Updated on 2025/03/26

Information

 

写真a

 
FUJIMOTO NOZOMU
 
Organization
Faculty of Engineering Department of Applied Quantum Physics and Nuclear Engineering Professor
School of Engineering (Concurrent)
Graduate School of Engineering (Concurrent)
Title
Professor
Contact information
メールアドレス
Tel
0928023505
Profile
高温ガス炉の設計研究・管理に携わってきた経験を元に、原子炉物理、原子炉プラントシステム等原子力システムに関する教育と、新型原子炉に関する研究を行っている。特に、黒鉛減速炉である高温ガス炉の炉心特性解析手法に関する研究、新しい原子炉システムの可能性の追求を行っている。
External link

Degree

  • Ph.D. ( 2006.9 Kyushu University )

Research History

  • 1987.4-2015.3 日本原子力研究開発機構   

    1987.4-2015.3 日本原子力研究開発機構

    1987.4 - 2015.3

Education

  • Kyushu University   工学府   エネルギー量子工学専攻 博士後期課程

    2005.10 - 2006.9

  • Kyushu University   大学院工学研究科   応用原子核工学専攻

    1985.4 - 1987.3

  • Kyushu University   工学部   応用原子核工学科

    1981.4 - 1985.3

Research Interests・Research Keywords

  • Research theme: High temperature gas-cooled reactors (HTGRs) is one of the promising reactor system in the future. To upgrade HTGR core evaluation method, researches of reactor physics and core thermo-hydraulic characteristics are carried out. Researches of innovative water cooled reactors are also carried out.

    Keyword: nuclear reactor, high temperature gas-cooled reactor, reactor physic, burnup characteristic, core dynamics

    Research period: 2015.4 - 2025.7

Awards

  • 表彰状

    2018.6   日本技術士会   長年にわたる原子力・放射線部会における活動に対して

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    Award type:Award from publisher, newspaper, foundation, etc. 

Papers

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

    Simanullang, IL; Fujimoto, N

    NUCLEAR ENGINEERING AND DESIGN   431   2025.1   ISSN:0029-5493 eISSN:1872-759X

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    Publisher:Nuclear Engineering and Design  

    Preliminary analysis of MgO-BeO composite material used as a moderator in a 50 MWt block-type high-temperature gas-cooled reactor (HTGR) was performed in our previous study. The target burnup of 80 GWd/t was achieved with a uniform fuel composition of 17 wt% 235U enrichment and 6 kg of heavy metal per fuel block. However, this resulted in high excess reactivity and a peak in axial power distribution at the core center. Therefore, this study aims to reduce excess reactivity by incorporating burnable poison (BP) material and optimize the axial power profile by introducing a nonuniform fuel composition in the core. Neutronic calculations were performed using the Monte Carlo MVP3.0 code developed by the Japan Atomic Energy Agency (JAEA). In this study, three fuel enrichments of 235U, ranging from 15 wt% to 20 wt%, were distributed across the core while maintaining a constant fuel packing fraction of 45 %. The results showed that the higher power density distribution shifted from the core's center to its upper part, leading to lower power density in the bottom region than the top. In addition, excess reactivity was reduced by inserting BP rods. Several parametric calculations were performed to achieve minimal excess reactivity without compromising the burnup target. The results showed that the BP rod with a radius of 0.7 cm and 12 wt% of Gd2O3 can reduce the excess reactivity from 25.5 %Δk/k to 13.47 % Δk/k.

    DOI: 10.1016/j.nucengdes.2024.113742

    Web of Science

    Scopus

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

    Simanullang, IL; Fujimoto, N

    NUCLEAR ENGINEERING AND DESIGN   420   2024.4   ISSN:0029-5493 eISSN:1872-759X

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    Publisher: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

    Web of Science

    Scopus

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

    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 417 (2024) 112795   2024.3

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    Language:English   Publishing type:Research paper (scientific journal)  

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

    Ho, HQ; Ishii, T; Nagasumi, S; Ono, M; Shimazaki, Y; Ishitsuka, E; Sawahata, H; Goto, M; Simanullang, IL; Fujimoto, N; Iigaki, K

    NUCLEAR ENGINEERING AND DESIGN   417   2024.2   ISSN:0029-5493 eISSN:1872-759X

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    Publisher: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

    Web of Science

    Scopus

  • Preliminary study of burnup measurement and relative power distribution in the HTTR using gamma-ray measurement Reviewed International journal

    Irwan Liapto Simanullang, Shohei Kawaguchi, Nozomu Fujimoto, Toshiaki Ishii, Satoru Nagasumi, Hai Quan Ho, Kunihiro Nakajima, Etsuo Ishitsuka, Kazuhiko Iigaki

    ICNC 2023 - The 12th International Conference on Nuclear Criticality Safety   2023.10

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  • Preliminary study of burnup measurement and relative power distribution in the HTTR using gamma-ray measurement Reviewed International journal

    Irwan Liapto Simanullang, Shohei Kawaguchi, Nozomu Fujimoto, Toshiaki Ishii, Satoru Nagasumi, Hai Quan Ho, Kunihiro Nakajima, Etsuo Ishitsuka, Kazuhiko Iigaki

    ICNC 2023 - The 12th International Conference on Nuclear Criticality Safety   2023.10

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  • High-temperature operation mode of HTTR for hydrogen production facility Invited Reviewed International journal

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

    2023.5

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  • Evaluation of burnable poison reactivity worth at the KUCA graphite moderated system Reviewed International journal

    Seiji Yamasaki, Soichiro Moriya, Irwan Liapto Simanullang, Nozomu Fujimoto, Atsushi Sakon, Tadafumi Sano, Takahashi Yoshiyuki

    2023.5

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  • High-temperature operation mode of HTTR for hydrogen production facility Invited Reviewed International journal

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

    2023.5

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  • Evaluation of burnable poison reactivity worth at the KUCA graphite moderated system Reviewed International journal

    Seiji Yamasaki, Soichiro Moriya, Irwan Liapto Simanullang, Nozomu Fujimoto, Atsushi Sakon, Tadafumi Sano, Takahashi Yoshiyuki

    2023.5

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

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

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2023.30 ( 0 )   1543   2023   eISSN:24242934

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    Language:English   Publisher:The Japan Society of Mechanical Engineers  

    <p>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.</p><p>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.</p><p>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.</p>

    DOI: 10.1299/jsmeicone.2023.30.1543

    CiNii Research

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

    Yamasaki Seiji, Moriya Soichiro, Simanullang Irwan L., Fujimoto Nozomu, Sakon Atsushi, Sano Tadafumi, Takahashi Yoshiyuki

    The Proceedings of the International Conference on Nuclear Engineering (ICONE)   2023.30 ( 0 )   1271   2023   eISSN:24242934

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    Language:English   Publisher:The Japan Society of Mechanical Engineers  

    <p>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.</p><p>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.</p><p>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.</p>

    DOI: 10.1299/jsmeicone.2023.30.1271

    CiNii Research

  • 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   ISBN:9784888982566

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    Publisher: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

  • Preparation Method of ORIGEN2 Library for High-Temperature Gas-Cooled Reactors (HTGRs) Reviewed International journal

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

    29th International Conference on Nuclear Engineering ICONE29   2022.8

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

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

    Nuclear Engineering and Design   396   2022.8   ISSN:0029-5493 eISSN:1872-759X

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    Language:English   Publisher: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

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  • Prediction of the Operating Control Rod Position of the HTTR with supervised Machine Learning Reviewed International journal

    Hai Quan Ho, Satoru Nagasumi, Youske Shimazaki, Toshiaki Ishii, Kazuhiko Iigaki, Minoru Goto, Irwan Liapto Simanullang, Nozomu Fujimoto, Etsuo Ishituka

    29th International Conference on Nuclear Engineering ICONE29   2022.8

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  • Prediction of the Operating Control Rod Position of the HTTR with supervised Machine Learning Reviewed International journal

    Hai Quan Ho, Satoru Nagasumi, Youske Shimazaki, Toshiaki Ishii, Kazuhiko Iigaki, Minoru Goto, Irwan Liapto Simanullang, Nozomu Fujimoto, Etsuo Ishituka

    29th International Conference on Nuclear Engineering ICONE29   2022.8

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  • Preparation Method of ORIGEN2 Library for High-Temperature Gas-Cooled Reactors (HTGRs) Reviewed International journal

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

    29th International Conference on Nuclear Engineering ICONE29   2022.8

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

    Irwan Liapto Simanullang, Naoki Nakagawa, Hai Quan Ho, Satoru Nagasumi, Etsuo Ishitsuka, Kazuhiko Iigaki, Nozomu Fujimoto

    Annals of Nuclear Energy   177   2022.7   ISSN:0306-4549 eISSN:1873-2100

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    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

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  • 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   ISBN:9784888982566

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    Publisher: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   ISBN:9784888982566

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    Publisher: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

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  • Nuclear data processing code FRENDY: A verification with HTTR criticality benchmark experiments Reviewed International journal

    Nozomu Fujimoto, Kenichi Tada, Hai Quan Ho, Shimpei Hamamoto, Satoru Nagasumi, Etsuo Ishitsuka

    Annals of Nuclear Energy   158 ( 108270 )   2021.4

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    DOI: 10.1016/j.anucene.2021.108270

  • Preparation for restarting the high temperature engineering test reactor: Development of utility tool for auto seeking critical control rod position Reviewed International journal

    Hai Quan Ho, Nozomu Fujimoto, Shimpei Hamamoto, Satoru Nagasumi, Minoru Goto, Etsuo Ishitsuka

    Nuclear Engineering and Design   377   2021.3

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    DOI: 10.1016/j.nucengdes.2021.111161

  • Promising Neutron Irradiation Applications at the High Temperature Engineering Test Reactor Invited Reviewed International journal

    Hai Quan Ho, Yuki Honda, Shimpei Hamamoto, Toshiaki Ishii, Shoji Takada, Nozomu Fujimoto, Etsuo Ishitsuka

    Journal of Nuclear Engineering and Radiation Science.   Vol.6   2020.6

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    DOI: 10.1115/1.4044529

  • Conceptual design of direct production of 99mTc at the high temperature engineering test reactor Reviewed International journal

    Hai Quan Ho, Hiroki Ishida, Shimpei Hamamoto, Toshiaki Ishii, Nozomu Fujimoto, Naoyuki Takaki, Etsuo Ishitsuka

    Nuclear Engineering and Design   ( 352 )   2019.7

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  • Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor Reviewed

    Hai Quan Ho, Yuki Honda, Shimpei Hamamoto, Toshiaki Ishii, Nozomu Fujimoto, Etsuo Ishitsuka

    Applied Radiation and Isotopes   140   209 - 214   2018.10

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    The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors (HTGRs) was investigated. A high-temperature engineering test reactor (HTTR), which is located in Japan at Oarai-machi Research and Development Center, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the
    125
    I production could be maximized and the
    126
    I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 1.8 × 10
    5
    GBq/y of
    125
    I production.

    DOI: 10.1016/j.apradiso.2018.07.024

  • Burn-up dependency of control rod position at zero power criticality in the High Temperature Engineering Test Reactor Reviewed International journal

    Yuki Honda, Nozomu Fujimoto, Hiroaki Sawahata, Shoji Takada, Kazuhiro Sawa

    American Society of Mechanical Engineering   2017 ( Vol.3 )   2017.1

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    The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor (HTGR) constructed in Japan, firstly. The operating data of the HTTR with burn-up is very important for developments of HTGRs. Many test data have been collected in the HTTR. Many tests are carried out in low power operation. On the other hand, the full power operation is not enough. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate the calculation code. Additionally, it is difficult to measure core temperature in HTTR. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle and the temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for confirm the characteristics of burn-up and validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.

    DOI: 10.1115/1.4033812

  • Benchmark study on realized random packing model for coated fuel particles of HTTR using MCNP6 Reviewed

    H. H. Quan, Koji Morita, Y. Honda, Nozomu Fujimoto, S. Takada

    2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017 2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings   2017.1

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    The Coated Fuel Particle plays an important role in the excellent safety feature of the High Temperature Gascooled Reactor. However, the random distribution of CFPs also makes the simulation of HTGR fuel become more complicated. The Monte Carlo N-particle (MCNP) code is one of the most well-known codes for validation of nuclear systems; unfortunately, it does not provide an appropriate function to model a statistical geometry explicitly. In order to deal with the stochastic media, a utility program for the random model, namely Realized Random Packing (RRP), has been developed particularly for High Temperature engineering Test Reactor (HTTR). This utility program creates a number of random points in an annular geometry. Then, these random points will be used as the center coordinate of CFPs in the MCNP6 input file and therefore the actual random arrangement of CFPs can be simulated explicitly. First, a pin-cell calculation was carried out to validate the RRP by comparing with Statistical Geometry (STG) model of MVP code. After that, the comparison between the RRP model (MCNP) and STG model (MVP) was shown in whole core criticality calculation, not only for the annular core but also for the fully-loaded core. The comparison of numerical results showed that the RRP model and STG model differed insignificantly in the multiplication factor as expected, regardless of the pin-cell or whole core calculations. In addition, the RRP model did not make the calculation time increase a lot in comparison with the conventional regular model (uniform arrangement).

  • Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor Reviewed International journal

    Yuki Honda, Nozomu Fujimoto, Hiroaki Sawahata, Shoji Takada, Kazuhiro Sawa

    American Society of Mechanical Engineering   January 2017 ( Vol. 3 )   2017.1

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    The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation
    than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

    DOI: 10.1115/1.4033813

  • Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR Reviewed International journal

    Daisuke Tochio, Nozomu Fujimoto

    Journal of Nuclear Science and Technology   53 ( 3 )   425 - 431   2016.3

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    The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.

    DOI: 10.1080/00223131.2015.1054910

  • Improvement of cell model for control rod in reflector region of high temperature test engineering reactor Reviewed

    Yuki Honda, Nozomu Fujimoto, Sawahata Hiroaki, Sawa Kazuhiro

    23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015 ICONE 2015 - 23rd International Conference on Nuclear Engineering Nuclear Power - Reliable Global Energy   2015-January   2015.1

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    The High Temperature Engineering Test Reactor (HTTR) [1] is a block type fuel High Temperature Gas-cooled Reactor. There are 32 control rods (16 pairs) in the HTTR. The 6 pairs of control rods are inserted into a core region and the others are inserted in a reflector region surrounding the core. The core temperature of the HTTR is too high to insert all control rods simultaneously at reactor scram near full power operation for keeping integrity of control rods metallic sleeve. Therefore, a two-step control rods insertion method for reactor scram is adopted. The reactivity inserted at the two-step control rod insertion method was measured at HTTR criticality tests. The calculated reactivity at the firststep showed larger underestimation than that of the second-step. On the other hand, calculated results of excess reactivity at the HTTR criticality tests showed good agree with tests. It is considered that a cell model for reflector region control rod is not suitable. Therefore, this paper focuses on a new cell model for control rods in a reflector region. In a previous control rod cell model, control rod is surrounded by fuel blocks only. The surrounding condition of the new cell model corresponds to the configuration around the reflector region control rod. The calculated reactivity at the first-step using the new cell model shows better results than previous calculation. It is considered that the new cell model brings appropriate neutron flux distribution around control rods in reflector region.

  • Benchmark evaluation of start-up and zero-power measurements at the high-temperature engineering test reactor Reviewed

    John D. Bess, Nozomu Fujimoto

    Nuclear Science and Engineering   178 ( 3 )   414 - 427   2014.11

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    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the high-temperature engineering test reactor (HTTR). Additional measurements of the subcritical configuration of the fully loaded core, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely detailed models of the HTTR as much of the design information is still proprietary. The uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. However, use of the benchmark critical configurations of the HTTR for nuclear data adjustment is not recommended as the impact of these biases has not been addressed with rigorous detail. The impact of any simplification biases, if any, is not expected to significantly impact evaluation of the other reactor physics measurement calculations. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that compose the HTTR. Monte Carlo calculations of kff are between ∼0.9&#37; and ∼2.7&#37; greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulations of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

    DOI: 10.13182/NSE14-14

  • Operation and maintenance experience from the HTTR database Reviewed International journal

    Atsushi Shimizu, Takayuki Furusawa, Fumitaka Homma, Hiroyuki Inoi, Masayuki Umeda, Masaaki Kondo, Minoru Isozaki, Nozomu Fujimoto, Tatsuo Iyoku

    Atomic Energy Society of Japan   51 ( 11-12 )   1444 - 1451   2014.8

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    The Japan Atomic Energy Agency has been establishing a database of operation and maintenance experience for the High Temperature Engineering Test Reactor. The objective of this database is to share information from operation and maintenance experience and make use of the knowledge gained in the design, construction, and operation of future High Temperature Gas-cooled Reactors (HTGRs). Between 1997 and 2012, more than 1000 events have been registered in this database system.
    This paper describes trends in operation and maintenance events recorded in this database, including experience gained from the Great East Japan Earthquake. The paper also identifies the following significant items that are expected to be useful in the design of future HTGRs: (1) performance degradation of helium gas compressors, (2) malfunction of the reserved shutdown system in the reactivity control system, (3) problems with emergency gas turbine generators, and (4) consequences of the Great East Japan Earthquake.

