加速器驱动的次临界快增殖堆中子学研究  

Neutronics Study of Accelerator driven Sub critical Fast Breeding Reactor

作　　者：杨永伟[1] Takah.,H 

机构地区：[1]清华大学核能技术设计研究院,北京100084

出　　处：《高技术通讯》1999年第8期49-53,共5页Chinese High Technology Letters

基　　金：JAERI基金

摘　　要：选取加速器驱动的快堆作为核燃料增殖堆。在堆芯， 燃料形式为（Ｕ－ Ｐｕ） Ｏｘ。在转换区，ＴｈＯ２ 被选为增殖材料以生产２３３Ｕ。由于Ｐｂ 的中子学性能和化学性能优于Ｎａ，因而被选为冷却剂。利用下列程序对所选方案进行中子学计算分析：ＬＡＨＥＴ——模拟质子与靶核的相互作用； ＭＣＮＰ４Ａ——模拟次临界包层内２０ＭｅＶ以下的中子与材料核的相互作用； ＯＲＩＧＥＮ２——利用ＭＣＮＰ４Ａ的输出提供的一群等效截面对包层进行多区燃耗计算。中子学计算分析的结果表明： 考虑到临界安全性、功率密度、燃耗等因素，Accelerator driven fast reactor is chosen as nuclear fissile material breeding reactor. In the core, the fuel type is (U Pu)Ox . ThO2 is chosen as fertile material in the converting region to produce 233 U. Lead is chosen as coolant because of its better neutronic and chemical characteristic over sodium. The neutronics calculation and analysis of the selected scheme have been done by using the following codes: LAHET, for the simulation of the interaction between the protons and the nuclei of the target; MCNP4A , for the simulation of interaction between neutrons with energy below 20MeV and the nuclei of materials in the sub critical blanket; ORIGEN2, for the multi region burnup calculation of the blanket by using the one group effective cross section provided in the output of MCNP4A. The neutronics calculation and analysis show that the selected scheme is feasible considering the factors such as criticality safety, power density , burnup , etc.

关 键 词：中子学 加速器 次临界 增殖堆 

分 类 号：TL501[核科学技术—核技术及应用]