2024/12/11 更新

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写真a

イルワン リアプト シマヌルラン
IRWAN LIAPTO SIMANULLANG
SIMANULLANG LIAPTO IRWAN
所属
工学研究院 エネルギー量子工学部門 助教
工学部 量子物理工学科(併任)
工学府 量子物理工学専攻(併任)
職名
助教
連絡先
メールアドレス
電話番号
0928023486
プロフィール
私の研究は小型高温ガスの研究を行ってきます。現在は黒鉛の減速材に代わる新しい複合材の減速材にコンセプトを当てています。小型高温ガス炉にて高い燃料度かつ使用済み燃料の削減を実現するには、新しい複合材の減速材を調査することが重要です。
外部リンク

経歴

  • March 2018 - September 2021: Postdoctoral Researcher at Criticality Safety Research Group, Japan Atomic Energy Agency (JAEA)

  • Oct 2017 - Mar 2018 : Researcher at Laboratory for Advanced Nuclear Energy, Tokyo Institute of Technology

研究テーマ・研究キーワード

  • 研究テーマ:高温ガス炉の原子力システムの研究をしています。

    研究キーワード:高温ガス炉、黒鉛の減速材、コンポジットの減速材

    研究期間: 2022年11月 - 2025年11月

論文

  • Study on potential burnable poison materials for a small modular block-type HTGR design using MgO-BeO as a composite-based moderators 査読

    Irwan L. Simanullang, Nozomu Fujimoto

    Nuclear Engineering and Design   431 ( 113742 )   2024年11月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.nucengdes.2024.113742

  • Burnable Poisons Utilization in the Small HTGR with MgO-BeO as a Composite-Based Moderator Material 査読

    Irwan L. Simanullang, Nozomu Fujimoto

    2024年11月

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    担当区分:筆頭著者, 責任著者   記述言語:英語   掲載種別:研究論文(国際会議プロシーディングス)  

  • Preliminary study of a small high-temperature gas-cooled reactor (HTGR) concept with MgO–BeO moderators

    Simanullang I.L., Fujimoto N.

    Nuclear Engineering and Design   420   2024年4月   ISSN:00295493

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    出版者・発行元:Nuclear Engineering and Design  

    Graphite plays a crucial role as a moderator and reflector in high-temperature gas-cooled reactors (HTGRs). However, under high-irradiation conditions, graphite exhibits microcracking within the operational period. Studies have been investigated the potential of composite-based materials for replacing graphite in HTGRs. This study focused on a magnesium oxide (MgO)-based composite material as the host matrix and a homogeneously distributed beryllium oxide (BeO) as the entrained moderating phase and investigated the feasibility of the MgO–BeO as the new moderator in a small HTGR to achieve high burnup performance. In this study, the conceptual design of the HTR50S was selected as one of the candidates for the small HTGR concept. Burnup calculations and safety evaluation in HTR50S design were performed. The Monte Carlo MVP 3.0 and MVP-BURN codes were used in this study for neutronic calculations. Results revealed that a high burnup of 80 GWd/t can be achieved using a fuel composition of 6 kg heavy metal per fuel block with 17 wt% of 235U enrichment. Furthermore, a negative temperature coefficient of reactivity was achieved during the operation period.

    DOI: 10.1016/j.nucengdes.2024.113036

    Scopus

  • Preliminary study of a small high-temperature gas-cooled reactor (HTGR) concept with MgO–BeO moderators 査読 国際誌

    420   2024年4月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.nucengdes.2024.113036

    リポジトリ公開URL: https://hdl.handle.net/2324/7172128

  • Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor 査読 国際誌

    @Hai Quan Ho, @Toshiaki Ishii, @Satoru Nagasumi, @Masato Ono, @Yosuke Shimazaki, @Etsuo Ishitsuka, @Hiroaki Sawahata, @Minoru Goto, #Irwan Liapto Simanullang, #Nozomu Fujimoto, @Kazuhiko Iigaki

    Nuclear Engineering and Design   2024年2月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.nucengdes.2023.112795

  • Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

    Ho H.Q., Ishii T., Nagasumi S., Ono M., Shimazaki Y., Ishitsuka E., Sawahata H., Goto M., Simanullang I.L., Fujimoto N., Iigaki K.

    Nuclear Engineering and Design   417   2024年2月   ISSN:00295493

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    出版者・発行元:Nuclear Engineering and Design  

    External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor (HTTR), 252Cf is used as the external neutron source. However, the 252Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of 252Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the 252Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains more than 10n.s−1.cm−2 even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

    DOI: 10.1016/j.nucengdes.2023.112795

    Scopus

  • Preliminary Study of Burnup Measurement and Relative Power Distribution in HTTR using Gamma-Ray Measurement 査読 国際誌

    Irwan L. Simanullang, #S.Kawaguchi, N. Fujimoto, @T. Ishii, @S. Nagasumi, @H.H.Quan, @K.Nakajima, @E.Ishitsuka, @K. Iigaki

    International Conference on Nuclear Criticality Safety   2023年10月

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    記述言語:英語   掲載種別:研究論文(国際会議プロシーディングス)  

  • EVALUATION OF BURNABLE POISON REACTIVITY WORTH AT THE KUCA GRAPHITE-MODERATED SYSTEM

    Yamasaki S., Moriya S., Simanullang I.L., Fujimoto N., Sakon A., Sano T., Takahashi Y.

    International Conference on Nuclear Engineering, Proceedings, ICONE   2023-May   2023年   ISBN:9784888982566

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    出版者・発行元:International Conference on Nuclear Engineering, Proceedings, ICONE  

    In a block-type High-Temperature Gas-Cooled Reactor (HTGR), a fixed number of burnable poisons are loaded at the beginning of operation time to control a large amount of excess reactivity. Therefore, it is necessary to evaluate the burnable poison (BP) reactivity worth to achieve the optimum design of the HTGR. In this study, an experiment to measure the burnable poison worth reactivity was conducted at the B-Core of Kyoto University Critical Assembly (KUCA). B-core is solid moderator materials such as polyethylene and graphite combined with fuel plates to form the fuel element. The experiment was performed to measure the reactivity worth of a small cadmium plate (15 × 15 × 0.5 mm) at the B-Core of KUCA. In the experiment, there are 8-unit cells in a fuel element. In this study, the unit cell position of cadmium is called the cadmium unit cell. The experiments were carried out by varying the cadmium plate position inside the cadmium unit cell. This study evaluated the cadmium reactivity worth using the Monte Carlo MVP3. The objective of this study was to evaluate the appropriate results between the measured and the calculation values. The simulation using MVP3 code was conducted by varying the number of batches in the calculation. The results showed that the maximum discrepancy between experimental and calculated results was 24% for 5,000 batches. However, the discrepancy decreased when the number of batches increased to 50,000. The cadmium reactivity worth difference between the experiment and simulation was approximately 18 % depending on the cadmium plate position in the cadmium unit cell.

    Scopus

  • HIGH-TEMPERATURE OPERATION MODE OF HTTR FOR HYDROGEN PRODUCTION FACILITY

    Ho H.Q., Shimazaki Y., Iigaki K., Goto M., Simanullang I.L., Nagasumi S., Fujimoto N., Ishii T., Ishitsuka E.

    International Conference on Nuclear Engineering, Proceedings, ICONE   2023-May   2023年   ISBN:9784888982566

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    出版者・発行元:International Conference on Nuclear Engineering, Proceedings, ICONE  

    The conceptual design of a demonstration hydrogen production facility using heat supply from the high-temperature engineering test reactor (HTTR) is being researched and developed at Japan Atomic Energy Agency (JAEA). This facility produces hydrogen with a thermochemical water-splitting Iodine-Sulphur (IS) process that requires high-temperature heat to extract hydrogen efficiently. The HTTR could supply the heat to the IS demonstration plant through the secondary helium loop, coupling the IS plant to the HTTR. The rated operation mode of the HTTR gives helium outlet temperature of 850οC with 660 effective full power days (EFPD). However, in order to achieve hydrogen with high efficiency, the helium outlet temperature should be as high as 950οC. Increasing the outlet temperature increases the reactor core temperature, and as a result the operation time decreases. If the operation time is reduced too much, it is not feasible to use the HTTR as a heat supply for the IS plant. Therefore, the purpose of this study is to estimate the operation time of the HTTR at high operation mode of 950οC to confirm whether it could supply long enough high-temperature heat for the demonstration hydrogen IS plant. The thermal-hydraulic model is also revised using the latest calculation method to improve the accuracy of the temperature distribution in the HTTR. As results, the core temperature increase by about 50 to 100οC when the outlet temperature increases from 850 to 950οC. Although the increase of core temperature makes keff decrease by about 0.3 %Δk/k, the HTTR can still operate approximately 660 EFPD. Therefore, it is possible to use the HTTR for long-term high-temperature heat supply to the demonstration hydrogen production IS plant.

