Development and Application of Multi-Physics Safety Analysis Code for Advanced Liquid Metal Cooled Reactor

2018 ◽  
Author(s):  
Chi Wang ◽  
Xuebei Zhang ◽  
Jingchao Feng ◽  
Muhammad Shehzad Khan ◽  
Minyou Ye ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Coolant Temperature ◽  
Medium Size ◽  
High Tech ◽  
Fuel Temperature ◽  
Flow Phenomena ◽  
Nuclear Reactor Safety ◽  
Cfd Method ◽  
Point Kinetics

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.

CORBA AND MPI-BASED “BACKBONE” FOR COUPLING ADVANCED SIMULATION TOOLS

2014 ◽  
Vol 3 (2) ◽  
pp. 83-90 ◽  
Author(s):  
M. Seydaliev ◽  
D. Caswell
Keyword(s):  
Fission Product ◽  
Computer Simulations ◽  
Message Passing ◽  
Message Passing Interface ◽  
Nuclear Reactor ◽  
Atmospheric Dispersion ◽  
Safety Analysis ◽  
Severe Accident ◽  
Nuclear Reactor Safety ◽  
Exchange Data

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.

Application of a new drag coefficient model at CFD-simulations on free surface flows relevant for the nuclear reactor safety analysis

2012 ◽  
Vol 39 (1) ◽  
pp. 70-82 ◽  
Author(s):  
Deendarlianto ◽  
Thomas Höhne ◽  
Pavel Apanasevich ◽  
Dirk Lucas ◽  
Christophe Vallée ◽  
...  
Keyword(s):  
Free Surface ◽  
Drag Coefficient ◽  
Nuclear Reactor ◽  
Safety Analysis ◽  
Free Surface Flows ◽  
Reactor Safety ◽  
Cfd Simulations ◽  
Surface Flows ◽  
Nuclear Reactor Safety

scholarly journals Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

2013 ◽  
Author(s):  
Dan Maljovec ◽  
Bei Wang ◽  
Valerio Pascucci ◽  
Peer-Timo Bremer ◽  
Diego Mandelli ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Reactor Safety ◽  
Nuclear Reactor Safety ◽  
Geometric Techniques

scholarly journals Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

2006 ◽  
Author(s):  
Richard W. Johnson ◽  
Richard R. Schultz ◽  
Patrick J. Roache ◽  
Ismail B. Celik ◽  
William D. Pointer ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Safety Analysis ◽  
Reactor Safety ◽  
Nuclear Reactor Safety

Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 419-436
Author(s):  
R. Kianpour ◽  
G. R. Ansarifar
Keyword(s):  
Neural Network ◽  
Nuclear Reactor ◽  
Volume Fraction ◽  
Reactor Core ◽  
Dynamical Analysis ◽  
Coolant Temperature ◽  
Optimal Size ◽  
Fuel Temperature ◽  
Volume Fractions ◽  
Reactivity Coefficients

Abstract The purpose of this study is to display the neutronic simulation of nanofluid application to reactor core. The variations of VVER-1000 nuclear reactor primary neutronic parameters are investigated by using different volume fraction of nanofluid as coolant. The effect of using nanofluid as coolant on reactor dynamical parameters which play an important role in the dynamical analysis of the reactor and safety core is calculated. In this paper coolant and fuel temperature reactivity coefficients in a VVER-1000 nuclear reactor with nanofluid as a coolant are calculated by using various volume fractions and different sizes of TiO2 (Titania) nanoparticle. For do this, firstly the equivalent cell of the hexagonal fuel rod and the surrounding coolant nanofluid is simulated. Then the thermal hydraulic calculations are performed at different volume fractions and sizes of the nanoparticle. Then, using WIMS and CITATION codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated. For doing optimization, an artificial neural network is trained in MATLAB using the observed data. The different sizes and various volume fractions are inputs, fuel and coolant temperature reactivity coefficients are outputs. The optimal size and volume fraction is determined using the neural network by implementing the genetic algorithms. In the optimization, volume fraction of 7% and size 77 nm are optimal values.

