A silicon carbide assembly of a thickness corresponding to about 10 mean free path for 14 MeV neutrons was irradiated at the Frascati Neutron Generator and the neutron and gamma-ray flux spectra were measured simultaneously, using a NE 213 scintillation spectrometer at four positions inside the block. The computational analysis of the benchmark experiment for the advanced structural material of fusion devices was performed with the Monte Carlo code MCNP-4C and nuclear data taken from the Fusion Evaluated Nuclear Data Library FENDL/MC-2.0, except for28Si, for which the new data evaluation of the European Fusion File EFF-3.0 was used. Measured and calculated flux spectra are compared and ratios of calculated-to-experimental values are derived for various energy ranges. © 2003 Elsevier Science B.V. All rights reserved.
All Science Journal Classification (ASJC) codes
- Civil and Structural Engineering
- Nuclear Energy and Engineering
- Materials Science(all)
- Mechanical Engineering
Seidel, K., Angelone, M., Batistoni, P., Chen, Y., Fischer, U., Freiesleben, H., ... Unholzer, S. (2003). Measurement and analysis of neutron and gamma-ray flux spectra in SiC. Fusion Engineering and Design, 69(1-4 SPEC), 379 - 383. https://doi.org/10.1016/S0920-3796(03)00077-2