Course Of Severe Accident Research Articles

On water-ingression during top-flooding of corium melts

In the course of severe accident in a nuclear power plant, the corium mixture may flow down in the reactor pit and start a thermal attack of the basemat concrete. A simple way to terminate the melt progression may be to add water on top of it. Among the various physical processes that may participate in the quenching, water may penetrate into the solidified corium through cracks generated by the thermal stress during the solidification.The present paper aims at providing a clarification of the process though an analysis of the few dedicated experimental data available using proto-typical corium, namely the SSWICS 1-7 experiments. At first, a general reminder and analysis of the data is given. Then, the thermal-hydraulic aspects are investigated through the construction of a dedicated heat flux correlation and its use via the experimentally measured post-test permeabilities. The analysis is supported by 1D and 2D evaluations with the multiphase flow code MC3D-PREMIX, slightly modified for the purpose. It is concluded that the heat flux drastically decreases with the amount of added concrete material. Furthermore, 2D border effects are investigated and their importance highlighted in view of the experimental results analysed. These effects should explain the absence of observation of water ingression heat flux in the cases with large concrete amounts. Following, the paper proposes a model for the created permeability. Due to the complexity of the process and to the large uncertainties of the needed material properties in the considered situation with very high temperatures, a semi-empirical model is derived.Lastly, the model is adapted to the situation with internal decay heat, although no open experimental data is available to precisely support any model. A complete modeling is out of the scope of the paper, hence the focus is to provide hints for a first analysis of the impact of decay heat on its coolability.

Analysis of different inert gas injection point's influence on hydrogen risk during post-inerting in nuclear power plant

Amounts of hydrogen is generated and released into the containment during the course of severe accidents in AP1000 nuclear power plant. It is very important to study the measures to reduce the risk of hydrogen in containment. The severe accident analysis code is used to study the hydrogen and water vapor source terms during hypothesized severe accident induced by LB-LOCA with gravity injection failure in advanced passive PWR containment. Then, assuming the two sets of hydrogen igniter are all out of power and the passive hydrogen recombiners are functioning normally, we use the three-dimensional computational fluid dynamics code GASFLOW to study the influence of inert gas injection point on mitigating hydrogen risk during post-inerting. It shows that gaseous CO2 is suitable to be used as inert gas. In the process of post-inerting, Inert gas injection point should be below the hydrogen source. Inert gas enhances the hydrogen diffusion and mixing, and reduces the local hydrogen risk in the containment.

Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment

A number of integral and separate effect experiments were performed in the last two decades for validation of containment computational tools. The main goal of these benchmark experiments was to assess the ability of turbulence models and computational fluid dynamics codes to predict hydrogen concentration distribution and steam condensation rate in a nuclear reactor containment in the course of severe accidents. It appears from the published literature that the predictive capability of the existing computational tools still needs to be improved.This work examines numerically the temporal evolution of helium concentration in the experiment called LOWMA-3, performed in the MISTRA facility of CEA-Saclay, France. In the experiment, helium is used to mimic hydrogen of a real-case accident. The aim of this separate effect experiment, where steam condensation was not involved, is to predict helium concentration field. The conditions of the experiment are such that both the momentum transport and molecular diffusion contributions to the mixing process are of the same order of magnitude (Fr∼1). A commercial CFD code, Fluent, and a CEA in-house code, Trio_U, are used for flow and helium concentration fields temporal evolution prediction in the present study.The preliminary separate effect studies provide guidance to an optimal modeling approach for the LOWMA-3 experiment. Temporal evolution of helium concentration in the stratification layer is shown, and a comparison to the experiment is discussed. It is shown that correct modeling of the round jet flowfield is essential for a reliable prediction of the mixing process.