A Study on Doppler Weighting Factor for Control Element Assembly Ejection Accident by Using Newly Developed Nuclear Design Code and Non-LOCA Methodology

2013 ◽  
Author(s):  
Kyungmin Yoon ◽  
Chansu Jang ◽  
Jooil Yoon
Keyword(s):  
Safety Analysis ◽  
Weighting Factor ◽  
Time Effect ◽  
Kinetics Model ◽  
Control Element ◽  
Fuel Temperature ◽  
Design Code ◽  
The Core ◽  
Point Kinetics ◽  
Reactivity Insertion

Among Reactivity Initiated Accidents (RIAs) for Pressurized Water Reactor (PWR), Control Element Assembly Ejection (CEAE) accident causes the rapid positive reactivity insertion to the core. It causes an asymmetric power distortion which results in the rising of local fuel temperature, fuel pellet thermal expansion and cladding ballooning or rupture. In the CEAE accident, Doppler feedback has a profound effect because the negative reactivity insertion due to the rise of fuel temperature reduces the core power after rapid power excursion. But the Doppler reactivity can’t be calculated properly in the safety analysis code, using point kinetics model, because the point kinetics model is not able to consider spatial-time effect of the sudden rise in local fuel temperature on Doppler feedback calculation during CEAE accident. And then the excessively high core power which results from the underestimated Doppler feedback would make more severe results such as PCMI fuel failure, fuel cladding rupture and serious DNB fuel failure. Therefore, Doppler Weighting Factor (DWF) is needed for the safety analysis of CEAE accident to compensate a missing spatial-time effect on Doppler feedback calculation. In this study, the adequacy of the application of DWF for APR1400 was evaluated by using nuclear design code called ASTRA (Advanced Static and Transient Reactor Analyzer)[1] and a methodology called ISAM (Integrated Safety Analysis Methodology)[2]. ASTRA is the 3D nuclear design code newly developed by KNF and has various functions such as the static core design, the transient core analysis and the operational support. ISAM is the methodology which is newly developed by KNF to perform the Non-LOCA safety analysis by using RETRAN[3] code which is widely used in the transient analysis and based on the point kinetics model.

scholarly journals Model of Reactivity Accident of the RBMK-1000 of the Chornobyl NPP 4th Power Unit

2021 ◽  
Vol 21 (2) ◽  
pp. 39-48
Author(s):  
V. I. Borysenko ◽  
◽  
V. V. Goranchuk ◽  
Keyword(s):  
Power Unit ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Fuel Temperature ◽  
Reactor Accident ◽  
Reactivity Coefficients ◽  
Accident Scenarios ◽  
Control Rods ◽  
Reactivity Accident

The article presents the results of modeling of the reactivity accident, which resulted in the destruction of reactor RBMK-1000 of the 4th power unit of the Chornobyl NPP on April 26, 1986. The RBMK-1000 reactivity accident model was developed on the basis of the kinetics of the nuclear reactor, taking into account the change in the reactivity of the reactor. Reactivity changes as a result of both external influence (movement of control rods; change in the reactor inlet coolant temperature (density)) and due to the action of reactivity feedback by the parameters of the reactor core (change in the fuel temperature, coolant temperature, concentration of 135Хе, graphite stack temperature, etc.). A similar approach was applied by the authors of the article for the study of transient processes with the operation of accelerated unit unloading mode on VVER-1000, and the validity of such model is confirmed. The study of the reactivity accident on RBMK-1000 was carried out for various combinations of values of the effectiveness of control rods; reactivity coefficients of the coolant temperature and fuel temperature; changes in the temperature of the coolant at the inlet to the reactor. In most of the studied RBMK-1000 reactor accident scenarios, the critical values of fuel enthalpy, at which the process of fuel destruction begins, are reached first. An important result of the research is the conclusion that it is not necessary to reach supercriticality on instantaneous neutrons, supercriticality on delayed neutrons is also sufficient to initiate fuel destruction.

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