Half-space albedo problem for the Anlı-Güngör scattering function

One speed, time-independent and homogeneous medium neutron transport equation is solved for second order scattering using the Anlı-Güngör scattering function which is a recently investigated scattering function. The scattering function depends on Legendre polynomials and the t parameter which is defined on the interval [−1, 1]. A half-space albedo problem is examined with the FN method and the recently developed SVD method. Albedo values are calculated with two methods and tabulated. Thus, the albedo values for the Anlı-Güngör scattering are compared with these methods. The behaviour of the scattering function is similar to İnönü’s scattering function according to calculated results.

Core Configuration Analysis for Modular Gas-cooled Fast Reactor (GFR) using OpenMC

The core configuration analysis of modular Gas-cooled Fast Reactor (GFR) has been done to understand GFR performance. The modular GFR used a fast neutron spectrum and high-temperature helium gas, providing higher thermal efficiency than the other generation IV reactor candidates. In this paper, the variation of core configuration and dimension for core design has been applied in radial, axial, and radial-axial directions. The Monte Carlo method, named OpenMC code, has been used for the criticality and isotope evaluation of design core GFR. The OpenMC code provides the probabilistic solution to solve the neutron transport equation in a 3D model and non-homogenous physical volumes using Evaluated Nuclear Data File (ENDF/B-VII.b5) and continuous energy spectrum. The neutronics parameters characterized are the value of keff, fission rate and neutron flux distribution, and fissile material evolution to know of GFR core design’s performance. The analysis showed that the core configuration in radial direction gives a good understanding of the feasibility of GFR core design.

Simulation of the burnup in cell calculation using the wimsd-5b code considering different nuclear data libraries

This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows determining the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups.