Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k0-INAA and absolute method. The average values of φth,φepi, and φfast by MCNP code were (2.19±0.03)×1012 cm−2 s−1, (1.26±0.02)×1011 cm−2 s−1 and (3.33±0.02)×1010 cm−2 s−1, respectively. These average values were consistent with the experimental results obtained by k0-INAA. The findings show a good agreement between MCNP code results and experimental results. - Highlights: ► We use 3-D neutronic model to enhance the utilization of the economical use of reactor.► The MCNP code is modified to analyze the neutronic parameters. ► The neutron flux distributions are homogeneous and consistent with experimental results. ► A complete simulation of the calculated neutron flux parameters in 40 RR irradiation channels is achieved.