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

Abstract

A review of the experimental data for nat C(n,y) and 12 C(n,y) was made to identify the origin of the nat C 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.l 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 12 C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the nat C 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.

Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

“…In contrast to the presented results, calculations (Díez et al 2013) for the BR-1 reactor show a 28 much more notable effect of (n,g) on the keff. The BR-1 reactor located at the Belgian Nuclear 29…”

“…The presented results of the HTTR calculation show also a trend of underestimation, 7 however higher, of JENDL-4 results with regard to ENDF/B-VII.0 or JEFF-3.1. The reason is 8 a more notable effect of (n,g) on the keff as reported in (Díez et al 2013 reported experiments with comparable fuel enrichment. 12…”

“…Graphite is now being 19 considered as the moderator for Fluoride-salt-cooled High temperature Reactors (FHR). The 20 critical moderator level was measured for various graphite block configurations in an 21 experimental dry assembly of the LR-0 reactor. Comparisons with experiments were 22 performed between Monte Carlo simulation tools for which satisfactory agreement was 23 obtained with the exception of some systematic discrepancies.…”