The paper presents some results of the ISTC (International Science and Technology Center)-financed project 'Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel' (METCOR). In the METCOR experiments the metallic phase of a two-liquid system was produced by the interaction between hot suboxidized corium and cooled VVER vessel steel, with the steel being corroded. Models of corrosion mechanisms in the considered conditions are used to systematize data on the limiting temperature of corrosion/(dissolution) of the vessel steel. A considerable influence of thermal gradient conditions is shown, which has to be taken into account in the analysis of molten pool behaviour.

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.

This study investigates heat transfer characters of a volumetrically heated melt pool in LWR lower plenum. Experimental restrictions on prediction reliability are discussed. These restrictions include cooling boundary conditions, vessel geometries, and simulant melt selection on general and localized heat transfer. A survey of existing heat transfer correlations derived from individual experimental definitions is presented. The inconsistency in parameter definitions in Nu–Ra correlations is discussed. Furthermore, the discrepancy of upward Nu depending on the existence of crust is stressed. Several serials of experiments with different combinations boundary condition of external cooling and top cooling were performed in LIVE3D and LIVE2D facilities. The experiments were conducted with simulants with and without crust formation. The influences of cooling boundary conditions, the vessel geometry, and the simulant material on overall heat transfer as well as on heat flux distribution are analyzed. This paper provides own explanations about the discrepancies among the exiting heat transfer correlations and recommends the most suitable descriptions of melt pool heat transfer under different accident management strategies.

This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

The simulation of severe accidents in nuclear power plants with system codes is a powerful tool to improve the safety measures to prevent severe accidents. The further development of severe accident codes is part of current research. MELCOR, as the leading nuclear safety code, provides the possibility to be coupled to other codes. A detailed knowledge of this coupling interface is necessary to use this possibility. Therefore, the software tool DINAMO, which contains the coupling routines and an interface to communicate with other programs, was developed. Using DINAMO it is possible to utilize new models for specific phenomena in MELCOR. In the present work the Phase-Change Effective Convectivity Model was coupled using the CFD-software OpenFOAM and DINAMO to MELCOR to improve the prediction of molten core material in the lower plenum of a reactor pressure vessel. The simulation results were compared to the experimental findings of the LIVE-facility.

The SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core-concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented. (C) 2015 Published by Elsevier B.V.

The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II-III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency. (C) 2014 Elsevier Ltd. All rights reserved.

This study was performed to obtain insight into the characteristics of contaminated cementitious materials which may result from a light water reactor core melt down accident. Such material arose in a huge amount from the Fukushima disaster. We analyzed the elemental and mineralogical composition of similar, heat-treated material and investigated its radionuclide retention properties.We present the radionuclide retention properties of concrete samples which originally were part of an experiment using a thick-walled concrete recipient that had been heated by simulating a reactor melt down. Batch sorption experiments have been performed with the elements Cs(I), Co(II), and Eu(III) in seawater under aerobic conditions. Sorption coefficients were measured: Rs(Eu) ∼5800mlg−1 and Rs(Co) ∼110mlg−1. A tentative value for Cs was determined, adulterated by the relatively high release of Cs from the concrete itself.

Using the source-based SIMPLE algorithm based on a fixed grid method, a two-dimensional numerical model for a convection-diffusion controlled mushy region phase-change problem was developed to investigate the heat transfer characteristics of LIVE 14 melt pool subjected to a partial solidification process in a Pressurized Water Reactor (PWR) lower plenum during a hypothetical severe accident. For the binary non-eutectic mixtures of L4 melt, a linear liquid fraction temperature relationship was implemented on the calculations of the velocity and enthalpy in the mushy zone. The effect of fluid flow in the melt pool was analyzed, and numerical results for the cases with and without phase change model were calculated to investigate the effects of solidification on the heat transfer characteristics of L4 melt pool. Numerical results indicated that the phase-change model could well predict the main parameters of melt pool, e.g. the melt pool temperatures, heat flux through the melt pool, and the crust thickness. Results also indicated that the predicted Nu number without solidification was overestimated by about 12%, compared with that with solidification. (C) 2014 Elsevier Ltd. All rights reserved.

Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations. (C) 2013 Elsevier B.V. All rights reserved.

The LIVE-L4 and LIVE-L5L experiments investigated the thermal-hydraulic behavior of the corium pool in the reactor pressure vessel lower head with the three-dimensional test vessel LIVE. The simulant material is a noneutectic binary mixture of 20% NaNO3-80% KNO3. Transient and steady-state parameters such as melt temperature and heat flux distribution through the vessel wall as well as crust formation characteristics were obtained. The two tests demonstrated that transient events like repeated melt relocation and change of decay power density facilitate crust deformation and change of crust thickness. Massive crust formation in a noneutectic melt pool leads to a change of melt pool composition and a decrease of melt-crust interface temperature. The melt temperature and heat flux at the same pool height and same power density can be roughly compared independent of heating history and initial melt pouring pattern. The dimensionless melt temperature as well as the dimensionless heat flux through the wall during the steady state are independent of power density if the pools have the same height. But, they are dependent on the pool height. For a low pool, the gradients with height of both melt temperature and heat flux through the vessel are larger than those for a high pool.

The LIVE-L4 test was conducted to investigate the transient and steady state behavior of the molten pool and the crust influenced by different heat generation rates. The main purpose of this work is to develop a simple novel model of the LIVE code to calculate the entire process of the LIVE-L4 test after the melt of KNO3-NaNO3 poured into the test vessel. The LIVE code is a transient code and can be used as a fast computational program to calculate the LIVE tests. Natural convection heat transfer in the melt pool, crust behavior, heat conduction in the vessel wall, and radiative heat transfer were all considered in the model of the LIVE code.In the LIVE code, Asfia-Dhir correlations were used to calculate average and local heat transfer co-efficients in the melt pool. With the assumption of no considering the composition change of local melt at melt/crust interface, many important parameters, including the melt pool temperature, heat flux distribution along the vessel wall, the thickness of the crust in steady state, and crust growth rate during the test, were calculated and compared with the LIVE-L4 experimental data.The melt pool Nu calculated by the LIVE code is larger than experimental data due to the use of Asfia-Dhir correlation in the LIVE code, which caused the average heat flux through the vessel wall larger than experiment data except the heating phase of 5 kW It is attributed that the temperature difference between the melt pool temperature and the interface temperature at melt/crust measured in the test is larger than that calculated by the LIVE code due to the constant interface temperature at melt/crust of 284 degrees C used in the LIVE code. Crust growth rate calculated by the LIVE code was consistent well with the experiment data. Calculation results indicated that the LIVE code could generally predict the main parameters of the melt and crust well during the LIVE-L4 test. (C) 2013 Elsevier Ltd. All rights reserved.

The LACOMECO project provides European organizations access to experimental facilities at Karlsruhe Institute of Technology (KIT) designed to study the remaining severe accident safety issues, including the coolability of a degraded core, corium coolability in RPV, melt dispersion to the reactor cavity, and hydrogen mixing and combustion in the containment. The facilities are unique in its specified field and the experiments are designed to be complementary to other European experimental platforms to form a coherent European nuclear experimental network. The LACOMECO platform includes: 1) QUENCH facility designed for the investigation of early and late phases core degradation in prototypical geometry for different reactor designs and cladding alloys; 2) LIVE facility, a large-scale 3D facility for the investigation of melt pool in the lower head of RPV; 3) DISCO facility, the only operating facility worldwide to investigate the melt dispersion to the reactor cavity and direct containment heating; 4) HYKA facility with a number of large and medium scale experimental vessels to investigate the hydrogen behaviour in containment under well controlled conditions. Eight experiments are to be performed in the LACOMECO project addressing the high and medium priority issues defined by the SARP group of SARNET. The paper addresses the objectives of the project and describes the main results obtained in the experiments performed up to now.