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Report No.

JENDL-4.0 benchmark for high temperature gas-cooled reactor, HTTR

Goto, Minoru  ; Shimakawa, Satoshi; Yasumoto, Takashi*

In the past, benchmark calculations of the HTTR criticality approach, which is a Japanese HTGR, have been performed by several countries, and almost of the calculations have overestimated its excess reactivity. In Japan, the benchmark calculations were performed by JAEA, and the calculations also resulted in overestimation. JAEA improved this overestimation by revising the problem geometry and replacing 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 until today. We performed the calculation of the HTTR criticality approach with several nuclear data libraries, and found the slightly difference of the capture cross section of carbon at thermal energy among the libraries causes the difference of the $$k$$$$_{eff}$$ values to be not negligible. This cross section value had not been concerned in reactor neutronics calculation because of its small value on the order of 10$$^{-3}$$ burn, and consequently the cross section value had not been revised for a long time even in the major nuclear data libraries: JENDL, ENDF/B and JEFF. We have thought the cross section value should be revised based on the latest measurement data to improve the accuracy of the neutronics calculations for HTTR. On May in 2010, the latest JENDL (JENDL-4) was released by JAEA, and the capture cross section of carbon was revised. Consequently, JENDL-4 yielded 0.4-0.9%$$Delta$$k smaller $$k$$$$_{eff}$$ values than JENDL-3.3 in the calculation for the HTTR critical approach, and then the problem of the overestimation of the excess reactivity in the HTTR benchmark calculation was resolved.



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