Calculation of the neutronic behavior of minor actinides burning in a thermal research reactor using the MCNPX 2.6 code
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S. A. H. Feghhi
Abstract
Due to the reduction of accessible uranium resources as well as waste proliferation issues, researchers are looking for more suitable approaches, such as replacement of uranium as breeding fuels. Among the practical fuel matrixes, the thorium fuel matrix is favored for its naturally abundant and minor actinide proliferation resistance. Monte Carlo computational methods are widely used to successfully simulate neutronic behavior of nuclear reactors. Calculation of some neutronic and dynamic parameters of a 37-assembly simulated research reactor consisting of thorium oxide fuel and 1 minor actinide pin have been carried out in the present work using the MCNPX 2.6 code.
Kurzfassung
Die Verringerung zugänglicher Uranvorräte wie auch Fragen der Weiterverbreitung von radioaktiven Abfällen haben Fachleute nach besseren Ansätzen suchen lassen, wie z.B. der Ersatz von Uran als Brennmaterial. Als praktische Brennstoffmatrix wird Thorium bevorzugt wegen seines natürlichen Vorkommens und weil minore Aktiniden zur Vermeidung der Verbreitung von Kernmaterial beitragen. Die Verwendung von Monte-Carlo-Rechenmethoden ist weit verbreitet bei der erfolgreichen Simulation des neutronsichen Verhaltens von Kernreaktoren. Berechnungen einiger neutronischer und dynamischer Parameter der Brennstoffanordnung eines Forschungsreaktors bestehend aus Thoriumoxid und einem Brennstab aus minoren Aktiniden werden in dieser Arbeit mit Hilfe des MCNPX 2.6 Codes durchgeführt.
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© 2013, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part II: R&D necessities and development across the world
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- Calculation of excitation functions for the production of Cu and Co medical isotopes
- Reliability assessment of the passive heat removal system of the VVER-1000 reactor at Kudankulam NPP
- Calculation of the neutronic behavior of minor actinides burning in a thermal research reactor using the MCNPX 2.6 code
- Steam drum level control studies of a natural circulation multi loop reactor
- Analyses in regulatory practice
- Technical Notes/Technische Mitteilungen
- Installation and measurement capacity of 3 × 592 GBq 241Am–Be neutron irradiation cell
- Calculation of the critical thickness for one-speed neutrons in a reflected slab with backward and forward scattering using modified TN method
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors: a review – Part II: R&D necessities and development across the world
- Analysis of radwaste management alternatives during dismantling of Ignalina NPP systems with low level contamination
- Calculation of excitation functions for the production of Cu and Co medical isotopes
- Reliability assessment of the passive heat removal system of the VVER-1000 reactor at Kudankulam NPP
- Calculation of the neutronic behavior of minor actinides burning in a thermal research reactor using the MCNPX 2.6 code
- Steam drum level control studies of a natural circulation multi loop reactor
- Analyses in regulatory practice
- Technical Notes/Technische Mitteilungen
- Installation and measurement capacity of 3 × 592 GBq 241Am–Be neutron irradiation cell
- Calculation of the critical thickness for one-speed neutrons in a reflected slab with backward and forward scattering using modified TN method
