Design Methods and Computer Codes

This chapter discusses the various methods and codes used in analyzing thermal (light-water reactors) and fast reactor (sodium-cooled fast reactors) cores. Multigroup nuclear data libraries and details of lattice physics calculations to account spatial and energy self-shielding effects in the resonance analysis are discussed. The whole-core calculations for estimating steady-state neutronics parameters for plant operation are also given. In addition, a brief outline of safety analysis and severe accident analysis in fast reactors is given at the end. The aim of reactor analysis is to obtain the neutron density distribution as a function of space, energy, scattering angle, and time. The neutron balance equation is formulated in the region of interest and solved for the fluxes. The problem is divided into two main sequences. First, the reactor core is considered to be an ensemble of small units. The neutron transport is treated in a hyperfine energy structure over this unit representative cell. The spatial and energy-dependent flux is then used to homogenize the unit cell properties to derive homogenized cell cross sections called lattice parameters. This representative cell is treated either in a one-dimensional (1-D) or two-dimensional (2-D) geometry. The inputs required for the lattice-level calculations are the energy-dependent cross-section set called as nuclear physics data and the geometry of the lattice. In the next step, the whole three-dimensional (3-D) reactor core is treated as a periodic arrangement of these unit representative cells with homogenized lattice properties. The global parameters are derived from the solution of flux over the 3-D core. The criticality problem is treated as a steady-state solution by formulating a time-independent neutron balance equation. The transient solutions are obtained by introducing a time-dependence in the balance equations.