In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc.In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool.This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool. (C) 2014 Elsevier B.V. All rights reserved.

The paper is devoted to the presentation of validation cases carried out for the models describing the corium behaviour in the "lower plenum" of the reactor vessel implemented in the V2.0 version of the ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany). In the ASTEC architecture, these models are grouped within the single ICARE module and they are all activated in typical accident scenarios. Therefore, it is important to check the validity of each individual model, as long as experiments are available for which a single physical process is involved. The results of ASTEC applications against the following experiments are presented: FARO (corium jet fragmentation), LIVE (heat transfer between a molten pool and the vessel), MASCA (separation and stratification of corium non miscible phases) and OLHF (mechanical failure of the vessel). Compared to the previous ASTEC V1.3 version, the validation matrix is extended. This work allows determining recommended values for some model parameters (e.g. debris particle size in the fragmentation model and criterion for debris bed liquefaction). Almost all the processes governing the corium behaviour, its thermal interaction with the vessel wall and the vessel failure are modelled in ASTEC and these models have been assessed individually with satisfactory results. The main uncertainties appear to be related to the calculation of transient evolutions. (C) 2013 Elsevier B.V. All rights reserved.

The accident at the Fukushima Dai-ichi Nuclear Power Plant has highlighted the vulnerability of nuclear fuels that are stored in spent fuel pools (SFP) before their evacuation and final disposal or possible reprocessing. This vulnerability is due to the potential loss of sufficient fuel cooling in case of internal events or of extreme external events such as earthquake or flooding. Further, the limited number of barriers containing the radioactive products, the fuel cladding is typically the only barrier in an SFP whereas there are three physical barriers when the fuel is in the reactor (the fuel cladding, the envelope of the primary circuit and the containment building), induces higher risks of radioactive releases to the environment if sufficient fuel cooling cannot be recovered.This paper presents the results of various assessments of SFP accidents performed with different severe accident codes for different SFP geometries. In these studies, both loss of cooling and loss of water transients were analyzed. These calculations have been performed in the framework of the SARNET2 project. The analysis of these different SFP scenarios was conducted to identify some limits and needs for improvement of the SA code developed and usually used for reactor applications. More specifically, some questions about the SA code capabilities to evaluate the impact of air flow on coolability of the fuel assemblies (FAs) and the rate and path of air/steam flow in these FAs, have arisen. This work has also enabled us to identify the need to improve the knowledge and the models related to the effect of mixed air/steam atmospheres, especially the role of nitrogen on the acceleration of the corrosion rate of Zircaloy claddings and on the degradation of their mechanical properties. (C) 2014 Elsevier Ltd. All rights reserved.

The loss of coolant accidents with core degradation e.g. TMI-2 and Fukushima demonstrated that the nuclear safety analysis has to cover accident sequences involving a late reflood activation in order to develop appropriate and reliable mitigation strategies for both, existing and advanced reactors. The reflood (injection of water) is possible if one or several water sources become available during the accident. In a late phase of accident, no well-defined coolant paths would exist and a large part of the core would resemble to a debris bed e.g. particles with characteristic length-scale: 1 to 5 mm, as observed in TMI-2. The French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation at a late stage of an accident. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core for ICARE-CATHARE code. The comparison of the calculations with PRELUDE experimental results is presented. It is shown that the quench front exhibits either a 1D behavior or a 2D one, depending on injection rate or bed characteristics. The PRELUDE data cover a rather large range of variation of parameters for which the developed model appears to be quite predictive.

Understanding of two-phase flow through porous medium with intense phase change is of interest in many situations, including nuclear, chemical or geophysical applications. Intense boiling occurs when the liquid is injected into a highly superheated medium. Under such conditions, the heat flux extracted by the fluid from the porous medium is mainly governed by the nucleation of bubbles and by the evaporation of thin liquid films. Both configurations are possible, depending on local flow conditions and on the ratio of bubble size to pore size. The present study is motivated by the safety evaluation of light water nuclear reactors in case of a severe accident scenario, such as the one that happened in Fukushima Dai-ichi plant in March, 2011. If water sources are not available for a long period of time, the reactor core heats up due to the residual power and eventually becomes significantly damaged due to intense oxidation of metals and fragmentation of fuel rods resulting in the formation of a porous medium where the particles have a characteristic length-scale of 1 to 5 mm. The coolability of the porous medium will depend on the water flow rate which can enter the medium under the available driving head and on the geometrical features of the porous matrix (average pore size, porosity). Therefore, it is of high interest to evaluate the conditions for which the injection of water in such porous medium is likely to stop the progression of the accident. The present paper addresses the issue of modelling two-phase flow and heat transfers in a porous medium initially dry, where water is injected. The medium is initially at a temperature well above the saturation temperature of water. In a first part, a summary of existing knowledge is provided, showing the scarcity of models and experimental data. In a second part, new experimental results obtained in an IRSN facility are analysed. The experiment consists in a bed of steel particles that are heated up to 700°Cbefore injecting water. The facility is briefly described. The velocity of the "quench front" (location where particles are quickly cooled down) and the total pressure drop across the medium are estimated. The dependencies of those quantities with respect to the inlet water flow rate, the initial temperature of the medium and the diameter of particles are obtained. In a third part, a model is proposed, based on a previously developed model which is improved in order to take into account intense boiling regimes (in particular nucleate boiling). The model includes a function that takes into account the contact area between water and the particles which depends on the temperature of particles and on the void fraction. That function affects the local intensity of phase change. The model involves a few parameters which cannot be evaluated analytically. Those parameters are bounded, following the analysis of experimental data. Finally, the model is assessed by comparison of calculations with those new experimental data. The satisfactory agreement shows that the model is almost predictive in the range of parameters studied. The experimental results also show that the quench front becomes unstable under certain conditions. This is also analysed and compared with the predictions of the model.

