in transient and steady states. The simulation results have been compared with the experimental results. The results have been demonstrated the applicability of ATHLET-CD to investigate the SAM strategy in-vessel melt retention by external cooling. Furthermore, on the basis of the experimental results the modelling of heat transfer between corium and coolant has been improved.

Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

ABSTRACT This study focuses on the in-vessel phase of severe accident management (SAM) strategy for a hypothetical 1000 MWe pressurized water reactor (PWR). To examine the effectiveness of SAM strategy, it is necessary to identify and assess epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is reactor coolant system (RCS) depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, sensitivity analyses are carried out to assess the impact of the cutoff porosity related to the flow area of core node (EPSCUT), the critical temperature (TCLMAX) and the minimum fraction of oxidized Zr (FZORUP) for cladding rupture, and the flag to divert gas flows in the core to the bypass channel (FGBYPA) on the core melting and relocation process. In this study, the effect of injection time on the integrity of RPV has been examined based on the quantification of the scenario and phenomenological uncertainties.

Prediction of in-vessel debris bed properties in BWR severe accident scenarios using MELCOR and neural networks

Severe accident management strategy in Nordic boiling water reactors (BWRs) employs ex-vessel corium debris coolability. In-vessel core degradation and relocation provide initial conditions for further accident progression. Outcomes of core relocation depend on the interplay between (i) accident scenarios, e.g. timing and characteristics of failure and recovery of safety systems and (ii) accident phenomena. Uncertainty analysis is necessary for comprehensive risk assessment. However, computational efficiency of system analysis codes such as MELCOR is one of the big obstacles.The goal of this work is to develop a computationally efficient surrogate model (SM) for prediction of main characteristics of corium debris in the vessel lower plenum of a Nordic BWR. The SM has been developed using artificial neural networks (ANNs). The networks were trained with a database of MELCOR solutions. The effect of the noisy data in the full model (FM) database was addressed by introducing scenario classification (grouping) according to the ranges of the output parameters. SMs using different number of scenario groups with/without weighting between predictions of different ANNs were compared. The obtained SM can be used for failure domain and failure probability analysis in the risk assessment framework for Nordic BWRs.

Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism.In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel.The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).

Modelling issues related to molten pool behaviour in case of In-Vessel Retention strategy

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000.The demonstration of IVR feasibility for high power reactors requires using less conservative models as the margins are reduced.In this paper modelling issues related to molten pool behaviour in case of IVR are highlighted based on sensitivity studies performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. In particular, the following physical processes are addressed:-Heat transfer coefficients in thin light metal layer and in heavy metal layer.-Kinetics of corium stratification and thermal exchanges between oxide and metal phases during inversion of stratification.-Relocation of steel coming from ablated vessel wall.-Characteristics of oxide crust.Analyses based on a simple configuration of corium in the lower head allow identifying potential impact of each modelling assumption on the heat flux distribution along the vessel wall, especially in transient situations. Based on this study, the most critical areas where the knowledge is too limited or where the current analytical methods should be improved are identified as being (i) the duration of situations with thin metal layer on the top of the oxide, associated to the knowledge of the kinetics of corium stratification and (ii) the possible super-heat of the metal layer at stratification inversion coupled with the risk of oxide crust failure.This work is performed in the frame of the European project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations. Its main objective is the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1000 MWe or higher.

Effects of orientation and structure geometry on boiling heat transfer for downward facing IGTAC surfaces

External Reactor Vessel Cooling has been a key severe accident management strategy to prevent the damage of the reactor vessel. Saturated pool boiling heat transfer coefficients and critical heat flux (CHF) were measured from the downward facing structured surface of interconnected-grooves with triangular array cavities (IGTAC surface) in deionized water to enhance the CHF. Four IGTAC surfaces were investigated at inclination angles of 5°, 30°, 45°, 60° and 90° (vertical). The experimental results indicated that the CHF increases with the increasing inclination angle. The highest CHF at inclination angle of 5° on downward facing IGTAC surface could reach 1258.4 kW/m2, which was over 2 times of that on a plain surface. The CHF on IGTAC surface did not appear at larger inclination angles when the heat flux reached 2400 kW/m2, which is the limit of the experimental apparatus. The IGTAC surfaces can form a liquid-vapor conversion path with the cavities as stable nucleation sites and the interconnected grooves as the cooling water supply pathways, which results in a significant boiling heat transfer enhancement and CHF increase. It is shown that deeper groove depth and larger cavity diameter will enhance the nucleate boiling heat transfer (NBHT) and CHF on IGTAC surfaces within the current dimension parameter range.

Modelling of debris bed reflooding in PEARL experimental facility with MC3D code

A hypothetical severe accident in a nuclear power plant has the potential for causing severe core damage, including a meltdown. To prevent or in the case of an already formed debris bed to limit the in-vessel core degradation, the basic severe accident management strategies consider the in-vessel reflooding to ensure the debris bed coolability.The purpose of our research was to understand the key processes and conditions related to the in-vessel debris bed coolability in the bottom reflooding conditions. Recently, experimental tests in the PEARL facility (IRSN, France) were performed to highlight the behaviour of the steam and water flow in a hot porous medium and to provide experimental data to validate 2-D and 3-D models for the debris bed reflooding. Our aim was to analyse chosen PEARL experiments performed at the atmospheric pressure. The objective was to analyse the importance of the uncertainties in the initial and boundary conditions on the simulation results and to assess the heat transfer modelling approaches. Simulations were performed using the MC3D code (IRSN, France).In general, the performed simulations are in good agreement with the experiments. The general features, in particular the water preferential entrainment in the bypass are recovered and the analysis of calculation gives further information on the mechanisms. In particular, the mechanism of water deviation in the bypass (2-D behaviour) is described. The hypothesis of water dragged by steam coming from the debris bed region cannot be supported. However, the simulation results are indicating a noticeable impact of the actual conditions as the water temperature and the initial support bed and bypass temperature. The simulations, varying the porosity of the test section, showed that this impact affects the flow configuration and is important for cases with the 2-D configuration. The reflooding capabilities in this configuration may depend strongly on the characteristics of the debris bed. Changes in the heat transfer modelling do not have greater effect on the simulation results.