Concept of a core cooling system and experiments performed

One of the problems which must be solved in severe accidents is the melt-concrete interaction which does occur when the core debris penetrates the lower pressure vessel head and contacts the basement. To prevent these accident consequences, a core catcher concept is proposed to be integrated into a new pressurized-water reactor design. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core melt based on the process of water inlet from the bottom.In order to justify the dominant process during flooding of the melt from the bottom, prototypic experiments with thermite melts in laboratory scale have been carried out. In these experiments flooding and early coolability of the melt is demonstrated. To obtain more detailed information on the important process of water penetration into the melt, a simulant experiment has been conducted using a transparent plastic melt with the typical viscosity behaviour of an oxidic corium melt and a temperature allowing evaporation of water. In every experiment the melt is flooded, and complete freezing in the form of a porous layer occurs within a few minutes only.     

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.     

Corium spreading and coolability: CSC Project

In the context of severe accidents, large R&D efforts throughout the world are currently directed towards ex-vessel corium behaviour. Among the mitigation means which can be envisaged, the European industries and utilities are considering the implementation of a core-catcher outside the reactor pressure vessel in order to prevent basemat erosion and to stabilize and control the corium within the containment. The CSC project focused on two key phenomena for external core-catcher efficiency, reliability and safety: spreading and coolability. An experimental programme, covering different scenarios and including both simulant and real materials provided a lot of results which now constitute a large database and which enabled the qualification of computer codes.     

Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt–water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb–Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed.     

Validation and application of the WABE code: Investigations of constitutive laws and 2D effects on debris coolability

The WABE-2D model aims at the problem of coolability of degraded core material during a severe accident in a light water reactor (LWR) and describes the transient boil-off and quenching behavior of debris beds. It is being developed in the frame of the KESS code system, which is considered to describe the processes of core heatup, melting, degradation and relocation to the lower plenum as well as the subsequent behavior. The models developed in this frame are being integrated in the German system code ATHLET-CD.An emphasis of the present contribution lies on multidimensional aspects of the cooling behavior. From multidimensional features a significant improvement of overall coolability is expected compared to what is concluded based on classical one-dimensional analyses. Such analyses – also mainly oriented at top cooling conditions – additionally miss the expected importance of interfacial drag which should support coolability in co-current flow situations due to bottom flooding. The latter situation plays a role in the multidimensional behavior expected under realistic conditions. Thus, a further emphasis in the present contribution lies on the constitutive drag laws and their effects in such configurations.Calculations comparing top and bottom flooding and the influence of interfacial friction are presented. An explanation for effects observed in related experiments at Forschungszentrum Karlsruhe is provided based on this influence. The significant increase of dryout heat flux with water inflow from below, driven by a lateral water column, is reproduced and understood. Enhanced cooling due to this and in general by lateral inflow is also demonstrated for reactor scenarios, considering particulate debris in the lower head of the reactor pressure vessel (RPV) of a LWR or in a deep water pool in the reactor cavity of a boiling water reactor (BWR). Cooling by steam flow through local dry zones can establish under lateral water supply to regions below and yield a further extension of coolability. Quenching of hot material is also analyzed. Finally, cases with loss of coolability, dry zone formation and melting are considered, especially in the perspective to analyze melt pool formation in the lower head of the RPV and the history of thermal interaction with the lower head wall. The latter will determine failure possibilities and modes of the RPV.     

Simulation of melt jet breakup and debris bed formation in water pools with IKEJET/IKEMIX

The objective of the development of the code system KESS is simulating the processes of core melting, relocation of core material to the lower head of the reactor pressure vessel (RPV) and its further heatup, modelling of fission product release and coolability of the core material. In the scope of the code development, IKEJET and IKEMIX were designed as key models for the breakup of a molten jet falling into a water pool, cooling of fragments and the formation of particulate debris beds. Calculations were performed with these codes, simulating FARO corium quenching experiments at saturated (L-28) and subcooled (L-31) conditions, as well as PREMIX experiments, e.g. PM-16. With the assumption of a reduced interfacial friction between water and steam as compared to usually applied laws, the melt breakup, energy release from the melt and pressurisation of the vessel observed in the experiments are reproduced with a reasonable accuracy. The model is further applied to reactor conditions, calculating the relocation of a mass of corium of 30 t into the lower plenum, its fragmentation and the formation of a particle bed.     

