Corium Coolability Research Articles

A scoping investigation on debris bed formation with high-temperature melt simulant Fe-Sn

In hypothetical severe accidents of Nordic boiling water reactors (BWRs), the core melt (corium) has the potential to relocate to the lower head, leading to the failure of the reactor pressure vessel (RPV). Consequently, the corium, a molten mixture of either UO2-ZrO2 or Fe-Zr, is anticipated to undergo fuel coolant interactions (FCI) in a deep-water pool within the reactor cavity which is supposed to be flooded during severe accidents. The characteristics of the resulting debris bed play an important role in corium coolability that is paramount to accident stabilization and termination. This study is dedicated to characterizing the debris bed resulting from the FCI using Fe-Sn as the simulant of Fe-Zr. Compared with other simulants in our previous studies, the Fe-Sn alloy has a a higher melting point of 1130 °C and the same element of Fe as in prototypical Fe-Zr. The Fe-Sn is melted in a clay graphite crucible with glass coating layer to minimize contamination of molten materials by graphite. Argon containing 1.26 % hydrogen is employed as cover gas to avoid the oxidation of Fe-Sn at high temperature. The process of debris bed formation, including melt jet breakup, debris fragmentation, and fragments’ sedimentation on the pool bottom, is meticulously recorded using high-speed cameras. The porosity of the debris bed is determined in the post-test analyses using a three-dimensional laser scanner and a graduated water tank. The laser scanner data are also used to reconstruct the geometric appearance of the final debris bed. The debris particulates are sieved to obtain their size distribution. The chemical interaction between water and melt is examined by the SEM analysis. The experimental results show that the debris bed of Fe-Sn has significant differences from those of low melting point simulants (Sn-Bi, Sn, Zn) in our previous studies.

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

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

Development of a formulation to predict molten core spreading in an LWR severe accident

Relocated corium into the Primary Containment Vessel (PCV) needs to be properly cooled to avoid or mitigate molten-core concrete interactions (MCCI) in the PCV in order to maintain its supporting capability for the reactor pressure vessel and to suppress combustible/non-condensable gas releases. To know how effective the cooling is, it became important to know the geometry of the relocated corium. The present study chooses to focus on the “Wet Cavity” strategy and to build a reliable tool to evaluate the corium coolability in such a case. To achieve this goal, a previously developed formulation built to predict the corium geometry under the “Dry Cavity” strategy was extended to the conditions used in the “Wet Cavity” strategy. This extension includes the effect of solidification and cooling from the water by using a newly developed expression for the dimensionless thickness s*, the water subcooling, and the melt’s super-heat. After the validation of the extended formulation was confirmed, potential restrictions and limitations were investigated.

Development of a semi-empirical model-based COrium COolability analysis tool (COCOA) validated against a large scale corium experiment, FARO

As molten corium is released from the reactor pressure vessel (RPV) under severe accidents, the molten corium is going to experience various complicated phenomena. If there is no steam explosion, the corium is going to be ejected through the reactor vessel, fragmented, and cooled through the interaction between the corium and water existing in the cavity during its falling stage.Our research group has recently developed a COrium COolability Analysis tool named COCOA. COCOA is the transient calculation tool for the evaluation of the phenomena from the molten corium release to debris formation through the corium falling stage of FCI. COCOA was corporately developed to calculate the transient results of the pressure, the released energy, the average water temperature, fragmented particle size distribution, particulate debris fraction etc., during the corium falling stage of the FCI phase.COCOA is validated against FARO, which is a large-scale corium experiment. Comparing to COMETA, which is a 2D severe accident analysis code, COCOA showed better accuracies on the prediction of the pressure, the released energy, and the particulate debris fraction during the corium falling stage of the FCI phase. COCOA gives 16.9% and 26.5% of NRMSD for the pressure and the released energy, while COMETA does 42.2% and 36.0% of NRMSD. Also, COCOA and COMETA predict the particulate debris fraction with 27.22% and 39.25% of RMSE, respectively. Since COCOA is a semi-empirical model-based simple 1-D tool, only small computational effort is required for the calculation.

