Preliminary Results of Neutron Transport in Blanket Module by MCNP with Profile Analysis Using Imaging Plate

Measurement of neutron flux and energy spectrum profile inside the blanket is required for fusion blanket design. An experiment using an imaging plate and activation materials (Dy, In, and Au) was performed to measure spatial distribution of neutron flux. Neutrons were generated by a discharge-type compact fusion neutron source whose neutron production rate was more than 10⁷ n/s. A linearity between the total number of active nuclides made by neutron and photo-stimulated luminescence per area on the activation material was confirmed for three orders of magnitude. The relationships between the total number of decay of activation in the materials and the flux of the neutron in a simplified breeder assembly was measured and compared with the computation by MCNP.