Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code
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M. Günay
Abstract
In this study, the molten salt-heavy metal mixtures 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion–fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.
Kurzfassung
In dieser Studie wurden Flüssigsalz-Schwermetall-Mischungen aus 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % UCO als Flüssigkeiten verwendet. Die Flüssigkeiten wurden an der ersten Flüssigwand, am Blanket- und Shield-Bereich des konzipierten hybriden Reaktorsystems eingesetzt. Als strukturelles Baumaterial wurde vier Zentimeter dicker 9Cr2WVTa ferritischer Stahl verwendet. In dieser Studie wurde untersucht, welche Einflüsse die Mischungsbestandteile auf den Neutronenfluss im konzipierten hybriden Reaktorsystem haben. Ausgehend von den Mischungsbestandteilen, der Verteilung der radialen Strömung und dem Energiespektrum wurde der Neutronenfluss im System berechnet. Dreidimensionale Analysen wurden mit Hilfe der neuesten Version des Monte Carlo Strahlungstransportcodes MCNPX-2.7.0 und der nuklearen Datenbibliothek ENDF/B-VII.0 gemacht.
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© 2014, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- A model-based approach to operational event groups ranking
- Application of the method for uncertainty and sensitivity evaluation to results of PWR LBLOCA analysis calculated with the code ATHLET. Part 2: sensitivity analysis
- Statistical Large-Break LOCA analysis for PWRs with combined ECC injection
- Analysis of the CSF model for simulated loss of coolant accident conditions
- Performance of two fluid model based numeric tool with pressure and enthalpy as independent variables for simulation of two phase flow in axial and radial direction
- Pre-decommissioning complex engineering and radiation inspection of the WWR-M reactor
- Analysis of the source term formation in a severe accident initiated by end fitting failure in CANDU type reactors
- Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code
- Analytical investigation of lignite and its ash samples taken from the Afşin-Elbistan coal basin in Turkey
- Explicit formulation of a nodal transport method for discrete ordinates calculations in two-dimensional fixed-source problems
- T1 and U1 approximations to neutron transport equation in one-dimensional spherical geometry
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- A model-based approach to operational event groups ranking
- Application of the method for uncertainty and sensitivity evaluation to results of PWR LBLOCA analysis calculated with the code ATHLET. Part 2: sensitivity analysis
- Statistical Large-Break LOCA analysis for PWRs with combined ECC injection
- Analysis of the CSF model for simulated loss of coolant accident conditions
- Performance of two fluid model based numeric tool with pressure and enthalpy as independent variables for simulation of two phase flow in axial and radial direction
- Pre-decommissioning complex engineering and radiation inspection of the WWR-M reactor
- Analysis of the source term formation in a severe accident initiated by end fitting failure in CANDU type reactors
- Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code
- Analytical investigation of lignite and its ash samples taken from the Afşin-Elbistan coal basin in Turkey
- Explicit formulation of a nodal transport method for discrete ordinates calculations in two-dimensional fixed-source problems
- T1 and U1 approximations to neutron transport equation in one-dimensional spherical geometry