In-Vessel Retention of PHWRs: Experiments at Prototypic Temperatures

Abstract In pressurized heavy water reactors (PHWRs), multiple failures of engineered safety features may cause a failure of core cooling eventually leading to core collapse. The failed fuel and fuel channels relocate to the bottom of the calandria vessel (CV) and form a terminal debris bed, which generates decay heat. With time, the moderator evaporates and the terminal debris bed ultimately melts and forms a molten pool of corium. If corium breaches the CV and enters the calandria vault, large amounts of hydrogen and other fission gases may be generated due to molten core concrete interaction, which may pressurize the containment leading to containment failure. In addition, the passive catalytic recombiner devices may be incapable of managing such large amounts of hydrogen. Hence, in-vessel retention of corium is the only option to the avert progression of the accident. The heat removal capability of the CV needs to be demonstrated in order to attain the goal of in-vessel retention, to contain the corium during severe accidents. A lot of numerical analysis of heat removal capability of the CV has been done. However, experimental demonstration of in-vessel retention has been rarely presented in the literature, especially for PHWRs. In this paper, in-vessel retention at prototypic temperatures has been presented. Experiments have been carried out in scaled CVs. Different corium simulants have been used at elevated temperatures and corium coolability has been demonstrated.

Modelling and capability of severe accident simulation code, AZORES to analyze In-Vessel Retention for a loop-type sodium-cooled fast reactor

Research and development of the physical models and codes of safety analyses have been promoted for loop-type Sodium-cooled Fast Reactors (SFRs). A severe accident code, AZORES has been developing for the evaluation of molten corium behavior and integrity of reactor and containment vessels. AZORES enables the simulation of plant equipment behavior under severe accident conditions. In recent years, much attention regarding the mitigation of severe accident progression has been attracted to In-Vessel Retention (IVR) of SFRs, which has relatively less existing experimental and analysis results than that of the Light Water Reactor (LWR). This paper describes the status of development for main physical models and preliminary verification for IVR simulation mainly with AZORES. For a verification process, the simulation has been conducted for representative Post-Accident Material Relocation (PAMR)/Post-Accident Heat Removal (PAHR) phases subsequent to an Anticipated Transients Without Scram (ATWS) to evaluate the impact of initial conditions on severe accident progression. From the study, it was found that the mass of molten corium and capability of decay heat removal were key parameters for the attainment of IVR. Furthermore, this paper describes the results of Computational Fluid Dynamics (CFD) simulations which have been conducted for molten corium relocation in a control rod assembly and for corium coolability with sodium coolant on the core catcher for the purpose of enhancing the related AZORES physical models.

Numerical Modeling and Experimental Validation of Melt Pool Coolability Under Bottom Flooding with Decay Heat Simulation

The nuclear accident at Fukushima has brought an increased research focus on nuclear safety and severe accidents. Radioactivity leakage into the environment has environmental, societal, and political impact, and further robustness of nuclear reactor design is essential. Thus, coolability and stabilization of corium within reactor containment in a severe accident scenario are important issues that need to be resolved. In this context, many new reactors have been envisaged with dedicated core catchers. In ex-vessel core catchers, corium coolability is one of the biggest concerns. Despite several efforts, melt pool coolability still needs to be understood in sufficient detail. Among the various cooling strategies, melt coolability using bottom flooding has been demonstrated to be one of the most efficient techniques so far. This paper presents the numerical and experimental study of melt pool coolability under bottom flooding with decay heat simulation. An experiment has been performed using a test section consisting of two parts: a lower part for melt retention and quenching, and an upper part for steam expansion and its outlet. To simulate the decay heat, ten radiative heaters equivalent to 10 kW were used to heat the lower part of the test section housing the melt. The experimental measurements showed that quenching of about 25 L of melt at 1200°C initial temperature took only a few minutes. The same phenomenon has been modeled using a mechanistic model. The model considers heat transfer in the melt pool, eruption in the melt pool by steam formed when water is injected at the bottom of the melt pool, and heat transfer from porous debris to a steam mixture. The model postulates the formation of crust below the melt pool when water is being inserted from the bottom. The model predicts the failure of this crust due to various stresses, resulting in an inverted cone–shaped melt eruption. The model captures the physics of this eruption cone along with the spatial variation in porosity. The model predictions have been compared with the measurements for the melt pool temperature during the cooling process. The results show that the model is able to capture quite accurately the multidimensional temperature fields in the melt pool during the cooling process.