Considerations on ex-vessel corium behavior: Scenarios, MCCI and coolability

The aim of this editorial article is to provide some structuring of the subject addressed in this part of the issue, which, in the view of the present authors, has not that clearly been reached by the specific contributions. Ex-vessel corium behavior is a wide field. Therefore, classification of the goals of investigations within a safety philosophy is especially required to get not lost in detailed aspects. The overall goals here are the coolability and retention options under ex-vessel conditions. Based on scenario considerations generally addressing risks and options, general principles of cooling and retention devices are outlined. Since the concrete erosion by melt yields a major risk and has to be considered in the concepts and devices, and also since several contributions to this part are dealing with specific aspects of this molten core-concrete interaction (MCCI), a large part of this editorial paper concerns the status of knowledge and modeling and the lines of research in this area. The status and the perspectives of codes is especially addressed by own contributions of one of the authors with the GRS code WEX and with MEDICIS, both codes included in the European integral code ASTEC. Finally, the coolability questions are discussed with respect to the different concepts in general and those addressed specifically in the present contributions. In some considerations, gas production by erosion plays a role to produce porous or particulate debris and thus to enhance coolability. Water injection from bottom is a more direct and probably more effective measure to reach this, specifically designed in the COMET core catcher concept. Specific contributions in this part deal with this concept, which is most closely related to the general subject of the present issue.     

A scoping study of debris bed formation in the DEFOR test facility

Motivated to understand the processes which govern the formation and characteristics of a debris bed and hence its coolability during a postulated severe accident of a light water reactor, a new research program called DEFOR (DEbris FORmation) was initiated at the Royal Institute of Technology (KTH). This paper presents results obtained in scoping experiments conducted during an initial phase of the DEFOR program. The DEFOR-E test campaign is concerned with the DEFOR test facility commissioning and exploratory study of phenomena occurred during a debris bed formation. Binary oxide mixtures at different superheat temperatures were used as the corium melt simulants. The scoping experiments revealed the effect of water pool depth and subcooling, melt mass and material properties on the debris bed characteristics. Insights gained from the DEFOR-E test campaign help guide the scaling, design and operation of the subsequent experiments in the DEFOR program.     

Ex-vessel core melt stabilization research (ECOSTAR)

The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical–chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO₂-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.     

Experiments for a core catcher concept based on water addition from below

A core catcher concept is proposed to be integrated into a new PWR design based on the standard German PWR. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core-melt based on the process of water inlet from the bottom through the melt.To ge more detailed information on the very important process of water penetrating into the melt, simulant experiments have been conducted using a transparent plastic and a solder melt representing the oxidic and metallic part of the core-melt. It appears from the results that fragmentation of the melts can be achieved by proper selection of water supply pressure and water feed cross-section.The important part of the transient medium scale experiments with thermite melts, conducted since mid 1993, is to get information on the process of evaporation of water by water ingression in hot melts from below and to investigate whether there is a possibility of strong melt-water interactions, or even steam explosions. The experimental set-up represents a section of the core catcher. A thermite melt is located on the catcher plate with water supply from the bottom. After ignition of the melt, the upper sacrificial layer is eroded until water penetrates into the melt from the bottom through the holes in the supporting plate and fragmentation and simultaneous solidification of the melt occurs. The experiments, up to now, show that flooding and early coolability of the melt by water addition from the bottom are achieved.These experiments serve also as pretests for the COMET-H experiments with sustained heating planned to be conducted in the BETA-facility at the beginning of next year.     

Development and verification of molten corium–concrete interaction code

During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes.     

Experimental results on the coolability of a debris bed with multidimensional cooling effects

Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations ( [Rashid et al., 2008], [Groll et al., 2008] and [0055]). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration.A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).     

In-vessel corium catcher of a nuclear reactor

A new concept of an in-vessel corium melt catcher is proposed. The lower part of an elongated reactor vessel, which is filled with a sacrificial material of a proper composition, porosity, and arrangement, is used as such a catcher. The concept accounts of the scientific and design experience with the development of the ex-vessel corium catcher for the Tyan’van NPP with VVER-1000 reactors.     

Experiments on in-vessel melt coolability in the EC-FOREVER Program

This paper reports the results from the experiments conducted on the coolability of corium melt during a severe accident scenario when the bottom head is full of the core melt, undergoing natural circulation. These experiments are part of the EC-FOREVER Program in which vessel failure experiments have also been performed. The experiments are performed in a 1/10th scale vessel (400 mm diameter and 15 mm wall thickness) and the oxidic melt employed is the mixture CaO + B₂O₃ at 1400 K, representing the corium melt mixture of UO₂ + ZrO₂.The experiments employed an initial phase, during which uniform volumetric heating of the melt was provided and the vessel was pressurised to 25 bar, for several hours, to generate maximum creep deformation of 5%, in order to provide the conditions for the formation of a gap between the melt-pool crust and the bottom head wall. After this phase, the vessel was flooded with water.Data were obtained on only the vessel and the melt pool temperatures in one of the EC-FOREVER experiments reported here. In the second experiment, however, besides the temperature data, additional data were obtained on the steam flow rate and the heat transfer to the water, at the upper face of the melt pool, as a function of time.It was found that the gap cooling mechanism was not effective in reducing the vessel wall temperatures after water flooding. Post-test examinations revealed that the water ingression extended to the depth of only 60 mm in the melt pool. The character of the heat transfer to the water from the melt pool upper surface was found to be similar to that observed in the MACE tests for the coolability of an ex-vessel melt pool flooded by water at the top.     