    DOI: 10.1080/00223131.2014.946568

  • Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor Reviewed International journal

    John D. Bess, Nozomu Fujimoto

    Nuclear Science and Engineering   178   414 - 427   2014.6

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    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zeropower
    measurements of the high-temperature engineering test reactor (HTTR). Additional measurements
    of the subcritical configuration of the fully loaded core, core excess reactivity, shutdown margins, six
    isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as
    acceptable benchmark experiments. Insufficient information is publicly available to develop finely
    detailed models of the HTTR as much of the design information is still proprietary. The uncertainties in
    the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias
    uncertainties incurred through the simplification process used to develop the benchmark models.
    However, use of the benchmark critical configurations of the HTTR for nuclear data adjustment is not
    recommended as the impact of these biases has not been addressed with rigorous detail. The impact of
    any simplification biases, if any, is not expected to significantly impact evaluation of the other reactor
    physics measurement calculations. Dominant uncertainties in the experimental keff for all core
    configurations come from uncertainties in the impurity content of the various graphite blocks that
    compose the HTTR. Monte Carlo calculations of keff are between *0.9&#37; and *2.7&#37; greater than the
    benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulations of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  • Establishment of floating support technology applied to high-temperature components and piping in HTTR Reviewed

    Masanori Shinohara, Yoshitomo Inaba, Shimpei Hamamoto, Nozomu Fujimoto

    journal of nuclear science and technology   51   1398 - 1406   2014.1

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    In the primary cooling system of the High Temperature Engineering Test Reactor (HTTR) with an outlet coolant temperature of 950°C, high-temperature components and piping such as an intermediate heat exchanger and coaxial double piping reach very high temperature, and large and complex thermal displacements arise in them. In order not only to absorb the thermal displacements but also to withstand earthquakes, the HTTR has adopted a new three-dimensional floating support system. In the limited space of the containment vessel, the support system can support the components and piping's own weights and follow the thermal displacements and have seismic capacity. On the other hand, the adoption of the support system was unprecedented in nuclear plants. Thus, the effectiveness of the support system was demonstrated through the HTTR operation. In this paper, by using the HTTR operation data, the thermal displacement behavior of the high-temperature components and piping is investigated, and the behavior and characteristics are simulated numerically. In addition, the aftermath of the Great East Japan Earthquake on the HTTR is confirmed. As a result, the effectiveness of the three-dimensional floating support system adopted by the HTTR is verified.

    DOI: 10.1080/00223131.2014.967734

  • High-temperature continuous operation of the HTTR Reviewed

    Kuniyoshi Takamatsu, Kazuhiro Sawa, Kazuhiko Kunitomi, Ryutaro Hino, Masuro Ogawa, Yoshihiro Komori, Toshio Nakazawa, Tatsuo Iyoku, Nozomu Fujimoto, Tetsuo Nishihara, Masayuki Shinozaki

    Transactions of the Atomic Energy Society of Japan   10 ( 4 )   290 - 300   2011.12

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    A high-temperature gas-cooled reactor (HTGR) is a graphite-moderated and helium gas-cooled reactor. It is particularly attractive due to its capability of producing high-temperature helium gas, and its passive and inherent safety features. To enable nuclear energy application to a wide range of heat process industries, Japan Atomic Energy Agency (JAEA) has continued extensive effort for the development of the HTGR using the high-temperature engineering test reactor (HTTR), which is the first HTGR in Japan with a thermal power of 30 MW, and operates it at the site of the JAEA's Oarai Research and Development Center. The HTTR has successfully completed a full-power high-temperature (950°C) continuous operation for 50 days from January to March in 2010. Through this operation, the potential of a stable high-temperature heat supply to heat application systems, such as a hydrogen production system, was demonstrated. This paper presents the operation results including reactor characteristics.

    DOI: 10.3327/taesj.J11.020

  • 高温工学試験研究炉(HTTR)の高温連続運転 Reviewed

    高松 邦吉, 沢 和弘, 國富 一彦, 日野 竜太郎, 小川 益郎, 小森 芳廣, 中澤 利雄, 伊与久 達夫, 藤本 望, 篠崎 正幸

    日本原子力学会和文論文誌   10 ( 4 )   290 - 300   2011.7

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  • Experimental validation of effectiveness of rod-type burnable poisons on reactivity control in HTTR Reviewed International journal

    Minoru Goto, Shusaku Shiozawa, Nozomu Fujimoto, Shigeaki Nakagawa, Yasuyuki Nakao

    Nuclear Engineering and Design   240   2994 - 2998   2010.10

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    In block-type high temperature gas-cooled reactors (HTGRs), insertion depth of control rods (CRs) into a core should be retained shallow to keep fuel temperature below 1495 ◦C through a burnup period, and hence excess reactivity should be reduced through a different method. Loading burnable poisons (BPs) into the core is considered as a method to resolve this problem as in case of light water reactors (LWRs).
    Effectiveness of BPs on reactivity control in LWRs has been validated by experimental data, however, this has not been done yet for HTGRs, because there was not enough burnup characteristics data for HTGRs required for the validation. The High Temperature Engineering Test Reactor (HTTR) is a block-type HTGRs and it adopts rod-type BPs to control reactivity. The HTTR has been operated up to middle burnup, and thereby the experimental data was expected to show effect of the BPs on the reactivity control. Hence, in order to validate effectiveness of rod-type BPs on reactivity control in the HTTR, we investigated on the HTTR results whether the BPs have functioned as designed. As a result, the CRs insertion depth has been retained shallow within allowable range, and then effectiveness of rod-type BPs on reactivity control in the HTTR was validated.

  • Analysis for HTTR burnup characteristics Reviewed

    Minoru Goto, Nozomu Fujimoto, Shigeaki Nakagawa

    International Congress on Advances in Nuclear Power Plants 2009, ICAPP 2009   1   309 - 314   2009.5

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    Comparison of burnup characteristics of the High Temperature Engineering Test Reactor (HTTR) were conducted between experimental data and analytical results, and then applicability of a three dimensional whole core burnup calculation method to burned block-type HTGRs was studied. The calculations were performed by the SRAC/COREBN based on diffusion theory. We focused on changes in control rods (CRs) position with burnup as one of the HTTR burnup characteristics, and analyzed that in rated power with 30MW and zero power, respectively. In the HTTR experiments, different tendencies were observed in the changes in CRs position with burnup between rated and zero power. CRs position in rated power decreased with burnup, namely, insertion depth of CRs into a core increased with burnup. Meanwhile that in zero power was almost constant through burnup. In the analysis, similar results to the above were obtained. Thus, applicability of the three dimensional whole core burnup calculation method to burned block-type HTGRs was confirmed by experimental data.

  • Evaluation of fuel temperature on high temperature test operation at high temperature gas-cooled reactor 'HTTR' Reviewed

    Daisuke Tochio, Junya Sumita, Eiji Takada, Nozomu Fujimoto, Shigeaki Nakagawa

    Transactions of the Atomic Energy Society of Japan   5 ( 1 )   57 - 67   2006.1

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    High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Agency (JAEA) achieved the reactor outlet coolant temperature of 950°C for the first time in the world at Apr. 19, 2004. To ensure the thermal integrity of fuel in high temperature test operation, it is necessary that fuel temperature is designed appropriately by fuel temperature designing method, and that estimated maximum fuel temperature is lower than the thermal limit temperature. In this report, by constructing newly a realistic core-shape representing model, the current fuel temperature estimation model is improved. Moreover fuel temperature in high-temperature test operation is estimated with the newly-constructed model, and it is confirmed that estimated maximum fuel temperature in high temperature test operation is lower than the thermal limit temperature.

    DOI: 10.3327/taesj2002.5.57

  • Annular core experiments in HTTR's start-up core physics tests Reviewed

    Nozomu Fujimoto, Kiyonobu Yamashita, Naoki Nojiri, Mituo Takeuchi, Shingo Fujisaki, Masaaki Nakano

    Nuclear Science and Engineering   150 ( 3 )   310 - 321   2005.1

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    Annular cores were formed in start-up core physics tests of the High Temperature Engineering Test Reactor (HTTR) to obtain experimental data for verification of design codes. The first criticality, control rod (CR) positions at critical conditions, neutron flux distribution, excess reactivity, etc., were measured as representative data. These data were evaluated with the MVP Monte Carlo code, which can consider directly the heterogeneity of coated fuel particles (CFPs) distributed randomly in fuel compacts. It was made clear that the heterogeneity effect of CFPs on keff's for annular cores is smaller than that for fully loaded cores. The measured and the calculated k eff's agreed with each other with differences <1&#37;Δk. The calculated neutron flux distributions agreed with the measured results. A revised method was applied for evaluation of excess reactivity to exclude the negative shadowing effect of CRs. The revised and calculated excess reactivity agreed with differences <1&#37;Δk/k.

    DOI: 10.13182/NSE03-79

  • Characteristic test of initial HTTR core Reviewed

    Naoki Nojiri, Satoshi Shimakawa, Nozomu Fujimoto, Minoru Goto

    Nuclear Engineering and Design   233 ( 1-3 )   283 - 290   2004.10

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    This paper describes the results of core physics test in start-up and power-up of the HTTR. The tests were conducted in order to ensure performance and safety of the high temperature gas cooled reactor, and was carried out to measure the critical approach, the excess reactivity, the shutdown margin, the control rod worth, the reactivity coefficient, the neutron flux distribution and the power distribution. The expected core performance and the required reactor safety characteristics were verified from the results of measurements and calculations.

    DOI: 10.1016/j.nucengdes.2004.08.015

  • Validation of the nuclear design code system for the HTTR using the criticality assembly VHTRC Reviewed

    Nozomu Fujimoto, Naoki Nojiri, Kiyonobu Yamashita

    Nuclear Engineering and Design   233 ( 1-3 )   155 - 162   2004.10

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    The high temperature engineering test reactor is the first block-type HTGR designed for a 950 °C outlet gas temperature which uses low-enriched uranium fuel with burnable poison rod. For validation of the nuclear design code system for the HTTR, a critical assembly of VHTRC had been constructed. The calculation uncertainties of effective multiplication factor, neutron flux distribution, burnable poison reactivity worth, and control rod worth, temperature coefficients were evaluated. Calculation accuracy of a Monte Carlo code is also evaluated.

    DOI: 10.1016/j.nucengdes.2004.08.005

  • Nuclear design Reviewed

    Nozomu Fujimoto, Naoki Nojiri, Hiroei Ando, Kiyonobu Yamashita

    Nuclear Engineering and Design   233 ( 1-3 )   23 - 36   2004.10

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    The high-temperature engineering test reactor (HTTR) has been designed for an outlet temperature of 950 °C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor (HTGR). The functions of the reactivity control system are determined considering the operational conditions, and the reactivity balance is planned so that the design requirements are fully satisfied. Moreover, the reactivity coefficients are evaluated to confirm the safety characteristics of the reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600 °C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. The original nuclear design model had to be modified based on the first critical experiments. The Monte Carlo code MVP was also used to predict criticality of the initial core. The predicted excess reactivities are now in good agreement with the experimental results.

    DOI: 10.1016/j.nucengdes.2004.07.008

  • Experience of HTTR construction and operation - Unexpected incidents Reviewed

    Nozomu Fujimoto, Yukio Tachibana, Akio Saikusa, Masayuki Shinozaki, Minoru Isozaki, Tatuo Iyoku

    Nuclear Engineering and Design   233 ( 1-3 )   273 - 281   2004.10

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    From a viewpoint of heat leakage, there were two incidents during HTTR power-rise-tests. One was a temperature rise of the primary upper shielding, and the other was a temperature rise of the core support plate. Causes of the both incidents were small amount of helium flow in structures. For the temperature rise of the primary upper shielding, countermeasures to reduce the small amount of helium flow, enhancement of heat release and installation of thermal insulator were taken. For the temperature rise of the core support plate, temperature evaluations were carried out again considering the small amount of helium flow and design temperature of the core support plate was revised. By these countermeasures, the both temperatures were kept below their limits.

    DOI: 10.1016/j.nucengdes.2004.08.014

  • Core thermal-hydraulic design Reviewed

    Eiji Takada, Shigeaki Nakagawa, Nozomu Fujimoto, Daisuke Tochio

    Nuclear Engineering and Design   233 ( 1-3 )   37 - 43   2004.10

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    The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88&#37; of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal-hydraulic design gives conservative results.

    DOI: 10.1016/j.nucengdes.2004.07.009

  • Safety shutdown of the high temperature engineering test reactor during loss of off-site electric power simulation test Reviewed

    Takeshi Takeda, Shigeaki Nakagawa, Fumitaka Honma, Eiji Takada, Nozomu Fujimoto

    Journal of Nuclear Science and Technology   39 ( 9 )   986 - 995   2002.1

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    The high temperature engineering test reactor (HTTR) is a graphite-moderated and helium-gas-cooled reactor, which is the first high temperature gas-cooled reactor in Japan. The HTTR achieved its first full power of 30 MW at rated operation on December 7 in 2001. In the rise-to-power test of the HTTR, simulation test of anticipated operational occurrence with scram was carried out by manual shutdown of off-site electric power from 30 MW operation. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, mass flow rates of helium and water decreased to the scram points. Sixteen pairs of control rods were inserted at two-steps into the core by gravity within the design criterion of 12 s. In 51 s after the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. In 40 min after the startup of the auxiliary cooling system, one of two auxiliary helium circulators stopped for reducing thermal stresses of core graphite components such as fuel blocks. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. Blackout sequences of the HTTR dynamic components were in accordance with the design. As a result of the loss of off-site electric power simulation test, it was confirmed that the HTTR shuts down safely after the scram.

    DOI: 10.1080/18811248.2002.9715285

  • Start-up core physics tests of high temperature engineering test reactor (HTTR), (II). First criticality by an annular form fuel loading and its criticality prediction method Reviewed

    Nozomu Fujimoto, Masaaki Nakano, Mitsuo Takeuchi, Shingo Fujisaki, Kiyonobu Yamashita

    Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan   42 ( 5 )   458 - 464   2000.1

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    The HTTP has achieved the first criticality on 11/10/98. The fuels were loaded into the core from outer region to inner region to obtain characteristic data of annular cores. The annular core is expected to be a core type of future HTGR. Fuel loading schedule was planned based on preliminary calculations by Monte Carlo method. These calculations predicted the first criticality at 16 ± 1 columns. However, the reactor achieved the first criticality at 19 columns. The first criticality was re-predicted by comparing measured and calculated 1/M curves. The calculated 1/M curves were obtained for different critical mass adjusting some parameters such as the amount of impurities, etc. The method is called "1/M sandwich method". This method well predicted the number of fuel columns at the first criticality. The combination of this method with Monte Carlo calculation was a rational method for predicting the first criticality. It was confirmed that Monte Carlo calculation could be used for the evaluation of HTTR with < 1&#37; of error for fully loaded core.

    DOI: 10.3327/jaesj.42.458

  • Startup core physics tests of High Temperature Engineering Test Reactor(HTTR), (I) test plan, fuel loading and nuclear characteristics tests Reviewed

    Kiyonobu Yamashita, Nozomu Fujimoto, Mituo Takeuchi, Shingo Fujisaki, Masaaki Nakano, Masayuki Umeda, Takeshi Takeda, Haruyoshi Mogi, Toshiyuki Tanaka

    Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan   42 ( 1 )   30 - 42   2000.1

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    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium-cooled reactor which has 30 MW of thermal power and 950°C of outlet coolant-gas temperature. The fuel loading of the HTTR was started on July 1, 1998, from the core periphery. The first criticality was attained in annular type core of 19 columns on Nov. 10, 1998. The startup core physics tests consisted mainly of tests for licensing and tests for establishing the technology bases necessary for HTGRs. It was confirmed in the former tests that all fuel blocks are loaded in certain positions and the excess reactivity is less than the limit. Experimental data for the annular core are obtained in the latter tests. Also, it was confirmed that the inverse kinetics method and delayed integral counting method are useful for the measurement of scram reactivity even if it takes about 10s for rod insertions. Furthermore, control rod worth curve, axial neutron flux distribution, etc. were measured to grasp the core performance. All tests planned in the startup core physics tests had been successfully performed and were completed on Jan. 21, 1999. It was confirmed from the tests that the HTTR was capable to step up to the power ascension tests.

    DOI: 10.3327/jaesj.42.30

  • Measuring method of reactivity worth of control rod with long falling time by IKRD technique. Reviewed

    Kiyonobu Yamashita, Mituo Takeuchi, Nozomu Fujimoto, Shingo Fujisaki, Masaaki Nakano, Naoki Nojiri, Seiji Tamura

    Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan   41 ( 1 )   35 - 38   1999.1

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    DOI: 10.3327/jaesj.41.35

  • Weapons-grade plutonium burning with high-temperature gas-cooled reactors using plutonium burner balls and thorium breeder balls Reviewed

    Kiyonobu Yamashita, Kazumi Tokuhara, Nozomu Fujimoto

    Nuclear Science and Engineering   126 ( 1 )   94 - 100   1997.1

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    A concept for a new reactor system is developed where weapons-grade plutonium can be made worthless for weapons use. It is a pebble bed-type high-temperature gas-cooled reactor that uses plutonium burner ball and thorium breeder ball fuels. The residual amount of 239Pu in spent plutonium balls becomes <1&#37; of the initial loading. The power coefficient is made negative by reducing the parasitic neutron absorption reaction rate of 135Xe.

    DOI: 10.13182/NSE97-A24460

  • Nuclear design of the high-temperature engineering test reactor (HTTR) Reviewed

    Kiyonobu Yamashita, Ryuichi Shindo, Isao Murata, So Maruyama, Nozomu Fujimoto, Takeshi Takeda

    Nuclear Science and Engineering   122 ( 2 )   212 - 228   1996.2

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    The high-temperature engineering test reactor has been designed whose outlet gas temperature is 950°C. That is the highest temperature in the world for a block-type high-temperature gas-cooled reactor. The power distribution in the core was optimized by changing the uranium enrichment to maintain the fuel temperature at less than the limit (1600°C). Deviation from the optimized distribution due to the burnup of fissile materials was avoided by flattening time-dependent changes in local reactivities. Flattening was achieved by optimizing the specifications of the burnable poisons. Control rod destruction of the optimized power distribution was avoided by limiting the depth of insertion. The insertion depth of the control rods is limited by reducing the excess reactivity of the whole core by the burnable poisons to the minimum value necessary for operations.

