Fuel depletion calculation in MTR-LEU NUR reactor

• Main Authors: Zeggar Foudil, Halilou Amrane, Meftah Brahim, Mokeddem Mohamed Y., Letaim Fathi, Bousbia-Salah Anis
• Format: Article
• Language: English
• Published: VINCA Institute of Nuclear Sciences 2008-01-01
• Series: Nuclear Technology and Radiation Protection
• Subjects: MTR-LEU fuel; burn up; NUR reactor; WIMSD-4
• Online Access: http://www.doiserbia.nb.rs/img/doi/1451-3994/2008/1451-39940801011Z.pdf

In this article, we present the results of a few energy groups calculations for the NUR reactor fuel depletion analysis up to 45 000 MWd/tU taken as the maximum fuel burn up. The WIMSD-4 cell code has been employed as a calculation tool. In this study, we are interested in actinides such as the uranium and plutonium isotopes, as well as fission products Xe-135, Sm-149, Sm-151, Eu-155, and Gd-157. Calculation results regarding the five energy groups are in a good agreement with those obtained with only two energy groups which can, therefore, be used in all subsequent calculations. Calculation results presented in this article can be used as a microscopic data base for estimating the amount of radioactive sources randomly dispersed in the environment. They can also be used to monitor the fuel assemblies inventory at the core level.