    Scopus

  • Preparation Method of ORIGEN2 Library for High Temperature Gas-Cooled Reactors 査読 国際誌

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

    ICONE29   2   2022年11月

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    記述言語:英語   掲載種別:研究論文(国際会議プロシーディングス)  

    DOI: https://doi.org/10.1115/ICONE29-90802

  • Prediction of the Operating Control Rod Position of the HTTR with supervised machine learning 査読 国際誌

    Hai Quan Ho, Satoru Nagasumi, Yosuke Shimazaki, Toshiaki Ishii, Kazuhiko Iigaki, Minoru Goto, Irwan L Simanullang, Nozomu Fujimoto, Etsuo Ishitsuka

    ICONE29   2022年11月

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    記述言語:英語   掲載種別:研究論文(国際会議プロシーディングス)  

    DOI: https://doi.org/10.1115/ICONE29-90826

  • Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code

    Simanullang I.L., Nakagawa N., Quan Ho H., Nagasumi S., Ishitsuka E., Iigaki K., Fujimoto N.

    Annals of Nuclear Energy   177   2022年11月   ISSN:03064549

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    出版者・発行元:Annals of Nuclear Energy  

    Power distribution plays a significant role in preventing the fuel temperature from exceeding the safety limit of 1600 °C in high-temperature gas-cooled reactors. Experiments for measuring the power distribution in the graphite-moderated system were performed at the Very High-Temperature Reactor Critical Assembly (VHTRC) facility. The power distribution was determined from the measured Cu activation rate for both the radial and axial distributions. In this study, the pin-wise power distribution of the VHTRC was evaluated with the Monte Carlo MVP3 code. The difference between the calculated and measured results was less than 1 % for the axial and radial distributions. The significant results were concerned with the area around the fuel and reflector regions in the axial direction, where the average discrepancy between the calculated and measured values was 0.8 %. This result showed improved agreement compared to the diffusion calculation that was conducted in the previous study.

    DOI: 10.1016/j.anucene.2022.109314

    Scopus

  • The effect of β on the consequence characteristics of a postulated criticality accident using 1/f<sup>β</sup> spectrum in randomized model of material distribution

    Simanullang I.L., Yamane Y., Ueki T., Tonoike K.

    Annals of Nuclear Energy   174   2022年9月   ISSN:03064549

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    出版者・発行元:Annals of Nuclear Energy  

    The distribution of the number of fissions has been studied as the consequence of a postulated criticality accident using a randomized model for the material distribution in fuel debris with the power law spectrum (1/fβ). The f is the frequency domain variable while the β value is the parameter related to randomness varying in a certain range. This study investigated the effect of β on the temperature coefficient of reactivity and number of fissions. The calculation results for the temperature coefficient of reactivity show that the shape of the histogram changes from symmetrical to positive skewness by increasing the β value. Moreover, large values of β increase the uncertainty of the number of fissions. The results of this study are worthwhile to realize the behavior of the temperature coefficient of reactivity and number of fissions in the randomized model of fuel debris using the 1/fβ spectrum model.

    DOI: 10.1016/j.anucene.2022.109167

    Scopus

  • Calculation of shutdown gamma distribution in the high temperature engineering test reactor

    Ho H.Q., Ishii T., Nagasumi S., Ono M., Shimazaki Y., Ishitsuka E., Goto M., Simanullang I.L., Fujimoto N., Iigaki K.

    Nuclear Engineering and Design   396   2022年9月   ISSN:00295493

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    出版者・発行元:Nuclear Engineering and Design  

    Estimation of decay gamma distribution in a reactor core is essential for safely conducting various works after reactor shutdown such as periodic maintenance, shuffling fuel, removing spent fuel at the end of cycle, etc. Because of the dependency on the complex operating history of the reactor, attempting to calculate the decay gamma rays distribution in the core remains a challenge. This study shows a method to calculate the shutdown gamma distribution in the HTTR core by coupling a Monte-Carlo transport calculation code MCNP6 and an activation code ORIGEN2 to take advantage of spatial dependence and transport abilities of MCNP6 and the detailed fission products tracking during burnup and cooling of ORIGEN2. As result, the three-dimensional shutdown gamma distribution in the HTTR core for different cooling times and spatial locations could be obtained accurately.