    DOI: 10.13182/NSE96-A24156

  • HTGR type minor actinide transmutation reactor Reviewed

    Nozomu Fujimoto, Kiyonobu Yamashita, Kazumi Tokuhara

    Nippon Genshiryoku Gakkaishi/Journal of the Atomic Energy Society of Japan   38 ( 4 )   304 - 306   1996.1

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    DOI: 10.3327/jaesj.38.304

  • Evaluation of core thermal and hydraulic characteristics of HTTR Reviewed

    So Maruyama, Nozomu Fujimoto, Yukio Sudo, Tomoyuki Murakami, Sadao Fujii

    Nuclear Engineering and Design   152 ( 1-3 )   183 - 196   1994.11

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    Japan Atomic Energy Research Institute has started the development of the high temperature engineering test reactor (HTTR), a graphite-moderated, helium gas-cooled reactor with 30 MW thermal power and maximum outlet coolant temperature of 950 °C. This paper describes the core thermal and hydraulic (T/H) design procedure, including the validation of the computer code system, design criteria pertaining to the fuel design limit and the evaluated core T/H charateristics. The core T/H design of the HTTR has been carried out considering the specific characteristics of the core structure and the fuel based on R&D results. The coolant flow rate and temperature distribution are evaluated by the flow network analysis code flownet. The fuel temperature distribution is evaluated by the fuel temperature analysis code temdim with multi-cylindrical model using hot spot factors. Fuel design limit for anticipated operational occurrences and fuel temperature limit for normal operation are specified at 1600°C and 1495°C, respectively based on experimental results. Several design considerations are also adopted to realize a high reactor outlet coolant temperature of 950°C. As a result of core T/H design, the effective core flow rate and maximum fuel temperature during the high temperature test operation are 88&#37; and 1492°C, respectively.

    DOI: 10.1016/0029-5493(94)90084-1

  • Design of High Temperature Engineering Test Reactor (HTTR) Reviewed

    Shinzo SAITO, Toshiyuki TANAKA, Yukio SUDO, Osamu BABA, Masami SHINDO, Shusaku SHIOZAWA, Haruyoshi MOGI, Minoru OKUBO, Noboru ITO, Ryuichi SHINDO, Noriaki KOBAYASHI, Ryoichi KURIHARA, Kimio HAYASHI, Kazuhiko HADA, Yuji KURATA, Kiyonobu YAMASHITA, Kozo KAWASAKI, Tatsuo IYOKU, Kazuhiko KUNITOMI, So MARUYAMA, Masahiro ISHIHARA, Kazuhiro SAWA, Nozomu FUJIMOTO, Isao MURATA, Shigeaki NAKAGAWA, Yukio TACHIBANA, Tetsuo NISHIHARA, Shinichi OSHITA, Masayuki SHINOZAKI, Takeshi TAKEDA, Shigeaki SAKABA, Akio SAIKUSA, Yujiro TAZAWA, Yoshio FUKAYA, Hiroshi NAGAHORI, Takayuki KIKUCHI, Satoshi KAWAJI, Minoru ISOZAKI, Shinjiro MATSUZAKI, Iwao SAKAMA, Kunio HARA, Noriaki UEDA, Shigeru KOKUSEN

    JAERI-1332   1994.9

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    Language:English  

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30MW in thermal output and outlet coolant temperature of 850 °C for rated operation and 950 °C for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests.
    JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R&D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation.

  • Evaluation Of Local Power Distribution With Fine-Mesh Core Model For High Temperature Engineering Test Reactor (HTTR) Reviewed

    Isao Murata, Kiyonobu Yamashita, So Maruyama, Ryuichi Shindo, Nozomu Fujimoto, Yukio Sudo

    journal of nuclear science and technology   31 ( 1 )   62 - 72   1994.1

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    In the high temperature gas-cooled reactors (HTGRs), the radial and axial heterogeneity resulted from a combination of fuel rods, burnable poison rods, block end graphite and so on causes local power peakings which increase the fuel temperature locally. An method was developed for calculating the local power and the fuel temperature distributions. This method deals with all heterogeneity effects of a whole core in the radial and axial directions with a design code system including a vectorized 3-dimensional diffusion code. The uncertainty of the method had been evaluated through the analyses of the power distribution obtained by critical experiments with the Very High Temperature Reactor Critical Assembly (VHTRC). The difference was less than 3&#37; between the calculated and measured power distributions. From the results, it was confirmed that this method could predict the local power distribution of the HTGR with high accuracy. This method was applied to the evaluation of the fuel temperature of the HTTR. It was shown that the maximum fuel temperature would be lower than the design limit of 1,495°C for the normal operation and that of 1,600°C for the anticipated operational transients.

    DOI: 10.1080/18811248.1994.9735115

  • Evaluation of hot spot factors for thermal and hydraulic design of HTTR Reviewed

    So Maruyama, Kiyonobu Yamashita, Nozomu Fujimoto, Isao Murata, Yukio Sudo, Tomoyuki Murakami, Sadao Fujii

    journal of nuclear science and technology   30 ( 11 )   1186 - 1194   1993.1

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    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950°C in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly.

    DOI: 10.1080/18811248.1993.9734606

  • Optimization of power distribution to achieve outlet gas-coolant temperature of 950°c for httr Reviewed

    Kiyonobu Yamashita, So Maruyama, Isao Murata, Ryuichi Shindo, Nozomu Fujimoto, Yukio Sudo, Tetsuo Nakata, Kazumi Tokuhara

    journal of nuclear science and technology   29 ( 5 )   472 - 481   1992.5

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    This report presents the optimization result with respect to the spatial power distribution of the High Temperature engineering Test Reactor (HTTR) core to achieve a high outlet coolant gas temperature of 950° C. At first, the power distribution optimization procedure was developed to achieve a high outlet coolant gas temperature while maintaining the fuel temperature as low as possible. Secondarily, the optimization procedure thus developed was applied for the power distribution design of the HTTR core. The maximum nominal fuel temperature was reduced about 300°C through the optimization and was 1,321°C. By the power distribution optimization, the maximum fuel temperature was maintained less than the fuel temperature design limit of 1,600°C, even accounting for the temperature increase at the hot spot and the anticipated operational occurrences.

    DOI: 10.1080/18811248.1992.9731553

  • CONTROL SYSTEM DESIGN OF VERY HIGH TEMPERATURE GAS COOLED REACTOR FOR START, STOP AND LOAD FOLLOW OPERATIONS. Reviewed

    Kazuhiko Kudo, Nozomu Fujimoto, Masao Ohta, Takaaki Ohsawa, Yasuyuki Nakao, Yoshikuni Shinohara

    Memoirs of the Kyushu University, Faculty of Engineering   47 ( 1 )   75 - 84   1987.3

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    Very High Temperature Gas Cooled Reactor (VHTR) is one of future reactors for multi-purpose use. The reactor core has large amount of graphite moderator with much heat capacity. So the response of the core outlet gas temperature has very long time constant for disturbances. In this report, the control system design has been studied for the start up, the shut down and the load follow operations using PID controllers.

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Presentations

  • UTR-KINKIにおけるCd試料の反応度価値測定

    #守屋 壮一郎,藤本 望,Irwan Lipto Simanullang, @左近 敦士,

    日本原子力学会 春の大会  2024.3 

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    Event date: 2024.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:近畿大学   Country:Japan  

  • KUCA黒鉛減速体系の拡散計算用エネルギー群構造に関する研究

    小林祐介,藤本望,Irwan L. Simanullang

    日本原子力学会九州支部第42回研究発表講演会  2023.12 

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    Event date: 2023.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • 高温ガス炉の詳細出力評価に関する予備検討

    楠木捷斗,藤本望,Irwan L. Simanullang

    日本原子力学会九州支部第42回研究発表講演会  2023.12 

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    Event date: 2023.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県福岡市   Country:Japan  

  • UTR-KINKIにおけるCdサンプルの反応度価値測定とモンテカルロ法による計算との比較

    守屋壮一郎,藤本望,Irwan L. Simanullang, 左近敦士

    日本原子力学会九州支部第42回研究発表講演会  2023.12 

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    Event date: 2023.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県福岡市   Country:Japan  

  • 高温ガス炉使用済み燃料の燃焼度測定手法の検討

    川口祥平,Irwan L. Simanullang,藤本 望

    日本原子力学会九州支部第42回研究発表講演会  2023.12 

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    Event date: 2023.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県福岡市   Country:Japan  

  • HTTRでの臨界近接解析と核データ

    藤本 望, 後藤 実

    核データ+PHITS合同研究会  2023.11 

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    Event date: 2023.11

    Language:Japanese  

    Venue:茨城県東海村   Country:Japan  

  • HTTRでの臨界近接解析と核データ

    藤本 望, 後藤 実

    核データ+PHITS合同研究会  2023.11 

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    Event date: 2023.11

    Language:Japanese  

    Venue:茨城県東海村   Country:Japan  

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  • Preliminary Study of Burnup Measurement and Relative Power Distribution in HTTR using Gamma-Ray Measurement International conference

    Irwan L. Simanullang, Shohei Kawaguchi, Nozomu Fujimoto, Toshiaki Ishii, Satoru Nagasumi, Hai Quan Ho, Kunihiro Nakajima, Etsuo Ishitsuka Kazuhiko Iigaki

    ICNC 2023 - The 12th International Conference on Nuclear Criticality Safety  2023.10 

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    Event date: 2023.10

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:Sendai, Japan   Country:Japan  

  • Preliminary Study of Burnup Measurement and Relative Power Distribution in HTTR using Gamma-Ray Measurement International conference

    Irwan L. Simanullang, Shohei Kawaguchi, Nozomu Fujimoto, Toshiaki Ishii, Satoru Nagasumi, Hai Quan Ho, Kunihiro Nakajima, Etsuo Ishitsuka Kazuhiko Iigaki

    ICNC 2023 - The 12th International Conference on Nuclear Criticality Safety  2023.10 

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    Event date: 2023.10

    Language:Japanese  

    Venue:Sendai, Japan   Country:Japan  

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  • Preliminary study of the composite moderator concept in small high-temperature gas-cooled reactors

    Irwan Simanullang, Nozomu Fujimoto

    日本原子力学会  2023.9 

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    Event date: 2023.9

    Language:English   Presentation type:Oral presentation (general)  

    Venue:愛知県名古屋市千種区 名古屋大学東山キャンパス   Country:Japan  

  • 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, 藤本 望, Irwan Liapto Simanullang

    日本原子力学会 2023春の年会  2023.3 

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    Event date: 2023.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:東京都   Country:Japan  

  • ⾼温ガス炉の燃料ブロック内における詳細出⼒分布の予備評価

    楠⽊捷⽃, 藤本望

    第10回炉物理専門研究会  2022.12 

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    Event date: 2022.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • ⾼温ガス炉の燃料ブロック内における詳細出⼒分布の予備評価

    楠⽊捷⽃, 藤本望

    第10回炉物理専門研究会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:オンライン   Country:Japan  

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  • 高燃焼度下での高温ガス炉燃料における燃焼挙動評価に関する研究

    園田翔太, Irwan Liapto Simanullang, 深谷裕司, 飯垣和彦, 藤本望

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • HTTR使用済み燃料の燃焼度測定方法の予備検討

    川口 祥平, Irwan Liapto Simanullang, 藤本 望

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県福岡市   Country:Japan  

  • KUCA黒鉛減速体系における中性子吸収体の反応度の評価手法に関する検討

    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第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県福岡市   Country:Japan  

  • 高温ガス炉の燃料ブロック内における詳細出力分布の予備評価

    楠木捷斗, Irwan Liapto Simanullang, 藤本望

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • KUCA黒鉛減速体系の解析用エネルギー群構造に関する研究 Invited

    小林祐介, Irwan Liapto Simanullang, 藤本望

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • KUCA黒鉛減速体系での拡散計算によるCd板反応度価値評価手法の検討

    守屋壮一郎, 山崎誠司, Irwan Liapt Simanullang, 藤本望, 左近淳士, 佐野忠史, 髙橋佳之

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • KUCA黒鉛減速体系の解析用エネルギー群構造に関する研究 Invited

    小林祐介, Irwan Liapto Simanullang, 藤本望

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

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  • HTTR使用済み燃料の燃焼度測定方法の予備検討

    川口 祥平, Irwan Liapto Simanullang, 藤本 望

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

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  • KUCA黒鉛減速体系での拡散計算によるCd板反応度価値評価手法の検討

    守屋壮一郎, 山崎誠司, Irwan Liapt Simanullang, 藤本望, 左近淳士, 佐野忠史, 髙橋佳之

    日本原子力学会九州支部第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

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  • KUCA黒鉛減速体系における中性子吸収体の反応度の評価手法に関する検討

    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第41回講演会  2022.12 

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    Event date: 2022.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

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  • HTTR使用済み燃料の燃焼度・物質量測定手法の予備検討 International conference

    川口 祥平, イルワン シマヌルラン, 藤本 望

    日本原子力学会 2022年秋の大会  2022.9 

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    Event date: 2022.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:茨城県日立市   Country:Japan  

  • KUCA黒鉛減速体系におけるCdサンプル反応度の評価 (2)拡散計算による解析

    守屋 壮一郎, 藤本 望, イルワン シマヌルラン, 山崎 誠司, 高橋 佳之, 左近 敦士, 佐野 忠史

    日本原子力学会 2022年秋の大会  2022.9 

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    Event date: 2022.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:茨城県日立市   Country:Japan  

  • KUCA黒鉛減速体系におけるCdサンプル反応度の評価 (1)モンテカルロ計算による解析

    山崎 誠司, 守屋 壮一郎, Simanullang Irwan, 藤本 望, 左近 敦士, 佐野 忠史, 高橋 佳之

    日本原子力学会 2022年秋の大会  2022.9 

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    Event date: 2022.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:茨城県日立市   Country:Japan  

  • KUCA黒鉛減速体系におけるCdサンプル反応度の評価 (2)拡散計算による解析

    守屋 壮一郎, 藤本 望, イルワン シマヌルラン, 山崎 誠司, 高橋 佳之, 左近 敦士, 佐野 忠史

    日本原子力学会 2022年秋の大会  2022.9 

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    Event date: 2022.9

    Language:Japanese  

    Venue:茨城県日立市   Country:Japan  

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  • HTTR使用済み燃料の燃焼度・物質量測定手法の予備検討 International conference

    川口 祥平, イルワン シマヌルラン, 藤本 望

    日本原子力学会 2022年秋の大会  2022.9 

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    Event date: 2022.9

    Language:Japanese  

    Venue:茨城県日立市   Country:Japan  

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  • KUCA黒鉛減速体系におけるCdサンプル反応度の評価 (1)モンテカルロ計算による解析

    山崎 誠司, 守屋 壮一郎, Simanullang Irwan, 藤本 望, 左近 敦士, 佐野 忠史, 高橋 佳之

    日本原子力学会 2022年秋の大会  2022.9 

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    Event date: 2022.9

    Language:Japanese  

    Venue:茨城県日立市   Country:Japan  

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  • 高温ガス炉用のORIGENライブラリ作成手法の検討

    福原 克樹, 藤本 望, Irwan Liapto Simanullang, 深谷 裕司, Hai Quan Ho, 長住 達, 石井 俊晃, 濱本 真平, 石塚 悦男

    日本原子力学会 2022年春の年会  2022.3 

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    Event date: 2022.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 高温ガス炉燃料の燃焼挙動評価に関する研究

    福原 克樹, 藤本 望, 深谷 裕司, 石塚 悦男, Ho Hai Quan, 石井 俊晃, 長住 達, 濱本 真平

    日本原子力学会九州支部第40回講演会  2021.12 

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    Event date: 2021.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 黒鉛減速体系におけるCd板の拡散計算による反応度価値評価手法の検討

    守屋 壮一郎, 山崎 誠司, 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第40回研究発表講演会  2021.12 

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    Event date: 2021.12

    Language:Japanese  

    Venue:オンライン   Country:Japan  

  • 高温ガス炉使用済燃料の物質量・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第40回研究発表講演会  2021.12 

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    Event date: 2021.12

    Language:Japanese  

    Venue:オンライン   Country:Japan  

  • 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第40回研究発表講演会  2021.12 

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    Event date: 2021.12

    Language:Japanese  

    Venue:オンライン   Country:Japan  

  • 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

    第9回炉物理専門研究会  2021.12 

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    Event date: 2021.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 圧力容器上部プレナム内の単相流混合挙動解析

    藤本 望, 功刀 資彰, 丸山 創, 数土 幸夫

    日本原子力学会  1998.10 

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    Event date: 2021.11

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  • 黒鉛減速体系での炉心規模のスペクトル評価に基づく燃焼挙動の概略評価

    福原 克樹, 藤本 望, 深谷 裕司, Hai Quan Ho, 長住 達, 石井 俊晃, 濱本 真平, 石塚 悦男

    日本原子力学会2021年秋の大会  2021.9 

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    Event date: 2021.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • Calculation of decay gamma spectrum of the HTTR after shutdown

    Hai Quan Ho, Shimpei Hamamoto, Nozomu Fujimoto, Satoru Nagasumi, Minoru Goto, Etsuo Ishitsuka

    日本原子力学会 2021年春の年会  2021.3 

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    Event date: 2021.3

    Language:English   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • MVP-BURNを用いた HTTRの燃焼挙動解析 -軸方向の詳細分割モデルによる解析-

    藤本 望, 池田 礼治, Hai Quan Ho, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会 2021年春の年会  2021.3 

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    Event date: 2021.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 詳細モデルを用いた HTTR臨界試験の再解析

