TRIGON, 2-D Homogeneous and Inhomogeneous Fixed Source Neutron Diffusion for Triangular or Hexagonal Mesh

1 - Nature of physical problem solved: The program solves the multigroup time-independent neutron diffusion equations with arbitrary group to group scattering in two space dimensions. In addition to homogeneous or K-effective calculations also inhomogeneous or fixed source calculations can be made. The problem may be specified with both external and internal logarithmic derivative boundary conditions. Using pointwise flux distribution the program calculates automatically the average flux and power distributions over equal hexagons of arbitrary size into which the area of solution is divided by the user. 2 - Method of solution: The diffusion theory equations are approximated by 7-point difference equations for the uniform triangular or hexagonal mesh grid. In the former case, the mesh points are situated at the cross sections of the mesh lines separating different materials and in the latter case at the centre points of the homogeneous mesh hexagons. The difference equations are solved iteratively by the line over-relaxation technique utilizing an exponential over-relaxation procedure. Coarse mesh re-balancing and special asymptotic flux extrapolations are used to improve the convergence. 3 - Restrictions on the complexity of the problem: Because a variable dimensioning technique is employed, the only restriction on problem size is the available core storage. The option of using auxiliary storage (drum, disk or tape) during iteration allows to run large problems even with rather small core storage