Preparation method of origen2 library for high temperature gascooled reactors

Irwan L. Simanullang, Katsuki Fukuhara, Keisuke Morita, Yuji Fukaya, Hai Quan Ho, Satoru Nagasumi, Kazuhiko Iigaki, Etsuo Ishitsuka, Nozomu Fujimoto

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

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 %.

Original languageEnglish
Title of host publicationNuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology
PublisherAmerican Society of Mechanical Engineers (ASME)
ISBN (Print)9784888982566
DOIs
Publication statusPublished - 2022
Event2022 29th International Conference on Nuclear Engineering, ICONE 2022 - Virtual, Online
Duration: Aug 8 2022Aug 12 2022

Publication series

NameInternational Conference on Nuclear Engineering, Proceedings, ICONE
Volume2

Conference

Conference2022 29th International Conference on Nuclear Engineering, ICONE 2022
CityVirtual, Online
Period8/8/228/12/22

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering

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