    山本 雄大, Hai Quan Ho, 石塚 悦男, 藤本 望

    日本原子力学会 2021年春の年会  2021.3 

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    Event date: 2021.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • MVP コードによる HTTR臨界試験の再解析

    山本 雄大, 藤本 望 Ho Hai Quan, 石塚 悦男

    日本原子力学会九州支部第39回研究発表講演会  2020.12 

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    Event date: 2020.12

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県福岡市   Country:Japan  

  • 黒鉛減速体系におけるモンテカルロコードを用いた炉内詳細中性子束分布評価

    中川 直樹, 藤本 望, Ho Hai Quan, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会九州支部第39回研究発表講演会  2020.12 

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    Event date: 2020.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • モンテカルロコードを用いた黒鉛減速体系における解析精度評価

    中川 直樹, 藤本 望, Quan Hai Ho, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会2020年秋の大会  2020.9 

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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 高温ガス炉核的予測精度高度化のための研究開発(6) 燃料濃縮度を高めた新臨界体系の検討

    佐野 忠史, 左近 敦士, 高橋 佳之, 深谷 裕司, 藤本 望, 橋本 憲吾

    日本原子力学会2020年秋の大会  2020.9 

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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 高温ガス炉核的予測精度高度化のための研究開発(5)KUCAに構築した黒鉛減速・黒鉛反射炉心における炉雑音解析実験

    左近 敦士, 中嶋 國弘, 佐野 忠史, 深谷 裕司, 藤本 望, 高橋 佳之, 橋本 憲吾

    日本原子力学会2020年秋の大会  2020.9 

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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • Development of a utility tool for auto-seeking critical control rod position of the high temperature engineering test reactor

    Hai Quan Ho, Nozomu Fujimoto, Shimpei Hamamoto, Satoru Nagasumi, Minoru Goto, Etsuo Ishitsuka

    日本原子力学会2020年秋の大会  2020.9 

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    Event date: 2020.9

    Language:English   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • モンテカルロ法によるHTTRの全炉心燃焼計算における出力分布の評価

    池田 礼治, Hai Quan Ho, 藤本 望, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会2020年秋の大会  2020.9 

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    Event date: 2020.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:オンライン   Country:Japan  

  • 黒鉛減速臨界集合体におけるモンテカルロコードの解析精度評価

    中川 直樹, 藤本 望, Ho Hai Quan, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会  2020.3 

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    Event date: 2020.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福島県福島市   Country:Japan  

  • Feasibility study of low power HTGR for long-term operation

    Hai Quan Ho, Nozomu Fujimoto, Shimpei Hamamoto, Satoru Nagasumi, Etsuo Ishitsuka

    Atomic Energy Society of Japan  2020.3 

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    Event date: 2020.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  • モンテカルロ法による HTTRの全炉心燃焼計算における炉内温度分布の影響評価 International conference

    池田 礼治, Ho Quan, 藤本 望, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会  2020.3 

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    Event date: 2020.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福島県福島市   Country:Japan  

  • 炉内温度分布を考慮したHTTR全炉心燃焼計算

    池田 礼治, 藤本 望, Ho Hai Quan, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会九州支部第38回研究発表講演会  2019.12 

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    Event date: 2019.12

    Language:Japanese  

    Country:Japan  

  • モンテカルロコードを用いた黒鉛減速体系における詳細熱中性子束分布評価

    中川 直樹, 藤本 望, Ho Hai Quan, 濱本 真平, 長住 達, 石塚 悦男

    日本原子力学会九州支部第38回研究発表講演会  2019.12 

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    Event date: 2019.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • Calculation of 3D neutron flux distribution in the HTTR using MCNP6

    Hai Quan Ho, Nozomu Fujimoto, Shimpei Hamamoto, Toshiaki Ishii, Satoru Nagasumi, Etsuo Ishituka

    日本原子力学会 2019年秋の大会  2019.9 

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    Event date: 2019.9

    Language:English   Presentation type:Oral presentation (general)  

    Venue:富山県富山市   Country:Japan  

  • MVPを用いたHTTR燃焼計算時の炉心内中性子スペクトルに対する検討

    松中 一朗, 藤本 望, 石井 俊晃, 長住 達, 石塚 悦男

    日本原子力学会 2019年春の年会  2019.3 

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    Event date: 2019.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:茨城大学   Country:Japan  

  • Feasibility study of 99mTc production at HTTR using sublimation method

    Hai Quan Ho, Hiroki Ishida, Shimpei Hamamoto, Toshiaki Ishii, Nozomu Fujimoto, Naoyuki Takaki, Etsuo Ishitsuka

    日本原子力学会 2019年春の年会  2019.3 

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    Event date: 2019.3

    Language:English   Presentation type:Oral presentation (general)  

    Country:Japan  

  • 原子力工学分野での教育と人材育成

    大野哲靖, 尾崎章, 藤本 望, 深田 智

    日本原子力学会  2018.9 

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    Event date: 2018.9

    Language:Japanese  

    Venue:岡山大学   Country:Japan  

  • 九州大学での原子力に関する新たな教育プログラム

    藤本 望, 安田 和弘, 前畑 京介

    日本保全学会  2018.7 

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    Event date: 2018.7

    Language:Japanese   Presentation type:Oral presentation (general)  

    Country:Japan  

  • 高温ガス炉における中性子エネルギー群構造の検討

    道野 雄大, 藤本 望, 本多 友貴

    日本原子力学会九州支部第36回研究発表講演会  2017.12 

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    Event date: 2017.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • 高温ガス炉における動特性パラメータの不確かさによる影響評価

    薄田 真歩, 藤本 望, 本多 友貴

    日本原子力学会九州支部第36回研究発表講演会  2017.12 

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    Event date: 2017.12

    Language:Japanese  

    Country:Japan  

  • 超臨界圧水冷却重水炉の概念の検討

    三津 有也, 藤本 望, Nguyen Duc Ha

    日本原子力学会九州支部第36回研究発表講演会  2017.12 

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    Event date: 2017.12

    Language:Japanese  

    Venue:福岡県福岡市   Country:Japan  

  • 高温ガス炉での燃焼挙動の温度及び濃縮度の感度解析

    守田 圭介, 藤本 望, 深谷 裕司, 本多 友貴

    日本原子力学会 2017年秋の大会  2017.9 

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    Event date: 2017.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:札幌   Country:Japan  

    高温ガス炉燃料の核種生成・崩壊挙動への影響評価の一つとして、高温ガス炉用のORIGENライブラリを作成し、温度及び濃縮度の影響評価を行った。その結果、核種の生成量などに対して濃縮度の影響は小さく、減速材温度の影響が大きいことを確認した。

  • Impact of truncated coated fuel particles on neutronic characteristic of statistical geometry model in MVP code

    Ho Hai Quan, Yuki Honda, Nozomu Fujimoto, Minoru Goto, Etsuo Ishituka

    日本原子力学会 2017年秋の大会  2017.9 

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    Event date: 2017.9

    Language:English   Presentation type:Oral presentation (general)  

    Country:Japan  

    This study investigated the impact of truncated coated fuel particles (CFPs) on neutronic characteristic of the fuel in a statistical geometry (STG) model. Calculation results showed that the truncated CFPs make the multiplication factor decrease by about 0.1 – 1.0 &#37;∆k/k depended on packing fraction, uranium enrichment, and particle size.

  • Burn-up dependency of control rod position at zero power criticality in the High Temperature Engineering Test Reactor International conference

    Yuki Honda, 藤本 望, Hiroaki Sawahata, Kazuhiro Sawa

    23rd International Conference on Nuclear Engineering(ICONE-23)  2015.5 

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    Event date: 2017.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Makuhari Messe Chiba   Country:Japan  

    The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor (HTGR) constructed in Japan, firstly. The operating data of the HTTR with burn-up is very important for developments of HTGRs. Many test data have been collected in the HTTR. Many tests are carried out in low power operation. On the other hand, the full power operation is not enough. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate the calculation code. Additionally, it is difficult to measure core temperature in HTTR. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle and the temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for confirm the characteristics of burn-up and validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.

  • Improvement of cell model for control rod in reflector region of High Temperature Engineering Test Reactor International conference

    Yuki Honda, 藤本 望, Hiroaki Sawahata, Kazuhiro Sawa

    23rd International Conference on Nuclear Engineering(ICONE-23)  2015.5 

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    Event date: 2017.5

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Makuhari Messe Chiba   Country:Japan  

    The High Temperature Engineering Test Reactor (HTTR) is a block type fuel High Temperature Gas-cooled Reactor. There are 32 control rods (16 pairs) in the HTTR. The 6 pairs of control rods are inserted into a core region and the others are inserted in a reflector region surrounding the core. The core temperature of the HTTR is too high to insert all control rods simultaneously at reactor scram near full power operation for keeping integrity of control rods metallic sleeve. Therefore, a two-step control rods insertion method for reactor scram is adopted. The reactivity inserted at the two-step control rod insertion method was measured at HTTR criticality tests. The calculated reactivity at the first- step showed larger underestimation than that of the second- step. On the other hand, calculated results of excess reactivity at the HTTR criticality tests showed good agree with tests. It is considered that a cell model for reflector region control rod is not suitable. Therefore, this paper focuses on a new cell model for control rods in a reflector region. In a previous control rod cell model, control rod is surrounded by fuel blocks only. The surrounding condition of the new cell model corresponds to the configuration around the reflector region control rod. The calculated reactivity at the first-step using the new cell model shows better results than previous calculation. It is considered that the new cell model brings appropriate neutron flux distribution around control rods in reflector region.

  • Benchmark Study on Realized Random Packing Model for Coated Fuel Particles of HTTR using MCNP6 International conference

    Ho Hai Quan, Keisuke Morita, Yuki Honda, Nozomu Fujimoto, Shoji Takada

    2017 International Congress on Advances in Nuclear Power Plants  2017.4 

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    Event date: 2017.4

    Language:English   Presentation type:Oral presentation (general)  

    Venue:Kyoto   Country:Japan  

    The Coated Fuel Particle plays an important role in the excellent safety feature of the High Temperature Gas-cooled Reactor. However, the random distribution of CFPs also makes the simulation of HTGR fuel become more complicated. The Monte Carlo N-particle (MCNP) code is one of the most well-known codes for validation of nuclear systems; unfortunately, it does not provide an appropriate function to model a statistical geometry explicitly. In order to deal with the stochastic media, a utility program for the random model, namely Realized Random Packing (RRP), has been developed particularly for High Temperature engineering Test Reactor (HTTR). This utility program creates a number of random points in an annular geometry. Then, these random points will be used as the center coordinate of CFPs in the MCNP6 input file and therefore the actual random arrangement of CFPs can be simulated explicitly. First, a pin-cell calculation was carried out to validate the RRP by comparing with Statistical Geometry (STG) model of MVP code. After that, the comparison between the RRP model (MCNP) and STG model (MVP) was shown in whole core criticality calculation, not only for the annular core but also for the fully-loaded core. The comparison of numerical results showed that the RRP model and STG model differed insignificantly in the multiplication factor as expected, regardless of the pin-cell or whole core calculations. In addition, the RRP model did not make the calculation time increase a lot in comparison with the conventional regular model (uniform arrangement).

  • 高温ガス炉での核種生成・消滅挙動の中性子スペクトルの影響評価

    守田 圭介, 藤本 望

    日本原子力学会 2017年春の年会  2017.3 

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    Event date: 2017.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:神奈川県平塚市 東海大学 湘南キャンパス   Country:Japan  

    高温ガス炉燃料の核種生成・崩壊挙動へ影響評価一つとして、中性子ペクトルを確認するために、高温ガス炉用ORIGENライブリを作成し解析行った。その結果 、Pu等の生成量に対する中性子スペクトルの影響がことが明らかになった 。

  • HTTRの燃焼特性におけるBPモデルの効果

    藤本 望, 本多 友貴, 福田 航大, 後藤 実, 栃尾 大輔, 高田 昌二

    日本原子力学会 2016年秋の大会  2016.9 

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    Event date: 2016.9

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:福岡県久留米市   Country:Japan  

    HTTRでは、燃焼による過剰反応度の変化を、BPを用いることにより小さく維持している。この燃焼による過剰反応度の変化を低温臨界状態での制御棒位置変化で評価した。解析により、臨界制御棒位置に対するBP周りのメッシュ分割の効果を明らかにした。

  • 高温ガス炉での核種生成・消滅挙動の予備評価

    藤本 望, 守田 圭介, 深谷 裕司

    日本原子力学会 2016年春の年会  2016.3 

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    Event date: 2016.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:宮城県 仙台市 東北大学   Country:Japan  

    高温ガス炉燃料の核種生成・消滅挙動の評価を目指して、高温ガス炉用ORIGENライブラリを作成し解析を行った。軽水炉用ライブラリの結果と比較すると、PuやFP量に差が見られた。

  • 高温ガス炉HTTRの1次ヘリウムガスの混合挙動

    栃尾大輔, 藤本望

    日本原子力学会2015春の年会  2015.3 

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    Event date: 2015.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:茨城県水戸市   Country:Japan  

  • 高温ガス炉の燃焼特性と崩壊熱の評価

    石井 俊晃, 橋本 憲吾, 杉山 亘, 奥田 遼平, 藤本 望, 高田 昌二, 島崎 洋祐, 佐野 忠史

    日本原子力学会2015春の年会  2015.3 

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    Event date: 2015.3

    Language:Japanese   Presentation type:Oral presentation (general)  

    Venue:茨城県水戸市   Country:Japan  

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MISC

  • HTTR の核的パラメータの計算 - 2021 年度夏期休暇実習報告- Reviewed

    五十川 浩希, 直井 基将, 山崎 誠司, Hai Quan HO, 片山 一成, 松浦 秀明, 藤本 望, 石塚 悦男

    2022.7

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    Language:Japanese   Publisher:JAEA Tehnology  

    DOI: 10.11484/jaea-technology-2022-015

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  • 2020年度夏期休暇実習報告;HTTR炉心を用いた原子力電池に関する予備的検討 -核設計のための予備検討(3)- Reviewed

    石塚 悦男, 満井 渡, 山本 雄大, 中川 恭一, Hai Quan Ho, 石井 俊晃, 濱本 真平, 長住 達, 高松 邦吉, Inesh Kenzhina, Yevgeni Chikhray, 松浦 秀明, 藤本 望

    2021.9

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    Language:Japanese  

    DOI: 10.11484/jaea-technology-2021-016

  • MVP-BURNを用いた軸方向詳細モデルによるHTTRの燃焼特性解析 Reviewed

    池田 礼治, Hai Quan Ho, 長住 達, 石井 俊晃, 濱本 真平, 中野 優美, 石塚 悦男, 藤本 望

    2021.9

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    Language:Japanese  

    DOI: 10.11484/jaea-technology-2021-015

  • HTTR燃料セルモデルにおける可燃性毒物周辺のメッシュ効果 Reviewed

    藤本 望, 福田 航大, 本多 友貴, 栃尾 大輔, Ho, Hai. Quan, 長住 達, 石井 俊晃, 濱本 真平, 中野 優美, 石塚 悦男

    2021.6

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    Language:Japanese  

    DOI: 10.11484/jaea-technology-2021-008

  • 2019 年度夏期休暇実習報告;HTTR 炉心を用いた原子力電池に関する予備的検討-核設計のための予備検討 (2)-

    石塚 悦男, 中島 弘貴, 中川 直樹, Hai Quan HO, 石井 俊晃, 濱本 真平, 高松 邦吉, Inesh Kenzhina, Yevgeni Chikhray, 松浦 秀明, 藤本 望

    2020.8

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    Language:Japanese  

    DOI: 10.11484/jaea-technology-2020-008

  • HTTR 炉心解析における制御棒モデルの検討

    長住 達, 松中 一朗, 藤本 望, 石井 俊晃, 石塚 悦男

    2020.5

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    Language:Japanese  

    DOI: 10.11484/jaea-technology-2020-003

  • HTTR の起動用中性子源の交換時期の推定

    小野正人, 小澤 太教, 藤本 望

    2019.9

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    Language:Japanese  

    DOI: 10.11484/jaea-technology-2019-012

  • 2018 年度夏期休暇実習報告;HTTR 炉心を用いた原子力電池に関する予備的検討-核設計のための予備検討-

    石塚 悦男, 松中 一朗, 石田 大樹, Hai Quan HO, 石井 俊晃, 濱本 真平, 高松 邦吉, Inesh Kenzhina, Yevgeni Chikhray, 近藤 篤, 高木 直行, 藤本 望

    2019.7

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    Language:Japanese  

  • Calculation of Decay Heat by ORIGEN Libraries for High Temperature engineering Test Reactor

    Irwan Liapto SIMANULLANG, Yuki Honda, Yuji Fukaya, Minoru Goto, Yoske Shimazaki, 藤本 望, Shoji Takada

    2016.9

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    Language:English  

    これまで高温工学試験研究炉の崩壊熱は、軽水炉のデータを基にしたShure の式やORIGEN計算で評価してきたが、厳密には軽水炉の中性子スペクトルと異なることから最適な評価方法を検討する必要がある。このため、黒鉛減速材量を変えた炉心の中性子スペクトルを用い、ORIGEN2コードで崩壊熱及び生成核種を計算して軽水炉の崩壊熱曲線と比較した。

  • Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2009

    Nozomu Fujimoto, H. Wang

    2010.7

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    Language:English  

    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2009.

