用平坦通量模型计算核燃料电池碰撞概率矩阵随中子能量群的函数

Q3 Engineering
M. Shafii, D. Fitriyani, S. H. Tongkukut, Z. Su'ud
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引用次数: 0

摘要

碰撞概率法是核燃料电池中子输运方程求解中广泛使用的方法之一。中子输运是一个非常重要的问题,因为中子分布与反应堆功率分布有关。CP方法中的重要内容是CP矩阵计算,众所周知,它在确定堆芯中子通量分布方面发挥着重要作用。本研究在具有白色边界条件的每个能量组的每个单元区域中使用线性平坦通量模型。尽管本研究中使用的反应堆类型是快堆,但矩阵计算仍在快堆和热群能量中进行。矩阵取决于每个单元区域中的网格数。由网格分布形成的矩阵将为每个能量组产生矩阵。因为系统的边界条件假设没有来自外部的贡献中子源,所以矩阵的和必须小于1。总的来说,本研究中的计算结果遵循理论
本文章由计算机程序翻译,如有差异,请以英文原文为准。
Calculation of Collision Probability Matrix of Nuclear Fuel Cell as a Function of Neutron Energy Group Using Flat Flux Model
One of the methods that widely used in solving neutron transport equations in the nuclear fuel cell is the collision probability (CP) method. The neutron transport is very important to solve because the neutron distribution is related to the reactor power distribution. The important thing in the CP method is the CP matrix calculation, better known as has an important role in determining the neutron flux distribution in the reactor core. This study uses a linear flat flux model in each cell region for each energy group with white boundary condition. Although the type of reactor used in this study is a fast reactor, the matrix calculation still carried out in fast and thermal group energy. The matrix depends on the number of mesh in each cell region. The matrix formed from the mesh distribution will produce a matrix for each energy group. Because the boundary condition of the system is assumed that there are no contributions neutron source from the outside, the sum of the matrix must be less than one. In general, the results of the calculations in this study are following the theory
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来源期刊
International Journal of Mechanics
International Journal of Mechanics Engineering-Computational Mechanics
CiteScore
1.60
自引率
0.00%
发文量
17
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