Preparation method of ORIGEN2 library for high temperature gas-cooled reactors

Simanullang, I. L.* ; Fukuhara, Katsuki*; Morita, Keisuke; Fukaya, Yuji   ; Ho, H. Q.   ; Nagasumi, Satoru; Iigaki, Kazuhiko ; Ishitsuka, Etsuo   ; Fujimoto, Nozomu*

The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the burnup characteristics and compared its results with Monte Carlo burnup calculation results. The high temperature test engineering reactor (HTTR) is a block type fueled HTGRs in Japan. The ORIGEN2 libraries for the HTTR were prepared with neutron spectrums evaluated by pin-cell burnup calculations. To validate the calculation results of the ORIGEN2 code, a comparison between the burnup calculation using the Monte Carlo MVP-BURN code was conducted. One of the nuclide isotopes evaluated in that study was Pu. The amount of Pu calculated by ORIGEN2 with a library prepared by pin-cell burnup calculation was higher about 35% than that calculated by MVP-BURN. It can be seen that the ORIGEN2 library based on the pin cell model was not enough for evaluating the burnup characteristics of the HTTR. Therefore, an improvement was conducted by generating the ORIGEN2 library for HTTR based on the fuel block system and the whole system (core and reflector regions). The comparison between the ORIGEN2 results and MVP-BURN results was investigated. In the case of the ORIGEN2 library being generated based on the fuel block system, the maximum difference was about 6% compared to the mass of Pu calculated by MVP-BURN. Furthermore, the difference of Pu amount between the ORIGEN2 library and MVP-BURN became 2.4% when the library for ORIGEN2 was generated based on the whole system of HTTR.