Main modelling features of the ASTEC V2.1 major version

A new major version of the European severe accident integral code ASTEC, developed by IRSN with some GRS support, was delivered in November 2015 to the ASTEC worldwide community.Main modelling features of this V2.1 version are summarised in this paper. In particular, the in-vessel coupling technique between the reactor coolant system thermal-hydraulics module and the core degradation module has been strongly re-engineered to remove some well-known weaknesses of the former V2.0 series. The V2.1 version also includes new core degradation models specifically addressing BWR and PHWR reactor types, as well as several other physical modelling improvements, notably on reflooding of severely damaged cores, Zircaloy oxidation under air atmosphere, corium coolability during corium concrete interaction and source term evaluation.Moreover, this V2.1 version constitutes the back-bone of the CESAM FP7 project, which final objective is to further improve ASTEC for use in Severe Accident Management analysis of the Gen.II–III nuclear power plants presently under operation or foreseen in near future in Europe. As part of this European project, IRSN efforts to continuously improve both code numerical robustness and computing performances at plant scale as well as users’ tools are being intensified.Besides, ASTEC will continue capitalising the whole knowledge on severe accidents phenomenology by progressively keeping physical models at the state of the art through a regular feed-back from the interpretation of the current and future experimental programs performed in the international frame.

Experimental investigation on melt coolability under bottom flooding with and without decay heat simulation

Investigations on severe accident phenomena help us in understanding the realistic accidental phenomena for the assessment of associated risk. The societal impact of radiological leakage to the environment has demanded further robustness in the line of defence of nuclear safety. Thus, to ensure the cooling and stabilization of corium within reactor containment in case of severe accident scenarios, many new reactors have been envisaged with core catcher. In this regard, corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. After studying the various cooling strategies, it has been demonstrated that melt coolability using bottom flooding of water is one of the most efficient techniques so far. To study the effect of decay heat on melt pool coolability under bottom flooding condition, two experiments have been performed in this paper; one without the decay heat and the other with decay heat. The test section used for carrying out these experiments consisted of two parts viz. lower part for retaining the melt from furnace, water inlet and melt quenching, and upper part for steam expansion and its outlet. The total height of the test section was 1400mm and was made of 33mm thick carbon steel. Total six stainless steel nozzles of diameter 12mm were used for injecting water at the bottom of the melt pool. The lower part was surrounded by 10 radiative heaters to simulate decay heat of 10kW which corresponds to 0.5MW/m3. The experiments showed that quenching of about 25l of melt at initial temperature of nearly 1200°C took only a few minutes in both the cases. The debris formed in the experiments ranged between 0.5mm fine particles to 50mm porous chunks. The measured porosities were nearly 51% without decay heat and nearly 67% with decay heat. This paper presents the comparative experimental study of melt coolability using bottom flooding without and with decay heat scenarios. The experimental measurements show that the quenching behaviour of melt pool is not affected by decay heat, although stabilization of debris temperatures to saturation temperature takes a bit longer duration as compared to without decay heat scenarios.

A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions.Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit.The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc.Important conclusions of this work are as follows:(a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel;(b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head;(c)Even if the vessel is externally cooled, transient 3D corium configurations in the lower head can induce high thermal loads on the vessel leading to a fast failure. Early in-vessel reflooding may prevent this risk. As a consequence, ex-vessel flooding should be accompanied by in-vessel reflooding;(d)In case of failure with external cooling, the breach is plausibly located on the lateral surface of the lower head. The failure is most likely local.The present work identifies effects that should be taken into account and modelled in numerical tools describing core degradation and vessel failure.

Quenching Behaviour of Top Flooded Molten Pool

During a severe accident in a nuclear reactor, the core can melt and the melt can be relocated into the lower head forming a melt pool. If the vessel fails, the molten corium can be relocated in the containment cavity forming a melt pool. Such a pool or bed, if not quenched in time, may interact with the concrete basement of the cavity causing its ablation resulting in generation of non-condensable gases and water vapor, which poses a threat of containment pressurization, explosion and ground contamination. In order to devise the strategy to retard the progression of severe accidents in stipulated time, understanding of the corium coolability is very much essential. Coolability of molten corium in such ex-vessel condition is limited depending on depth of water ingression. But the question arises about to what extent the water will ingress? Thermal and physical properties of corium change drastically with increase in concrete percentage which strongly affects the coolability. The phenomenon of corium coolability is not fully understood owing to its complexity involving multi phase multi component heat and mass transfer. In order to better understand the phenomenon, experiments at present are the only answer; since modeling the complex phenomenon is still difficult. In order to gain some insights into melt coolability, the authors have carried out an experiment on quenching behavior of top flooded molten pool. A simulant material (sodium borosilicate glass) of about 25 liters at temperature 1200 oC was poured into the test section and was flooded from top with water. The transient temperature of the molten pool was measured. The experiment highlighted that, under adiabatic conditions, water ingression occurred only upto 10 mm depth, below which a stable solid crust was formed which limited the heat transfer. Also, no gap between crust and vessel was formed.

