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.

Backbone: A Multiphysics Framework for Coupling Nuclear Codes Based on CORBA and MPI

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Yu Liu ◽  
Michael Nishimura ◽  
Marat Seydaliev ◽  
Markus Piro
Keyword(s):  
Pilot Study ◽  
Computer Simulations ◽  
Message Passing ◽  
Message Passing Interface ◽  
Nuclear Reactor ◽  
Neutron Transport ◽  
Temporal Integration ◽  
Heuristic Method ◽  
Reactor Performance ◽  
Time Step

Recent trends in nuclear reactor performance and safety analyses increasingly rely on multiscale multiphysics computer simulations to enhance predictive capabilities by replacing conventional methods that are largely empirically based with a more scientifically based methodology. Through this approach, one addresses the issue of traditionally employing a suite of stand-alone codes that independently simulate various physical phenomena that were previously disconnected. Multiple computer simulations of different phenomena must exchange data during runtime to address these interdependencies. Previously, recommendations have been made regarding various approaches for piloting different design options of data coupling for multiphysics systems (Seydaliev and Caswell, 2014, “CORBA and MPI Based “Backbone” for Coupling Advanced Simulation Tools,” AECL Nucl. Rev., 3(2), pp. 83–90). This paper describes progress since the initial pilot study that outlined the implementation and execution of a new distribution framework, referred to as “Backbone,” to provide the necessary runtime exchange of data between different codes. The Backbone, currently under development at the Canadian Nuclear Laboratories (CNL), is a hybrid design using both common object request broker architecture (CORBA) and message passing interface (MPI) systems. This paper also presents two preliminary cases for coupling existing nuclear performance and safety analysis codes used for simulating fuel behavior, fission product release, thermal hydraulics, and neutron transport through the Backbone. Additionally, a pilot study presents a few strategies of a new time step controller (TSC) to synchronize the codes coupled through the Backbone. A performance and fidelity comparison is presented between a simple heuristic method for determining time step length and a more advanced third-order method, which was selected to maximize configurability and effectiveness of temporal integration, saving time steps and reducing wasted computation. The net effect of the foregoing features of the Backbone is to provide a practical toolset to couple existing and newly developed codes—which may be written in different programming languages and used on different operating systems—with minimal programming effort to enhance predictions of nuclear reactor performance and safety.

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.

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

scholarly journals Perspectives on a Severe Accident Consequences—10 Years after the Fukushima Accident

2021 ◽  
Vol 2 (4) ◽  
pp. 398-411
Author(s):  
Jinho Song
Keyword(s):  
Nuclear Reactor ◽  
Severe Accident ◽  
Accident Analysis ◽  
Fukushima Accident ◽  
The Public ◽  
Analysis Methodology ◽  
International Attention ◽  
Nuclear Reactor Safety ◽  
Term Monitoring

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.

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

EU Funded Research Activities on NPPs Operational Safety

2002 ◽  
Author(s):  
P. Manolatos ◽  
G. Van Goethem
Keyword(s):  
Nuclear Reactor ◽  
Research Programme ◽  
Severe Accident ◽  
Life Extension ◽  
The European Union ◽  
Research Activities ◽  
Plant Life ◽  
Accident Management ◽  
Operational Safety ◽  
Nuclear Reactor Safety

The 5th framework programme (FP-5), the pluriannual research programme of the European Union (EU), covers the period 1998–2002. Research on nuclear energy, fusion and fission, is covered by the Euratom part of the FP-5. An overview of the Euratom’s research on Nuclear Reactor Safety, managed by the DG-RTD of the European Commission (EC), is presented. This concerns 70 multi-partner projects of approximately € 82.5 million total contract value that have been selected and co-financed during the period 1999–2001. They form the three clusters of projects dealing with the “Operational Safety of Existing Installations”. “Plant Life Extension and Management” (PLEM), “Severe Accident Management” (SAM) and “Evolutionary concepts” (EVOL). Emphasis is given here to the projects of the PLEM cluster.

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