New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
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Abstract
The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.
Kurzfassung
Der vom Helmholtz-Zentrum Dresden-Rossendorf entwickelte Reaktordynamik-Code-DYN3D wird derzeit auf die Anwendung für schnelle Reaktoren erweitert. Dieser Artikel gibt einen Überblick über die neue Version von DYN3D, die für Kernrechnungen von natriumgekühlten Reaktoren angewendet werden soll. Der vorliegende Artikel beschreibt kurz die neu implementierten Modelle für die Modellierung der thermischen Ausdehnungseffekte des Reaktorkerns. Weiterhin ist die Methodik der Erzeugung von homogenisierten Wenig-Gruppenwirkungsquerschnitten für natriumgekühlte schnelle Reaktoren zusammengefasst. Die durchgeführten und geplanten Verifikations- und Validierungsstudien werden kurz vorgestellt. Zur Vervollständigung des Überblicks werden die entsprechenden ausführlichen Publikationen kurz dargestellt.
References
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© 2017, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- SIMULATE-3 K coupled code applications
- Application of the new IAPWS Guideline on the fast and accurate calculation of steam and water properties with the Spline-Based Table Look-Up Method (SBTL) in RELAP-7
- Simulation of water hammer phenomena using the system code ATHLET
- New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses
- Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR: verification of the simulation results
- Calculation of the fuel temperature coefficient of reactivity considering non-uniform radial temperature distribution in the fuel rod
- The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes
- Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes
- Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR
- A comparison study for mass attenuation coefficients of some amino acids using MCNP code
- Validation of radioactive isotope activity measurement in homogeneous waste drum using Monte Carlo codes
- Study of the response reduction of LiF:Mg, Ti dosimeter for high dose dosimetry
- Non-contact micro mass evaluation method using an X-ray microscope