Comparison of simple design of sodium and lead cooled fast reactor cores

• Autorzy: Mazgaj P. , Darnowski P. , Gurgacz S. , Lipka M. , Dziubanii K.
• Języki publikacji: EN

Abstrakty

EN

This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid metal cooled fast reactor core, combined with simple neutron population computing for an infinite pin cell lattice. Two types of coolant were studied: liquid sodium and liquid lead, with all requirements regarding safety conditions observed. Temperature distributions along the cooling channel and distributions in the radial direction were prepared, then criticality calculations were performed for MOX fuel using MCNP Monte Carlo code.

Słowa kluczowe

EN

LFR; lead fast reactor; SFR; sodium fast reactor; pin design; MCNP; Monte Carlo N-Particle Transport Code

PL

LFR; reaktor prędki chłodzony sodem; SFR; reaktor prędki chłodzony ołowiem; projektowanie reaktorów; MCNP; metoda Monte Carlo

• Wydawca: Institute of Heat Engineering, Warsaw University of Technology
• Czasopismo: Journal of Power Technologies
• Rocznik: 2014
• Tom: Vol. 94, nr 5
• Strony: 16--25
• Opis fizyczny: Bibliogr. 33 poz., tab., wykr.

Twórcy

autor

Mazgaj P.

• Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

autor

Darnowski P.

• Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

autor

Gurgacz S.

• Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

autor

Lipka M.

• Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

autor

Dziubanii K.

• Institute of Heat Engineering, Warsaw University of Technology, 21/25 Nowowiejska Street, 00-665 Warsaw, Poland

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