ICONE19-44028 A study of applicability of JENDL-4.0 to the HTTR criticality analysis

Abstract

In the past, benchmark calculations of criticality approach for the HTTR, which is a Japanese High Temperature Gas-cooled Reactor (HTGR), were performed by research institutes in several countries, and almost all of the calculations overestimated the excess reactivity. In Japan, the benchmark calculations performed by JAEA also resulted in overestimation. JAEA improved the calculations by revising the geometric model and replacing the nuclear data library with JENDL-3.3, which was the latest JENDL at that time. However, the overestimation remained and this problem had not been resolved. We performed calculations of the HTTR criticality approach with several nuclear data libraries, and found that slight difference in the neutron capture cross section of carbon at thermal energy among the libraries causes significant difference in the k_<eff> values. The cross section value of carbon was not a concern in reactor neutronics calculation because of its small value of the order of 1E-3 burn, and consequently the cross section value was not revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We thought that the cross section should be revised based on the latest measurement data to improve the accuracy of the HTGR criticality analysis. In May 2010, the latest JENDL: JENDL-4.0, was released by JAEA, and the capture cross section of carbon was revised. JENDL-4.0 yielded 0.4-0.9%Δk smaller k_<eff> values than JENDL-3.3 in the criticality calculations for the HTTR critical approach, and consequently the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved by replacing the nuclear data libraries with JENDL-4.0.