Development and verification of a transport solver for DYN3D

The computer code DYN3D is widely used for safety estimation of nuclear power plants. It has been developed in Helmholtz-Zentrum Dresden-Rossendorf, Germany. The last version of the DYN3D code can be used for investigations of transients in light water reactors cores with hexagonal or quadratic fuel assemblies. Moreover, the multigroup SP3-approximation can be used for reactors with quadratic and hexagonal fuel assemblies.
The DYN3D code is based on the nodal expansion method. It applies for the calculations fuel assembly wise homogenized cross-sections and determines fluxes homogenized for each node in the core. But in many cases for safety calculations good knowledge of power and temperature distributions on nodal level is not enough. In most cases the ability to predict accurately the local pin powers in nuclear reactors is necessary.
In the work a new methodology for pin-power calculations in chosen nodes has been proposed. The main idea is to determine the pin wise power distribution inside fuel assemblies applying a transport solver using unstructured mesh with boundary conditions extracted from the 3D full core nodal diffusion solution. In this technique the global calculation is used to define partial incoming currents as boundary conditions for a local problem.
In the work, advanced method for solution of transport equation based on the current coupling collision probability method with orthonormal flux expansion was proposed and tested for the single cell . In the present work the proposed transport methodology is extended from the level of single cell to full assembly level. The developed transport solver has been used for the solution of a one group fixed source problem as well as for a multigroup eigenvalue problem. The results of the calculations demonstrate good agreement with the results of Monte-Carlo calculations as well as with the results of HELIOS calculations.