In the framework of research activities on fusion reactors a great effort is dedicated by the scientific community to the development of tritium breeding blankets. One of the main goals is to assess the neutronic behaviour of such devices to analyse their tritium breeding performance and to evaluate the required data for their thermal–mechanic and thermal–hydraulic design. Many papers have been published on this topic considering some stationary condition to calculate such important quantities as heating power, gas production and dpa rates, tritium breeding ratio, etc., but not much attention has been focussed to neutronic transport analyses in transient conditions. The present paper proposes a simple model based on the point kinetics approximation, which has been set up deriving an alternative formulation of the time-dependent neutron transport equation. This approach allows to define some physical characteristics that can be interpreted in a statistical way, making possible to calculate these quantities numerically by the Monte Carlo method. The adoption of the aforementioned numerical method has the great advantage that complex geometries (as the fusion reactor’s blankets are) can be analysed with acceptable computational times. Some simple cases have been investigated to implement the theoretical model proposed with MCNP5 code and to show its potentiality. Then, applications to fusion reactor ITER blanket module and to the Helium Cooled Test Blanket Module, to be tested in ITER, have been taken into account in order to assess their neutronic time-dependent behaviour and the results obtained have been critically discussed.
|Numero di pagine||6|
|Rivista||Fusion Engineering and Design|
|Stato di pubblicazione||Published - 2012|
All Science Journal Classification (ASJC) codes
- Civil and Structural Engineering
- Nuclear Energy and Engineering
- Mechanical Engineering