    DOI: 10.1016/j.nucengdes.2022.111913

    Scopus

  • Calculation of shutdown gamma distribution in the high temperature engineering test reactor 招待 査読 国際誌

    H.Q. Ho, T. Ishii, S. Nagasumi, M. Ono, Y. Shimazaki, E. Ishitsuka, M. Goto, I.L. Simanullang, N. Fujimoto, K. Iigaki

    Nuclear Engineering and Design   2022年8月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.nucengdes.2022.111913

  • Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code 査読 国際誌

    Irwan L. Simanullang, #N. Nakagawa, H.Q.Ho, S. Nagasumi, E.Ishitsuka, K. Iigaki, N. Fujimoto

    Annals of Nuclear Energy   2022年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.anucene.2022.109314

  • The effect of β on the consequence characteristics of a postulated criticality accident using 1/f#Uβ#DR spectrum in randomized model of material distribution 査読 国際誌

    174   2022年4月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.anucene.2022.109167

    その他リンク: https://www.sciencedirect.com/science/article/pii/S030645492200202X

  • Prediction of the operating control rod position of the httr with supervised machine learning

    Ho H.Q., Nagasumi S., Shimazaki Y., Ishii T., Iigaki K., Goto M., Simanullang I.L., Fujimoto N., Ishitsuka E.

    International Conference on Nuclear Engineering, Proceedings, ICONE   2   2022年   ISBN:9784888982566

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    出版者・発行元:International Conference on Nuclear Engineering, Proceedings, ICONE  

    During operation of the HTTR, hundreds of technical signals and operating conditions must be observed and evaluated to ensure safe operation of the reactor. The accumulated experiment data of the HTTR is not only important for the HTTR operation, but also for the basic development of the HTGR in general. Artificial intelligence (AI) and particularly machine learning (ML) could give the ability to make predictions as well as allow the extraction of key information about physical process from large datasets. Hence, there is a lot of potential to apply AI and ML to predict the operating and safety parameters of the HTTR. In this study, the control rod position of the HTTR is predicted based on ML without using the conventional neutronic codes. The ML with a linear regression algorithm finds a functional relationship between the input dataset and a reference dataset, constructing a function that predicts control rod position from the other operation conditions. As result, the ML gives a good prediction of the HTTR control rod position with less than 5% difference compared to that in the experiment. With increasingly complicated experiments that create a large amount of data, ML is also expected to improve the design and safety analysis of the HTTR in the future.

    DOI: 10.1115/ICONE29-90818

    Scopus

  • Preparation method of origen2 library for high temperature gascooled reactors

    Simanullang I.L., Fukuhara K., Morita K., Fukaya Y., Ho H.Q., Nagasumi S., Iigaki K., Ishitsuka E., Fujimoto N.

    International Conference on Nuclear Engineering, Proceedings, ICONE   2   2022年   ISBN:9784888982566

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    出版者・発行元:International Conference on Nuclear Engineering, Proceedings, ICONE  

    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 %.

    DOI: 10.1115/ICONE29-90755

    Scopus

  • Consequence analysis of a postulated nuclear excursion in BWR spent fuel pool using 1/fβ spectrum model of randomization 査読 国際誌

    147   2020年7月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.anucene.2020.107675

  • Burnup performance of MOX and Pu-ROX fuels in PBR with accumulative fuel loading scheme 査読 国際誌

    Irwan Liapto Simanullang, Toru Obara

    Annals of Nuclear Energy   2018年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.anucene.2018.06.014

  • Potential for the use of a liquid activation circulation material to measure the neutron flux in a high gamma-ray background 査読 国際誌

    Irwan Liapto Simanullang, Jun Nishiyama, Toru Obara

    Nuclear Instrumentations and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment   903   109 - 113   2018年9月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.nima.2018.06.066

  • Burnup performance of a PBR with an accumulative fuel loading scheme utilizing burnable poison particles in UO2 and ROX fuels 招待 査読 国際誌

    Irwan Liapto Simanullang, Toru Obara

    Energy Procedia   131   61 - 68   2017年10月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.egypro.2017.09.476

  • Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme 査読 国際誌

    Irwan Liapto Simanullang, Toru Obara

    Annals of Nuclear Energy   107   110 - 118   2017年9月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.anucene.2017.04.019

  • Improvement of core design of small pebble bed reactor with accumulative fuel loading scheme 査読 国際誌

    Irwan Liapto Simanullang, Toru Obara

    Annals of Nuclear Energy   94   87 - 92   2016年8月

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    記述言語:英語   掲載種別:研究論文(学術雑誌)  

    DOI: https://doi.org/10.1016/j.anucene.2016.02.027

  • Burnup analysis of accumulative fuel loading scheme pebble bed reactor with optimum fuel composition 査読 国際誌

    Simanullang Irwan Liapto, Toru Obara

    American Nuclear Society   113 ( 1 )   2015年10月

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    記述言語:英語   掲載種別:研究論文(国際会議プロシーディングス)  

  • Improvement of burnup analysis for pebble bed reactors with an accumulative fuel loading scheme 査読 国際誌

    Irwan Liapto Simanullang, Toru Obara

    International Conference of Nuclear Engineering (ICONE)   2015年5月

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    記述言語:英語   掲載種別:研究論文(国際会議プロシーディングス)  

▼全件表示

講演・口頭発表等

  • Introducing the Next Generation of Nuclear Power Plant: Small Modular Reactor High Temperature Gas Cooled Reactors 招待 国際会議

    Irwan Liapto Simanullang

    Kyushu University Forum "Kyudai Now" in Thailand  2024年3月  Kyushu University Institute for ASIAN and OCEANIAN STUDIES

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    開催年月日: 2024年3月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Crown Plaza Bangkok Lumpini Park   国名:タイ王国  

  • Machine Learning Implementation for Improving Parameter PRediction in High Temperature Gas Cooled Reactors (HTGR) 招待

    Irwan Liapto Simanullang

    2024年12月 

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    開催年月日: 2024年12月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Core Geometry and Reflector Optimization of 10 MWt Micro-Peluit Pebble Bed HTGR 国際共著 国際会議

    F. Miftasani, N. Widiawati, N. Trianti, T. Setiadipura, D. Irwanto, Z. Su'ud, Irwan Simanullang

    International Conference on Small Modular Reactors and Their Applications  Internaitonal Atomic Energy Agency

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    開催年月日: 2024年10月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Vienna   国名:オーストリア共和国  

  • Power Distribution Optimized in a Prismatic Type HTGR using MgO-BeO as a Composite Moderator

    Irwan Liapto Simanullang

    2024年9月  Atomic Energy Society of Japan

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    開催年月日: 2024年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Tohoku University Kawauchi Kita Campus   国名:日本国  

  • Cd sample reactivity measurements at UTR-KINKI

    Nozomu Fujimoto, Soichiro Moriya, Irwan Liapto Simanullang, Atsushi Sakon

    Atomic Energy Society of Japan  2024年3月 

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    開催年月日: 2024年3月

    記述言語:日本語  

    国名:日本国  

  • Introducing the Next Generation of Nuclear Power Plant: Small Modular High Temperature Gas Cooled Reactors 招待 国際会議

    Irwan L. Simanullang

    Kyushu University Forum  2024年3月 

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    開催年月日: 2024年3月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:タイ王国  

  • Research Activities on Advanced Nuclear Reactors Based on HTGR in Kyushu University 招待 国際会議

    Irwan L. Simanullang

    Bandung Institute of Technology  2024年3月 

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    開催年月日: 2024年3月

    記述言語:英語   会議種別:シンポジウム・ワークショップ パネル(公募)  

    国名:インドネシア共和国  

  • Preliminary Study of The Composite Moderator Concept in Small High Temperature Gas Cooled Reactors

    Irwan L. Simanullang, N. Fujimoto

    Atomic Energy Society of Japan  2023年9月 

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    開催年月日: 2023年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Preliminary Study of Burnup Measurement and Relative Power Distribution in HTTR using Gamma-ray Measurement 国際会議

    Irwan L. Simanullang, #S. Kawaguchi, N. Fujimoto, T. Ishii, S. Nagasumi. .H.H.Quan. K. Nakajima, E. Ishitsuka, K. Iigaki

    2023年9月 

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    開催年月日: 2023年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Pengembangan Reaktor Nuklir Generasi IV Skala Kecil Berbasis HTGR 招待 国際会議

