Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc.; Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion.; The behaviour of the system is determined by composition of resonant and predominantly scattering isotopes. When the concentration of resonant isotopes is small, its presence does not affect the flux shape which is smooth. But when the concentration becomes high, there will be dips in the flux where resonances of the isotopes occur. This will affect the reaction rate, which is a product of cross section and flux. The reaction rate will thus be lower than that when one does not consider the flux dip. This is the phenomenon of self shielding. Self shielding treatment is thus a very important aspect of reactor lattice analysis code. This needs to be correctly modelled to obtain a physically sound and acceptable solution. In this research we will be looking into behaviour of the advanced self shielding models that have been incorporated in the code DRAGON Version4. The self shielding models are primarily classified into two broad groups, which are based on "equivalence in dilution" and "subgroup approach". These self shielding models will be tested against a variety of lattices which include Canada Deuterium Uranium (CANDU-6), CANDU-New Generation (CANDU-NG), Light Water Reactor (LWR), and High Conversion Light Water Reactor (HCLWR). The fuel composition will vary from natural uranium oxide to enriched uranium oxide and plutonium-uranium mixed oxide (MOX). We will also consider the presence of strong neutron absorbers like gadolinium and dysprosium in the lattice. The coolant/moderator chosen for the analysis will be light water/heavy water or a combination. The lattice geometry will vary from square, hexagonal and annular. Thus a broad spectrum of lattices will be analysed to assess the behaviour of advanced self shielding models. The results obtained using DRAGON will be validated against that obtained using Monte Carlo code MCNP5. The reference solutions for all situations will be provided by MCNP5.; The depletion behaviour of any lattice will depend on the power or flux normalization that is considered. In general the flux in vario...