Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux Model
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G. Baghban
Abstract
In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.
Kurzfassung
Im Hinblick auf die Bedeutung der Untersuchung des Transientenverhaltens in einem Reaktor ist dieser Beitrag der thermohydraulischen Analyse von transienten Strömungsvorgängen in einem WWER-1000-Reaktor gewidmet. Eine Reihe von geeigneten mathematischen und physikalischen Modellen auf der Grundlage des Drift-Flux-Modells wurde angewendet. Basierend auf einem Multikanal-Ansatz wurde der Reaktorkern in verschiedene Regionen eingeteilt mit jeweils unterschiedliche Charakteristika, repräsentiert durch einen Brennstab mit dem dazugehörigen Kühlkanal. Geeignete Anfangs- und Randbedingungen wurden betrachtet und 2 Szenarien mit Abschaltung von 4 und 2 Hauptkühlmittelpumpen bei abgeschaltetem Reaktor und zusätzlich ein Szenario mit Abschaltung aller 4 Pumpen bei nicht abgeschaltetem Reaktor wurden analysiert. Für jede Transiente wurde ein vollständiger Satz thermohydraulischer Parameter erhalten. Zur Verifikation des vorgeschlagenen Modells wurden die Ergebnisse mit denen des RELAP5/MOD3-Codes und dem Sicherheitsbericht (FSAR) des Kernkraftwerks Bushehr verglichen. Für die Transienten wurde eine gute Übereinstimmung der Ergebnisse erhalten.
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© 2017, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- CANDU pressure tube leak detection by annulus gas dew point measurement: a critical review
- Multiple regression approach to predict turbine-generator output for Chinshan nuclear power plant
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- Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor
- Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux Model
- Sensitivity analysis for CORSOR models simulating fission product release in LOFT-LP-FP-2 severe accident experiment
- Analysis of the optimal fuel composition for the Indonesian experimental power reactor
- Radiogenic lead from poly-metallic thorium ores as a valuable material for advanced nuclear facilities
- The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident
- Font Attributes based Text Steganographic algorithm (FATS) for communicating images: A nuclear power plant perspective
- Size control synthesis and characterization of ZnO nanoparticles and its application as ZnO-water based nanofluid in heat transfer enhancement in light water nuclear reactor
- Nuclear characteristics of epoxy resin as a space environment neutron shielding
- Exact solution of the neutron transport equation in spherical geometry
- Technical Notes/Technische Mitteilungen
- Determination of self-attenuation correction factor for lichen samples by using gamma-ray spectrometry
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- CANDU pressure tube leak detection by annulus gas dew point measurement: a critical review
- Multiple regression approach to predict turbine-generator output for Chinshan nuclear power plant
- 10.3139/124.110675
- Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor
- Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux Model
- Sensitivity analysis for CORSOR models simulating fission product release in LOFT-LP-FP-2 severe accident experiment
- Analysis of the optimal fuel composition for the Indonesian experimental power reactor
- Radiogenic lead from poly-metallic thorium ores as a valuable material for advanced nuclear facilities
- The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident
- Font Attributes based Text Steganographic algorithm (FATS) for communicating images: A nuclear power plant perspective
- Size control synthesis and characterization of ZnO nanoparticles and its application as ZnO-water based nanofluid in heat transfer enhancement in light water nuclear reactor
- Nuclear characteristics of epoxy resin as a space environment neutron shielding
- Exact solution of the neutron transport equation in spherical geometry
- Technical Notes/Technische Mitteilungen
- Determination of self-attenuation correction factor for lichen samples by using gamma-ray spectrometry
