Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

Gunay M.

KERNTECHNIK, cilt.79, sa.2, ss.145-149, 2014 (SCI-Expanded) identifier identifier

  • Yayın Türü: Makale / Tam Makale
  • Cilt numarası: 79 Sayı: 2
  • Basım Tarihi: 2014
  • Doi Numarası: 10.3139/124.110408
  • Dergi Adı: KERNTECHNIK
  • Derginin Tarandığı İndeksler: Science Citation Index Expanded (SCI-EXPANDED), Scopus
  • Sayfa Sayıları: ss.145-149
  • İnönü Üniversitesi Adresli: Evet

Özet

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.