ONE STEP METHOD FOR MULTIGROUP ADJOINT NEUTRON FLUX THROUGH CONTINUOUS ENERGY MONTE CARLO CALCULATION

被引：0

作者：

Shang, Xiaotong ^{[1
]}

Shi, Guanlin ^{[1
]}

Wang, Kan ^{[1
]}

机构：

[1] Tsinghua Univ, Beijing, Peoples R China

来源：

PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2018, VOL 3 | 2018年

基金：

中国国家自然科学基金;

关键词：

EIGENVALUE SENSITIVITY COEFFICIENTS; NUCLEAR-DATA;

D O I：

暂无

中图分类号：

TH [机械、仪表工业];

学科分类号：

0802 ;

摘要：

The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability (IFP) method is introduced to calculate the neutron importance which is able to represent the adjoint neutron flux in continuous energy problem and have been applied to the calculation of kinetics parameters. However, the adjoint neutron flux can't be obtained directly and applied to both Monte Carlo transient event analysis and deterministic methods. In this study, a method based on IFP is studied and implemented in Monte Carlo code RMC. The multi-group adjont neutron flux can be obtained directly through the discretization of energy and space with the modification of fission neutrons through continuous energy Monte Carlo calculations. The obtained multi-group adjoint neutron flux can be used in both Monte Carlo transient analysis and deterministic methods.

引用

页数：6

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