Miniature neutron source reactor burnup calculations using IRBURN code system

被引：4
|
作者：
Feghhi, S. A. H. [1 ]
Jafarikia, S. [1 ]
Abtin, F. [2 ]
机构：
[1] Shahid Beheshti Univ, Dept Radiat Applicat, Tehran, Iran
[2] Nucl Sci & Technol Res Inst, Tehran, Iran
来源：
ANNALS OF NUCLEAR ENERGY | 2012年 / 47卷
关键词：
Miniature neutron source reactor; IRBURN; Monte Carlo method; Burnup calculation;
D O I：
10.1016/j.anucene.2012.04.016
中图分类号：
TL [原子能技术]; O571 [原子核物理学];
学科分类号：
0827 ; 082701 ;
摘要：
Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that k(eff) decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%. (C) 2012 Elsevier Ltd. All rights reserved.
引用
收藏
页码：242 / 248
页数：7
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