The COMET-L3 experiment on long-term melt-concrete interaction and cooling by surface flooding

The COMET-L3 experiment considers the long-term situation of corium/concrete interaction in an anticipated core melt accident of a light water reactor after the metal melt is layered beneath the oxide melt. The experimental focus is on the cavity formation in the basemat and the risk of a long-term basemat penetration by the metallic part of the melt. The experiment investigates the two-dimensional concrete erosion in a cylindrical crucible of 60 cm in diameter fabricated from siliceous concrete in the first phase of the test, and the influence of surface flooding in the second phase. The initial mass of the melt was 425 kg steel and 211 kg oxide. Decay heating in the two-component metal and oxide melt is simulated by sustained induction heating of the metal phase that is overlaid by the oxide melt.In the initial phase of the test, the overheated, highly agitated metal melt causes intense interaction with the concrete, which leads to fast decrease of the initial melt overheat and reduction of the initially high concrete erosion rate. Thereafter, the erosion by the metal melt slows down to about 0.07 mm/s into the axial direction. Lateral erosion is by a factor of 3 smaller. Surface flooding of the melt is initiated at 800 s. Flooding does not lead to strong melt/water interactions and to penetration of water into the melt. Concrete erosion continues with about 0.040 mm/s until the melt reaches the maximum erosion limit of the crucible. Post-test analysis of the solidified melt was performed after the crucible was sectioned. The solidified melt shows no indication of water ingression from the upper surface. Tight surface crusts explain poor heat removal to the flooding water and the ongoing concrete erosion also after the top flooding.Details of the experiment are reported. The experiment shall be used for validation of models and computer codes for safety assessment.     

A numerical and experimental study of water ingression phenomena in melt pool coolability

During a postulated severe accident, the core can melt and the melt can fail the reactor vessel. Subsequently, the molten corium can be relocated in the containment cavity forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question that arises is to what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using the computer code MELCOOL. The code considers the heat transfer behaviour in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, thermal stresses built-in the crust, disintegration of crust into debris, natural convection heat transfer in debris and water ingression into the debris bed. To validate the computer code, experiments were conducted in a facility named as core melt coolability (COMECO). The facility consists of a test section (200 mm × 200 mm square cross-section) and with a height of 300 mm. About 14 L of melt comprising of 30% CaO + 70% B₂O₃ (by wt.) was poured into the test section. The melt was heated by four heaters from outside the test section to simulate the decay heat of corium. The melt was water flooded from the top, and the depth of water pool was kept constant at around 700 mm throughout the experiment. The transient temperature behaviour in the melt pool at different axial and radial locations was measured with 24 K-type thermocouples and the steam flow rate was measured using a vortex flow meter. The melt temperature measurements indicated that water could ingress only up to a certain depth into the melt pool. The MELCOOL predictions were compared with the test data for the temperature distribution inside the molten pool. The code was found to simulate the quenching behaviour and depth of water ingression quite well.     

Boiling experiments for the validation of dryout models used in reactor safety

The coolability of fragmented corium is a major issue in reactor safety. Since the long-term coolability of such particle beds is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the pressure field inside the debris has a strong effect on the cooling potential in multi-dimensional cases as expected in severe accidents in light water reactors (LWR). Therefore, the determination of the pressure field for two-phase flows in porous media is one central point of interest.In this context simulation models and in particular dryout models were developed for reactor safety analyses which have to be validated by reliable experimental data. Therefore, basic experimental investigations have been carried out with inductively heated steel balls of 6 or 3 mm diameter to provide a database for the validation and modification of the friction laws included in these dryout models.The performed boiling and dryout experiments show clearly that models without the explicit consideration of the interfacial drag cannot predict the pressure distribution inside a boiling particle bed, not even qualitatively. Against it, models with an explicit consideration of the interfacial drag can describe the distribution of pressure inside a boiling particle bed.     

Corium crust strength measurements

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO₂, ZrO₂, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.     

Core cooling in pressurized-water reactor during water injection

In this paper,the reactor core cooling and its melt progression terminating is evaluated,and the initiation criterion for reactor cavity flooding during water injection is determined.The core cooling in pressurized-water reactor of severe accident is simulated with the thermal hydraulic and severe accident code of SCDAP/RELAP5.The results show that the core melt progression is terminated by water injection,before the core debris has formed at bottom of core,and the initiation of reactor cavity flooding is indicated by the core exit temperature.