| Home | E-Submission | Sitemap | Contact Us |  
J Radiat Prot > Volume 14(1); 1989 > Article
Journal of Radiation Protection 1989;14(1):46-0.
XSDRN, ONEDANT및 MCNP에 의한 사용후 핵연료 용기의 중성자 차폐 해석
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;
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.
Editorial Office
#319, Hanyang Institute of Technology Bldg., 222 Wangsimni-ro, Seongdong-gu,Seoul, Republic of Korea
Tel: +82-2-2297-9775   Fax: +82-2-2297-9776
Email: managing.editor@jrpr.org
About |  Browse Articles |  Current Issue |  For Authors and Reviewers
Copyright © by Korean Association for Radiation Protection. Developed in M2PI