In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods may result in the formation of a 'debris bed'. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. Themore » PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)« less

A generalized non local-equilibrium, three-equation model was developed for the macroscopic description of two-phase flow heat and mass transfer in porous media subjected to phase change. Six pore-scale closure problems were proposed to determine all the effective transport coefficients for representative unit cells. An improved model is presented in this paper with the perspective of application to intense boiling phenomena. The objective of this paper is to present application of this model to the simulation of reflooding of severely damaged nuclear reactor cores. In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of the core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a "debris bed". The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), which corresponds to a high permeability porous medium. The proposed two-phase flow model is implemented in the ICARE-CATHARE code, developed by IRSN to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN has set up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, with the objective to validate safety models. The PRELUDE program studies the complex two phase flow of water and steam in a porous medium (diameter 180 mm, height 200 mm), initially heated to a high temperature (400 degrees C or 700 degrees C). The series of PRELUDE experiments achieved in 2010 constitute a significant complement to the database of high temperature bottom reflood experimental data. They provide relevant data to understand the progression of the quench front and the intensity of heat transfer. Modeling accurately these experiments required improvements to the reflooding model, especially in terms of the existence of various saturation regimes. The improved two-phase flow model shows a good agreement with PRELUDE experimental results.

In case of severe accident in a nuclear reactor, the fuel rods may be highly damaged and oxidized and finally collapse to form a debris bed. Removal of decay heat from a debris bed is a challenging issue because of the difficulty for water to flow inside. From a safety point of view, it is important to evaluate chances of coolability of the reactor core during a severe accident. This is in line with the safety philosophy of defence in depth which requires to foresee and to analyse all options to stop an accident at any stage. Currently, the French IRSN has started two experimental programs, PRELUDE and PEARL, to investigate the physical phenomena during a reflood process at high temperature and to provide relevant data in order to improve predictive models. From a selection of experimental results, a reflooding model was improved and validated. The model is implemented in the code ICARE-CATHARE developed by IRSN which is used for severe accident reactor analysis.

In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a bed. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1-5 mm). The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN (Institut de radioprotection et de surete nucleaire), to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow (water/steam), in a porous medium (diameter 180 mm, height 200 mm), initially heated to a high temperature (400 °C or 700 °C). On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results.

The understanding of the reflood process of a severely damaged reactor core represents a challenge in the prediction of safety margin of existing and future pressurized water reactors. After the TMI-2 accident, the understanding of coolability of severely damaged reactor core became an objective of many theoretical and experimental studies. Currently, the French Institute of Radioprotection and Nuclear Safety (IRSN) has started two experimental programs, PRELUDE and PEARL, to investigate the physical phenomena during a reflood process at high temperature and to provide relevant data in order to improve predictive models. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core. The presented model is based on the theory of heat transfer and two-phase flow in porous media and in small hydraulic diameter channels. The proposed model is implemented into the European computer code for severe accident analysis ICARE-CATHARE. The comparison of the calculations with PRELUDE experimental results is presented. Finally, the issue of transposition to the reactor scale is discussed and some answers are proposed using calculation results for a debris bed in a configuration similar to what could be expected in a severely damaged reactor core.

Severe accidents in PWRs are characterized by a continuously changing geometry of the core due to chemical reactions, melting and mechanical failure of the rods and other structures. These local variatio ns of the porosity and other parameters lead to multidimensional flows and heat transfers. In this paper , a comprehensive set of multi-dimensional models describing heat transfers, thermal-hydraulics and m elt relocation in a reactor vessel is presented. Those models are suitable for the core description during a seve re accident transient. A series of applications at the reactor scale shows the benefits of using such model.

Abstract For several years the French Institut de Radioprotection et de Sûreté Nucléire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH have been jointly developing a system of calculation codes—the integral Accident Source Term Evaluation Code (ASTEC)—to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment, i.e., the source term. ASTEC has progressively reached a larger European dimension through projects of the European Commission Framework Programme. In particular, in the frame of the European Severe Accident Research NETwork of Excellence (SARNET), jointly executed research activities were performed with the ultimate objectives of providing physical models for integration into ASTEC and making the code the European reference. This effort will go on in the frame of the SARNET2 next network. The ASTEC models are today at the state of the art, except for reflooding of a degraded core. Many applications have been performed by IRSN for significant safety studies, including the probabilistic safety analysis level 2 on a French pressurized water reactor. The first version V2.0 of the new ASTEC series, released in spring 2009, will allow simulation of the European Pressurized Reactor (EPR) and will include advanced core degradation models. Then, ASTEC will remain the repository of knowledge gained from international research and development. Other long-term objectives are on one hand extension of the scope of application to boiling water reactors and CANada Deuterium Uranium (CANDU) reactors, to accidents in the ITER Fusion facility, and to Very High Temperature Reactor (VHTR) Generation IV reactors, and on the other hand to the use for emergency response tools and for severe accident simulators.

Severe accidents in PWRs are characterized by a continuously changing geometry of the core due to chemical reactions, melting and mechanical failure of the rods and other structures. These local variations of the porosity and other parameters lead to multi-dimensionnal flows and heat transfers. In this paper, a comprehensive set of multi-dimensionnal models describing heat transfers, thermal-hydraulics and melt relocation in a reactor vessel is presented. Those models are suitable for the core description during a severe accident transient. A series of applications at the reactor scale shows the benefits of using such models.