Neutron Shielding Analysis for a Spent Fuel Container Using XSDRN, ONEDANT and MCNP Codes
Kim, Kyo-Youn;Lee, Tae-Young;Ha, Chung-Woo;Kim, Jong-Kyung;
ABSTRACT
Neutron shielding for a spent fuel container was analized using the Monte Carlo code MCNP coupled with discrete ordinates codes, XSDRN and ONEDANT. The ORIGEN-S code was used to determine the fixed neutron source, and the spectral source information for MCNP were obtained from a 10 group XSDRN calculation and a 27 group ONEDANT calculation. When the depleted uranium shield was used, the results from ONEDANT and XSDRN calculations agreed with the MCNP results within 10% and 20%, respectively. Depleted uranium appears more effective than lead or steel as a neutron shield.
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