Loss of flow Accident (LOFA) analyses using LabView-based NRR simulator
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A. Arafa
Abstract
This paper presents a generic Loss of Flow Accident (LOFA) scenario module which is integrated in the LabView-based simulator to imitate a Nuclear Research Reactor (NRR) behavior for different user defined LOFA scenarios. It also provides analyses of a LOFA of a single fuel channel and its impact on operational transactions and on the behavior of the reactor. The generic LOFA scenario module includes graphs needed to clarify the effects of the LOFA under study. Furthermore, the percentage of the loss of mass flow rate, the mode of flow reduction and the start time and transient time of LOFA are user defined to add flexibility to the LOFA scenarios. The objective of integrating such generic LOFA module is to be able to deal with such incidents and avoid their significant effects. It is also useful in the development of expertise in this area and reducing the operator training and simulations costs. The results of the implemented generic LOFA module agree well with that of COBRA-IIIC code and the earlier guidebook for this series of transients.
Kurzfassung
Dieser Beitrag beschreibt ein Modul für ein Loss-of-Flow-Accident-(LOFA)-Szenario, eingebunden in einen LabView-basierten Simulator zur Untersuchung des Verhaltens eines Forschungsreaktors unter verschiedenen Anwender-definierten LOFA-Szenarien. Das Modul beinhaltet auch graphische Darstellungen zur Verdeutlichung der Auswirkungen der untersuchten Strömungsstörfälle. Außerdem sind der Anteil der Massenströmungsrate, die Art der Strömungsreduzierung, sowie Startzeit und Ausregelzeit des LOFA-Anwender-definiert zur Erhöhung der Flexibilität bei verschiedenen LOFA-Szenarien. Ziel der Einbindung solch generischer LOFA-Module ist es, mit dieser Art Störfälle fertig zu werden und ihre Auswirkungen zu vermeiden. Sie sind außerdem wichtig bei der Kompetenzentwicklung auf diesem Gebiet. Die Ergebnisse der Untersuchungen mit Hilfe der LOFA-Module stimmen gut überein mit den Ergebnissen des COBRA-IIIC-Codes.
References
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© 2016, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Use of molybdenum as a structural material of fuel elements for improving nuclear reactors safety
- Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code
- Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling
- Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor
- Dependence of neutron rate production with accelerator beam profile and energy range in an ADS-TRIGA RC1 reactor
- Effects of the wallpaper fuel design on the neutronic behavior of the HTR-10
- Loss of flow Accident (LOFA) analyses using LabView-based NRR simulator
- Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies
- Simulation of polycarbonate-CNT nanocomposite dosimeter based on electrical characteristics
- Thermoluminescence properties of micro and nano structure hydroxyapatite after gamma irradiation
- Equilibrium based analytical model for estimation of pressure magnification during deflagration of hydrogen air mixtures
- Polynomial approach method to solve the neutron point kinetics equations with use of the analytic continuation
- The slab albedo problem for the triplet scattering kernel with modified FN method
- Calculation of the fuel composition and the deterministic reloading pattern in the second cycle of the BUSHEHR VVER-1000 reactor using the weighting factor method
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Use of molybdenum as a structural material of fuel elements for improving nuclear reactors safety
- Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code
- Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling
- Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor
- Dependence of neutron rate production with accelerator beam profile and energy range in an ADS-TRIGA RC1 reactor
- Effects of the wallpaper fuel design on the neutronic behavior of the HTR-10
- Loss of flow Accident (LOFA) analyses using LabView-based NRR simulator
- Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies
- Simulation of polycarbonate-CNT nanocomposite dosimeter based on electrical characteristics
- Thermoluminescence properties of micro and nano structure hydroxyapatite after gamma irradiation
- Equilibrium based analytical model for estimation of pressure magnification during deflagration of hydrogen air mixtures
- Polynomial approach method to solve the neutron point kinetics equations with use of the analytic continuation
- The slab albedo problem for the triplet scattering kernel with modified FN method
- Calculation of the fuel composition and the deterministic reloading pattern in the second cycle of the BUSHEHR VVER-1000 reactor using the weighting factor method
