Analysis of the neutronic data in infinite medium using fusion neutron source for various material compositions


Ipek O.

ENERGY CONVERSION AND MANAGEMENT, cilt.44, ss.439-458, 2003 (SCI İndekslerine Giren Dergi) identifier identifier

  • Cilt numarası: 44 Konu: 3
  • Basım Tarihi: 2003
  • Doi Numarası: 10.1016/s0196-8904(02)00059-6
  • Dergi Adı: ENERGY CONVERSION AND MANAGEMENT
  • Sayfa Sayıları: ss.439-458

Özet

The main purpose of this study is to calculate the integral neutronic data, which are integral heat release, U-233 and Pu-239 fissile fuels breeding rate, tritium breeding ratio, integral neutron multiplication and net neutron gain through (n,x) and fission (when applicable) reactions in an infinite medium using a fusion neutron source for different mixtures. The most important materials in fusion technology, namely tritium, beryllium, lead, thorium and uranium, have been investigated in an infinite medium. Fusion neutrons, which have a high first wall flux value of 2.22 x 10(14) (14.1 MeV) neutron/cm(2)s, are released in an inertial confinement fusion (ICF). ThO2 and UC, composed with Be and Pb, are mixed with natural lithium or Li-6 for volume fractions from 0% to 100%. However, the variable ThO2-Li-6 (ThO2 mixed with Li-6), UC-Li-6 (UC mixed with Li-6), ThO2-Nat.Li (ThO2 mixed with Nat.Li) and UC-Nat.Li (UC mixed with Nat.Li) compositions will be mixed with Be and Pb for the volume fractions mentioned above. Integral neutron data calculations have been conducted with the help of the 3-D Monte Carlo Code. In a fusion reactor blanket with finite dimension, the integral quantities will be more or less different from the infinite medium results, depending on the neutron leakage fraction. Design studies foresee reduction of the neutron leakage out of the blanket to very low levels in order to prevent nuclear heating in the superconducting fusion magnets and to keep all neutrons primarily in the coolant. (C) 2002 Elsevier Science Ltd. All rights reserved.