  • 高温工学試験研究炉(HTTR)の高温機器・配管における熱変位挙動の評価

    篠原 正憲, 濱本 真平, 藤本 望

    2009.12

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    HTTRの機器配管の支持方式には、3次元の浮動支持方式を採用しているため、その熱変位挙動・特性を把握することは、3次元浮動支持方式の熱変位特性を調べるうえでも重要である。HTTRの定格運転では、1次冷却設備の熱変位特性を知るために、高温配管の熱変位測定試験を実施した。本報は、その試験結果及び解析評価結果を示したものである。熱変位挙動は温度変化に対して1次的に変化することを確認した。解析評価では、支持構造物の抵抗力が機器の熱変位挙動に影響を与えることが明らかとなり、また、その抵抗力を最適化することで実測熱変位挙動を再現できることを確認した。

  • Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2008

    Nozomu Fujimoto, Yukio Tachibana, Y. Sun

    2009.7

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    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2008.

  • HTTR長期連続運転の結果の概要 -定格・並列30日運転-

    栃尾 大輔, 野尻 直喜, 濱本 真平, 猪井 宏幸, 関田 健司, 近藤 雅正, 七種 明雄, 亀山 恭彦, 齋藤 賢司, 藤本 望

    2009.5

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  • HTTR長期連続運転の結果の概要; 定格・並列30日運転

    栃尾 大輔, 野尻 直喜, 濱本 真平, 猪井 宏幸, 関田 健司, 近藤 雅明, 七種 明雄, 亀山 恭彦, 齋藤 賢司, 藤本 望

    2009.5

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    HTTRは1998年の初臨界達成後、定格運転及び高温試験運転の出力上昇試験を経て、現在、供用運転を行っている。今回、HTTRでは長期にわたって熱利用系に安定な熱供給ができることを実証するために、定格・並列運転で30日連続運転を行った。本報はその運転で得られたHTTRの長期連続運転の特性をまとめたものである。

  • HTTRの燃焼を通じた燃料温度の評価; 850℃運転の場合

    栃尾 大輔, 篠原 正憲, 藤本 望

    2009.1

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  • Information exchange on HTGR and nuclear hydrogen technology between JAEA and INET in 2007

    Nozomu Fujimoto, Yukio Tachibana, Nariaki Sakaba, Ryutaro Hino, S. Yu

    2008.6

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    The worldwide interests in the HTGR (High Temperature Gas-cooled Reactor) have been growing because the high temperature heat produced by the reactor can be utilized not only for efficient power generation but also for broad process heat applications, especially for thermo-chemical hydrogen production to fuel a prospective hydrogen economy in future. Presently only two HTGR reactors are operational in the world, including the HTTR (High Temperature Engineering Test Reactor) in Japan Atomic Energy Agency (JAEA) and the HTR-10 in the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. JAEA and INET have cooperated since 1986 in the field of HTGR development, particularly on the HTTR and HTR-10 projects. This report describes the cooperation activities on HTGR and nuclear hydrogen technology between JAEA and INET in 2007.

  • HTTR運転データベース,2; HTTR炉心特性データベース等の具体例

    野尻 直喜, 大和田 博之, 藤本 望

    2007.6

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    「HTTR運転データベース」は将来型高温ガス炉の実用化開発や高温工学試験研究炉(HTTR)の運転管理に資することを目的にHTTRの運転データを蓄積・整理したデータベースである。対象データは基本的にHTTRの運転より得られた過剰反応度,炉心又はプラント内の各部温度,冷却材中不純物濃度等の測定値を整理・評価したデータである。データベースは重要度の高い運転データを長期的及び系統的に管理する目的で、将来型高温ガス炉の実用化開発やHTTRの運転管理に関する目的別のデータベースから成る構造とした。本報ではHTTR運転データベースのうちHTTR共通データベース,HTTR核特性データベース,ヘリウム純度管理データベース,炉内構造物健全性データベース及びその他データベースの具体例を示す。

  • 高温工学試験研究炉(HTTR)の炉心支持黒鉛構造物のサーベイランス試験のための基礎データ

    角田 淳弥, 柴田 大受, 菊地 孝行, 石原 正博, 伊与久 達夫, 沢 和弘, 藤本 望

    2007.2

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    我が国初の高温ガス炉であるHTTRでは炉心支持黒鉛構造物の健全性及び特性等を確認するために、供用期間中検査(ISI)としてTVカメラを用いた炉心支持黒鉛構造物の目視検査及びサーベイランス試験片を用いた物性値の測定を行うこととしている。このうちサーベイランス試験では、HTTRの炉内支持黒鉛構造物について高速中性子照射,酸化等による物性値,強度等の経年変化を調べることにしており、ここで得られるデータは、HTTRの炉心支持黒鉛構造物の健全性確認に用いられるほか、第4世代原子炉システムの1つとして国際的に検討を行っている超高温ガス炉(VHTR)の黒鉛構造物の設計等に活用することのできる貴重なものとなる。本報は今後実施することになるHTTRのサーベイランス試験片の炉内への装荷位置及び使用前の状態における物性データをまとめたものである。

  • 高温工学試験研究炉(HTTR)の核特性評価手法の改良に関する研究(学位論文)

    藤本 望

    2007.1

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    HTTRの核特性評価手法は、設計に用いた解析手法をもとに臨界試験で得られた結果に基づいて開発が行われてきた。開発では、まず反応度調整材(BP)反応度価値の評価手法の改良,制御棒挿入孔からの中性子ストリーミング効果の考慮,燃料セルモデルの改良が行われた。これらの改良により、過剰反応度の測定値に対して1&#37;&#36;&#36;Delta&#36;&#36;k/k以下の誤差で一致するモデルを作成することができた。この改良をもとに、出力運転解析のためのモデルの拡張を行った。ここでは、おもにBPの燃焼変化の挙動について濃縮度依存性,温度効果等を検討した。燃焼挙動についての測定データとの比較では、ほぼ一致する結果を得ることができた。また、モンテカルロコードによる出力分布の比較では、両者はほぼ同じ分布を示すことを明らかにした。また、HTTRの燃料体からの&#36;&#36;gamma&#36;&#36;線測定結果との比較では、改良したモデルは測定結果と良い一致を示すことを明らかにした。さらに、設計時に用いられた核設計手法と比較することにより、今回のモデルの改良が、実効増倍率だけでなく出力分布の改善にも効果があることを明らかした。

  • HTTR運転データベース,1; 全体概要及び作成方針

    野尻 直喜, 栃尾 大輔, 濱本 真平, 梅田 政幸, 藤本 望, 伊与久 達夫, 武田 哲明

    2006.10

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    「HTTR運転データベース」は将来高温ガス炉の実用化開発や高温工学試験研究炉(HTTR)の運転管理に資することを目的にHTTRの運転データを蓄積・整理したデータベースである。対象データは基本的にHTTRの運転より得られた過剰反応度,炉心又はプラント内の各部温度,冷却材中不純物濃度等の測定値を整理・評価したデータである。データベースは重要度の高い運転データを長期的及び系統的に管理する目的で、将来高温ガス炉の実用化開発やHTTRの運転管理に関する目的別のデータベースから成る構造とした。本報ではHTTR運転データベースの全体概要及び作成方針について報告する。また、本データベースのうちHTTR共通データベース,HTTR核特性データベース及びヘリウム純度管理データベースの一部を例として示す。

  • HTTRの補機冷却水設備冷却塔の伝熱性能に関する評価

    栃尾 大輔, 亀山 恭彦, 清水 厚志, 猪井 宏幸, 山崎 和則, 清水 康彦, 新垣 悦史, 太田 幸丸, 藤本 望

    2006.9

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  • HTTRの反応度調整材の燃焼挙動と炉心特性の評価

    藤本 望, 野尻 直喜

    2006.1

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  • 炉容器冷却設備冷却器の伝熱性能の変化とその回復作業について

    濱本 真平, 渡辺 周二, 小山 直, 太田 幸丸, 栃尾 大輔, 藤本 望

    2005.7

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    HTTRの炉容器冷却設備(Vessel cooling system: VCS)は工学的安全施設の一つであり、配管破断や減圧事故時のような強制循環による炉心の冷却ができない場合に、炉心の熱を圧力容器の周りに設置した水冷管パネルで輻射及び自然対流によって除去する設備である。また、通常運転時には1次遮蔽体のコンクリート温度を制限値以下に保つ機能を有している。これまでの運転で、VCSによる除熱量の変化はないものの、VCSを流れる冷却水の温度レベルが徐々に上がりはじめた。さらに冷却水の温度上昇に伴い、遮蔽体であるコンクリート温度の上昇も懸念された。このまま運転を続けると、コンクリート温度が制限値に近づくことも予想されたため、VCS冷却水の温度レベルを低下させる対策を行うこととした。VCS冷却水の温度を管理している機器の1つにVCS冷却器がある。このVCS冷却器の伝熱性能を評価しその低下の程度を明らかにした。その結果、伝熱性能の低下が認められたため、伝熱管の洗浄を行い伝熱性能の回復を図った。これにより、VCS除熱量を変化させることなく、VCS冷却器の伝熱性能を大きく改善し、VCS冷却水の温度を低下させることができた。

  • 高温ガス冷却炉・格子燃焼特性解析コード「DELIGHT-8」(共同研究)

    野尻 直喜, 藤本 望, 毛利 智聡, 小幡 宏幸

    2004.10

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    「DELIGHT」は、炉心計算等に必要な群定数を作成する高温ガス冷却炉用格子燃焼特性解析コードである。円環状または球状の高温ガス炉燃料を対象に衝突確率法による格子計算を行う。高温ガス炉燃料特有の被覆燃料粒子による燃料格子の二重非均質性を考慮した燃焼計算が可能なことが特徴として挙げられる。今回、従来のDELIGHTコードをより燃焼度の高い炉心の解析に対応させることを目的に、核データライブラリのJENDL-3.3への更新,燃焼チェーンを詳細化する等の改良を行った。また、可燃性毒物(BP)格子計算モデルにおいて、BP棒周辺の物質領域を多領域化し、BP格子計算の計算精度の向上を図った。本報は、改良DELIGHTコード(DELIGHT-8)の改良点と使用方法について説明するものである。

  • HTTR制御棒引抜き試験の動特性解析(受託研究)

    高田 英治, 中川 繁昭, 高松 邦吉, 島川 聡司, 野尻 直喜, 藤本 望

    2004.6

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    高温ガス炉の固有の安全性を定量的に実証するため、高温工学試験研究炉(High Temperature Engineering Test Reactor: HTTR)において、反応度投入及び炉心除熱量減少を試験として実機の原子炉で生じさせる安全性実証試験を実施している。安全性実証試験の1つである制御棒引抜き試験について、1点炉近似モデルにより試験時の動特性解析を実施した。実測値と解析値の比較から、1点炉近似モデルが試験の結果を再現できることを確認した。また、添加反応度,温度係数,物性値等の各パラメータについて、制御棒引抜き事象に対する原子炉動特性への感度を明らかにした。

  • HTTR出力密度分布評価における拡散計算モデルの検討 Reviewed

    高松 邦吉, 島川 聡司, 野尻 直喜, 藤本 望

    2003.10

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    HTTR炉心の燃料最高温度の評価においては、炉心出力密度分布の予測精度向上が重要であり、炉心管理コードとしても用いられる拡散燃焼計算モデルの改良を図る必要がある。拡散計算によるHTTR炉心の出力密度分布解析について、可燃性反応度調整材(BP)を燃料体内に均質に分布させたモデル(BP混合モデル)とBP領域を分離したモデル(BP分離モデル)の解析結果を、グロス&#36;&#36;gamma&#36;&#36;線による出力密度分布測定結果及び連続エネルギーモンテカルロ計算コードMVPの計算値と定量的に比較した。その結果、BP混合モデルでは、炉心の軸方向出力密度分布に対する予測精度が不十分であること、BP分離モデルを用いることにより、予測精度が大幅に改善されることがわかった。

  • HTTR高温試験運転の出力上昇試験計画

    坂場 成昭, 中川 繁昭, 高田 英治, 野尻 直喜, 島川 聡司, 植田 祥平, 沢 和弘, 藤本 望, 中澤 利雄, 足利谷 好信, 川崎 幸三, 伊与久 達夫

    2003.3

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    HTTRは、原子炉出口冷却材温度950&#36;&#36;^{circ}C&#36;&#36;の達成を目指した高温試験運転による出力上昇試験を平成15年度に計画している。高温試験運転の実施にあたっては、被覆粒子燃料を使用し、ヘリウムガス冷却を行う我が国初の高温ガス炉であることを念頭に、これまで実施してきた出力上昇試験(定格運転30MW及び原子炉出口冷却材温度850&#36;&#36;^{circ}C&#36;&#36;までの試験)での知見をもとに計画する。高温試験運転においては、温度の上昇に従ってより厳しくなる、原子炉の核熱設計,放射線遮へい設計及びプラント設計が適切であることを確認しながら実施する。本報では、HTTRの安全性確保に重要な燃料,制御棒及び中間熱交換器について、定格運転モードでの運転データに基づき、高温試験運転時の安全性の再確認を行った結果を示すとともに、これまでに摘出された課題とその対策を示した。加えて、高温試験運転における試験項目摘出の考え方を示し、実施する試験項目を具体化した。その結果、原子炉施設の安全を確保しつつ、原子炉熱出力30MW,原子炉出口冷却材温度950&#36;&#36;^{circ}C&#36;&#36;の達成の見通しを得た。

  • HTTR原子炉スクラム時の制御棒温度解析; 商用電源喪失試験の実測データに基づく評価

    高田 英治, 藤本 望, 松田 淳子, 中川 繁昭

    2003.3

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    HTTRは1次冷却材の温度が最高約950&#36;&#36;^{circ}C&#36;&#36;に達するため、制御棒の金属材料には特殊合金としてAlloy800Hが使用されている。この制御棒の使用に関しては、Alloy800Hの強度データにより、制限温度を900&#36;&#36;^{circ}C&#36;&#36;以下と定め、これを超えるような環境で使用された場合には必要に応じて制御棒を交換することになっている。制御棒温度が900&#36;&#36;^{circ}C&#36;&#36;を超える可能性のある事象として高温試験運転からの商用電源喪失に伴う原子炉スクラムが挙げられる。本書では、出力上昇試験で得られた商用電源喪失試験時の実測データを用いて制御棒温度解析を実施した結果を示す。解析の結果、繰り返し使用を考慮した事象の中で制御棒温度が最も高くなる商用電源喪失が発生したとしても、制御棒温度は制限値を上回ることはなく、健全性が確保されることを確認した。

  • Data on test results of vessel cooling system of High Temperature Engineering Test Reactor

    Akio Saikusa, Shigeaki Nakagawa, Nozomu Fujimoto, Yukio Tachibana, Tatsuo Iyoku

    2003.2

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    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28,1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is a first Reactor Cavity Cooling System applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it is confirmed that the VCS heat removal at 30 MW power operation is higher than 0.3MW. This paper shows outline of the VCS and test results on the VCS performance.

  • 高温工学試験研究炉の出力上昇試験; 試験経過及び結果の概要

    中川 繁昭, 藤本 望, 島川 聡司, 野尻 直喜, 竹田 武司, 七種 明雄, 植田 祥平, 小嶋 崇夫, 高田 英治, 齋藤 賢司, 橘 幸男, 足利谷 好信, 川崎 幸三, 中澤 利雄, 沢 和弘, 伊与久 達夫

    2002.8

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    高温工学試験研究炉の出力上昇試験; 試験経過及び結果の概要

    中川 繁昭; 藤本 望; 島川 聡司; 野尻 直喜; 竹田 武司; 七種 明雄; 植田 祥平; 小嶋 崇夫; 高田 英治*; 齋藤 賢司; et al.

