Neutron Flux Interpolation with Finite Element Method in the Nuclear Fuel Cell Calculation using Collision Probability Method

Abstract

Articles you may be interested inA novel method for modeling the neutron time of flight detector response in current mode to inertial confinement fusion experiments (invited)a) Rev. Sci. Instrum. 83, 10D915 (2012); Abstract. Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the eff k and other parameters are investigated.

Neutron Distribution in the Nuclear Fuel Cell using Collision Probability Method with Quadratic Flux Approach

“…The detail of mesh composition, the radius of each region cell and number of nuclide in each region are shown in [4].…”

“…Corrected background cross section is required to calculate macroscopic cross section for width of cell. The collision probability matrix ij p is calculated for all regions that divided into several mesh uses equation (10) of [4]. Equation (1) is inserted to equation (2) to get the k-eff and neutron flux distribution.…”

“…The geometry design of cylindrical nuclear fuel cell is created based on composition of three regions: region 1 is fuel, region 2 is cladding and region 3 is coolant [4]. The composition of region number or mesh in each region follows [5].…”

“…In this concept, spatial function distribution is used to govern neutron collision probabilities in real condition of inside cell. Neutron flux in each region is different each other and flux transformation is governed by linear interpolation scheme [4][5] and quadratic interpolation investigation [6]. As neutron flux is non uniform within the regions of the cell therefore this concept is commonly called as non-flat flux (NFF) approximation [5].…”