Development of burnup analysis system for fast reactor (Contract research)

Hyodo, Hideaki*; Tatsumi, Masahiro*

In order to utilize the measured burnup data for improvement on accuracy in reactor core design, it is important to minimize the methodological errors to retrieve physical meanings from experimental data. The system for neutronics analyses that has been developed as the JUPITER standard analysis method assumes geometry of critical assembly, thus the system has not been maintained in functionalities for analysis of composition change of fuel materials. Therefore, there is a potential restriction for the purpose of detail analysis due to extreme inefficiency that comes from variety of limitation on its functionalities. It is not sufficient to follow a predefined analysis sequence in burnup analysis for reactor core; it is also needed to change analysis sequence to examine modeling error in analysis or to retrieve calculated values in an intermediate computation step for interpretation of physical meanings. Therefore it is not complete with a simple join of each function; it is needed to develop a new system for burnup analysis of reactor cores with flexibility on composition and decomposition of analysis components such as cell and core calculations. In this work, necessary conditions are examined for a new burnup analysis system targeted to actual reactor cores from the results of a research on the current working set in burnup analysis. With results in the research, a set of conceptual and fundamental designs were done.