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JAEA Reports

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2021)

Department of HTTR

JAEA-Review 2023-016, 82 Pages, 2023/09

JAEA-Review-2023-016.pdf:2.31MB

The High Temperature Engineering Test Reactor (HTTR) is the first Japanese High Temperature Gas-cooled Reactor (HTGR) with 30MW in thermal power and 950$$^{circ}$$C of maximum outlet coolant temperature that is constructed by the Japan Atomic Energy Agency located at Oarai-machi, Higashiibaraki-gun, Ibaraki-ken, Japan. The purpose of the HTTR is establishment of basic HTGR technologies, demonstration of HTGR safety characteristics and so on. The HTTR has had a lot of experience of HTGRs' operation and maintenance throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2021, as the HTTR completed activities to conform to the New Regulatory Requirements of Nuclear Regulation Authority, The HTTR restarted since the 2011 off the Pacific coast of Tohoku Earthquake and carried out the Loss-of-forced cooling test without Vessel Cooling System (VCS) operational at 9MW (Three gas circulators trip and VCS is stopped.) as the safety demonstration test. This report summarizes the activities carried out in the fiscal year 2021, which were the situation of the New Regulatory Requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.

JAEA Reports

Study of fabrication of SiC-matrixed fuel compact for HTGR

Kawano, Takahiro*; Mizuta, Naoki; Ueta, Shohei; Tachibana, Yukio; Yoshida, Katsumi*

JAEA-Technology 2023-014, 37 Pages, 2023/08

JAEA-Technology-2023-014.pdf:2.35MB

Fuel compact for High Temperature Gas-cooled Reactor (HTGR) is fabricated by calcinating a matrix consisting of graphite and binder with the coated fuel particle. The SiC-matrixed fuel compact uses a new matrix made of silicon carbide (SiC) replacing the conventional graphite. Applying the SiC-matrixed fuel compact for HTGRs is expected to improve their performance such as power densities. In this study, the sintering conditions for applying SiC as the matrix of fuel compacts for HTGR are selected, and the density and thermal conductivity of the prototype SiC are measured.

Journal Articles

Reactor physics experiment on a graphite-moderated core to construct integral experiment database for HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Research on improvement of HTGR core power-density, 4; Feasibility study for a reactor core

Okita, Shoichiro; Mizuta, Naoki; Takamatsu, Kuniyoshi; Goto, Minoru; Yoshida, Katsumi*; Nishimura, Yosuke*; Okamoto, Koji*

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

Journal Articles

Validation of evaluation model for analysis of steam reformer in HTGR hydrogen production plant

Ishii, Katsunori; Aoki, Takeshi; Isaka, Kazuyoshi; Noguchi, Hiroki; Shimizu, Atsushi; Sato, Hiroyuki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

Journal Articles

Improvement of cooling performance of reactor pressure vessel using passive cooling

Banno, Masaki*; Funatani, Shumpei*; Takamatsu, Kuniyoshi

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05

A fundamental study on the safety of a passive cooling system for the RPV with radiative cooling is conducted. The object of this study is to demonstrate that passive RPV cooling system with radiative cooling is extremely safe and reliable even in the event of natural disasters. Therefore, an experimental apparatus, which is about 1/20 scale of the actual cooling system, was fabricated with several stainless steel containers. The surface of the heating element in the experimental apparatus simulates the surface of the RPV, and the heating element generates natural convection and radiation. A comparison of the Grashof number between the actual cooling system and the experimental apparatus confirmed that both were turbulent, and the experimental results as a scale model are valuable. Moreover, the experimental results confirmed that the heat generated from the surface of the RPV during the rated operation can be removed.

Journal Articles

Comparison on safety features among HTGR's Reactor Cavity Cooling Systems (RCCSs)

Takamatsu, Kuniyoshi; Funatani, Shumpei*

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 17 Pages, 2023/04

The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Therefore, the authors concluded that the proposed RCCS based on atmospheric radiation has the advantage that the temperature of the RPV can be stably maintained against disturbances in the outside air (ambient air). Moreover, methodology to utilize all the heat emitted from the RPV surface for increasing the degree of waste-heat utilization was discussed.

