Since the nodal method can calculate the eigenvalue and the node-averaged power distribution in a reactor core accurately with much smaller computation time than the finite difference method, it becomes a powerful tool for nuclear fuel management study. However, the nodal method cannot give any ifformation of the flux distribution in a node. Therefore, dehomogenization is necessary for obtaining such information after a global reactor core analysis using a nodal method. The flux distribution, equivalently the power distribution, in a node is required to deplete the fuel in the node and thus to calculate node-homogenized group constants for the node which are essential for the global calculation using a nodal method.
Various schemes had been studied for reconstructing local flux distribution from the node-averaged quantities obtained by the global reactor core calculation using a nodal method. Most common scheme is to approximate the local flux distribution in a node as the flux distribution obtained by the node calculation for the node-homogenized group constants with multiplying by so called a form function. First, the mode of the form function is assumed such as quadratic, sinuous or exponential mode. Then, the unknown coefficients of the form function are determined using the node-averaged quantities obtained of two by the global core calculation using the nodal method. The success of this method depends highly on the choice of appropriate mode for the form function since the most resulting error in the reconstructed local flux depends on the mode.
To eliminate the difficulty in choosing an appropriate mode of form function for reconstructing local flux distribution in a node and to improve the accuracy of the reconstruction, a new method has been developed through this study. The new method is to interpolate the flux distribution on the surfaces of nodes in the reactor core and then to solve the diffusion equation for a node of interest using the flux distribution on the surfaces of the node as a boundary condition. It is proved in this thesis that this new method can successfully reconstruct the flux distribution in a node from the node-averaged values obtained by a reactor core calculation using a nodal method.
본 논문의 목적은 원자로 해석을 위해 소격격자 계산으로부터 관심있는 집합체별 중성자속 분포를 구하는 것이다.
본 논문에서는 집합체별 유한 차분 계산으로 균질화 상수를 구했으며 ADF/AXS 를 이용한 소격격자 계산이 UDF/AXS 계산보다 유효함을 예시 하였다.
이를 토대로 몇개의 Benchmark Problem 에서의 집합체별 중성자속 분포를 계산하였는데, 먼저 고전적인 계산 방법으로서 Form Function을 이용한 방법과 본 논문에서 새로이 개발한 방법의 결과를 비교하였다. 이 새로운 방법이 중성자속 분포의 정확도 면에서 기존의 방법보다 월등히 우수함을 보였다.
