Assessment of the neutronic performance of some alternative fluids in a fusion-fission hybrid reactor by using Monte Carlo method

Küçük Resim Yok

Tarih

2013

Dergi Başlığı

Dergi ISSN

Cilt Başlığı

Yayıncı

Pergamon-Elsevier Science Ltd

Erişim Hakkı

info:eu-repo/semantics/closedAccess

Özet

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVFa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, or 99-95% Li20Sn80-1-5% SFG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as neutron flux, tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fissile fuel breeding in liquid first wall, blanket and shield zones and investigates effects of spent fuel grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.O. Published by Elsevier Ltd.

Açıklama

Anahtar Kelimeler

Hybrid reactor, TBR, M, Fissile fuel, MCNPX-2.7.0

Kaynak

Annals of Nuclear Energy

WoS Q Değeri

Q2

Scopus Q Değeri

Q1

Cilt

60

Sayı

Künye