Journal Articles

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

Simanullang, I. L.*; Nakagawa, Naoki*; Ho, H. Q.; Nagasumi, Satoru; Ishitsuka, Etsuo; Iigaki, Kazuhiko; Fujimoto, Nozomu*

Annals of Nuclear Energy, 177, p.109314_1 - 109314_8, 2022/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Present status of JAEA's R&D toward HTGR deployment

Shibata, Taiju; Nishihara, Tetsuo; Kubo, Shinji; Sato, Hiroyuki; Sakaba, Nariaki; Kunitomi, Kazuhiko

Nuclear Engineering and Design, 398, p.111964_1 - 111964_4, 2022/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Japan Atomic Energy Agency (JAEA) has been promoting the research and development (R&D) of High Temperature Gas-cooled Reactor (HTGR). R&D on reactor technologies is carried out by using High Temperature engineering Test Reactor (HTTR). The HTTR was resumed without significant reinforcements in 2021. On January 2022, a safety demonstration test under the OECD/NEA LOFC project was carried out. JAEA is promoting R&D on a carbon-free hydrogen production by thermochemical water splitting Iodine-Sulfur process (IS process). JAEA conducts design study for various HTGR systems toward commercialization. A new test program about demonstration of hydrogen production by the HTTR was launched. Steam methane reforming hydrogen production system was selected for the first demonstration by 2030.

Journal Articles

Study on heat transfer characteristics of reactor cavity cooling system using radiation

Banno, Masaki*; Funatani, Shumpei*; Takamatsu, Kuniyoshi

Yamanashi Koenkai 2022 Koen Rombunshu (CD-ROM), 6 Pages, 2022/10

A fundamental study on the safety of a passive cooling system for the reactor pressure vessel (RPV) with radiative cooling is conducted. The object of this study is to demonstrate that passive RPV cooling system with radiative cooling is extremely safe and reliable even in the event of natural disasters. Therefore, an experimental apparatus, which is about 1/20 scale of the actual cooling system, was fabricated with several stainless steel containers. The surface of the heating element in the experimental apparatus simulates the surface of the RPV, and the heating element generates natural convection and radiation. As a result of the experiments, we succeeded in visualizing the natural convection in the experimental apparatus in detail.

Journal Articles

MCNP6 calculation of neutron flux map in the HTTR during normal operation

Ho, H. Q.; Ishitsuka, Etsuo; Iigaki, Kazuhiko

Recent Contributions to Physics, 82(3), p.16 - 20, 2022/09

Journal Articles

Study on the effect of long-term high temperature irradiation on TRISO fuel

Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sakaba, Nariaki

Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08

 Times Cited Count:4 Percentile:88.09(Nuclear Science & Technology)

Journal Articles

Preliminary experiment in a graphite-moderated core to avoid full mock-up experiment for the future first commercial HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05

Journal Articles

Design of a portable backup shutdown system for the high temperature gas cooled reactor

Hamamoto, Shimpei; Ho, H. Q.; Iigaki, Kazuhiko; Goto, Minoru; Shimazaki, Yosuke; Sawahata, Hiroaki; Ishitsuka, Etsuo

Nuclear Engineering and Design, 386, p.111564_1 - 111564_8, 2022/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The experience of Fukushima Daiichi Nuclear Power Plant accident caused by the great earthquake that occurred in eastern Japan in 2011 showed the importance of preparing for the loss of function of the engineered safety features. Increasing the strength of equipment to prevent loss of function in an accident is effective, but the possibility of loss of function remains. Therefore, it is important to have an alternative to lost functions in order to put the accident under control early. Thus, this study designed an alternative shutdown system, namely a portable backup shutdown system (PBSS), to make countermeasures in the event of a loss of shutdown function more robust without impairing economic efficiency of the High Temperature Gas-cooled Reactor (HTGR). The PBSS is portable and capable of being installed manually so that it can operate in a total loss of off-site electricity. Various neutron absorber materials for the PBSS were also considered from the viewpoints of technical and cost-effective properties. As results of optimization, the boron nitride (BN) was selected as it shows a good neutronic property as well as a reasonable cost in comparison with other materials.

Journal Articles

Seismic classification of high temperature engineering test reactor

Ono, Masato; Shimizu, Atsushi; Ohashi, Hirofumi; Hamamoto, Shimpei; Inoi, Hiroyuki; Tokuhara, Kazumi*; Nomoto, Yasunobu*; Shimazaki, Yosuke; Iigaki, Kazuhiko; Shinozaki, Masayuki

Nuclear Engineering and Design, 386, p.111585_1 - 111585_9, 2022/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In the late 1980s during the design stage, the seismic classification of the high temperature engineering test reactor (HTTR) was formulated. Owing to the lack of operation experiences of the HTTR to sufficiently understand the safety characteristics of high temperature gas cooled reactors (HTGR) at that time, the seismic classification of commercial light water reactors (LWR) was applied to HTTR. However, the subsequent operation experiences and test results using HTTR made it clear that the seismic classification of commercial LWR was somewhat too conservative for the HTGR. As a result, Class S facilities were downgraded compared to the commercial LWR. Moreover, the validity of the new seismic classification is confirmed. In June 2020, the Nuclear Regulatory Authority approved that the result of the seismic classification conformed to the standard rules of the reactor installation change.