A simple model to understand physics of melt coolability under bottom flooding

Following a severe accident in a nuclear reactor, stabilization and termination of molten pool/debris is necessary to ensure safety of the reactor. In this regard, ex-vessel corium coolability still remains an unresolved issue in spite of several efforts being taken towards its understanding. Extensive research has been carried out in the field of corium coolability phenomenon, but the outcome is far from establishing a certainty of complete coolability of melt under all possible scenarios. Major thrust so far was on top flooding condition, which often results in uncoolable condition due to inability of water ingression deep into the solid crust. The research on coolability under bottom flooding condition started long back, but has gained lot of momentum recently. However, mainly the focus is on development of engineering system to enhance the coolability using bottom flooding. The understanding of the phenomenon is not yet totally brought out. In this paper, we have postulated the mechanism of melt coolability under bottom flooding. We have developed a simplified mechanistic model for prediction of coolability of molten corium under bottom flooding condition based on the postulations. The model has been validated with the experimental data from literature. Comparison of top flooding vis-à-vis bottom flooding has also been presented.

Three-Dimensional Modeling of Debris Mixing and Sedimentation in Severe Accidents Using the Moving Particle Semi-Implicit Method Coupled with Rigid Body Dynamics

Particle mixing and sedimentation, related to corium debris bed formation and coolability in severe accidents, is investigated using a new computational fluid dynamics tool: the Analysis of Debris Dynamics and Agglomeration (ADDA) code. ADDA was developed based on an enhanced numerical method combining the moving particle semi-implicit algorithm with a rigid body dynamic model. The analysis successively simulates the entire process of debris bed formation, including corium jet breakup, mixing, and sedimentation. The methodology allows identification of key characteristics in the formation of the corium debris bed. Two-dimensional (2-D) and three-dimensional (3-D) simulations were utilized to model the detailed flow structures and mixing phenomena, along with the final sedimentation process, and were compared to the Q21 QUEOS test performed at Forschungszentrum Karlsruhe, Germany. For the analysis of debris bed formation, it is recommended that full 3-D simulations be utilized to provide enhanced accuracy related to corium debris field prediction. The 2-D simulations were found to be insufficient because of the debris field dependence on particle agglomeration and mixing, prior to debris settling.

The DEFOR-S Experimental Study of Debris Formation with Corium Simulant Materials

Characteristics of corium debris beds formed in a severe core melt accident are studied in the Debris Bed Formation-Snapshot (DEFOR-S) test campaign, in which superheated binary-oxidic melts (both eutectic and noneutectic compositions) as the corium simulants are discharged into a water pool. Water subcooling and pool depth are found to significantly influence the debris fragments’ morphology and agglomeration. When particle agglomeration is absent, the tests produced debris beds with porosity of ~60 to 70%. This porosity is significantly higher than the ~40% porosity broadly used in contemporary analysis of corium debris coolability in light water reactor severe accidents. The impact of debris formation on corium coolability is further complicated by debris fragments’ sharp edges, roughened surfaces, and cavities that are partially or fully encapsulated within the debris fragments. These observations are made consistently in both the DEFOR-S experiments and other tests with prototypic and simulant corium melts. Synthesis of the debris fragments from the DEFOR-S tests conducted under different melt and coolant conditions reveal trends in particle size, particle sphericity, surface roughness, sharp edges, and internal porosity as functions of water subcooling and melt composition. Qualitative analysis and discussion reaffirm the complex interplay between contributing processes (droplet interfacial instability and breakup, droplet cooling and solidification, cavity formation and solid fracture) on particle morphology and, consequently, on the characteristics of the debris beds.

