Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data

The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized andlaunched the UAM benchmark. Its main objective is to perform uncertainty analysis in lightwater reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopiccross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code inorder to perform uncertainty analysis on kinf and 2-group homogenized macroscopic cross-sections. The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which isa 15*15 pressurized water reactor (PWR) fuel assembly segment with poison and at fullpower conditions. A statistical methodology is employed for the uncertainty assessment, wherecross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIBlibrary format are considered as normal random variables. Two libraries were created for suchpurposes, one based on JENDL-4 data and the other one based on the recently releasedENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries needed to be computed for the different isotopic reactions by means of ERRORJ. Theuncertainty assessment performed on kinf and macroscopic cross-sections, that is based onJENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It wasfound that the computed Uranium 235 fission covariance matrix based on JENDL-4 is muchlarger at the thermal and resonant regions than, for instance, the covariance matrix based onENDF/B-VII.1 data. This can be the main cause of significant discrepancies between differentuncertainty assessments