    Irwan L. Simanullang

    Indonesian Diaspora Scientific Forum  2023年8月 

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    開催年月日: 2023年8月

    記述言語:その他  

    国名:インドネシア共和国  

  • 高温ガス炉の燃料体内における詳細出力分布の予備評価

    #楠木捷斗, Irwan L. Simanullang, 藤本望

    日本原子力学会  2023年3月 

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    開催年月日: 2023年3月

    記述言語:日本語   会議種別:口頭発表(一般)  

    開催地:Tokyo University   国名:日本国  

  • Evaluation of Cadmium sample reactivity worth at the KUCA graphite moderated system core (2) Evaluation by diffusion theory

    #Moriya Soichiro, #Seiji Yamasaki, Nozomu Fujimoto, Irwan liapto Simanullang, Yoshiyuki Takahashi, Atsushi Sakon, Tadashi Sano

    2022年9月 

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    開催年月日: 2022年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  • The preliminary study of methods for burn-up and mass measurements of spent fuel of HTTR

    2022年9月 

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    開催年月日: 2022年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Evaluation of Cadmium sample reactivity worth at the KUCA graphite moderated system core (1) Evaluation by Monte Carlo simulation

    2022年9月 

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    開催年月日: 2022年9月

    記述言語:日本語   会議種別:口頭発表(一般)  

    国名:日本国  

  • PREPARATION METHOD OF ORIGEN2 LIBRARY FOR HIGH TEMPERATURE GAS-COOLED REACTORS 国際会議

    Irwan L. Simanullang, #K. Fukuhara, K. Morita, Y. Fukaya, H.Q. Ho, S. Nagasumi, K. Iigaki, E. Ishitsuka, N. Fujimoto

    ICONE29  2022年8月 

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    開催年月日: 2022年8月

    記述言語:英語  

    国名:中華人民共和国  

  • KUCAでの黒鉛減速体系におけるCdサンプルの反応度価値測定

    #山崎誠司, #守屋壮一郎, Irwan Liapto Simanullang, 藤本望, 左近敦士, 佐野忠史, 高橋佳之

    日本原子力学会九州支部  2021年12月 

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    開催年月日: 2021年12月

    記述言語:日本語  

    開催地:Online Presentation   国名:日本国  

  • 物質量・燃焼度測定方法についての予備検討

    #川口祥平, Irwan Liapto Simanullang, 藤本望

    日本原子力学会九州支部  2021年12月 

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    開催年月日: 2021年12月

    記述言語:日本語  

    開催地:Online Presentation   国名:日本国  

  • Burnup analysis of accumulative fuel loading scheme pebble bed reactor with optimum fuel composition 国際会議

    Irwan Liapto Simanullang, Toru Obara

    American Nuclear Society  2015年11月 

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    開催年月日: 2021年10月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:アメリカ合衆国  

  • Consequence analysis of postulated criticality in SFP using the randomized model of fuel debris

    Irwan Liapto Simanullang, Yuichi Yamane, Takeo Kikuchi, Kotaro Tonoike

    Atomic Energy Society of Japan (AESJ)  2020年3月 

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    開催年月日: 2020年3月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Effect of Gd2O3 to a fission number in the first pulse of a postulated nuclear excursion

    Irwan Liapto Simanullang, Yuichi Yamane, Kotaro Tonoike

    Atomic Energy Society of Japan (AESJ)  2019年3月 

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    開催年月日: 2019年3月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Introduction of MOX and Pu-ROX fuels in PBR with accumulative fuel loading scheme

    Irwan Liapto Simanullang, Toru Obara

    Atomic Energy Society of Japan (AESJ)  2017年9月 

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    開催年月日: 2017年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Introduction of rock-like oxide fuel in PBR with an accumulative fuel loading scheme

    Irwan Liapto Simanullang, Jun Nishiyama, Toru Obara

    Atomic Energy Society of Japan (AESJ)  2016年9月 

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    開催年月日: 2016年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Optimization of the initial fuel composition for small pebble bed reactor with an accumulative fuel loading scheme: (2) use of burnable poison

    Irwan Liapto Simanullang, Toru Obara

    Atomic Energy Society of Japan (AESJ)  2016年3月 

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    開催年月日: 2016年3月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Optimization of the initial fuel composition for small pebble bed reactor with an accumulative fuel loading scheme: (1) reduction of fuel enrichment

