Investigation of (n,γ) reaction in hybrid reactor zones
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M. Günay
Abstract
In this study, the fluids were composed with increased mole fractions of a mixture of molten salt: heavy metals 99 – 95 % Li20Sn80−1 – 5 % SFG-Pu, 99 – 95 % Li20Sn80−1 – 5 % SFG-PuF4, 99 – 95 % Li20Sn80−1 – 5 % SFG-PuO2. In this study, the effect on conversion of each isotope (238–242Pu) in spent fuel grade plutonium by (n,γ) reactions was investigated in liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) is the neutron multiplier by (n,2n) reactions. The Be zone used was 3 cm thick. 9Cr2WVT, a ferritic steel, is used as structural material. Three-dimensional nucleonic calculations were performed by using the most recent versions of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.0.
Kurzfassung
In dieser Studie wurden mit den erhöhten Mol-Anteilen der Salz-Schwermetallmischungen 99 – 95 % Li20Sn80−1 – 5 % SFG-Pu, 99 – 95 % Li20Sn80−1 – 5 % SFG-PuF4, 99 – 95 % Li20Sn80−1 – 5 % SFG-PuO2 Fluide hergestellt. In dieser Arbeit wurde der Einfluß eines jeden Isotops (238–242Pu) auf die Umwandlung in Spent-Fuel-Grade Plutonium mit Hilfe der (n,γ) Reaktion in der ersten Flüssigwand, im Blankett und in den Schutzzonen des im Rahmen dieser Studie konzipierten Reaktors untersucht. Berilyum (Be) ist hierbei mit der (n,2n) Reaktion der Neutronen-Multiplikator. Der Be-Bereich hat eine Dicke von 3 cm. Als Konstruktionsmaterial wurde 9Cr2WVTa ferritischer Stahl verwendet. Die dreidimensionalen nukleonischen Berechnungen wurden mit Hilfe der neuesten Versionen des MCNPX-2.7.0 Monte Carlo-Codes und der nuklearen Daten-Bibliothek ENDF/B-VII.0 durchgeführt.
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© 2014, Carl Hanser Verlag, München
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Evaluated gas production cross-section data for natural titanium irradiated with protons at energies up to 3 GeV
- Investigation of (n,γ) reaction in hybrid reactor zones
- Burnup analysis of the VVER-1000 reactor using thorium-based fuel
- Equivalent thermal conductivity of the storage basket with spent nuclear fuel of VVER-1000 reactors
- A study on empirical systematic for the (d,n) reaction cross sections at 8.6 MeV
- Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations
- Thermoluminescent characteristics of nano-structure hydroxyapatite:Dy
- Performance evaluation of anion exchange resins Purolite NRW-5050 and Duolite A-611 by application of radioisotopic techniques
- A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading
- Development of an educational nuclear research reactor simulator
- Technical Notes/Technische Mitteilungen
- Obtaining the neutronic and thermal hydraulic parameters of the VVER-1000 Bushehr nuclear reactor core by coupling nuclear codes
