Review of the natC(n,γ) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for (nat)c(n,gamma) and C-12(n,gamma) was made to identify the origin of the C-nat capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK.CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of C-12 at thermal energy which is recommended by Firestone and Revay. Additional criticality calculations were carried out in order to illustrate that the C-nat thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data. (C) 2013 Elsevier Ltd. All rights reserved.