Lumped-parameter simulations of wall condensation and sump evaporation under typical thermal-hydraulic conditions of nuclear reactor containment severe accident

2014 ◽  
Vol 77 ◽  
pp. 11-19 ◽  
Author(s):  
J. Malet ◽  
T. Gélain
2013 ◽  
Vol 34 (4) ◽  
pp. 257-266
Author(s):  
Magdalena Orszulik ◽  
Adam Fic ◽  
Tomasz Bury ◽  
Jan Składzień

Abstract Passive autocatalytic recombiners (PAR) is the only used method for hydrogen removal from the containment buildings in modern nuclear reactors. Numerical models of such devices, based on the CFD approach, are the subject of this paper. The models may be coupled with two types of computer codes: the lumped parameter codes, and the computational fluid dynamics codes. This work deals with 2D numerical model of PAR and its validation. Gaseous hydrogen may be generated in water nuclear reactor systems in a course of a severe accident with core overheating. Therefore, a risk of its uncontrolled combustion appears which may be destructive to the containment structure.


2012 ◽  
Vol 2012 ◽  
pp. 1-7 ◽  
Author(s):  
Pavan K. Sharma ◽  
B. Gera ◽  
R. K. Singh ◽  
K. K. Vaze

In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.


Author(s):  
HaoMin Sun ◽  
Shinichi Machida ◽  
Yasuteru Sibamoto ◽  
Yuria Okagaki ◽  
Taisuke Yonomoto

During a severe accident of a nuclear reactor, radioactive aerosols may be released from degraded nuclear fuels. Pool scrubbing is one of the efficient filters with a high aerosol removal efficiency, in other words a high decontamination factor (DF). Because of its high performance, many pool scrubbing experiments have been performed and several pool scrubbing models have been proposed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption. In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations under the same boundary conditions. The test section is a transparent polycarbonate pipe with an inner diameter of 0.2 m. 0.5 μm SiO2 particles were used as aerosols. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was not because of our experimental system error including measurement instruments but a real phenomenon of the pool scrubbing.


2021 ◽  
Vol 2 (4) ◽  
pp. 398-411
Author(s):  
Jinho Song

Scientific issues that draw international attention from the public and experts during the last 10 years after the Fukushima accident are discussed. An assessment of current severe accident analysis methodology, impact on the views of nuclear reactor safety, dispute on the safety of fishery products, discharge of radioactive water to the ocean, status of decommissioning, and needs for long-term monitoring of the environment are discussed.


Author(s):  
Heriberto Sánchez-Mora ◽  
Carlos Chávez-Mercado ◽  
Chris Allison ◽  
Judith Hohorst

RELAP/SCDAPSIM is a nuclear reactor simulator and accident analysis code that has been used in the nuclear energy industry for many years. Currently, Innovative Systems Software is developing a new tool that will show the behavior of the core components during a simulation of an accident. The addition of contour plots for the SCDAP components showing different properties: temperature, hydrogen production, etc. will allow a better understanding of core behavior during a severe accident in a nuclear reactor. The SCDAP components are fuel rods, electrically heated simulator rods, such as those used in the CORA experiments, control rods, a shroud and a BWR blade/box. This paper describes the progress in the development of the contour plot tool based on the OpenGL and FORTRAN90 libraries. The purpose of this tool is help to the user analyze the simulation of an accident and to debug an input file.


Author(s):  
Zhong Lei ◽  
Jian Deng ◽  
Wei Li ◽  
Xiaoli Wu ◽  
Chunrui Deng

Abstract Core melting and molten migration behavior are hot and difficult issues in the field of nuclear reactor severe accident research. The Moving Particle Semi-implicit (MPS) meshless method has potential to simulate free-surface and multiphase flows. In this study, the MPS method was utilized to simulate the melting process of UO2-Zr rod-type fuel elements. The models of heat conduction with phase change, simplified UO2-Zr eutectic reaction, viscous flow and surface tension were implemented with the framework of standard MPS method. Then, the improved MPS code was used to simulate and analyze the process of high-temperature melting and characteristics of molten migration and solidification in the coolant channel, aiming at revealing the severe accidents for light water reactors (LWR), particularly the early core damage. The results showed that compared with the case of higher initial temperature, when the initial temperature of molten UO2 is lower, more molten UO2 will solidify on the surface of rod cluster, and the blockage of upper flow channel caused by molten UO2 is more serious. In addition, this study also demonstrated the potential of the MPS method for the study of complicated severe accident phenomena in not only traditional LWR but also advanced nuclear reactors in the future.


Author(s):  
F. L. Cho

This paper reveals a paradigm of analyzing the consequential effects of severe nuclear reactor accident, radionuclides fraction and source terms release, that will influence the MACCS2 codification [1], by coupling with the results of SAPHIA-PSA Levels l & 2 quantification process [2], MELCORE [3], STCP [4], PST [5], and XSOR [6]. Those codes are mutually exclusive and useful. However, it lacks of the closed interface and linkage for addressing Plant Damage States (PDS), Severe Accident Sequences, and Risk Consequence. Thus, it is imperative to formulate the consistent baseline information for MACCS2, PSA Levels 1, 2 and 3, and then linking to a new algorithm of NCM.


2014 ◽  
Vol 3 (2) ◽  
pp. 83-90 ◽  
Author(s):  
M. Seydaliev ◽  
D. Caswell

There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the “Backbone,” to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different CORBA and MPI configurations examined to date, as well as the preliminary configuration selected for coupling 2 existing safety analysis programs used for modeling thermal–mechanical fuel behavior and fission product behavior respectively. In addition, preliminary work in hosting both the Backbone and the associated safety analysis programs in a cluster environment are discussed.


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