The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system
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M. Günay
Abstract
In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF4, 90 % Flibe-10 % UF4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion–fission hybrid reactor system. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.
Kurzfassung
Bei dieser Studie wurde unter Anwendung der Fluide 100 % Flibe, 90 % Flibe-10 % ThF4, 90 % Flibe-10 % UF4, der Nukleardaten-Bibliotheken ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 und der Baumaterialien Ferritic Steel, 9Cr2WVTa, V4Cr4Ti und SiC eine hybride Reaktor-Anlage konzipiert. Bei der ersten Flüssigwand, der zweiten Flüssigwand (Blanket) und den Panzer-Bereichen des hybriden Fissions-Reaktors wurden Fluide eingesetzt. Die nuklearen Parameter des hybriden Fusion-/Fissions-Reaktors, wie z.B. das Tritium-Brutverhältnis (TBR), der Energieanreicherungs-Faktor (M) und die Wärmeenergie, wurden an der ersten Flüssigwand, am Blanket und am Panzer-Bereich berechnet. Mit Hilfe der aktuellen Version des MCNPX-2.7.0 Monte Carlo-Codes wurden dreidimensionale nukleonische Berechnungen durchgeführt.
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© 2015, Carl Hanser Verlag, München
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Theoretical study of steam condensation induced water hammer phenomena in horizontal pipelines
- Estimation of experimental uncertainty for physical measurements based on the start-up data of the latest VVER-1000 units
- Analysis of SBO ATWS for Maanshan PWR
- Subchannel analysis of Al2O3 nanofluid as a coolant in VMHWR
- The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system
- Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW
- International assessment of application of the Code of Conduct on the Safety of Research Reactors
- 15 MeV proton irradiation effects on Bi-based high temperature superconductors
- Estimation of radiation damage of iron by a reactor gamma spectrum
- Measuring U concentration in solution product of UF6 hydrolysis using a gamma ray densitometer
- Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask
- Application of UN method to neutron transport equation in slab geometry using HG phase function
- Preparation of human resources for future nuclear energy using FBNR as the instrument of learning
- Technical Note
- Interface network groups