B 4C oxidation modelling in severe accident codes: Applications to PHEBUS and QUENCH experiments

Boron carbide is used in many nuclear power plants like BWR, VVER, some PWR, and EPR as a neutron absorber material. Consequently, it is important to assess its role in the core degradation phenomena during a severe accident (SA), as well as that of the carbon gas released from its degradation on the fission products behaviour. This paper describes the progresses achieved in the frame of the Network of Excellence SARNET concerning the B 4C control rod degradation modelling in Severe Accident codes, such as ATHLET-CD, ICARE2, ASTEC MELCOR and MAAP. These new developments complete improvements made during the European Union 5th Framework COLOSS project. Starting from basic modelling derived from available tests reported in the literature, large improvements of the kinetic correlation for B 4C oxidation were obtained from analytical experiments performed at FzK (Germany) and IRSN (France), mostly in the temperature range above 1400 K. The new modelling was considered in the analysis of experiments involving a B 4C control rod in small fuel rod assemblies, such as Phebus FPT3 in-pile experiment, as well as out of pile experiments Quench 07 and 09, aimed at studying the course of severe accidents. Regarding the hydrogen generation, the results given by different code simulations are consistent with the experimental values. Concerning the control rod degradation, SA codes such as ICARE2 and ATHLET-CD, using suitable modelling of B 4C oxidation, predicted rather well the total carbon release. The results of the MELCOR code, involving initially a B 4C oxidation model designed to be used for BWR control blades, have been largely improved in the most recently released version, with a model extended for PWR B 4C control rods. Codes still have some difficulties to reproduce the final degradation of fuel bundles involving B 4C rods. Spreading of molten materials from the control rods onto fuel rods of the bundle is suspected, suggesting that the main effect of the B 4C control rod materials on the bundle behaviour during degradation is connected with B 4C-Stainless Steel (SS) eutectics formation and B 4C-SS-Zry liquid mixture relocation. These phenomena are not accounted for in the SA codes. The need for further code developments of the early phase of core degradation is recognized, involving the absorber rod material behaviour. The BECARRE experiments, on-going in the framework of the International Source Term Program, are designed to provide in-depth understanding of these phenomena and help improving their modelling.

Improvement of Core Modeling in ICARE/CATHARE: Application to the Calculation of a Six-Inch-Break LOCA Leading to a Severely Degraded Situation

In order to analyze the course of a hypothetical severe accident, the French “Institut de Radioprotection et de Sûreté Nucléaire” in the last decade has developed computer codes that have been extensively used for supporting the Level 2 Probabilistic Safety Assessment (PSA2) and, in general, for the safety analysis of French pressurized water reactors (PWRs).In particular, the computer code ICARE/CATHARE V1 is a tool that has been widely validated and intensively used within the framework of the PSA2 of the 900-MW(electric) French PWR. This code has been tested on many accident scenarios, and the results obtained have been considered to be satisfactory and reliable up to the end of the early degradation phase. But, severe accidents in PWRs are characterized by a continuous evolution of the core geometry 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 multidimensional flows and heat transfers. So, the lack of a multidimensional two-phase thermal-hydraulic model appeared to be prejudicial to achieve best-estimate reactor studies with ICARE/CATHARE V1 in the case of large core blockages and/or in the case of large cavity appearance. In accordance, a full multidimensional modeling (covering both the fluid flow and the corium behavior) was developed and introduced in a new ICARE/CATHARE version referenced as V2, which includes two options for the thermal-hydraulic modeling: either one-dimensional (1D) or two-dimensional (2D).The first part of this paper demonstrates that without activating the new V2 models, ICARE/CATHARE V2(1D) is able to reproduce the results obtained with ICARE/CATHARE V1 on the basis of a 6-in.-break loss-of-coolant accident. Then, in order to illustrate some of the new V2 modeling improvements, the last part is focused on the results obtained with ICARE/CATHARE V2(2D), and a preliminary comparison is made with ICARE/CATHARE V2(1D).This 1D-2D comparison points out in particular the important role that could be played in the course of a severe accident by the multidimensional flow pattern.

