TY - GEN
T1 - Preparation method of origen2 library for high temperature gascooled reactors
AU - Simanullang, Irwan L.
AU - Fukuhara, Katsuki
AU - Morita, Keisuke
AU - Fukaya, Yuji
AU - Ho, Hai Quan
AU - Nagasumi, Satoru
AU - Iigaki, Kazuhiko
AU - Ishitsuka, Etsuo
AU - Fujimoto, Nozomu
N1 - Funding Information:
This research was funded by National Natural Science Foundation of China (No. 12005140) and Guangdong Basic and Applied Basic Research Foundation (No. 2019A1515110318).
Publisher Copyright:
© 2022 American Society of Mechanical Engineers (ASME). All rights reserved.
PY - 2022
Y1 - 2022
N2 - 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 fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-To-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35 % than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pincell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4 %.
AB - 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 fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-To-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35 % than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pincell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4 %.
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U2 - 10.1115/ICONE29-90755
DO - 10.1115/ICONE29-90755
M3 - Conference contribution
AN - SCOPUS:85143140941
SN - 9784888982566
T3 - International Conference on Nuclear Engineering, Proceedings, ICONE
BT - Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology
PB - American Society of Mechanical Engineers (ASME)
T2 - 2022 29th International Conference on Nuclear Engineering, ICONE 2022
Y2 - 8 August 2022 through 12 August 2022
ER -