    JAERI-Tech 2002-069, 87 Pages, 2002/08

    JAERI-Tech-2002-069.pdf:10.12MB

    高温工学試験研究炉(High Temperature engineering Test Reactor : HTTR)の出力上昇試験は、30MW運転時に原子炉出口冷却材温度が850&#36;&#36;^{circ}C&#36;&#36;となる「定格運転」モードでの試験として、平成12年4月23日から原子炉出力10MWまでの出力上昇試験(1)を行い、その後、原子炉出力20MWまでの出力上昇試験(2),30MW運転時に原子炉出口冷却材温度が950&#36;&#36;^{circ}C&#36;&#36;となる「高温試験運転」モードにおいて原子炉出力20MWまでの出力上昇試験(3)を行った。定格出力30MW運転達成のための試験として平成13年10月23日から出力上昇試験(4)を開始し、平成13年12月7日に定格出力30MWの到達及び原子炉出口冷却材温度850&#36;&#36;^{circ}C&#36;&#36;の達成を確認した。出力上昇試験(4)については、平成14年3月6日まで実施し、定格出力30MWからの商用電源喪失試験をもって全ての試験検査を終了して使用前検査合格証を取得した。「定格運転」モードにおける原子炉出力30MWまでの試験結果から、原子炉、冷却系統施設等の性能を確認することができ、原子炉を安定に運転できることを確認した。また、試験で明らかとなった課題を適切に処置することで、原子炉出力30MW,原子炉出口冷却材温度950&#36;&#36;^{circ}C&#36;&#36;の達成の見通しを得た。

  • Data on loss of off-site electric power simulation tests of the High Temperature Engineering Test Reactor

    Takeshi TAKEDA, Shigeaki NAKAGAWA, Nozomu FUJIMOTO, Tatsuo IYOKU

    2002.7

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  • HTTRでの出力分布測定時の線量当量率測定及び放射線モニタリング結果

    高田 英治, 藤本 望, 野尻 直喜, 梅田 政幸, 石仙 繁, 足利谷 好信

    2002.5

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  • 出力上昇試験におけるHTTR炉心支持板温度上昇の原因と対策

    藤本 望, 高田 英治, 中川 繁昭, 橘 幸男, 川崎 幸三, 七種 明雄, 小嶋 崇夫, 伊与久 達夫

    2002.1

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  • HTTR出力上昇試験での臨界制御棒位置と温度係数; 中間報告

    藤本 望, 野尻 直喜, 高田 英治, 齋藤 賢司, 小林 正一, 澤畑 洋明, 石仙 繁

    2001.3

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    現在HTTRでは出力上昇試験を進めており、これまで50&#37;出力を達成している。HTTRの出口温度は950&#36;&#36;^{circ}C&#36;&#36;と高いため、出力上昇の過程で炉心内の温度変化が大きい。このような炉心の解析精度の向上を目的として各出力での臨界制御棒位置及び温度係数について測定を行い、解析との比較を行った。解析は、熱流動解析コードと拡散計算のくり返しにより求めた炉内温度分布を用いて、モンテカルロ計算と拡散計算により行った。その結果、臨界制御帽位置はモンテカルロ計算により50mm以下の誤差で一致し、100&#37;出力では2900mm程度になると予想された。温度係数は拡散計算の結果とよく一致した。今後、出力100&#37;までの測定を行い、解析結果と比較することにより解析精度の向上を目指す。

  • 高温工学試験研究炉の燃料体からの&#36;&#36;gamma&#36;&#36;線測定; 方法と結果

    藤本 望, 野尻 直喜, 高田 英治, 山下 清信, 菊地 孝行, 中川 繁昭, 小嶋 崇夫, 梅田 政幸, 星野 修, 金田 誠, 小林 正一, 石仙 繁, 川崎 幸三, 國富 一彦

    2001.2

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    HTTRの炉心内の情報を得ることを目的として、炉心から燃料体を取り出し再装荷する過程での燃料体からの&#36;&#36;gamma&#36;&#36;線の測定を行った。測定は、燃料体が通過する床上ドアバルブに設置したGM管及びCZT半導体検出器と、スタンドパイプ室に設置したエリアモニタで行い、炉内のウラン濃縮度配分の対称性を考慮して4カラムの燃料体計20体について行った。測定の結果GM管及びCZT検出器による測定では、各カラムでの軸方向の相対分布は解析とほぼ一致したが、炉心上部では解析値が高く、炉心下部では低くなった。エリアモニタによる測定でも軸方向の分布を測定することができた。さらにカラム間の比較も行った。今後は測定結果について詳細な解析・評価を行い、炉内出力密度分布等の評価精度の向上に役立てる予定である。

  • 高温工学試験研究炉炉心解析モデルの改良;過剰反応度に関する検討

    藤本 望, 山下 清信

    1999.11

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    これまで、HTTRの炉心解析モデルについて、VHTRCの実験結果を用いて検証が行われてきた。またモンテカルロコードとの比較に基づき、ゼブラ型反応度調整材の形状及び位置の効果、中性子ストリーミング効果を考慮できるようモデルの改良が進められてきた。さらにこの改良モデルを用いて臨界試験の予備解析が行われてきた。しかしながら臨界試験の結果から、予備解析に用いたモデルでも過剰反応度を過大に評価することが明らかとなった。検討の結果、燃料セルの外径が過大で実際より減速材の黒鉛が多いため柔らかい中性子スペクトルとなり、&#36;&#36;^{235}&#36;&#36;Uの核分裂断面積を大きく評価していることが原因であると考えられた。そこで、燃料セルの外径をこれまでより小さい、燃料棒のピッチによる値とすることにより、臨界試験結果とよく一致する結果を得ることができた。

  • 燃料体内の反応度調整材位置を考慮した反応度価値評価手法のVHTRC実験データによる検討

    藤本 望, 山下 清信, 秋濃 藤義

    1999.9

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    VHTRC炉心に複数本の反応度調整材(BP)棒を装荷した実験結果について、BP反応度価値の解析精度を評価した。その結果、HTTRの核設計に用いている、ブロック内を均質としたモデルではBP反応度を20&#37;程度過小評価することが明らかとなった。この結果は、BP反応度を系統的に過小評価しているため、過剰反応度を高めに評価するという観点からは保守的であり安全上問題ない。しかしながら、高温ガス炉の設計の合理化、将来炉の設計、HTTRの運転管理等のためには評価精度の向上が必要である。そこで、BP位置をモデル化し炉心内のインポータンス分布をより詳細に考慮すれば精度が向上すると考え、燃料体内でのBP棒の位置を考慮できるよう燃料体のメッシュ分割数を増やしたモデルを作成した。このモデルとともに、炉心計算のBP棒領域に対応する範囲で均質化することにより作成した実効断面積を用いることにより、10&#37;以下の誤差でBP反応度を評価できることが明らかとなった。

  • Analysis of the HTTR's benchmark problems and comparison between the HTTR and the FZJ code systems

    Nozomu FUJIMOTO, Ursula OHLIG, Hans BROCKMANN, Kiyonobu YAMASHITA

    1998.11

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    The first Research Coordination Meeting for the Coordinated Research Program on the HTTR benchmark problems were held in August 1998. The results and calculation models of JAERI and Forshcungszentrum Jiilich GmbH (FZJ) by diffusion calculation were compared. Both results showed a good agreement at fully-loaded core but the results of JAERI showed about 1 &#37;Ak higher value during fuel loading state. To investigate the cause of the difference, effects of energy group number, neutron streaming from control rod insertion holes and cell models of burnable poison (BP) were studied. As the results, we found that the difference caused by energy group number and neutron streaming were small. The effect of BP cell model was evaluated by sensitivity analysis of dimension of BP cell. Improvements for each calculation model were proposed.

  • 高温工学試験研究炉(HTTR)臨界試験の予備解析結果; モンテカルロコードMVPに基づく解析

    野尻 直喜, 中野 正明, 安藤 弘栄, 藤本 望, 竹内 光男, 藤崎 伸吾, 山下 清信

    1998.8

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    高温工学試験研究炉(HTTR)の臨界試験の事前評価として、連続エネルギー法に基づくモンテカルロ計算コードMVPにより核特性解析を行った。拡散理論による炉心計算では直接モデル化が困難であった、燃料コンパクト、燃料棒、燃料棒挿入孔、反応度調整材等の燃料体内の非均質構造、制御棒及び制御棒挿入孔、後備停止系ほう素ペレット落下孔、炉心構成要素間の間隙等を詳細にモデル化した。解析により、初回臨界は16カラム前後燃料を装荷した状態で到達する見込みであること、その際第1,2,3リング制御棒を全引き抜きし中心制御棒だけを操作することで臨界調節が可能であることを確認した。また、臨界時の制御棒位置、過剰反応度、炉停止余裕等を求めた。これらの解析結果を臨界試験の計画策定に用いた。

  • 高温工学試験研究炉(HTTR)臨界試験の予備解析結果; HTTR核特性解析コードシステムに基づく解析

    藤本 望, 野尻 直喜, 中野 正明, 竹内 光男, 藤崎 伸吾, 山下 清信

    1998.6

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    本報は、HTTR核特性解析コードシステムの炉心解析モデルの改良と、このモデルを用いて行った臨界試験の予備解析結果について報告するものである。解析モデルは、BPの軸方向装荷パターンがゼブラ状であること並びに燃料体内での径方向位置をモデル化できるよう及び制御棒挿入孔等からのストリーミングを考慮できるよう改良した。予備解析では、燃料装荷に伴う実効増倍率の変化、中性子検出器の応答確認、逆増倍係数、制御棒反応度価値、炉停止余裕、動特性パラメータ、中性子束分布及び出力換算係数に関する解析を行った。本報に示した結果は、既に試験計画及び使用前検査に用いている。今後は、この結果と臨界試験結果を比較し、モデル及び試験結果の妥当性の確認を行う計画である。

  • 高温工学試験研究炉(HTTR)の過剰反応度測定での制御棒干渉効果の解析評価

    中野 正明, 山下 清信, 藤本 望, 野尻 直喜, 竹内 光男, 藤崎 伸吾, 徳原 一実, 中田 哲夫

    1998.5

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    高温工学試験研究炉(HTTR)の過剰反応度を燃料追加法によって測定する場合について、制御棒の干渉効果が過剰反応度に与える影響を評価した。制御棒が全引き抜き状態の実効増倍率から求める過剰反応度に比べて、制御棒操作を考慮することによって、-10&#37;~+50&#37;程度の測定値が変化することがわかった。また、干渉効果の影響を小さくするためには、被測定制御棒、補償制御棒とも複数の制御棒を用いればよく、(1)被測定制御棒として第3リング制御棒を除く13対を用い、そのうちの1対の反応度測定の際にその他の12対を補償制御棒として用いる組合わせ、(2)第1リング制御棒6対を(1)と同様に用いる組み合わせ、が過剰反応度測定に適していることが明らかになった。

  • Study on temperature coefficients of actinide burning HTGRs

    Nozomu Fujimoto, Kiyonobu Yamashita, H. J. Rütten

    1997.11

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  • VHTRC炉物理実験の解析によるモンテカルロコードMVPの精度評価; 臨界時の実効増倍率、反応度調整材反応度、ボイド反応度

    野尻 直喜, 山下 清信, 藤本 望, 中野 正明, 山根 剛, 秋濃 藤義

    1997.11

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    本報は、高温ガス炉臨界実験装置(VHTRC)の臨界時の実効増倍率、反応度調整材反応度、ボイド反応度の実験結果を汎用中性子・光子輸送計算モンテカルロコード(MVP)により評価し、MVPを高温ガス炉の核特性評価使用する場合の解析精度の評価を行ったものである。解析の結果、臨界時の実効増倍率、反応度調整材反応度、ボイド反応度の解析誤差は最大で、それぞれ0.8&#37;&#36;&#36;Delta&#36;&#36;k,7&#37;,25&#37;以下であった。臨界時の実効増倍率を十分な精度で予測できることを明らかにした。よって、HTTRの炉心特性評価にMVPを適用することが可能であることがわかった。

  • 高温工学試験研究炉(HTTR)の高性能炉心概念の設計

    山下 清信, 中野 正明, 野尻 直喜, 藤本 望, 沢 和弘, 中田 哲夫, 渡部 隆

    1997.8

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    高温工学試験研究炉(HTTR)の照射性能の向上及び炉心性能を実規模高温ガス炉のもと同等にすることを目的とし、核熱設計の観点から炉心の高性能化の検討を行った。本検討より、ダルマ落とし燃料交換方式を採用することにより、90GWd/tという高い燃料燃焼度を達成できる見通しを得た。また、一体型燃料コンパクトを用いることにより燃料の徐熱性能を向上でき、炉心平均出力密度を7.1W/cm&#36;&#36;^{3}&#36;&#36;まで大きくしても、原子炉出口冷却材温度950&#36;&#36;^{circ}&#36;&#36;Cを達成できる見通しを得た。高速中性子束及び熱中性子束の最大値は、HTTRの約2.2倍及び約1.5倍まで増大できる見込みを得た。炉心有効流量の低下を防止するため、炭素複合材を被覆管に用いた制御棒の使用及び上部遮蔽体ブロックの側面にはめあい構造を設けカラム間の漏れ流れを低減することが、有効であることが分かった。

  • &#36;&#36;^{239}&#36;&#36;Pu高濃度のプルトニウムを装荷したペブルベッド型高温ガス炉の温度係数の検討

    徳原 一実, 山下 清信, 新藤 隆一, 藤本 望

    1996.6

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    ペブルベッド型高温ガス炉に兵器級Puを装荷した場合、温度係数が正になるという問題があった。そこで、温度係数を負にする方法を、兵器級Puを劣化U等の親物質と混合せずに、全炉心に兵器級Puの燃料のみを装荷する炉心を対象として検討した。検討の結果、熱領域に共鳴捕獲反応を持つ核種(Er)を炉心に添加すれば、温度係数が負になることを確認した。また、この場合には燃料球の燃焼度が低下するが、Erを添加せずとも、Puの装荷量を増大して熱中性子束のピークを小さくすれば、燃焼度の低下なしに温度係数を負にできることを明らかにした。燃焼度の増大とともに熱中性子束のピークが再び大きくなる可能性があるが、ペブルベッド型炉では燃料球の最大燃焼度の1/2程度で原子炉が平衡状態になるため、初期の装荷量が多い限り熱中性子束のピークが再び大きくなることはない。

  • 高温工学試験研究炉における原子炉スクラム時の炉停止余裕の評価

    村田 勲, 山下 清信, 丸山 創, 藤本 望, 新藤 隆一, 数土 幸夫

    1991.10

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    HTTRは原子炉出口温度が950&#36;&#36;^{circ}&#36;&#36;Cと高く、このためスクラムに伴う原子炉停止においては、制御棒の高温における繰り返し使用による寿命の低下を避けるため、まず反射体領域の制御棒を挿入して原子炉を未臨界にし、ついで炉心温度が所定の温度(原子炉出口温度が750&#36;&#36;^{circ}&#36;&#36;C)に下がるのを待って、あるいは所定の時間(2400秒)をおいて燃料領域へ制御棒を挿入して常温で未臨界を維持する制御棒2段階挿入方式を採用している。本報告では、2段階挿入方式を用いたスクラム時において燃料領域の制御棒が挿入されるまでの間、原子炉を未臨界に維持できることの確認を行った。この結果、もっとも厳しい条件となる原子炉出口温度950&#36;&#36;^{circ}&#36;&#36;Cからのスクラム時でも0.7&#37;&#36;&#36;Delta&#36;&#36;k/k(制御棒1対のスタックを考慮した場合)の炉停止余裕を確保できることがわかった。

  • 高温工学試験研究炉(HTTR)の冷却材の流れ; 炉心支持板下面における冷却材の流動特性の解析

    稲垣 嘉之, 藤本 望, 元木 保男, 伊与久 達夫, 丸山 創, 塩沢 周策

    1990.12

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    高温工学試験研究炉(HTTR)の炉心支持板は、炉心及び炉心支持黒鉛構造物の鉛直方向の荷重を直接支持する機能を有し、その上部にある炉床部断熱層により、炉心内の高温冷却材(約950&#36;&#36;^{circ}&#36;&#36;C)からの熱伝導を低減するとともに、その下面を低温冷却材(約400&#36;&#36;^{circ}&#36;&#36;C)で冷却して、制限温度を超えない構造としている。炉心支持板下面の冷却材流路には、1次ヘリウム配管、補助ヘリウム配管及び多数の支持板支持柱等の構造物がある。これらの構造物は、冷却材を偏流させる可能性があり、その結果として炉心支持板にホットスポットが生じる可能性がある。炉心支持板下面の冷却材の流動を明らかにするために、3次元熱流体解析コードSTREAMを用いて解析を行なった。更に、その解析結果から得られた流速分布より、炉心支持板の温度分布を解析した結果、ホットスポットが発生するような偏流が生じないことを確認した。

  • 高温工学試験研究炉(HTTR)の制御棒温度解析

    丸山 創, 西口 磯春, 藤本 望, 小倉 健志, 塩沢 周策, 数土 幸夫

    1990.7

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    高温工学試験研究炉では、原子炉スクラム時に制御棒被覆管の高温における繰り返し使用による寿命の低下を避けるため、まず反射体領域の制御棒を挿入し、その後炉心温度が所定の温度以下となった時点で燃料領域の制御棒を挿入し、低温まで未臨界を維持する2段階挿入法を採用している。炉心領域制御棒の挿入は、タイマーによる設定時間または原子炉出口冷却材温度の設定値に達した時点で行う。本報は、種々のスクラム条件下での制御棒被覆管温度解析の手法、条件及び結果についてまとめたものである。

  • 高温工学試験研究炉用制御動特性解析コードASURAの検証解析

    藤本 望, 中川 繁昭, 露崎 典平, 丹治 幹雄, 島川 佳郎, 数土 幸夫

    1989.11

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    本報は、高温工学試験研究炉(HTTR)の動特性解析コードASURAについて、その概要と検証解析についてまとめたものである。ASURAコードはHTTRプラントシステムの制御動特性解析を目的としているため、各種制御系を含めたプラントシステム全体をモデル化している。検証条件は、各種パラメーターサーベイ及びBLOOST-J2コード、THYDE-HTGRコードとのクロスチェック解析を行った。その結果、ASURAコードの特徴及び妥当性が確認された。また、Fort St.Vrain炉での実験データによる3検証も行い、妥当性の確認を行った。

  • 高温工学試験研究炉の炉心入口冷却材温度の評価

    藤本 望, 丸山 創, 数土 幸夫

    1989.5

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    本報は、高温工学試験研究炉(HTTR)について、原子炉圧力容器入口から、炉心入口部までの冷却材の熱流動解析についてまとめたものである。HTTRでは、原子炉圧力容器に流入した冷却材は、炉心と原子炉圧力容器の間を上方へ流れて上部プレナムへ至り、上部プレナム内で反転して下降流となり炉心へ流入する。本報では、冷却材が炉心と原子炉圧力容器の間を上昇する際の冷却材温度上昇及び温度上昇誤差の評価、上部プレナム内における冷却材の3次元熱流動解析による冷却材温度混合の評価についてまとめたものである。また、炉心入口温度の燃料最高温度評価に及ぼす影響についても検討を加えた。

  • 高温工学試験研究炉炉心燃料最高温度計算用工学的安全係数の評価 Reviewed

    丸山 創, 山下 清信, 藤本 望, 村田 勲, 新藤 隆一, 数土 幸夫

    1988.12

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    Language:Japanese  

    本報は、高温工学試験研究炉(HTTR)の炉心熱流力設計の主要項目である燃料最高温度評価の用いる工学的安全係数についてまとめたものである。工学的安全係数は、系統的性質を有するシステマティック因子と統計的性質を有するランダム因子とに分け、それぞれHTTRの特徴を踏まえて、考慮すべき主項目と値を決定した。

