BURNUP SENSITIVITY CALCULATIONS WITH CBZ FOR LIGHT WATER REACTOR ASSEMBLY PROBLEMS

Determination ofresearch reactor fuel burnup

A New Assembly-level Monte Carlo Neutron Transport Code for Reactor Physics Calculations

This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title “Probabilistic Scattering Game”, or PSG. The PSG code uses a method known as Woodcock tracking to simulate neu...

Residual Life Assessment of Major Light Water Reactor Components - Overview

........................................................ iii EXECUTIVE SUMMARY ....................................................... iv ACKNOWLEDGMENTS .. ix

Light Water Reactor Sustainability Program Risk - Informed Safety Margin

Application of Sensitivity/Uncertainty Methods to Burnup Credit Validation

The responsible use of calculational methods in nuclear criticality safety includes a determination of bias and bias uncertainty that may exist between the calculated results and reality. Such biases exist due to approximations used to model the real world, uncertainties in nuclear data, and approximations associated with the calculational method (e.g., Monte Carlo method). The bias and bias un...