Calculation of safety parameters for VVER - 1000 MWe reactor of KK project

Full text: Steady state fuel cycle analyses of VVER-1000 MWe reactor of Kudankulam project have been carried out with the indigenous code system EXCEL- TRIHEX-FA. Improvement in the calculation of 3D power distribution is achieved when the realistic temperature and power dependent feedback phenomena are modeled in TRIHEX-FA code. The flux distribution which is the main unknown of the problem determines the power distribution which in turn leads to a certain coolant and fuel temperature distribution. The saturated fission products Xe and Sm are also proportional to the mesh power. All these phenomena cause perturbations to the cross sections of a mesh of a given fuel type, burn up and soluble boron. A flux-power feedback iterative loop is incorporated in TRIHEX-FA to obtain internally consistent 3D distributions of flux, power and temperature. In order to facilitate the modeling of any reactor state a wide parametric few group lattice database is generated by EXCEL code. For every fuel assembly type, few group cross sections are obtained as a function of burnup (0 to 60 GWD/T) and soluble boron concentration (0 to 3000ppm) at nominal values of fuel and coolant temperatures and mean power level with saturated xenon and samarium. At coarser burnup steps, four other parameters are varied around their mean values, i) the coolant density from 0.05, to 1.0 g/cc (mean=0.7148g/cc), ii) coolant temperature from 276 to 326 degC (mean=306 degC), iii) fuel temperature from 250 to 1500 degC (mean=693 degC), iv) linear heat rating from 0 to 1667 w/cm (mean=166.7w/cm). All these four types of perturbations are converted into equivalent change in the absorption cross section in resonance (fuel temp.) or thermal groups (other perturbations). For evaluating the reactivity coefficients as a function of burnup during the fuel cycles, the core is followed up at nominal power (3000MWt). 3D distributions of power, coolant and fuel temperature values are obtained by power-flux iterations. Anyone of these distribution is then perturbed by a small fraction. For the perturbed 3D profile, the k{sub eff} values are obtained with no power-flux iterations. The reactivity difference between the reference and perturbed state is used to determine a given type of reactivity coefficient. The core average delayed neutron fraction ({beta}{sub eff}), and prompt neutron lifetime ('l') is obtained by first order perturbation theory. Adjoint and direct flux weighting is used in evaluating these parameters. Detailed results are given in Ref. 2. Sample results will be given in the paper.