  • 高温工学試験研究炉炉心熱流力設計

    丸山 創, 藤本 望, 山下 清信, 村田 勲, 新藤 隆一, 数土 幸夫

    1988.12

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    本報は、現在日本原子力研究所が設計を進めている高温工学試験研究炉(熱出力30MW、原子炉出口冷却材温度950&#36;&#36;^{circ}&#36;&#36;C)の炉心熱流力設計の概要と設計結果についてまとめたものである。炉心熱流力設計では、通常運転時及び運転時の異常な過渡変化時において、それぞれ燃料最高温度の制限値を定め、これを超えないように設計することとしている。このため、炉心及び燃料の構造上の特徴を踏まえ、十分な炉心冷却材流量を確保するとともに燃料最高温度を極力低くするように設計する。本設計において得られた燃料最高温度は、基準炉心の燃焼330日における原子炉出口冷却材温度950&#36;&#36;^{circ}&#36;&#36;C運転時の1495&#36;&#36;^{circ}&#36;&#36;Cである。

  • 高温工学試験研究炉の燃料温度評価に影響を及ぼす要因とその評価

    藤本 望, 丸山 創, 藤井 貞夫, 仁熊 義則, 数土 幸夫

    1988.10

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    Language:Japanese  

  • 高温工学試験研究炉の炉心内流量配分計画と評価

    丸山 創, 藤本 望, 木曽 芳広, 村上 知行, 多喜川 昇, 早川 均, 数土 幸夫

    1988.9

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    Language:Japanese  

    本報は、高温工学試験研究炉(HTTR)の炉心熱流力設計の基礎となる炉心内冷却材流量配分計画と評価の結果を、解析用データとともにまとめたものである。HTTRの炉心は、黒鉛ブロックを積重ねた積層構造となっており燃料体ブロック及び制御棒案内ブロック内の計画された流路以外に冷却材の流れる流路が構成される、そのため、炉心の有効な冷却の確保のために、このような計画外の流量で極力低減し、冷却材出口温度950&#36;&#36;^{circ}&#36;&#36;C達成のため適切な流量配分を定めている。

  • 熱流動・熱伝導連成解析コードFLOWNET/TRUMPの検証

    丸山 創, 藤本 望, 木曽 芳広, 村上 知行, 数土 幸夫

    1988.9

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    Language:Japanese  

    本報は、高温工学試験研究炉(HTTR)の設計において、炉心の伝熱流動、特に燃料ブロック内の冷却材流路間の流量配分、燃料ブロック応力解析用熱的境界条件の決定並びに燃料ブロック内の冷却材流路閉塞事故時の温度評価に使用する熱流動・熱伝導連成コードFLOWNET/TRUMPの検証結果について報告するものである。

  • 燃料温度解析コードTEMDIMの検証

    丸山 創, 藤本 望, 藤井 貞夫, 渡部 隆, 数土 幸夫

    1988.9

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    Language:Japanese  

    本報は、高温工学試験研究炉の炉心熱流力設計において、燃料温度解析に使用する計算コードTEMDIMの検証結果についてまとめたものである。検証解析は、HENDEL燃料体スタック実証試験部1チャンネル試験装置による伝熱流動試験結果を用いて行い、燃料温度評価手法の妥当性、保守性が確認された。

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Professional Memberships

  • The Institution of Professional Engineer, Japan

  • Atomic Energy Society of Japan

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  • The Japan Society of Mechanical Engineers

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Committee Memberships

  • 九州大学   エネルギー科学科学科長  

    2023.4 - 2024.3   

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    Committee type:Other

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  • 日本原子力学会   教科書調査WGグループ員  

    2022.10 - 2025.3   

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  • 九州大学   核燃料物質取扱施設運営委員会委員  

    2022.4 - 2024.3   

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    Committee type:Other

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  • 九州大学   エネルギー科学科 副学科長  

    2022.4 - 2023.3   

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    Committee type:Other

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  • 日本原子力学会   オープンスクール小委員会委員   Domestic

    2021.4 - 2026.3   

  • 日本原子力学会   代議員   Domestic

    2021.4 - 2023.3   

  • 日本原子力学会   代議員  

    2021.4 - 2023.3   

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  • 九州大学   九州大学放射線等障害防止委員会委員  

    2021.4 - 2023.3   

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    Committee type:Other

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  • 九州大学   九州大学核燃料物質管理専門部会委員  

    2021.4 - 2023.3   

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    Committee type:Other

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  • 日本原子力学会   九州支部長   Domestic

    2020.5 - 2021.5   

  • 九州大学   工学府学務委員  

    2020.4 - 2022.3   

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    Committee type:Other

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  • 副支部長   副支部長   Domestic

    2019.5 - 2020.5   

  • 九州大学   放射線等障害防止委員会 核燃料物質管理専門部会委員  

    2016.4   

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    Committee type:Other

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  • 九州大学   工学部核燃料計量管理責任者  

    2016.4   

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    Committee type:Other

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  • 日本原子力学会九州支部   Organizer   Domestic

    2015.4 - 2017.3   

  • 関東支部茨城ブロック 商議員   関東支部茨城ブロック 商議員   Domestic

    2012.4 - 2014.3   

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Academic Activities

  • Screening of academic papers

    Role(s): Peer review

    2024

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:15

  • その他 International contribution

    The 21th International Conference onNuclear Criticality Safety ICNC2023  ( Japan ) 2023.10

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    Type:Competition, symposium, etc. 

  • The 21th International Conference onNuclear Criticality Safety ICNC2023 International contribution

    ( 仙台 Japan ) 2023.10

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    Type:Competition, symposium, etc. 

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  • 高温ガス炉実証炉開発事業に係る評価委員

    資源エネルギー庁  2023.2 - 2024.3

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  • Screening of academic papers

    Role(s): Peer review

    2023

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:12

    Proceedings of International Conference Number of peer-reviewed papers:3

  • ANS working group member for ANSI/ANS standard titled “Initial Fuel Loading and Startup Tests for FOAK Advanced Reactors (ANSI/ANS-19.13-2023)”.

    American Nuclear Society  2022.10 - 2023.10

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  • Screening of academic papers

    Role(s): Peer review

    2022

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:2

  • 玄海原子力発電所の再稼働に関して広く意見を聴く委員会 原子力安全専門部会委員

    佐賀県  2021.7 - 2022.3

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  • 放射線利用技術等国際交流(講師育成)の審議・評価に係る国内運営委員会 委員長

    Role(s): Review, evaluation

    日本原子力研究開発機構  2021.5 - 2025.3

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    Type:Scientific advice/Review 

  • 放射線利用技術等国際交流(講師育成)の審議・評価に係る国内運営委員会 委員長

    日本原子力研究開発機構  2021.5 - 2022.3

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  • Screening of academic papers

    Role(s): Peer review

    2021

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:2

  • 放射線利用技術等国際交流(講師育成)専門部会 部会長

    Role(s): Review, evaluation

    日本原子力研究開発機構  2020.5 - 2021.3

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    Type:Scientific advice/Review 

  • Screening of academic papers

    Role(s): Peer review

    2020

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:5

  • Screening of academic papers

    Role(s): Peer review

    2019

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:8

  • 放射線利用技術等国際交流(講師育成)専門部会 部会長

    Role(s): Review, evaluation

    日本原子力研究開発機構  2018.4 - 2020.3

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    Type:Scientific advice/Review 

  • Screening of academic papers

    Role(s): Peer review

    2018

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    Type:Peer review 

    Number of peer-reviewed articles in foreign language journals:8

  • 放射線利用技術等国際交流(講師育成)専門部会 部会長

    Role(s): Review, evaluation

    日本原子力研究開発機構  2017.5 - 2018.3

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    Type:Scientific advice/Review 

  • その他 International contribution

    2017 International Congress on Advances in Nuclear Power Plants(ICAPP2017)  ( Japan ) 2017.4

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    Type:Competition, symposium, etc. 

  • Other International contribution

    2017 International Congress on Advances in Nuclear Power Plants (ICAPP2017)  ( kyoto Japan Japan ) 2017.4

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  • 日本原子力研究開発機構 研究嘱託

    日本原子力研究開発機構 大洗研究開発センター  2017.4 - 2025.3

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  • その他

    日本原子力学会 2016年秋の大会  ( Japan ) 2016.9

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    Type:Competition, symposium, etc. 

    Number of participants:700

  • その他

    日本原子力学会 春の年会  ( Japan ) 2014.3

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Research Projects

  • 高温ガス炉の燃焼挙動等に関する研究

    2015.12 - 2018.3

    Joint research

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    Authorship:Principal investigator  Grant type:Other funds from industry-academia collaboration

Class subject

  • 原子炉物理学Ⅱ

    2023.12 - 2024.2   Winter quarter

  • 教養の放射線科学と原子力Ⅱ

    2023.12 - 2024.2   Winter quarter

  • 連続体力学

    2023.12 - 2024.2   Winter quarter

  • 原子炉物理学Ⅰ

    2023.10 - 2023.12   Fall quarter

  • 教養の放射線科学と原子力Ⅰ

    2023.10 - 2023.12   Fall quarter

  • 原子炉システム工学Ⅱ

    2023.6 - 2023.8   Summer quarter

  • エネルギー科学卒業研究

    2023.4 - 2024.3   Full year

  • 課題集約演習

    2023.4 - 2024.3   Full year

  • 核エネルギーシステム学研究計画演習 A

    2023.4 - 2024.3   Full year

  • 核エネルギーシステム学実験 A

    2023.4 - 2024.3   Full year

  • 核エネルギーシステム学発表演習 A

    2023.4 - 2024.3   Full year

  • 産学連携演習Ⅰ

    2023.4 - 2024.3   Full year

  • 産学連携演習Ⅱ

    2023.4 - 2024.3   Full year

  • 産学連携演習Ⅲ

    2023.4 - 2024.3   Full year

  • 核エネルギーシステム学講究 A

    2023.4 - 2024.3   Full year

  • 産学連携実習

    2023.4 - 2024.3   Full year

  • Research Study in Industrial Fields

    2023.4 - 2024.3   Full year

  • 工学倫理(Ⅵ群)

    2023.4 - 2023.9   First semester

  • データ解析概論

    2023.4 - 2023.9   First semester

  • データ解析概論

    2023.4 - 2023.9   First semester

  • 原子力工学概論

    2023.4 - 2023.9   First semester

  • 量子物理工学演習Ⅱ

    2023.4 - 2023.9   First semester

  • 量子理工学演習Ⅱ

    2023.4 - 2023.9   First semester

  • 工学倫理(Ⅲ群)

    2023.4 - 2023.9   First semester

  • 原子炉システム工学Ⅰ

    2023.4 - 2023.6   Spring quarter

  • 量子物理工学A

    2023.4 - 2023.6   Spring quarter

  • 原子炉物理学Ⅱ

    2022.12 - 2023.2   Winter quarter

  • 教養の放射線科学と原子力Ⅱ

    2022.12 - 2023.2   Winter quarter

  • 連続体力学

    2022.12 - 2023.2   Winter quarter

  • Nuclear Reactor System Engineering

    2022.10 - 2023.3   Second semester

  • 量子物理工学演習Ⅰ

    2022.10 - 2023.3   Second semester

  • 原子炉物理学

    2022.10 - 2023.3   Second semester

  • 量子理工学演習Ⅲ

    2022.10 - 2023.3   Second semester

  • 原子炉物理学Ⅰ

    2022.10 - 2022.12   Fall quarter

  • 教養の放射線科学と原子力Ⅰ

    2022.10 - 2022.12   Fall quarter

  • 原子炉システム工学Ⅱ

    2022.6 - 2022.8   Summer quarter

  • Research Study in Industrial Fields

    2022.4 - 2023.3   Full year

  • 核エネルギーシステム学研究計画演習 A

    2022.4 - 2023.3   Full year

  • 核エネルギーシステム学実験 A

    2022.4 - 2023.3   Full year

  • 核エネルギーシステム学発表演習 A

    2022.4 - 2023.3   Full year

  • 産学連携演習Ⅰ

    2022.4 - 2023.3   Full year

  • 産学連携演習Ⅱ

    2022.4 - 2023.3   Full year

  • 産学連携演習Ⅲ

    2022.4 - 2023.3   Full year

  • 核エネルギーシステム学講究 A

    2022.4 - 2023.3   Full year

  • 産学連携実習

    2022.4 - 2023.3   Full year

  • 原子炉システム工学Ⅱ

    2022.4 - 2022.9   First semester

  • 原子力工学概論

    2022.4 - 2022.9   First semester

  • 量子理工学演習Ⅱ

    2022.4 - 2022.9   First semester

  • 量子理工学演習Ⅱ

    2022.4 - 2022.9   First semester

  • 工学倫理(Ⅲ群)

    2022.4 - 2022.6   Spring quarter

  • 量子物理工学A

    2022.4 - 2022.6   Spring quarter

  • 原子炉システム工学Ⅰ

    2022.4 - 2022.6   Spring quarter

  • 量子物理工学A

    2022.4 - 2022.6   Spring quarter

  • 原子炉システム工学Ⅰ

    2022.4 - 2022.6   Spring quarter

  • 量子物理工学A

    2022.4 - 2022.6   Spring quarter

  • 工学倫理(Ⅵ群)

    2022.4 - 2022.6   Spring quarter

  • 教養の放射線科学と原子力Ⅱ

    2021.12 - 2022.2   Winter quarter

  • 原子炉物理学Ⅱ

    2021.12 - 2022.2   Winter quarter

  • Nuclear Reactor System Engineering

    2021.10 - 2022.3   Second semester

  • 原子炉物理学

    2021.10 - 2022.3   Second semester

  • 量子理工学演習Ⅲ

    2021.10 - 2022.3   Second semester

  • 量子理工学演習Ⅰ

    2021.10 - 2022.3   Second semester

  • 教養の放射線科学と原子力Ⅰ

    2021.10 - 2021.12   Fall quarter

  • 量子物理工学A

    2021.10 - 2021.12   Fall quarter

  • 原子炉物理学Ⅰ

    2021.10 - 2021.12   Fall quarter

  • 量子物理工学A

    2021.10 - 2021.12   Fall quarter

  • 原子炉システム工学Ⅱ

    2021.6 - 2021.8   Summer quarter

  • 原子物理学

    2021.6 - 2021.8   Summer quarter

  • Research Study in Industrial Fields

    2021.4 - 2022.3   Full year

  • 核エネルギーシステム学研究計画演習 A

    2021.4 - 2022.3   Full year

  • 核エネルギーシステム学実験 A

    2021.4 - 2022.3   Full year

  • 核エネルギーシステム学発表演習 A

    2021.4 - 2022.3   Full year

  • 産学連携演習Ⅰ

    2021.4 - 2022.3   Full year

  • 産学連携演習Ⅱ

    2021.4 - 2022.3   Full year

  • 産学連携演習Ⅲ

    2021.4 - 2022.3   Full year

  • 産学連携実習

    2021.4 - 2022.3   Full year

  • 量子理工学演習Ⅱ

    2021.4 - 2021.9   First semester

  • 原子炉システム工学Ⅰ

    2021.4 - 2021.6   Spring quarter

  • 工学倫理(Ⅵ群)

    2021.4 - 2021.6   Spring quarter

  • 工学倫理(Ⅲ群)