    Irwan Liapto Simanullang, Toru Obara

    Atomic Energy Society of Japan (AESJ)  2015年9月 

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    開催年月日: 2015年9月

    記述言語:英語   会議種別:口頭発表(一般)  

    国名:日本国  

  • Machine Learning Applications in Nuclear Reactors 招待

    Irwan Liapto Simanullang

    Indonesian Diaspora Scientific Forum in Japan  2024年7月  International Indonesian Scholars Association

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    記述言語:インドネシア語   会議種別:口頭発表(一般)  

    開催地:Consulate General of The Republic of Indonesia in Osaka   国名:日本国  

  • Burnable Poisons Utilization in the Small HTGR with MgO-BeO as a Composite-Based Moderator Material 国際会議

    Irwan Liapto Simanullang, Nozomu Fujimoto

    The 11th International Topical Meeting on High Temperature Reactor Technology  Institute of Nuclear and New Energy Technology, Tsinghua University

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    記述言語:英語   会議種別:口頭発表(一般)  

    開催地:Beijing   国名:中華人民共和国  

▼全件表示

所属学協会

  • Atomic Energy Society of Japan

  • International Indonesian Scholars Association

学術貢献活動

  • 学術論文等の審査

    役割:査読

    2024年

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    種別:査読等 

    外国語雑誌 査読論文数:2

  • Session Chair 国際学術貢献

    International Conference on Nuclear Criticality Safety  ( Sendai International Center Japan ) 2023年10月

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    種別:大会・シンポジウム等 

  • 学術論文等の審査

    役割:査読

    2023年

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    種別:査読等 

    国際会議録 査読論文数:4

  • 学術論文等の審査

    役割:査読

    2022年

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    種別:査読等 

    国際会議録 査読論文数:8

共同研究・競争的資金等の研究課題

  • Implementation of Machine Learning to improve and optimize parameter prediction in High Temperature Gas Cooled Reactor (HTGR)

    2024年7月 - 2025年3月

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    担当区分:研究代表者  資金種別:学内資金・基金等

  • Kyushu University Institute for Asian and Oceanian Studies (Q-AOS)/ Introducing the next generation nuclear power plant: HTGR conceptual design.

    2024年

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    資金種別:寄附金

教育活動概要

  • 教育活動について:
    量子理工学実験
    創造科学工学基礎実験
    課題集約演習

担当授業科目

  • 創造科学工学基礎実験

    2024年10月 - 2025年1月   後期

  • 量子物理工学実験

    2024年4月 - 2025年3月   通年

  • 課題演習

    2024年4月 - 2024年7月   前期

  • 創造科学工学基礎実験

    2023年10月 - 2024年3月   後期

  • 課題演習

    2023年10月 - 2023年12月   秋学期

  • 量子物理工学実験

    2023年4月 - 2024年3月   通年

  • 課題演習

    2022年10月 - 2022年12月   秋学期

  • 創造科学工学基礎実験

    2022年4月 - 2022年9月   前期

  • 量子理工学実験

    2022年4月 - 2022年6月   春学期

  • 創造科学工学基礎実験

    2021年10月 - 2021年12月   秋学期

▼全件表示

社会貢献活動

  • Study on Nuclear Energy and Hydrogen Production in Indonesia

    Embassy of The Republic of Indonesia in Japan  Japan (online)  2022年12月

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    対象: 社会人・一般, 学術団体, 企業, 市民団体, 行政機関

    種別:その他

  • Virtual Course: Introduction to Nuclear Reactor Safety

    Siwabessy Initiative  Online  2022年7月

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    対象: 社会人・一般, 学術団体, 企業, 市民団体, 行政機関

    種別:セミナー・ワークショップ

  • Nuclear Energy Technology, Challenge, Opportunity, and Competencies Needs

    Department of Nuclear Engineering and Engineering Physics, University of Gadjah Mada, Indonesia  Online Lecture  2022年5月

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    対象: 社会人・一般, 学術団体, 企業, 市民団体, 行政機関

    種別:講演会

外国人研究者等の受け入れ状況

  • Indonesia Nuclear Energy Regulatory Agency

    受入れ期間: 2024年10月 - 2024年12月   (期間):1ヶ月以上

    国籍:インドネシア共和国

    専業主体:文部科学省