Radioactive release from VVER-1000 reactors after a terror attack

One of the terror scenarios for nuclear power plants is a severe damage of the reactor containment caused by a plane crash or a missile. Owing to the loss of electric power the cooling of the core would not be maintained, leading to a core melt accident. Normally in the course of severe accidents an intact containment has the ability to retain a large part of the radioactive inventory. The goal of this work is the investigation of the behaviour of the radioactive releases from a VVER-1000 reactor during a severe accident with a large containment leak from the beginning of the accident. The results are compared with the release in a severe accident via a very small leakage due to the looseness of the containment.

Ergodynamics in the Reliability of Power Plant Operators and Prospective Hybrid Intelligence Systems.

Based on ergodynamics and the hybrid intelligence theory, an analysis of the nuclear power plant operator's performance is given at the levels of strategies, tactics, and actions. Special attention is paid to the strategies used in the course of severe accidents at nuclear power plants. Data from Ukrainian and Russian power plants and training centres, and from accidents around the world were collected and processed. It is shown that in an emergency it is essential for the human operator to be flexible. This flexibility includes two main training and personal factors: a large set of strategies and tactics the operator manages to use, and quick transformations between the strategies (tactics). It was also found that some emergency tasks are too complicated: They require simultaneous use of different strategies, with time strictly limited by nuclear power plant dynamics. Those tasks cannot be successfully solved by any individual operator. Hybrid intelligence systems involving different specialists should be used in those cases in order to avoid failures in emergency problem solving and macroergonomic organizational design.

LWR behavior under severe core damage conditions

The accident in the Three Mile Island-2 (TMI-2) Reactor has clearly demonstrated that severe accidents can be terminated successfully also far beyond the regulatory guidelines. It also has shown, that one needs to be prepared to analyze them on a non-hypothetical basis. As a result, various experimental and theoretical programs have been initiated to analyze the fuel rod and core behavior during degraded core cooling conditions. During the course of severe accidents in light water reactors, the heat transfer conditions between the fuel rods or steam generator tubes and the fluid are mainly affected by the uncovery of the fuel rods, tubes or the dryout of the flow channels. For the case of the uncovery transients, a relevant parameter is the level of the two-phase mixture. The height of the so-called mixture level is generally marked as a sharp increase of the void fraction and the surface temperature. After uncovery, the fuel rods will heat-up because of the relatively low heat transfer rates from the rods to the steam. At temperatures above approximately 1500 K, the metal-water reaction predominates the fuel rod behavior. It also reduces the steam to hydrogen ratio, which possibly may limit the chemical reaction. The processes described are typical for the situations analyzed in the frame of risk studies. Especially in small break accidents and transients, the system response on core behavior is substantial. The analyses contribute to the following questions: • - Find the minimum availability of safety components to avoid uncoolable conditions. Calculations with the severe accident code MELSIM3 have bound the degree of core uncovery which leads to coolable core conditions [1,2]. The core temperatures of the TMI-2 reactor were much higher than the regulatory limits. • - Determine the amount of time available for operators to take actions to bring water into a reactor during a severe accident. • - Provide a better understanding of the plant behavior leading to severe core damage conditions. • - Improve judgement and reduce conservatism in the risk assessment.

Numerical Modeling Of Heat-up And MeltingProcesses In Internally Heated Debris Bed AndReactor Vessel Lower Plenum

The paper considers the two-dimensional thermal processes in internally heated oxidic debris beds in the lower plenum of a light water reactor vessel during the course of severe accidents with core meltdown and relocation. Heat transfer and phase change processes in debris beds and vessel lower head wall are investigated numerically with a particular emphasis on the dynamic characteristics of thermal transients in question. The main purpose of the work is to develop an appropriate modeling approach, which enables the analysis of complicated physical processes (involving dynamic changes of bed porosity, anisotropic heat transfer in remelting beds with internal energy sources) in complex domains of debris bed and vessel lower head wall. The natural convection heat transfer in the forming melt pools are described by means of an originally developed effective conductivityconvectivity approach. A multi-block control-volume numerical scheme for two-dimensional cartesian or cylindrical non-orthogonal geometries has been developed and applied to predict heat transfer in the solid or porous core debris beds and forming melt pools.