    2021.4 - 2021.6   Spring quarter

  • 原子炉物理学Ⅱ

    2020.12 - 2021.2   Winter quarter

  • Nuclear Reactor System Engineering

    2020.10 - 2021.3   Second semester

  • 量子理工学演習Ⅲ

    2020.10 - 2021.3   Second semester

  • 量子理工学演習Ⅰ

    2020.10 - 2021.3   Second semester

  • 原子炉物理学

    2020.10 - 2021.3   Second semester

  • 原子炉物理学Ⅰ

    2020.10 - 2020.12   Fall quarter

  • 原子物理学

    2020.6 - 2020.8   Summer quarter

  • 原子物理学

    2020.6 - 2020.8   Summer quarter

  • 原子炉システム工学Ⅱ

    2020.6 - 2020.8   Summer quarter

  • Research Study in Industrial Fields

    2020.4 - 2021.3   Full year

  • 核エネルギーシステム学研究計画演習 A

    2020.4 - 2021.3   Full year

  • 核エネルギーシステム学実験 A

    2020.4 - 2021.3   Full year

  • 核エネルギーシステム学発表演習 A

    2020.4 - 2021.3   Full year

  • 産学連携演習Ⅰ

    2020.4 - 2021.3   Full year

  • 産学連携演習Ⅱ

    2020.4 - 2021.3   Full year

  • 産学連携演習Ⅲ

    2020.4 - 2021.3   Full year

  • 産学連携実習

    2020.4 - 2021.3   Full year

  • 量子理工学演習Ⅱ

    2020.4 - 2020.9   First semester

  • 原子力工学概論Ⅰ

    2020.4 - 2020.6   Spring quarter

  • 原子炉システム工学Ⅰ

    2020.4 - 2020.6   Spring quarter

  • 原子炉物理学Ⅱ

    2019.12 - 2020.2   Winter quarter

  • 原子炉物理学Ⅱ

    2019.12 - 2020.2   Winter quarter

  • 量子理工学演習Ⅲ

    2019.10 - 2020.3   Second semester

  • 原子炉物理学

    2019.10 - 2020.3   Second semester

  • Nuclear Reactor System Engineering

    2019.10 - 2020.3   Second semester

  • 量子理工学演習Ⅰ

    2019.10 - 2020.3   Second semester

  • 原子炉物理学Ⅰ

    2019.10 - 2019.12   Fall quarter

  • 原子炉物理学Ⅰ

    2019.10 - 2019.12   Fall quarter

  • 原子炉システム工学Ⅱ

    2019.6 - 2019.8   Summer quarter

  • 原子炉システム工学Ⅱ

    2019.6 - 2019.8   Summer quarter

  • 核エネルギーシステム学発表演習 A

    2019.4 - 2020.3   Full year

  • 核エネルギーシステム学研究計画演習 A

    2019.4 - 2020.3   Full year

  • 核エネルギーシステム学実験 A

    2019.4 - 2020.3   Full year

  • 原子炉システム工学Ⅰ

    2019.4 - 2019.6   Spring quarter

  • 原子炉システム工学Ⅰ

    2019.4 - 2019.6   Spring quarter

  • エネルギー量子工学基礎

    2018.10 - 2019.3   Second semester

  • 原子炉物理学

    2018.10 - 2019.3   Second semester

  • Nuclear Reactor System Engineering

    2018.10 - 2019.3   Second semester

  • 原子炉システム工学Ⅱ

    2018.6 - 2018.8   Summer quarter

  • 核エネルギーシステム学講究 A

    2018.4 - 2019.3   Full year

  • 核エネルギーシステム学研究計画演習 A

    2018.4 - 2019.3   Full year

  • 核エネルギーシステム学実験 A

    2018.4 - 2019.3   Full year

  • 核エネルギーシステム学発表演習 A

    2018.4 - 2019.3   Full year

  • 原子力工学概論

    2018.4 - 2018.9   First semester

  • 原子炉システム工学Ⅰ

    2018.4 - 2018.6   Spring quarter

  • 原子炉物理学

    2017.10 - 2018.3   Second semester

  • Nuclear Reactor System Engineering

    2017.10 - 2018.3   Second semester

  • エネルギー量子工学基礎

    2017.10 - 2018.3   Second semester

  • Nuclear Reactor System Engineering

    2017.10 - 2018.3   Second semester

  • 量子理工学演習Ⅰ

    2017.10 - 2018.3   Second semester

  • 量子理工学演習Ⅲ

    2017.10 - 2018.3   Second semester

  • 原子炉システム工学Ⅱ

    2017.6 - 2017.8   Summer quarter

  • Laboratory and Presentation for Nuclear Energy SystemsⅠ

    2017.4 - 2018.3   Full year

  • エネルギー量子工学研究企画演習

    2017.4 - 2018.3   Full year

  • エネルギー量子工学指導演習

    2017.4 - 2018.3   Full year

  • エネルギー量子工学特論

    2017.4 - 2018.3   Full year

  • Res. Plan. on Appl. Quantum Physics and Nuclear Engineering

    2017.4 - 2018.3   Full year

  • Teach.Practice in Appl.QuantumPhysics and NuclearEngineering

    2017.4 - 2018.3   Full year

  • Adv. Topics of Appl. Quantum Physics and Nuclear Engineering

    2017.4 - 2018.3   Full year

  • 核エネルギーシステム学実験 A

    2017.4 - 2018.3   Full year

  • Nuclear Energy Systems LaboratoryⅠ

    2017.4 - 2018.3   Full year

  • 核エネルギーシステム学発表演習 A

    2017.4 - 2018.3   Full year

  • 原子炉工学概論

    2017.4 - 2017.9   First semester

  • 原子力工学概論

    2017.4 - 2017.9   First semester

  • 原子炉システム工学Ⅰ

    2017.4 - 2017.6   Spring quarter

  • 原子炉物理

    2016.10 - 2017.3   Second semester

  • 量子理工学演習Ⅲ

    2016.10 - 2017.3   Second semester

  • Nuclear Reactor System Engineering

    2016.10 - 2017.3   Second semester

  • 原子炉システム工学Ⅰ

    2016.4 - 2016.9   First semester

  • 原子炉工学概論

    2016.4 - 2016.9   First semester

  • 原子炉システム工学Ⅱ

    2016.4 - 2016.9   First semester

  • 原子炉物理

    2015.10 - 2016.3   Second semester

  • 量子理工学演習Ⅰ

    2015.10 - 2016.3   Second semester

  • Nuclear Reactor System Engineering

    2015.10 - 2016.3   Second semester

  • 原子炉工学概論

    2015.4 - 2015.9   First semester

▼display all

FD Participation

  • 2022.3   Role:Participation   Title:メンタルヘルス講演会

    Organizer:University-wide

  • 2020.5   Role:Participation   Title:オンサイト授業 vs. オンライン授業:分かったこと,変わったこと

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2019.11   Role:Participation   Title:メンタルヘルス講演会

    Organizer:University-wide

  • 2019.3   Role:Participation   Title:平成33年度入学者選抜改革 【一般選抜における主体性等評価について】

    Organizer:University-wide

  • 2018.6   Role:Participation   Title:平成33年度入学者選抜改革【九州大学新入試QUBE】~各学部での検討に向けた情報提供~

    Organizer:University-wide

  • 2018.1   Role:Participation   Title:M2B(みつば)学習支援システム講習会および研究分析ツール「SciVal」及び研究者プロファイリングツール「Pure」に関する説明会(応用編)

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2017.8   Role:Participation   Title:本格的な共同研究に向けた取組み/東北大学工学研究科・工学部の現状と課題

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2017.2   Role:Participation   Title:QRECの教育プログラムの活用法

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2016.10   Role:Participation   Title:九州大学における男女共同参画の取り組み

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2016.8   Role:Participation   Title:ELITEプログラム研修報告

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2016.6   Role:Participation   Title:ハラスメント防止のための研修会

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2016.1   Role:Participation   Title:アイソトープ統合安全管理センターFD 国際規制物質の利用に関する教育講習会

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2016.1   Role:Participation   Title:東京工業大学の教育改革

    Organizer:[Undergraduate school/graduate school/graduate faculty]

  • 2015.11   Role:Participation   Title:第3回全学FD 自殺防止メンタルヘルス研修会 大学全体で行う自殺防止対策の実践に向けて

    Organizer:University-wide

  • 2015.4   Role:Participation   Title:新任教員の研修

    Organizer:University-wide

  • 2015.4   Role:Participation   Title:九州大学の国際戦略

    Organizer:University-wide

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Other educational activity and Special note

  • 2025  Lecture at Education Method and Practice  九大フィルハーモニーオーケストラ

     詳細を見る

    顧問

  • 2024  Coaching of Students' Association  九大フィルハーモニーオーケストラ

     詳細を見る

    顧問

  • 2023  Coaching of Students' Association  九大フィルハーモニーオーケストラ

     詳細を見る

    顧問

  • 2022  Coaching of Students' Association  九大フィルハーモニーオーケストラ

     詳細を見る

    顧問

  • 2020  Special Affairs  令和2年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

     詳細を見る

    令和2年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

  • 2019  Special Affairs  平成31年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

     詳細を見る

    平成31年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

  • 2018  Special Affairs  平成30年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

     詳細を見る

    平成30年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

  • 2017  Special Affairs  Lecture and discussion at a seminar of Soomkmyung Women's University in present and future status of energy of world and Japan, utilization of nuclear energy.

     詳細を見る

    Lecture and discussion at a seminar of Soomkmyung Women's University in present and future status of energy of world and Japan, utilization of nuclear energy.

  • 2017  Special Affairs  平成29年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

     詳細を見る

    平成29年度原子力規制人材育成事業「多角的思考力の要請と規制を加味した九州大学カリキュラムの充実」事業代表者

▼display all

Social Activities

  • エネルギーに関する講演

    Role(s):Lecturer

    熊本工業高校  2025.3

  • 放射線と原子力災害時の対応に関する講演

    Role(s):Lecturer

    鹿児島県  鹿児島県姶良市串木野  2025.2

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    Audience:General

    Type:Lecture

  • 放射線と原子力災害時の対応に関する講演

    Role(s):Lecturer

    鹿児島県  鹿児島県姶良市姶良  2025.2

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    Audience:General

    Type:Lecture

  • エネルギーに関する講演

    Role(s):Lecturer

    愛媛県松前町防災委員会  愛媛県松前町  2025.2

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    Type:Lecture

    愛媛県松前町の防災委員会総会において、世界のエネルギー需給とGXへの取り組みについて講演した。

  • エネルギーに関する講演

    Role(s):Lecturer

    九州エネルギー問題懇話会  大分県立中津東高等学校  2024.12

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    Audience:High school students

    Type:Lecture

    工業高校(電気科)の生徒向けにエネルギー供給と消費の現状について講演した。

  • エネルギーに関する講演

    Role(s):Lecturer

    マイナビ  探求型校内ガイダンス  筑前高校  2024.10

  • 放射線に関する講演と実験

    那珂川市立那珂川北中学校  2024.9

  • エネルギー・環境教育に関する意見交換会講師

    Role(s):Lecturer

    九州経済連合会 九州エネルギー問題懇話会  2024.8

  • マイナビ進学ライブでの霧箱の紹介と講義(鹿児島、別府、熊本、福岡)

    Role(s):Lecturer

    マイナビ  マイナビ進学イベント  2024.7

  • 電気主任技術者会第107回総会でのエネルギーに関する講演

    電気主任技術者会  埼玉県さいたま市  2024.6

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 鹿児島県原子力防災訓練での放射線に関する基礎知識と原子力災害時の対応についての講演

    Role(s):Lecturer

    電源地域振興センター  鹿児島県阿久根市  2024.2

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 福岡県消防学校での放射線防護に関する講義と実習

    福岡県消防学校  福岡県嘉麻市  2023.12

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Seminar, workshop

  • 電気主任技術者会 令和5年度施設研修会でのエネルギーに関する講演

    電気主任技術者会  鹿児島県鹿児島市  2023.11

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    Type:Lecture

    researchmap

  • 長崎県消防学校での放射線防護に関する講演と実習

    長崎県消防学校  長崎県大村市  2023.10

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    Type:Lecture

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  • 九州大学同窓会東京九機械総会での高温ガス炉に関する講演

    Role(s):Lecturer

    東京九機会  東京都千代田区  2023.9

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    Type:Lecture

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  • マイナビ進学ライブでの霧箱の紹介と講義(鹿児島、別府、熊本、福岡)

    マイナビ  2023.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 放射線に関する講義及び実験

    沖縄県竹富町立大原中学校  2023.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • マイナビ進学ライブでの霧箱の紹介と講義(鹿児島、別府、熊本、福岡)

    マイナビ  2023.7

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    Type:Seminar, workshop

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  • エネルギーに関する講演

    熊本県立東陵高校  2023.6

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギーに関する講演会

    株式会社 新出光  鹿児島県鹿児島市  2023.4

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • エネルギーに関する講演会

    古河電工  福岡県福岡市  2023.3

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Seminar, workshop

  • エネルギーに関する講演会

    日本建築材料協会九州支部  福岡県福岡市  2022.12

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • エネルギーに関する授業

    長崎県立五島高校  2022.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 放射線に関する講義及び実験

    福岡県立明善高等学校  2022.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 出前講義

    島根県立出雲高等学校  2022.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギーに関する授業

    長崎県立五島高校  2022.7

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    Type:Seminar, workshop

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  • 福岡エレコン交流会総会での講演

    福岡エレコン交流会  福岡県福岡市  2022.5

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 長崎県工業連合会 令和4年度定期総会 講演

    長崎県工業連合会  長崎県佐世保市  2022.5

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 長崎県工業連合会 令和4年度定期総会 講演

    長崎県工業連合会  長崎県佐世保市  2022.5

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    Type:Lecture

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  • 福岡エレコン交流会総会での講演

    福岡エレコン交流会  福岡県福岡市  2022.5

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    Type:Lecture

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  • 長崎県工業連合会全体交流会でのカーボンニュートラルに関する講演

    長崎県工業連合会  長崎県長崎市  2021.12

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 私たちの社会とエネルギーの現在と将来に向けての講演

    福岡県立筑紫丘高等学校  2021.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 長崎県工業連合会全体交流会でのカーボンニュートラルに関する講演

    長崎県工業連合会  長崎県長崎市  2021.12

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    Type:Lecture

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  • 私たちの社会とエネルギーの現在と将来に向けての講演

    福岡県立筑紫丘高等学校  2021.12

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    Type:Seminar, workshop

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  • 福岡女性管理職・リーダのためのオンライン勉強会での講演

    株式会社ビスネット  オンライン  2021.9

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 出前講義

    熊本県立八代高等学校  2021.9

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 福岡女性管理職・リーダのためのオンライン勉強会での講演

    株式会社ビスネット  オンライン  2021.9

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    Type:Lecture

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  • エネルギーと環境についての講演

    西日本政経懇話会  福岡県北九州市小倉北区  2021.4

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • エネルギーと環境についての講演

    西日本政経懇話会  福岡県久留米市  2021.4

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • エネルギーと環境についての講演

    西日本政経懇話会  福岡県飯塚市  2021.2

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • エネルギーと環境についての講演

    西日本政経懇話会  福岡県大牟田市  2021.1

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 低炭素電源としての原子力についての講演

    電気設備学会九州支部  福岡県福岡市  2020.12

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 低炭素電源としての原子力についての講演

    電気設備学会九州支部  福岡県福岡市  2020.12

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    Type:Lecture

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  • 私たちの生活とエネルギーについての講演

    福岡市立玄界中学校  2020.11

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギー消費と環境についての講演

    有明工業高等専門学校  2020.11

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 私たちの生活とエネルギーについての講演

    福岡市立玄界中学校  2020.11

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    Type:Seminar, workshop

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  • エネルギー消費と環境についての講演

    有明工業高等専門学校  2020.11

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    Type:Seminar, workshop

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  • スーパーサイエンスハイスクール講師

    福岡県立城南高校  2020.10

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Other

  • 出前講義

    島根県立松江北高等学校  2020.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギー政策に関する講義

    長崎県立大学  2019.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギー政策に関する講義

    長崎県立大学  2019.12

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    Type:Seminar, workshop

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  • 出前講義

    鹿児島県立鶴丸高校  2019.11

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギーに関する講演

    大分県立安心院高等学校  2019.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • スーパーサイエンスハイスクール講師

    福岡県立城南高等学校  2019.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Other

  • エネルギーと環境に関する講演

    大分県立鶴崎工業高校  2018.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 放射線に関する講演と実験

    鹿児島県立古仁屋高等学校  2018.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 放射線に関する講演と実験

    北九州工業高等専門学校  2018.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギーと環境に関する講演

    有明高専  2018.11

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 出前講義

    明治学園中学高等学校  2018.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 出前講義

    宮崎県立延岡高等学校  2018.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 日本のエネルギーの現状と課題

    九電福岡商友会  福岡県福岡市  2018.4

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 日本のエネルギーの現状と課題

    愛知時計電機株式会社LPガス事業者研修会  福岡県福岡市  2018.3

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 放射線に関する講義と演習

    北九州工業高等専門学校  2017.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 放射線に関する講義と演習

    北九州工業高等専門学校  2017.12

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    Type:Seminar, workshop

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  • 佐賀県原子力防災訓練での講師

    佐賀県  佐賀県みやき町  2017.9

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Lecture

  • 出前講義

    愛媛県立松山東高校  2017.9

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 第47回サイエンスカフェ福岡での講師

    公益財団法人九州経済調査会  福岡県福岡市  2017.6

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Other

  • スーパーサイエンスハイスクール講師

    福岡県立城南高等学校  2017.6

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Other

  • 第47回サイエンスカフェ福岡での講師

    公益財団法人九州経済調査会  福岡県福岡市  2017.6

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    Type:Other

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  • 放射線とエネルギーに関する講演

    熊本高等専門学校  2016.11

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • 放射線とエネルギーに関する講演

    熊本高等専門学校  2016.11

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    Type:Seminar, workshop

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  • スーパーサイエンスハイスクール講師

    福岡県立城南高等学校  2016.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Other

  • エネルギーと放射線に関する講演

    熊本高等専門学校  2015.12

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Seminar, workshop

  • エネルギーと放射線に関する講演

    熊本高等専門学校  2015.12

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    Type:Seminar, workshop

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  • 佐賀県原子力防災訓練での講師

    佐賀県  佐賀県有田町  2015.11

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    Audience:General, Scientific, Company, Civic organization, Governmental agency

    Type:Other

  • スーパーサイエンスハイスクール講師

    福岡県立城南高校  2015.7

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    Audience:Infants, Schoolchildren, Junior students, High school students

    Type:Other

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Media Coverage

  • 原子力の歩みとエネルギー

    九州エネルギー問題懇話会  教えて!エネルギー  2019.3

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    Author:Myself 

Activities contributing to policy formation, academic promotion, etc.

  • 2022.10 - 2023.10   American Nuclear Society

    ANS working group member for ANSI/ANS standard titled “Initial Fuel Loading and Startup Tests for FOAK Advanced Reactors (ANSI/ANS-19.13-2023)”.

  • 2022.4 - 2024.3   資源エネルギー庁

    高温ガス炉実証炉開発事業に係る評価委員

  • 2021.7 - 2022.3   佐賀県

    玄海原子力発電所の再稼働に関して広く意見を聴く委員会 原子力安全専門部会委員

  • 2020.8 - 2020.9   アカデミーオブフィンランド

    フィンランドフラッグシッププログラム2020 審査員

  • 2020.4 - 2026.3   九州電力

    原子力に係る安全性・信頼性向上委員会 委員

  • 2017.4 - 2027.4   日本原子力研究開発機構 大洗研究開発センター

    日本原子力研究開発機構 研究嘱託

  • 2016.4 - 2027.3   日本原子力研究開発機構

    放射線利用技術等国際交流(講師育成)の審議・評価に係る国内運営委員会 委員長

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Acceptance of Foreign Researchers, etc.

  • Nuclear Regulatory Agency of Indonesia

    Acceptance period: 2024.9 - 2024.12   (Period):1 month or more

    Nationality:Indonesia

    Business entity:Ministry of education

Travel Abroad

  • 1996.3 - 1997.3

    Staying countory name 1:Germany   Staying institution name 1:Forschungszentrum Juelich