Calculation of neutron flux characteristics of Dalat reactor using MCNP4A code

摘要

The neutron flux characteristics of the Dalat reactor such as energy spectra, absolute neutron flux and neutron flux distribution along an irradiation channel were calculated by using MCNP4P code. All computations were done on a personal computer with the running time about 2 days for every case. The calculation configuration is similar to real one of reactor at the nominal power 500 kW. The calculated results of neutron flux and a spectrum fitting parameter, (alpha), are in good agreement with experimental ones within 5%. By using the calculated energy spectra, the cadmium ratios (R(sub cd)) and effective cross-section of (sup 197)Au in the irradiation channels were calculated. In this calculation, were used the multi-group cross-sections of (sup 197)Au from Japanese Evaluated Nuclear Data Library (JENDL) and International Reactor Dosimetry File 82 (IRDF82). The comparison of the calculated results shows that: (1) the difference of R(sub cd) values between the experiment and calculation using cross-section of (sup 197)Au(n,(gamma)) (sup 198)Au reaction is 1 to 6% for IRDF82 and 4 to 8% for JENDL, and (2) the effective cross-sections of (sup 197)Au(n,(gamma)) (sup 198)Au reaction from IRDF82 and JENDL dosimetry files are completely in good agreement with each other. (author)