JAEA Reports

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2019)

Department of HTTR

JAEA-Review 2021-017, 81 Pages, 2021/11

JAEA-Review-2021-017.pdf:2.53MB

The High Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas cooled Reactor (HTGR) constructed in Japan at the Oarai Research and Development Institute of the Japan Atomic Energy Agency with 30MW in thermal power and 950$$^{circ}$$C of outlet coolant temperature. The purpose of the HTTR is to establish and upgrade basic technologies for HTGRs. The HTTR has accumulated a lot of experience of HTGRs' operation and maintenance up to the present time throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2019, we continued to make effort to restart operations of the HTTR that stopped since the 2011 off the Pacific coast of Tohoku Earthquake. It is necessary for the HTTR reoperation to prove conformity with the new regulatory requirements for research reactors enacted in December 2013. So we might cope with government agency to pass the inspection of application document for the HTTR licensing. This report summarizes the activities carried out in the fiscal year 2019, which were the situation of the new regulatory requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.

Journal Articles

Comparisons between passive RCCSs on degree of passive safety features against accidental conditions and methodology to determine structural thickness of scaled-down heat removal test facilities

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 162, p.108512_1 - 108512_10, 2021/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Next, simulations on accidental conditions, such as increasing average heat-transfer coefficient via natural convection due to natural disasters, were performed with STAR-CCM+, and methodology to control the amount of heat removal was discussed. As a result, a new RCCS based on atmospheric radiation is recommended because of the excellent degree of passive safety features/conditions, and the amount of heat removal by heat transfer surfaces which can be controlled. Finally, methodology to determine structural thickness of scaled-down heat removal test facilities for reproducing natural convection and radiation was developed, and experimental methods by using pressurized and decompressed chambers was also proposed.

Journal Articles

Study on chemical form of tritium in coolant helium of high temperature gas-cooled reactor with tritium production device

Hamamoto, Shimpei; Ishitsuka, Etsuo; Nakagawa, Shigeaki; Goto, Minoru; Matsuura, Hideaki*; Katayama, Kazunari*; Otsuka, Teppei*; Tobita, Kenji*

Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), 5 Pages, 2021/10

Impurity concentrations of hydrogen and hydride in the coolant were investigated in detail for the HTTR, a block type high-temperature gas reactor owned by Japan. As a result, it was found that CH$$_{4}$$ was 1/10 of H$$_{2}$$ concentration, which was under the conventional detection limit. If the ratio of H$$_{2}$$ to CH$$_{4}$$ in the coolant is the same as the ratio of HT to CH$$_{3}$$T, the CH$$_{3}$$T has a larger dose conversion factor, and this compositional ratio is an important finding for the optimal dose evaluation. Further investigation of the origin of CH$$_{4}$$ suggested that CH$$_{4}$$ was produced as a result of a thermal equilibrium reaction rather than being released as an impurity from the core.

JAEA Reports

Impact assessment for internal flooding in HTTR (High temperature engineering test reactor)

Tochio, Daisuke; Nagasumi, Satoru; Inoi, Hiroyuki; Hamamoto, Shimpei; Ono, Masato; Kobayashi, Shoichi; Uesaka, Takahiro; Watanabe, Shuji; Saito, Kenji

JAEA-Technology 2021-014, 80 Pages, 2021/09

JAEA-Technology-2021-014.pdf:5.87MB

In response to the new regulatory standards established in response to the accident at TEPCO's Fukushima Daiichi Nuclear Power Station in March 2011, measures and impact assessments related to internal flooding at HTTR were carried out. In assessing the impact, considering the characteristics of the high-temperature gas-cooled reactor, flooding due to assumed damage to piping and equipment, flooding due to water discharge from the system installed to prevent the spread of fire, and flooding due to damage to piping and equipment due to an earthquake. The effects of submersion, flooding, and flooding due to steam were evaluated for each of them. The impact of the overflow of liquids containing radioactive materials outside the radiation-controlled area was also evaluated. As a result, it was confirmed that flooding generated at HTTR does not affect the safety function of the reactor facility by taking measures.

Journal Articles

Concepts and basic designs of various nuclear fuels, 5; Fuels for high temperature gas-cooled reactor and molten salt reactor

Ueta, Shohei; Sasaki, Koei; Arita, Yuji*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(8), p.615 - 620, 2021/08

no abstracts in English

661 (Records 1-20 displayed on this page)