An approach to numerical simulation and analysis of molten corium coolability in a boiling water reactor lower head

This paper discusses an approach for application of the computational fluid dynamics (CFD) method to support development and validation of computationally effective methods for safety analysis, on the example of molten corium coolability in a BWR lower head. The approach consists of five steps designed to ensure physical soundness of the effective method simulation results: (i) analysis and decomposition of a severe accident problem into a set of separate-effect phenomena, (ii) validation of the CFD models on relevant separate-effect experiments for the reactor prototypical ranges of governing parameters, (iii) development of effective models and closures on the base of physical insights gained from relevant experiments and CFD simulations, (iv) using data from the integral experiments and CFD simulations performed under reactor prototypic conditions for validation of the effective model with quantification of uncertainty in the prediction results and (v) application of the computationally effective model to simulate and analyze the severe accident transient under consideration, including sensitivity and uncertainty analysis. Implementation of the approach is illustrated on a so-called effective convectivity model for simulation of turbulent natural convection heat transfer and phase changes in a decay-heated corium pool. It is shown that detailed information obtained from the CFD simulations are instrumental to ensure the effective models capture safety-significant local phenomena, e.g. the enhanced downward heat flux in the vicinity of a cooled control rod guide tube.

The effects of debris bed's prototypical characteristics on corium coolability in a LWR severe accident

This paper is concerned with coolability assessment of a debris bed formed in fuel coolant interactions (FCIs) during a hypothetical severe accident in a light water reactor (LWR). The focus is placed the potential effect of the bed's prototypical characteristics on its coolability, in terms of (i) porosity range, (ii) multi-dimensionality, (iii) inhomogeneity, (iv) particle morphology, and (v) heat generation method (e.g. volumetric heating vs. local heaters). The analysis results indicate availability of substantial coolability margins compared to previous assessments based on models and experiments using an idealized bed configuration (e.g. 1D homogenous debris layer). Notably, high porosity (up to 70%) of debris beds, obtained in experiments and expected to be the case of prototypical debris beds, could increase the dryout heat flux by 100% and more, depending on particle size, compared with the dryout heat flux predicted for debris beds with traditionally assumed porosity of approximately 40%. Bed inhomogeneity represented by micro-channels in a mini bed is predicted to enhance the dryout heat flux by up to ∼50%, even if the micro-channels occupy only a small volume fraction (e.g., less than 4%) of the bed. The effect of coolant side ingress into a multidimensional bed is predicted to enhance the dryout heat flux by up to 40% for the beds analyzed.

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation

This paper, and its companion paper [Tran C.T., Dinh, T.N. The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application. Progress in Nuclear Energy (companion paper), in preparation] document the development, validation and applications of a simulation platform for computationally-effective, sufficiently-accurate numerical predictions of core melt-structure-water interactions in the light water reactor lower head during a postulated severe core-melting accident. The centerpiece of this work is the Effective Convectivity Model (ECM) for description of energy splitting in a core melt pool. Built on the concept of characteristic velocities in Effective Convectivity Conductivity Model and supported by the key findings obtained from Computational Fluid Dynamics (CFD) simulations of turbulent natural convection, heat transfer and phase changes in volumetrically heated liquid pools, the ECM is refined and extended to three-dimensions and phase changes to enable simulations of melt pool formation and corium coolability in complex geometry such as a Boiling Water Reactor (BWR) lower plenum.

CORIUM COOLABILITY UNDER EX-VESSEL ACCIDENT CONDITIONS FOR LWRs

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+ LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

Results of the QUENCH-L2, DISCO-L2, and COMET-L2 experiments performed within the LACOMERA project at the Forschungszentrum Karlsruhe

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany (FZK) is a 4-year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU Member Countries and Associated States access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments: QUENCH-L2: boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L2: fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor. COMET-L2: investigation of long-term melt–concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.

ICONE15-10227 RESULTS OF THE LIVE-L1 EXPERIMENT ON MELT BEHAVIOUR IN RPV LOWER HEAD PERFORMED WITHIN THE LACOMERA PROJECT AT THE FORSCHUNGSZENTRUM KARLSRUHE

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a four year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU Member Countries and Associated States access to four large scale experimental facilities QUENCH, LIVE, DISCO, and COMET which can be used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper briefly summarises the objectives of the LACOMERA project and presents the main results obtained in the LIVE-L1 experiment which was designed to investigate the core melt behaviour in the lower plenum of the reactor pressure vessel and the influence of the cooling of the vessel outer surface with water in the conditions that may occur during core meltdown accident in Pressurized Water Reactors (PWR).